Text

Exam 50-369, 370/2000-301 Draft Submittal SRO Written Exam Answer Key Version
	ML020450389
Person / Time
Site:	McGuire, Mcguire
Issue date:	02/12/2002
From:	Michael B; Division of Reactor Safety II
To:	;
References
	-nr 50-369/00301, 50-370/00301
	Download: ML020450389 (139)
	v • d • e

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McGuire Exam 50-369, 370/2000-301 DATE OF REQUEST February 12, 2002 DATE NEEDED DRAFT SubmDittal KU Wrvten ExamFr Answer Key Version February 15, 2002 SENSITIVITY:

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NRC Official Use Only Nuclear Regulatory Commission Senior Reactor Operator Licensing Examination McGuire Nuclear Station This document is removed from Official Use Only category on Date of examination NRC Official Use Only

SRO Exam McGuire Nuclear Station Bank Question: 51 Answer: C I Pt(s)

A large break LOCA is in progress and the operators are responding in E-I (Reactor Trip or Safety Injection). Given the following conditions:

ND pump IA is tagged out of service for maintenanc * Containment pressure is 14 psi * FWST level is below the swap over setpoin When shifting to cold leg recirc using ES-1.3 (Transfer to Cold Leg Recirc),

valve 1NI-184B (RB Sump to Train lB ND & NS) fails to open. The operators implement ECA-1.1 (Loss of Emergency Coolant Recirculation).

FR-Z. 1 (Response to High Containment Pressure) requires both NS pumps to be in operation. ECA-1.1 limits the operators to only one NS pump in step 11. Which of these two procedures takes priority under these conditions and what is the basis for this requirement? FR-Z.1 takes priority because a total loss of ND causes the NS system to become relatively more important to reduce containment pressur FR-Z.1 takes priority because it was implemented in response to a red path and FRPs always have priority over ECA procedure ECA-1.1 takes priority because it conserves FWST water level as long as possible for injection while providing sufficient NS flow to mitigate containment pressur ECA-1.1 takes priority because ECA procedures always have priority over FRP....

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Distracter Analysis: Incorrect: ECA-1.1 takes priority over FR-Plausible: Although a loss of ND and containment sump recirc causes a loss of the containment heat sink, the supply for NS comes from the FWST which will be drawn down until containment sump recirculation can be establishe Incorrect: ECA-1.1 takes priority over FR-Z. I Plausible: FRPs normally take priority over most EOPs Correct answer Incorrect: ECAs do not always have priority over FRPs.

Page 1 For Official Use Only Question #1 QuesLO51

SRO Exam McGuire Nuclear Station Plausible: Some ECAs take priority e.g. ECA-0.0 has priority over FRPs in that F-0 is not applicable until transition out of ECA-For Official Use Only Page 2 QuesO051 Question #1

SRO Exam McGuire Nuclear Station Bank Question: 60 Answer: A 1 Pt(s)

Unit 2 was operating at 100% power when an electrical fire started inside the auxiliary building cable spreading room. What type of fire suppression system is installed inside the cable spreading area and what are the hazards to personnel if they enter this room? A manual deluge (Mulsifyre) System is installed. An electrical shock hazard exists due to the use of water to combat an electrical fir An automatic sprinkler system is installed. An electrical shock hazard exists due to the use of water to combat an electrical fir An automatic Halon system is installed. An asphyxiation hazard exists due to the presence of Halon ga A manual Cardox system is installed. An asphyxiation hazard exists due to the presence of carbon dioxide ga...

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Distracter Analysis: Correct Answer: Incorrect: A manual deluge Mulsifyre system is installed Plausible: an electrical shock hazard exists Incorrect: A manual deluge Mulsifyre system is installed Plausible: Halon gas is generally used in areas in which electrical fires are the predominant risk because it does not create a shock hazard Incorrect: A manual deluge Mulsifyre system is installed Plausible: Cardox gas is a personnel hazard - although all the CARDOX systems have been replaced with HALON, the pull switches still say CARDOX in some areas (like the diesel generators)

C.r f^lrilI lice Onlv Page 3 QueS._OUU Question #2 171JI VIIl*lll*ll V*'*l" vnJ"

SRO Exam McGuire Nuclear Station Bank Question: 82 Answer. D I Pt(s)

Unit 2 is recovering from a loss of 120 VAC instrument bus 2EKVA due to the loss of the 2EVIA static inverter. 2EKVA has been reenergized from the alternate supply. After repairs to inverter 2EVIA are completed, the operator is directed to restore the 2EKVA bus to the normal line u Which one of the following actions is necessary to restore the electrical lineup to a normal operating configuration after tags are cleared? Manually transfer bus power from static inverter 2EVIB back to static inverter 2EVI Enable the automatic transfer of power from static inverter 2EVIB back to 2EVI Enable the automatic transfer of power from regulated power center 2KRP back to static inverter 2EVI Manually transfer power from regulated load center 2KRP back to static inverter 2EVI..........

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Distracter Analysis: Incorrect: Power is not transferred to 2EVIB - this is the normal power supply to 2EKVB Plausible: If the candidate thinks that 2EKIB can be used to supply 2EKVA Incorrect: There is no automatic transfer associated with these static inverters Plausible: There are automatic bus transfers for some of the 120 VAC power supply breakers Incorrect: There is no automatic transfer between 2KRP and 2EVIA Plausible: 2KRP is the correct alternate supply for 2EKVB if 2EVIA is not operating and there is an auto transfer switch between the normal and alternate power supplies for 2KR Correct answer Page 4 For Official Use Only Question #3 Ques_082

SRO Exam McGuire Nuclear Station Bank Question: 97 Answer: B I Pt(s)

Unit 1 is in the process of preparing to conduct a plant cooldown in Mode 3 in preparation for a refueling period. The OSM denied a request from maintenance to tag shut IND-30A (TRAIN A ND TO HOT LEG ISOL) for valve stroke time testin What is the reason for his decision? Although this would be permitted in mode 3, closing 1ND-30A would cause one ND train to be inoperable and is prohibited in Mode This action would isolate the cross tie between the ND trains which is assumed to be open in the FSAR ensuring one ND pump can inject into all four NC loop IND-15B is interlocked with 1NI-136B (B NI PUMP SUCTION FROM ND) to provide an alternate boration path. Closure of IND-30A will defeat this interloc ND-30A is interlocked with 1ND-58A (TRAIN A ND TO NV &

NI PUMPS) and this will defeat the interlock and prevent IND 58A from opening and establishing an alternate boration path from the NV system

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Distracter Analysis: Incorrect: permitted in mode 4 Plausible: based on transposition of operating requirements Correct Answer: Incorrect: IND-I 58 is not interlocked with INI-136B Plausible: based on misunderstanding of the 1NI-136B interlock Incorrect: IND-30A is not interlocked with 1ND-58 Plausible: based on misunderstanding of the IND-58A interlock Page 5 For Official Use Only Question #4 Ques_097

SRO Exam McGuire Nuclear Station Bank Question: 120 Answer. C I Pt(s)

Unit 1 was operating at 100% power. Given the following motor driven auxiliary feedwater pump operating parameters:

0200Q 0210 0220

Discharge Pressure (ft water) 3325 3325 3010 2950 Suction Pressure (ft water)

75

75 Pump flow rate (gpm)

420 480 520 560 What is the onset (earliest time) of pump cavitation conditions?

REFERENCES PROVIDED Curve 8.4 of enclosure 4.3 to OP/1/A/6100/22 Distracter Analysis: Incorrect: - cavitation has not yet occurred: 3325 - 75 = 3250 in Plausible: - below pump characteristic curve but above NPSH requirement Incorrect: - cavitation has not yet occurred: 3325-75 = 3250 Plausible: - point is above the pump characteristic curve - but still below NPSH curve Correct answer: below NPSH curve 3010 - 75 = 2935 in Incorrect: cavitation has already occurred Plausible: - - if the candidate does not remember to subtract the suction pressure from the discharge pressure: 3010 from answer C is above the NPSH curve

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SRO Exam McGuire Nuclear Station Question #6 Bank Question: 125 Answer. C 1 Pt(s)

A worker is preparing to enter a high radiation area to work on a valve in the reactor building. During the pre-job briefing, RP states that the expected whole body radiation level are as follows:

"* Dose rate in the center of the room 20 ft away = 200 mrem/hr

"* Dose rate 18 inches from valve = 700 mrem/hr

"* Contact reading = 1100 mrem/hr How should the area around the valve be classified? The room is a radiation area; the valve is a hot spot The room is a high radiation area; valve is NOT a hot spot The room is a high radiation area; the valve is a hot spot The room is an extra high radiation area; the valve is NOT a hot spot

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Distracter Analysis: A hot spot is an area where the dose rate on contact is

> 5x general area radiation but > 100 mremnhr. In this case 5 x 200 = 1000 mrem/hr < 1100 mrem/hr on contac Incorrect: 200 mrem/hr general area dose rate > 100 mrem/hr =

high radiation area Plausible: if the candidate does not know that the lower limit for a high radiation area is 100 mrem/hr - and the valve is a hot spot Incorrect: The valve is a hot spot Plausible: the room is a high radiation area - if the candidate thinks that the definition of a hot spot is > 5x general area dose rate when measured 18 inches from the contact reading Correct Answer: Incorrect: The room is not an extra high radiation area Plausible: if the candidate thinks that the definition of a hot spot is

> 5x general area dose rate when measured 18 inches from the contact reading.s. Lb l

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Question #7 McGuire Nuclear Station SRO Exam Bank Question 152 Answer. C I Pt(s)

The operators are conducting a reactor startu Given the following indications on the source range (SR) and intermediate range (IR) excore nuclear instruments:

Time 0200 0205 0210 0215 SR "A" (cps)

1.5x10 4 2.5x10 4 2.8xl 04 1.0x10 5 SR "B" (cps)

1.4x10 4 2.3xl 04 2.7x104 9.8x10 4 IR "A" (amps)

7.6x10'

L.1xl0-10 1.5x10-`

7.0xl0-° IR "B" (amps)

7.9x10-&"

9.0xl0-1 1.lxl0-1° 7.5x10 0 What is the earliest time that the operators should block the source range nuclear instruments?...

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Distracter Analysis:

The objective behind this question is to determine if the candidate can differentiate between when they CAN block SR high flux (because P-6) is in and when they are ALLOWED to block SR high flux - after observing

"proper" overlap between IR and SR - i.e. one decade. They will not observe one decade of overlap until both IR NI channels are > I E-10 amps because they come on scale at IE-I I amp Incorrect: < IE-10 amps in both IR channels Plausible: Source range nuclear instruments should be blocked by the time the level is 10 CPS by the training material Incorrect: only IR A channel has reached 1 decade of observed overlap with the SR Plausible: this is when P6 is in and they can physically block SR high flux Correct answer: proper overlap has been observed on IR B Page 8 For Official Use Only Ques_152

Question #7 Ques_152 SRO Exam McGuire Nuclear Station Incorrect: - will reach the SR high flux trip setpoint at IEl0 cps Plausible: - if the candidate does not know P6 or if he is confusing the high flux trip setpoint with the P6 setpoint For Official Use Only Page 9

SRO Exam McGuire Nuclear Station Bank Question: 191 Answer: B I Pt(s)

Unit I was operating at 100% power when a total loss of feedwater occurred. The operators reached step 35 of FR-H.I(Response to Loss of Secondary Heat Sink) which states:

IFA TANY TIME while in this procedure any SIG W/R level goes below 12% (17% A CC), THEN G M9 Enclosure 10 (Hot/Dry Steam Generator Limits)

Given the following conditions:

Loo Lo LioopC Loop D S/G (WR)[%]

15

10 NC T1.0, [TF]

150 555 530 545

"* Containment pressure is 3.4 psig

"* The TD CA pump is available to feed the S/Gs Which one of the following statements correctly describes the bases for the restrictions for restoring feedwater flow following feed and bleed in FR-H. I? Restore flow to the A S/G because loop A T-hot is the lowest of the loops and this will reduce the chance of thermal shocking the S/G tube sheet. Flow should not be restored to the B and C S/Gs because they will be reserved for use later to provide a steam supply for the TD CA pum Restore flow to the B S/G because B S/G level is the highest and this will reduce the chance of thermal shocking the S/G tube sheet. Flow should be preferentially restored to the B or C S/G to maintain the TD CA pump steam suppl Restore flow to the C S/G because loop C T-hot is less than loop B T-hot and this will reduce the chance of thermal shocking the S/G tube sheet. Flow should be preferentially restored to the B or C S/G to maintain the TD CA pump steam suppl Restore flow to the D S/G because the D S/G is higher than A S/G level, which will reduce the risk of thermal shock. Flow should not be restored to the B and C S/Gs because they will be reserved for use later to provide a steam supply for the TD CA pum....

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Page 10 For Official Use Only Question #8 Ques_191

SRO Exam McGuire Nuclear Station Distracter Analysis: There was a change in this procedure since the last NRC exam. The previous guidance was not to feed a S/G when T hot > 550 'F and to select the B and C S/Gs for restoration of flo Now the guidance is to select the S/G that has the highest apparent level and to preferentially select the B or C S/ Incorrect: T-hot should not be used to determine which S/G should receive flow. It is not a reliable means of determining S/G shell temp in a dry stagnant loo Plausible: The apparent temp of the A loop is the lowest and it may appear that the chance of thermal shock is lessene Correct answer: feed the S/G that has the highest level and preferentially feed B & C S/Gs to maintain steam supply to the TD CA pum Incorrect: C S/G has a lower S/G level than B S/G Plausible: C S/G has a lower T-hot than B S/G Incorrect: No basis for reserving the B & C S/Gs for restoring flow Plausible: There is a high probability that restoring feed to a dry S/G could rupture the tube sheet due to thermal stress. It makes sense to select a S/G that is NOT used to supply steam to the TD CA pump for the initial restoration of the heat sink.

Page 11 For Official Use Only Question #8 Ques_191

SRO Exam McGuire Nuclear Station Bank Question: 207 Answer. B I Pt(s)

Unit I is preparing for a reactor start up following a refueling outage. Given the following conditions:

"* Tavg = 515 OF

"* Plant heatup in progress using NCPs At 0200, a Station Engineer reports that a mistake had been made in analyzing the containment Appendix J Leak Rate Test results that were conducted prior to exceeding 200 0F in Mode 5. Reanalysis indicated that the combined containment leak rate (Type A) had exceeded 1.0 L Which one of the following actions is required by Tech Specs in response to this situation?

REFERENCES PROVIDED Tech Spec 3.6.1PAGES 1, 2 Tech Spec Bases B3.6.1-3, 4 Commence a plant cooldown to reach Mode 5 within 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months Commence a plant cooldown to reach Mode 5 within 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Commence a plant cooldown to reach Mode 5 within 37 hour4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months Commence a plant cooldown to reach Mode 5 within 43 hour4.976852e-4 days 0.0119 hours 7.109788e-5 weeks 1.63615e-5 months ...

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Distracter Analysis:

This question tests the application of ITS completion times. Tech Spec 3.6.1 allows 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore containment to an operable status, 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to be in mode 3 and 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to be in mode 5. However, the plant is currently in mode 3 - but this does not change the 36 hour4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months completion time to enter mode 5. This represents a change over the old method of computing completion time Incorrect: 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is allowable Plausible: if the candidate deducts the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to be in mode 3 from the total 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to enter mode 5 - would have been a correct method of determining completion time under old tech specs Incorrect: 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is allowable Plausible: if the candidate takes the total 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to enter mode 5 and adds in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore containment to operability - would have Page 12 For Official Use Only Question #9 Ques 207

SRO Exam McGuire Nuclear Station been a correct method of determining completion time under old tech specs Correct answer ITS completion times are different than old Tech Specs D Incorrect: 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is allowable Plausible: if the candidate adds the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to enter mode 3 to the 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to reach mode 5 and adds 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore containment to operability. This would have been the old Tech Spec method of doing completion times if the candidate made the mistake of NOT recognizing he/she was already in mode 3.

Page 13 For Official Use Only Question #9 Ques_207

SRO Exam McGuire Nuclear Station Bank Question: 216 Answer: B 1 Pt(s)

Which one of the following describes a responsibility associated with the Fuel Handling SRO during Fuel Handling operations? Operate Fuel Handling Equipment, in accordance with approved procedure(s). Directly observe Fuel Handling activities from the reactor building operating dec Physically latch/unlatch each fuel assembly, in accordance with approved procedure(s). Supervise reactor vessel in-service inspections immediately after core off-loa...

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Distracter Analysis: Incorrect: Fuel Handling SRO does not operate the FH equipment Plausible: The FH SRO is responsible to supervise the operation of the FH equipment Correct Answer: Incorrect: Latching/unlatching fuel assemblies conducted by fuel handlers Plausible: FH SRO supervises this operation Correct answer:

Page 14 For Official Use Only Question #10 Ques_216

SRO Exam McGuire Nuclear Station Bank Question: 217 Answer. A I Pt(s)

Unit 1 was operating at 100% power in Mode 1. Given the following conditions:

2 main steam safety valves (MSSVs) on the 1D S/G have been gagged shut to prevent chattering Which one of the following statements describes the required action(s) and a basis for these actions?

REFERENCES PROVIDED Tech Spec 3.7.1 pages 1-3 Reduce power below 39% to ensure that the reactor coolant pressure boundary is not over-pressurize Reduce power below 39% to ensure that the positive reactivity effect on NCS cooldown associated with the operation of the main steam system safety valves is minimize Reduce power below 19% to ensure that the reactor coolant pressure boundary is not over-pressurize Reduce power below 19% to ensure that the positive reactivity effect on NCS cooldown associated with the operation of the main steam system safety valves is minimize...................................................------------------------------.........

Distracter Analysis: Correct Answer: Incorrect: incorrect basis Plausible: power level is correct Incorrect: power level too low - 39% is correct Plausible: basis is correct, will select 19% power if they look up tech spec actions required for 2 safety valves operable (rather than 2 inoperable) Incorrect: power level too low - 39% is correct, basis is incorrect Plausible: opening MSSVs will cause a positive reactivity transient

- will select 19% power if they look up tech spec actions required for 2 safety valves operable (rather than 2 inoperable)

Page 15 For Official Use Only Question #11 Ques_217

Question #12 McGuire Nuclear Station SRO Exam Bank Question: 241 Answer: D I Pt(s)

Unit 1 is operating at 100% power when the supply breaker from ILXG to Control Rod Drive MG set #2 opens. Which one of the following sequence of events will occur to the reactor trip breakers A or B (RTA/B) and the reactor trip bypass breakers A or B (BYA/B)? RTA and BYB will open RTB and BYA will open BYA and BYB will open No breakers will open

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Distracter Analysis: Incorrect: the rod drive MG sets are run in parallel - losing one MG set will not cause any reactor trip breakers to ope Plausible: If the candidate thinks that the lB rod drive MG set provides control power RTA and BY Incorrect: the rod drive MG sets are run in parallel - losing one MG set will not cause any reactor trip breakers to ope Plausible: If the candidate thinks that the lB rod drive MG set provides control power RTB and BY Incorrect: the rod drive MG sets are run in parallel - losing one MG set will not cause any reactor trip breakers to ope Plausible: If the candidate thinks that the lB rod drive MG set provides control power BYA and BY Correct answer: the rod drive MG sets are run in parallel - losing one MG set will not cause any reactor trip breakers to open.

Page 16 For Official Use Only Ques-241

SRO Exam McGuire Nuclear Station Bank Question: 242 Answer: A 1 Pt(s)

Unit I is operating at 100% power. Given the following conditions on the 1A NCP:

Time 0200 0210 0220 0230 Motor winding temp (F0):

312 315 320 324 Pump shaft vibration (mils):

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1 seal AP (psid):

201 196 223 235

1 seal outlet temp (F0):

201 226 236 240 What is the earliest time that the operators are required to trip NCP-IA?....

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Distracter Analysis: Objective - to determine if the candidate can analyze the above conditions and select the correct time to trip the NCP. Parameters are selected to plausibly distract on the basis of the different set point Correct: Must trip when motor winding temperature exceeds 311 OF Incorrect:

Plausible:

psid Incorrect:

Plausible: Incorrect:

Plausible:

reached trip set point at 0200 NCP #1 seal differential pressure is less than limit of 200 reached trip set point at 0200 NCP #1 seal outlet temp exceeds the limit of 235 0F reached trip set point at 0200 pump shaft vibration limit reached at 20 mils Page 17 For Official Use Only Question #13 Ques_242

SRO Exam McGuire Nuclear Station Bank Question: 243 Answer: C I Pt(s)

Unit I is conducting a plant startup in Mode 1. The operators have reached 8% power when a momentary electrical transient occurs resulting in the following conditions:

Bus ITA 11B ITC 1TD Frequency (Hz)

60

60 Voltage (VAC)

6410 6900 6410 6900 Which one of the following sequences would occur? A reactor trip does NOT occur and NCPs 1A and IC trip on under-frequency while NCPs IB and ID continue runnin A reactor trip occurs and NCPs IA and 1C trip on under voltage while NCPs lB and 1D continue runnin A reactor trip does NOT occur and all four NCPs trip on under frequenc A reactor trip occurs and all four NCPs trip on under-frequency

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Distracter Analysis: Incorrect: all 4 NCPs trip due to the NC pump monitor system action Plausible: only 2 pumps have a low frequency condition Incorrect: all 4 NCPs trip due to the NC pump monitor system action - the reactor dies not trip as power is below P-7 (10%)

Plausible: only 2 pumps have a low voltage condition Correct answer Incorrect: the reactor does not trip below P-7 (10%)

Plausible: all four NCPs trip due to under-frequency on 2/4 NCPs Page 18 For Official Use Only Question #14 Ques_2143

SRO Exam McGuire Nuclear Station Bank Question: 298 Answer: D 1 Pt(s)

Unit I was operating at 100% power. Given the following conditions:

"* Pressurizer pressure controller is selected to "1-2"

"* Pressurizer pressure controls are in AUTO

"* Pressurizer pressure channel I detector fails LOW Which one of the following describes the plant response with no operator action? High pressurizer pressure reactor trip will occu PORV INC-34A will maintain NC system pressure 80 to 100 psig above norma PORV 1NC-34A will maintain NCS pressure from 100 psig above normal to 50 psig below norma PORVs 1NC-32B and 1NC-36B maintain NC system pressure 80 to 100 psig above norma..............................................

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Distracter Analysis: This question was modified from a question on the Catawba NRC Exam from 1997. The stem was changed from position 3-2 to position 1-2 and distracter C was changed and the correct answer was change Incorrect: no trip will occur Plausible: would be the correct answer for pressure control in the 3 2 position - this was the correct answer on the 1997 Catawba NRC exam Incorrect: NC-34A will not open Plausible: the plant pressure control band is correct but NC-34A only opens if pressurizer pressure channel I fails high, not low right pressure, wrong PORV Incorrect: NC-34A will not open Plausible: NC-34A opens if pressurizer pressure channel I fails high and the pressure control band is correct for NC-34A Correct answer Page 19 For Official Use Only Question #15 Ques_298

Question #16 McGuire Nuclear Station SRO Exam Bank Question: 307 Answer: C 1 Pt(s)

Unit 1 was operating at 100% power when a crud burst occurred. Given the following events and conditions:

"* EMF-48 (Reactor Coolant Hi Rad) trip 2 alarm

"* IEMF-18 (Reactor Coolant Filter IA) trip 2 alarm Which one of the following actions is required to reduce coolant activity due to a crud burst in the NC system? Purge the VCT with nitrogen Place/ensure both mixed bed demineralizers are in service Increase letdown flow Add hydrogen to the reactor coolant

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Distracter Analysis: Incorrect: Will not correct a high NC activity from a crud burst Plausible: One of the subsequent actions in AP/1 8 is to purge the VCT to the waste gas system with Hydrogen. In addition, Nitrogen is used to purge the VCT for shutdown. It is likely that a candidate could mix up these purge Incorrect: Do not want to load crud particles into BOTH mixed bed demineralizers Plausible: Mixed bed demins will filter crud particles and remove fission product ionic impurities - this action required for fuel element failure/high fission product activity in AP/I 8 - but not for crud burst Correct: Will increase removal rate of crud particles by increased filtratio Incorrect: Will not remove crud burst particulate activity Plausible: Used to scavenge Oxygen from the NC coolant and thus reduce the corrosion rates and crud production in the RC However, this does not affect crud burst particulates that are already in the NC system coolant.

Page 20 For Official Use Only Ques_307

SRO Exam McGuire Nuclear Station Bank Question: 308 Answer: D 1 Pt(s)

Unit I is operating at 100% power. Given the following conditions:

"* Rod control is in manual

"* Control Bank D is at 200 steps If the rods in control bank D start stepping out at 8 steps per minute, what one of the following actions is required at this time? Select Control Bank D on the rod selector switch and manually insert Control Bank D Select "AUTO" on the Bank Select Switch and see if rod motion stops Commence emergency boration Trip the reactor

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Distracter Analysis: Incorrect: The correct response is to trip the reactor for a rod withdrawal Plausible: this action could stop the rod withdrawal, as the rods in signal should over-ride the rods out signal Incorrect: Trip the reactor is the correct respons Plausible: If the malfunction was in the manual section of the rod control circuitry, this could stop the rods. If the rod control was in auto - then going to manual would be the correct answer. This reverses that thought proces Incorrect: Trip the reactor is the correct response Plausible: This action would be required to insert negative reactivity if the trip did not work Correct answer: Immediate action in step 3 of AP-14 Page 21 For Official Use Only Question #17 Ques_308

Question #18 McGuire Nuclear Station SRO Exam Bank Question: 311 Answer. A 1 Pt(s)

Unit 1 is operating at 50% power. Given the following conditions:

"* Pressurizer pressure is 2235 psig

"* Pressurizer Relief Tank (PRT) pressure is 20 psig

"* PRT temperature is 125 0F

"* PRTlevelis 81%

"* The PRT is being cooled by spraying from the RMWST

"* A pressurizer code safety valve is suspected of leaking by it's seat What temperature would be indicated on the associated safety valve discharge RTD if the code safety were leaking by?

REFERENCES PROVIDED: Steam Tables OF OF OF F

Distracter Analysis: Correct answer Incorrect: Temp is too low - the correct temp is 260 OF Plausible: If the candidate makes the mistake of not correcting for atmospheric pressure by failing to adding 14.6 psi to the PRT pressure and uses 20 psi Incorrect: Temp is too low - the correct temp is 260 0F Plausible: If the candidate reverses the correction for atmospheric pressure by subtracting 14.6 psi from PRT pressure of 20 psig to get 5 psi Incorrect: Temp is too low - the correct temp is 260 0F Plausible: If the candidate thinks that the discharge temperature will be at the same temperature as the PRT fluid.

Page 22 For Official Use Only Ques_311

SRO Exam McGuire Nuclear Station Bank Question: 338 Answer: C I Pt(s)

Which one of the following statements complies with the requirements of OMP 4-3 regarding the rules of usage for abnormal procedures (APs) when the EOPs have been implemented? APs may not be implemented when EOPs have been entere Only one AP at a time may be implemented when EOPs have been implemented. Concurrent implementation of APs when EOPs are in use is not allowe APs may be implemented concurrently with EOPs. However, the APs were written assuming that SI has not actuated and operators must be careful when using APs if SI has occurre APs may be implemented concurrently with EOPs with the exception of events where SI has actuated. APs were written assuming the SI had not occurred and cannot be used if SI has actuate...

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.

..

.

..

.

..

.

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.

..

.

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-

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Distracter Analysis: Incorrect: APs may be entered after EOPs have been started Plausible: Many plants have this provision - symptomatic EOPs should address all significant safety challenges without requiring APs Incorrect: No limitation on the number of APs Plausible: Makes sense to limit the number of concurrent procedures in use Correct answer Incorrect: No explicit prohibition against use of APs when SI has actuated BUT there is a caution and the APs were written for the situation where SI has NOT occurre Plausible: APs were written for the situation where SI has NOT occurred.

Page 23 For Official Use Only Question #19 Ques_338

SRO Exam McGuire Nuclear Station Bank Question: 353 Answer: D I Pt(s)

A male worker needs to repack a valve in an area that has the following radiological characteristics:

"* The worker's present exposure is 1800 mrern for the yea "* General area dose rate = 65 mrem/hr

"* Airborne contamination concentration = 20 DAC The job will take 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months with a mechanic wearing a full-face respirator. It will only take 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months if the mechanic does NOT wear the respirato Which of the following choices for completing this job would maintain the workers exposure within the Station ALARA requirements? The worker should wear the respirator otherwise he will exceed 25% of the DAC limi The worker should NOT wear the respirator because the dose received will exceed neither NRC nor site dose limit The worker should wear the respirator because the total TEDE dose received will be less than if he does not wear on The worker should NOT wear the respirator because the total TEDE dose received will be greater than if he wears on....

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Distracter Analysis:

Radiation exposure comparison:

Without respirator DDE = 65 mrem/hr x 2 hr = 130 mrem From airborne contamination:

CEDE = 20 DAC 2 hr x 2.5 mrem/DAC-hr = 100 mrem TEDE = 130 + 100 = 230 mrern from job Total exposure for year = 1800 + 230 = 2030 mrem With respirator DDE = 65 mrem/hr x 4 hr = 260 mrem CEDE = 0 TEDE = 260 mrem Total exposure for year = 260 + 1800 = 2060 mrem Page 24 For Official Use Only Question #20 Ques_35351

SRO Exam McGuire Nuclear Station (with respirator)

(without respirator)

TEDE = 2060 mrem > 2030 mrem = do NOT use a respirator Incorrect: Will not exceed 25% the DAC limit - this is not how DAC is applied to exposure limits Plausible: 25% DAC is the limit at which an area requires posting as a high airborne contamination are Incorrect: The dose will exceed station admin limits of 2000 mrem Plausible: if the candidate does not know the station admin limit or miscalculates the dose received Incorrect: The exposure will be greater if you wear the respirator Plausible: If the candidate incorrectly computes the exposure - this was the correct answer on the 1997 Catawba NRC exam Correct answer Page 25 For Official Use Only Question #20 Ques_353

SRO Exam McGuire Nuclear Station Bank Question: 372 Answer: C I Pt(s)

Unit 2 was responding to a faulted steam generator event. The operators entered FR-P. 1 (Response to Imminent PTS) and reach step 15 where they are directed to isolate cold leg accumulators (CLAs).

What is the EOP basis for isolating the CLAs in FR-P. I? To prevent injecting the CLA nitrogen bubble into the reactor and creating a gas bubble in the vessel head regio To prevent repressurizing the reactor vessel and adding pressure stress to thermal stres To prevent adding more cold water to the reactor vessel and increasing the thermal stres To prevent depleting CLA volume and to preserve a source of highly borated water to prevent recriticallity during cooldow....

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Distracter Analysis: Incorrect: - the CLAs are isolated to prevent adding to the thermal stress. The gas bubble would not be a limiting concern in FR-Plausible: - this is a valid limiting condition for isolating the CLAs during LOCA depressurizations - good answer, wrong even Incorrect: - adding the CLA volume would not cause an increase in pressure because the addition of the CLA volume is caused by pressure in the system decreasing below CLA pressur Plausible: - the basis for terminating SI is to prevent adding water to the system and increasing pressure thereby adding pressure stress to thermal stress. Not applicable to the CLA Correct answer Incorrect: - the CLAs are not require for Boron addition for this scenari Plausible: - re-criticality this is a valid concern for scenarios that involve adding large quantities of unborated water. Good answer wrong event.

Page 26 For Official Use Only Question #21 Ques_372

SRO Exam McGuire Nuclear Station Bank Question: 390 Answer: B 1 Pt(s)

Unit 1 is recovering from a LOCA. The operators started the process of terminating safety injection at 2:00 AM. Given the following indications at the following times:

Paramett

2_05 2.'0 2:15 1)

Pressurizer level (%)

29

11 2)

NC pressure (psig)

280 285 290 295 3)

ND Flow 1000 1025 1075 1085 4)

Core exit T/Cs (0F)

690 702 695 685 5)

FWST level (inches)

183 179 149 113 6)

Containment Pressure (psig).3.1 What is the earliest time that the operators should transition to ES-1.3, transfer to cold leg recirculatio Distracter Analysis: This question is designed to test the candidate's ability to identify the criteria for switchover to cold leg recirc from a list of plant parameters. A change to the switchover criteria was changing the FWST level from 150 inches to 180 inches. In addition, he meets the S/I reinitiation criteria at 0205 and this will check if he recognizes that ES-1.3 has priority Incorrect: - does not meet criteria for switchover, FWST > 180 inches Plausible: - if candidate does not know foldout criteria for switchover Correct answer - FWST level < 180 inches - reaches ACC value for S/I reinitiation criteria - needs to decide if ES-1.3 needs to be delayed until S/I reinitiation is completed Incorrect: - too late Plausible: - if candidate does not know foldout criteria for switchover - recent change (since last exam) to the switchover criteria - used to be 150 inches - now is 180 inches Page 27 For Official Use Only Question #22 Ques_390

SRO Exam McGuire Nuclear Station Incorrect: - too late Plausible: - if candidate thinks that SI termination must be competed before switchover to cold leg recirc or if he thinks that S/I reinitiation takes priority over switchover - reaches non-ACC value for S/I reinitiation.

Page 28 For Official Use Only Question #22 Ques_39o

SRO Exam McGuire Nuclear Station Bank Question: 401 Answer: D I Pt(s)

Unit 2 was operating at 100% when a single control rod in control bank D drops into the core due to a failed CDRM. The SRO directs that the dropped rod be recovere Which one of the following prevents the remaining rods in the control bank from being withdrawn while the dropped rod is being recovered? The rod control non-urgent failure alarm will actuate when the dropped rod is withdrawn blocking all rod motio C-t1 actuated when the rod dropped and will prevent outward rod motion by control bank D The Lift Coil Disconnect Switch is opened on the dropped rod to electrically isolate it from control bank D The Lift Coil Disconnect Switches are opened on control bank D rods that did not drop

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Distracter Analysis: Incorrect: the non-urgent failure alarm does not actuate unless you lose a power supply to a logic or power cabinet Plausible: - the urgent failure alarm actuates when the rod is withdrawn and this would block rod motion for all rods on the opposite power cabinet to the dropped ro Incorrect: - C-1I will not allow any auto rod motion but allows manual motion - C-I l has not actuated under these conditions (bank D at top of core)

Plausible: - C-Il will prevent outward rod motion in bank D Incorrect: - will not be able to pick up the dropped rod Plausible - if the candidate was not familiar with the actions of the lift coil disconnect switch - it will disconnect the rod from bank D but the rod cannot then be withdrawn with the switch open Correct answer Page 29 For Official Use Only Question #23 QuesL401

SRO Exam McGuire Nuclear Station Bank Question: 404 Answer. A I Pt(s)

Unit 1 was responding to a small break LOCA. Containment pressure reached 3.5 psig. The Subcooling Margin Monitor currently indicated +35 OF. Which of the following statements best describes the status of subcooling in the core? The core is subcooled by 35 0F The core is superheated by 35 0F The core is superheated by more than 35 0F due to the effects of adverse containment conditions The core is subcooled by more than 35 OF due to the effects of adverse containment conditions

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Distracter Analysis: This is a modified question from a previous NRC exam. The original question asked what the core conditions were if ICCM was reading -35 0F. The original answer was "C". Note: the upper limit for measuring superheat is -35 0F. The lower limit is +200 Although the ICCM was designed for ACC inputs, this option was never used because the pressure transmitters were located outside of containment Correct: Incorrect: - subcooling is 35 0F Plausible: - if the candidate reverses the meaning of the indication (i.e. - means subcooled, + means superheated) Incorrect: - subcooling is 35 0F Plausible: - if the candidate reverses the meaning of the indication this was the answer on the NRC exam in 199 Incorrect: - subcooling is 35 0F Plausible: - if the candidate reverses the meaning of the indication.

Page 30 For Official Use Only Question #24 Ques_4O4

SRO Exam McGuire Nuclear Station Question #25 Bank Question: 407 Answer: C 1 Pt(s)

Unit I has a liquid radioactive waste release in progress from the Ventilation Unit Condensate Drain Tank (VUCDT) through the RC system. All lineups and authorizations have been properly made in accordance with OP/0/B1/6200/35 using the normal path. 2 RC pumps are the minimum required under LWR documen Given the following initial conditions:

2 RC pumps are running

Controlling EMF properly adjusted for trip 1 and trip 2 settings

No other releases are in progress What automatic actions would terminate the release? WM-46 will close automatically if I RC pump trips WM-46 will close automatically when EMF-44 (VUCDT)

reaches the trip 2 setpoint WL-320 and WP-35 will close automatically if 1 RC pump trips WL-320 and WP-35 will close automatically when EMF-49 (Liquid Waste) reaches the trip 2 setpoint

Distracter Analysis: Used a similar question on the last NRC exam - but modified the stem and the answer. The answer from the previous exam was

"D". IEMF-44 is the controlling EMF, not IEMF-4 Incorrect: WM-46 is isolated and not used anymore as a release pat Plausible: - RC pump interlock will actuate - set at 2 pumps (minimum required on LWR document). - WM-46 was formerly the normal release path Incorrect: - WM-46 receives a closing signal from EMF-44 but this is not the normal path for a release. WM-46 is isolated and not used anymor Plausible: - this was formerly the normal release path - EMF-44 sends a closing signal to WM-46 but the valve is no longer in servic Correct: - RC pump interlock will actuate - set at 2 pumps (minimum required on LWR document).

Page 31 For Official Use Only Ques_407

SRO Exam McGuire Nuclear Station Incorrect: - EMF-49 does not trip WL-320 is not used to monitor the release from the VUCDT Plausible: - EMF-49 would monitor and isolate a liquid release from the Waste Monitor Tank (WMT) This was the correct answer from the last NRC exam - except the monitor referenced was 1 EMF-44 instead of 1 EMF-49.

Page 32 For Official Use Only Question #25 Ques_407

SRO Exam McGuire Nuclear Station Bank Question: 412 Answer. A I Pt(s)

Unit 1 is responding to a LOCA. Given the following initial conditions:

"* A reactor trip and safety injection actuation occurred at 0150

"* MSIVs are shu "* Phase B containment isolation has occurred The operators reach step 2 in ES-1.1 (SI Termination) requiring a reset of the safety injection signa Given the following parameter trends at 0200:

"* NC pressure = dropped to 1850 psig then stabilized at 1951 psig

"* Steamline pressure = 771psig - decreasing slowly

"* Containment pressure = 2.2 psig - decreasing slowly Given the following sequence of operator actions:

0202 Blocks the low steam line pressure MSI signal 0203 Blocks the low PZR pressure SI signal 0204 Resets the phase B isolation signal What is the earliest time that depressing the SI reset pushbuttons (trains A and B) would reset safety injection? Distracter Analysis:

This question will test if a candidate understands that safety injection can be reset even with valid SI actuation signals still present (not blocked). The only restrictions are the 60-second timer and P-4 (RTBs open). Correct answer - safety injection can be reset after a 60 second timer has elapsed and the train related reactor trip breaker has opened (P-4). None of the SI signals being present will prevent reset of SI. Once reset, only manual SI is available Page 33 For Official Use Only Question #26 Ques_412

SRO Exam McGuire Nuclear Station Incorrect: - SI already reset at 0200 Plausible: - the steam line low pressure MSI can be blocked < P-1I

- doesn't effect SI Incorrect: - SI already reset at 0202 Plausible: - this will block the low pressurizer pressure SI signal Incorrect: - SI already reset at 0202 Plausible: - this action will block hi-hi containment pressure SI signal Page 34 For Official Use Only Question #26 Ques__412

SRO Exam McGuire Nuclear Station Bank Question: 430 Answer: A I Pt(s)

Unit I is responding to a large break LOCA into containment. Given the following events and conditions:

Containment spray auto-actuated and reduced containment pressure

Containment pressure is now at 0.24 psig and continues to increase

NS actuation logic has been reset by the operators Which one of the following describes the NS system response to an increase in containment pressure? NS pumps will start and discharge valves will open after containment pressure reaches 3.0 psi NS pumps will start and discharge valves will open after containment pressure exceeds 0.8 psi NS pumps will start when containment pressure reaches 0.35 psig and discharge valves will open when containment pressure exceeds 0.8 psi Discharge valves will open when containment pressure exceeds 0.35 psig and NS pumps will start when containment pressure exceeds 0.8 psi...

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.

Distracter Analysis: This is a modification of a question used in the 1999 exam. The changes were to modify the condition that NS spray was reset. The previous correct answer was D. Answer A was also modified. The normal expected behavior of NS is as stated in D. If Containment pressure did not fall below 0.35 psig, then D would still be correct because the CPCS interlock would not be satisfie Correct answer - with containment spray actuation reset, the valves will open and the pumps will start at 3.0 psig Incorrect: - valves do not open at 0.8 psig Plausible: - pumps start at 0.8 psig by CPCS Incorrect: - pumps do not start at 0.35 psig and valves do not open at 0.8 psig Plausible: - this is backwards to what happens Incorrect: - containment spray actuation was reset Plausible: - this is how the system would function if containment spray actuation was NOT reset.

Page 35 For Official Use Only Question #27 Ques__430

SRO Exam McGuire Nuclear Station Bank Question: 447 Answer. D 1 Pt(s)

Unit 1 is shutdown, Mode 6, in a refueling outage. Given the following conditions:

"* Containment airlock doors are both open

"* A full shift of qualified maintenance personnel are available inside containment

"* The Refueling SRO is in the control room

"* The Fuel Handling Supervisor is inside containment Refueling has been completed and the Fuel Handling Supervisor (who is not a qualified SRO) requests permission to latch all control rods to prepare for the reactor startup. What additional requirements must be met (if any) to proceed with latching rods? Latching rods may proceed at the discretion of the Fuel Handling Superviso Latching rods may not proceed until after containment integrity has been restore Latching control rods may not proceed until after the Refueling SRO arrives inside containment to supervis Latching control rods may not proceed until after the Refueling SRO arrives inside containment and containment integrity has been restore...................................................------------------------------.........

Distracter Analysis: Incorrect: - the Refueling SRO is required to supervise this evolution and containment integrity must be restored Plausible: - if the candidate does not recognize that latching rods is a core alteration or doesn't recognize that this requires containment integrity to be established Incorrect: - the Refueling SRO is required to supervise this evolution Plausible: - if the candidate does not recognize that latching rods is a core alteration Incorrect: - containment integrity must first be established Plausible: - core alterations requires SRO coverage and containment integrity Correct answer Page 36 For Official Use Only Question #28 Ques_447

SRO Exam McGuire Nuclear Station Bank Question: 451 Answer. B I Pt(s)

Unit I is shutdown in a refueling outage. Given the following events and conditions:

"* A VI header rupture occurs

"* The VI system completely depressurize "* VI-820 was open at the time of the ruptur "* The VS system was in a normal lineup What effect does a total loss of the VI system have on the VS system? VI-820 will auto-close as VI header pressure decreases below 90 psig and the VS air compressor will start automatically at 82 psig to maintain VS header pressure VI-820 will auto-close as VI header pressure decreases below 82 psig and the VS air compressor must be manually started to maintain VS header pressure Check valves in the VI - VS cross-connect line will close to isolate VS system pressure before it drops below 90 psig VS pressure in the Fire Protection Pressurizer Tank will be lost until a VS air compressor can be starte Distracter Analysis: Incorrect: - the VS air compressor does not automatically start to maintain pressure - VI-820 auto-closes at 82 psig not 90 psig Plausible: - The VI system is safety significant, VI-820 does close at 82 psig and there is a separate VS air compressor which has an automatic startup feature -but it just is normally in "off' and requires operator action to star Correct answer Incorrect: - there are no check valves in this line Plausible: - this is another possible method to prevent depressurizing the VS header at some plant Incorrect: - the RF system tank is pressurized with VS air - but is maintained isolated from the VI header Plausible: - if the candidate does not know that the RF system air tank is isolated from the VS header.

Page 37 For Official Use Only Question #29 Ques_451

SRO Exam McGuire Nuclear Station Bank Question: 465 Answer. A I Pt(s)

Unit I is operating at 15% power going to 100% power. The operators just completed synchronizing the main generator on the power grid. Which one of the following sequences describes the correct operator actions for increasing the main generator load? Select MW IN Raise the GV limit from 17% to 120%

Depress LOAD RATE pushbutton and enter desired load rate Depress the REFERENCE pushbutton and enter the load Depress the GO pushbutton Select MW IN Raise the GV limit from 17% to 120%

Depress STANDARD pushbutton and enter desired load and load rate using the keypad Depress the GO pushbutton Select MW OUT Raise the GV limit from 17% to 100%

Depress the REFERENCE pushbutton and enter the load Depress LOAD RATE pushbutton and enter desired load rate Depress the GO pushbutton

Distracter Analysis: Correct answer: Incorrect: - do not use the STANDARD button Plausible: - this is designed to work correctly but is not used at McGuire Incorrect: - MW must be IN - do not use the STANDARD button Plausible: - if the candidate does not understand the MW IN feedback loop Incorrect: - MW must be IN for feedback loop Plausible: - if the candidate does not understand the MW IN feedback loop

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EIn. flmriml IUs Onlv Page 38 Ques_4o6 Question #30 Wu lllkll V

SRO Exam McGuire Nuclear Station Bank Question: 471 Answer. D 1 Pt(s)

Unit I is responding to a LOCA. Given the following events and conditions:

Completed E-0 (Reactor Trip or Safety Injection)

Entered E-I (Loss of Reactor or Secondary Coolant)

The STA reported the following valid critical safety functions:

"* Subcriticality - orange path

"* Integrity - red path

"* Heat Sink - red path

"* All other CSFs are green or yellow What procedure should be operator select? Remain in E-1 (Loss of Reactor or Secondary Coolant) Transition immediately to FR-S.I (Response to Nuclear Generation /ATWS) Transition immediately to FR-P.I (Response to Imminent Pressurized Thermal Shock Condition) Transition immediately to FR-H.I (Response to Loss of Secondary Heat Sink)

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Distracter Analysis: Incorrect: - must transition to CSFs Plausible: - if candidate does not know restrictions and applicability of F-0 Incorrect: - Orange path does not have priority over red paths Plausible: - if candidate does not know rules of usage Incorrect: - Integrity does not have priority over Heat Sink Plausible: - if candidate does not know CSF rules of usage Correct answer: - Heat sink does not have priority over core cooling Page 39 For Official Use Only Question #31 Ques_471

SRO Exam McGuire Nuclear Station Bank Question: 479 Answer. A 1 Pt(s)

Unit 1 is in the process of making a radioactive gaseous waste release from the waste gas decay tank in accordance with OP/0/A/6200/18. Given the following conditions:

0 MRIRR=21 CFM 0 MOSRR =40 CFM

I EMF-50 trip 1 setpoint = 1.0E5 CPM

IEMF-50 trip 2 = 2.0E5 CPM

1EMF-36 is in service Time 0200 0215 0230 0245 Release rate (CFM)

25

37 EMF-50 (CPM)

1.8E5 2.2E5 2.1E5 3.2E5 If the operators reset IEMF-50 whenever allowed by procedure, what is the earliest time (if any) that the operators are required to terminate the gaseous releas Distracter Analysis: Correct answer - the release rate (21 CFM) > MRIRR (20 CFM)

(most restrictive instantaneous release rate) Incorrect: - exceeded MRIRR at 0200 Plausible: - EMF-50 tripped WG-160 for the first time Incorrect: - exceeded MRIRR at 0200 Plausible: - exceeded MOSRR (maximum observed system release rate) - the operator is allowed to reset EMF-36 and restart the release 3 times before being required to terminate the release - this is the 2nd time EMF-36 has reached trip 1 Incorrect: - exceeded MRIRR at 0200 Plausible: - exceeded trip 2 on EMF-50 for the 3rd time - allowed to reset this trip 3 times before terminating release

... flwRri I I1Ue flnlv Page 40 Ques__4/y Question #32 WO I

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SRO Exam McGuire Nuclear Station Bank Question: 501 Answer. D 1 Pt(s)

Unit 2 was operating at 100% power when a terrorist attack in the control room caused the operators to rapidly evacuate to the Auxiliary Shutdown Panel. The operators were not able to perform AP/17 (Loss of Control Room) actions prior to evacuation at 020 The terrorists tripped the turbine but did not operate any other control There are no other local operator actions taken. Given the following steam generator narrow range levels:

0200 0202 0204 0206 0208 2A S/G NR 65%

37%

22%

15%

25%

2B S/G NR 64%

38%

23%

18%

26%

2C S/G NR 63%

39%

25%

16%

24%

2D S/G NR 65%

38%

26%

20%

27%

Which one of the following statements describes the complete list of running feedwater pumps when the operators first arrive at the ASP at 0210 to take local control of the plant? Both motor driven CA pumps Both motor drive CA pumps and the turbine drive CA pump Both motor driven CA pumps and both CF pumps (in roll-back hold) Both motor driven CA pumps, the turbine driven CA pump and both CF pumps (in roll-back hold)

Distracter Analysis: The bo-lo setpoint for SGWL is 17%. This causes:

Reactor trip - on 1 of 4 S/Gs in 2 of 4 channels

MD CA pumps auto-start - on 1 of 4 S/Gs in 2 of 4 channels

TD CA pump auto-start - on 2 of 4 S/Gs in 2 of 4 channels Incorrect: CF pumps will not trip - this is done by a local operator action in AP-17, TD CA pump auto-start Plausible: MD CA pumps will start when S/G levels < 17% on 1/4 S/Gs Incorrect: The CF pumps will continue to run until tripped by local operator action in AP-17 Plausible: The MD and TD CA pumps auto start

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Car fulf-riI IUse Onlv Page 41 QuesjUl1 Question #33 U1 llldlll Vm~

SRO Exam McGuire Nuclear Station Incorrect: The TD CA pump will auto star Plausible: The MD CA pumps auto start and the CF pumps remain running Correct answer:

For Official Use Only Page 42 Ques_501 Question #33

SRO Exam McGuire Nuclear Station Bank Question: 504 Answer. C I Pt(s)

Unit I was operating at 100% power when a reactor trip occurred due to a feedwater control valve malfunction. Given the following events and conditions:

"* Both motor-driven CA pumps started

"* The operators have entered E-0 (Reactor Trip or Safety Injection)

"* Feedwater flow to each generator is greater than 450 gpm

"* Given the following steam generator levels Time 0200 0201 0202 0203 0205*

Steam Generator Level 1A S/G (% NR)

12

39

1B S/G (% NR)

12

39

1C S/G (% NR)

13

40

IDS/G(%NR)

16

39

Containment pressure (psig).5.5 *At 0205, the operators reach step 16.c of E-0 that reads:

WHEN N/IR level in any S/G greater than 11% (32%

ACC), J control CA flow to maintain N/IR levels between 11%(32%) and 50%.

Which one of the following statements correctly describes when the operators are allowed to reset and reduce CA flow to the steam generators? Any time after 0201 Any time after 0202 Any time after 0203 Any time after 0205

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Distracter Analysis: OMP 4-3 allows reset and control of CA flow when any S/G level reaches its normal setpoint (39%). Step 16 of E-0 directs the operators to control CA flow between 11% (32% ACC) and 50% S/G NR level. Until the operators reach step 16, they are allowed to control S/G level between 39% and 55% per OMP 4-..,. iwFrinil I1Use Onlv Page 43 Ques95u4 Question #34 Wu1 lI I

I C

='

SRO Exam McGuire Nuclear Station Note: This question was significantly modified - in the previous version of this question, the correct answer was A, can control S/G levels immediatel Incorrect: no S/Gs have reached their normal control bands Plausible: S/G levels exceed 11%, the "normal" setpoint for the controlling band after reaching step 16. If the candidate thinks that the S/Gs are within their control band or if the candidate thinks that OMP 4-3 allows control of levels at any tim Incorrect: no S/Gs have reached their normal control band Plausible: OMP 4-3 allows reset and control of CA flow when any S/G level reaches its normal control band which is 39%. However, the candidate may become confused with the control bands given in step 16. S/G levels exceed 11% but ACC values are now in effect and the minimum band for S/G level is 32O/o-even though containment pressure < 3.0 psi Correct Answer: OMP 4-3 allows control of S/G levels when they reach their "normal level setpoint" - in this case > 39% NR level Incorrect: can reset CA at 0203 when S/G NR level reaches its control ban Plausible: If candidates do not know OMP 4-3 guidance that allows resetting CA flow prior to reaching step 1 For Official Use Only Page 44 Ques_504 Question #34

Question #35 MivCuuire rNucear.LIUII-U Bank Question: 507 Answer: B I Pt(s)

Unit 2 is responding to a LOCA into the Auxiliary Building in ECA-I.2 (LOCA Outside of Containment). Upon completion of ECA-I.2, NC system pressure continues to decrease. Which one of the following statements correctly describes the correct mitigating strategy to assure continued removal of decay heat under these conditions? Transition back to E-1 (Loss of Reactor or Secondary Coolant). Transition to ECA-t.1 (Loss of Emergency Coolant Recirc). Transition to ES-1.2 (Post LOCA Cooldown and Depressurization) Transition to ES-1.3 (Transition to Cold Leg Recirc).

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Distracter Analysis: Incorrect: Not the correct procedural transition if the NC system pressure continues to decrease (ie leak path not isolated)

Plausible: This IS the correct procedure if the NC system pressure was stable or increasin Correct answer Incorrect: Transition to ES-1.2 not allowed as the leak is not isolated Plausible: The name of the procedure is appropriate for the situatio Incorrect: Transition to ES-1.3 not in accordance with the major action steps Plausible: Although many actions are the same, it is not the correct procedur Ques_507 For Official Use Only Page 45 SRO Exam

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SRO Exam McGuire Nuclear Station Bank Question: 512 Answer. B I Pt(s)

Unit 2 was in the process of starting up the reactor following a refueling outage. Given the following plant conditions and events:

"* Reactor trip breakers are closed

"* Shutdown bank rod withdrawal has commenced

"* Train A of Wide Range Shutdown Monitoring is inoperable If source range N-32 fails, which one of the following actions is required? Startup may continue with train B of the Gamma-Metrics Shutdown Monitor System substituting for the failed N-32 source range channel Immediately stop withdrawal of shutdown banks Immediately open the reactor trip breakers Immediately reinsert shutdown banks and open the reactor trip breakers

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Distracter Analysis: Incorrect: Cannot substitute gamma metrics for SR channel Plausible: Allowed to substitute SR channels for gamma metrics Correct answer: immediate action per AP/16 case I Incorrect: Not required unless 2 SR channels fail Plausible: If candidate does not know tech spec requirements Incorrect: Not a tech spec action Plausible: If candidate does not know tech spec requirements Ques_512 For Official Use Only Page 46 Question #36

SRO Exam McGuire Nuclear Station Bank Question: 531 Answer: C 1 Pt(s)

Unit 2 is operating at 75% power when a load rejection occurs. Which one of the following statements correctly describes the response of 2CM-420 (Load Rej Byp) to this transient? CM-420 closes to prevent condensate water from being diverted to the suction of the hotwell booster pumps from the condensate booster pumps to assure minimum flow to the CF pump CM-420 closes to prevent diversion of water from the "C" heater drain tank back to the UST thereby ensuring sufficient CF pump suction pressur CM-420 opens to divert condensate flow directly to the condensate booster pump suction to ensure that CF pumps have sufficient suction pressur CM-420 opens to divert condensate flow, bypassing around the condensate booster pumps, directly to the CF pumps to assure minimum flow requirement.

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Distracter Analysis: Incorrect: CM-420 opens - does not close. Does not prevent water from being recirculated around the hotwell pump Plausible: this function is performed by CM-407 - which opens to assure minimum flow around the hotwell pumps to prevent water hammer on the CM system during startu Incorrect: CM-420 opens - does not close. Does not prevent a loss of water to the condensate booster pump suctio Plausible: CM-227 opens to recire condensate from the C feedwater heater to the USTs to assure minimum recirc flow on the CBPs Correct answer Incorrect: CM-420 does not provide a flow path around the condensate booster pumps directly to the CF pumps to meet minimum flow requirements Plausible: CM-420 opens to provide bypass flow - but directly to the CBPs - not the CF pumps

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efr.Rinl Use Onlv Page 47 Ques-pa Question #37 I VI VIIl*lllil vv*

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SRO Exam Question #38 Mcuuire rNuucai aLaru"

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Bank Question: 538 Answer: C 1 Pt(s)

Unit I is operating at 28% power during a plant startup to 100%. Given the following conditions on the IC steam generator:

"* Main feedwater regulating valve (FRV) is in AUTO control at 25% open

"* Bypass FRV is in MANUAL control at 100% open

"* Steam flow channel I fails high Which one of the following statements correctly describes the plant response for the IC steam generator FRVs?? Main FRV modulates open to increase feedwater flow and steam generator water level increases to the high level alarm setpoin Main FRV modulates shut to reduce feedwater flow and steam generator level decreases to the low level alarm setpoin Main FRV modulates open to increase feedwater flow but sufficient level error signal develops to restore CF flow to normal without reaching the high level alarm setpoin Main FRV modulates shut to reduce feedwater flow but sufficient level error signal develops to restore CF flow to normal without reaching the low level alarm setpoin....

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Distracter Analysis: Incorrect: S/G water level does not increase to the high level alarm as level error quickly overcomes flow mismatch Plausible: CF control valves open to 120% Incorrect: FRVs do not modulate shut and SG water level does not fall to the low level alarm Plausible: If the candidate reverses the effect of the instrument failure - this is what happens for a steam flow transmitter failing low Correct answer Incorrect: FRVs do not modulate shut Plausible: level error does overcome flow mismatch and level will be restored Ques_538 For Official Use Only Page 48 Cý,+.i

SRO Exam McGuire Nuclear Station Bank Question: 591 Answer. D I Pt(s)

Unit 2 was operating at 5% power during a plant startup when the following sequence of actions occurre "* Opened 2NV-265B

"* Started Boric Acid Transfer pump #2A If no other operator actions occurred, which of the following statements correctly describes the response of reactor power and control rods? Power remains at 5%

Control rods drive in Power remains at 5%

Control rods do not move Power decreases Control rods drive in Power decreases Control rods do not move

Distracter Analysis: The candidate must recognize that:

"* The sequence of operations amounts to emergency boration of the reacto "* Control rods are in manual at this point during the startup

"* Power decreases due to boron addition Incorrect: power will decrease Plausible: control rods would drive in - IF they were in auto control Incorrect: power will decrease Plausible: control rods will not move Incorrect: control rods are in manual and will not move Plausible: power will decrease Correct answer r.,. f*lFfrinl II*le Onlv Page 49 Question #39 Ques._b91 I~'Wl IIllll Vi

SRO Exam McGuire Nuclear Station Bank Question: 592 Answer: B 1 Pt(s)

Unit 1 was cooling down in Mode 4 when the 1AI KC pump trips. Given the following conditions:

"* Both trains of KC were initially in operation

"* 1A2 KC pump was secured due to high KC flow

"* Both trains of ND were aligned for RHR shutdown cooling

NCS temperature was 205 'F If train A KC pumps cannot be restarted, which one of the following list of actions is the compltelist of actions that must be taken to prevent damage to equipment? Stop ND pump 1A Stop ND pump IA Isolate ND flow through the IA ND heat exchanger Cross-connect KC flow to the IA ND heat exchanger Cross-connect KC flow through the IA ND Pump mechanical seal heat exchanger Stop ND pump IA Isolate KC flow through the letdown heat exchanger

Distracter Analysis: Upon a loss of KC to an operating ND train, AP/21 requires two actions (per Foldout page):

"* Stop the associated ND pump

"* Isolate flow to the associated ND HX Incorrect: Must also stop flow to the ND HX per AP/21 Plausible: action to stop the IA ND pump is correct. There is a separate operating precaution to maintain flow through the ND HX >

2000 gpm to prevent water hammer - but it does not apply to this cas Correct answer Incorrect: cannot cross-connect B train KC flow to the A train ND HX under these conditions - AP/21 specifies that flow must be stopped to the ND H Plausible: There is a precaution to ensure that KC flow is maintained to ND mechanical seal HX for all operating ND pumps Incorrect: no need to secure flow the letdown HX

.ar ft-Ii I1Ice flniv Page 50 Question #40 Q~ues-59-z

[11l,/I VIII511511 VVV vnlw*

SRO Exam McGuire Nuclear Station Plausible: this would be required if KC was lost when the plant was at power and NCS temp was higher to prevent flashing in the letdown lin For Official Use Only Page 51 Ques_592 Question #40

SRO Exam McGuire Nuclear Station Question #41 Bank Question: 593 Answer. B I Pt(s)

Unit 2 was operating at 99% power when a steamline rupture occurre Given the following events and conditions:

0200 The operators enter AP/01 (Steam Leak)

0200 The operators reduce turbine load to match Tave and Tref 0201 The operators start a second NV pump and isolate letdown 0202 NLOs start investigating for the location of the steam leak 0203

"P/R OVER POWER ROD STOP" alarm - the RO reports that power has turned and is decreasin STA reports pressurizer level is decreasing and cannot be maintained 0205 The turbine building operator reports that the line to the atmospheric dump valves has a steam leak and cannot be isolated If no safety injection has occurred, pressurizer pressure is maintained and no reactor trip signals are received prior to 0205, which one of the following operator responses is correct? Manually trip the reactor at 0203 Manually trip the reactor at 0204 Manually trip the reactor at 0205 Commence a rapid down power using AP/04 at 0205

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Distracter Analysis: Incorrect: no requirement to trip the reactor because reactor power has turned and is decreasing. Not approaching the overpower automatic reactor trip at 109% in 2 of 4 channel Plausible: shows a power mismatch - reactor power reaches 103%

on 1 of 4 PR channels to cause C-2. OMP 4-3 requires the operator to trip when an automatic safeguards action setpoint is approached to avoid challenging the automatic safeguards functio Correct answer required to trip under AP/01 (and many other procedures) if you cannot maintain pressurizer level with 2 NV pumps Incorrect: required to trip when PZR level cannot be maintained Plausible: if the candidate thinks that a reactor trip is required because the steam leak was not isolate Incorrect: required to trip when PZR level cannot be maintained Cnr nffrciil Ulse Onlv Page 52 Ques_-u, 1 VI Vlllt*lB1 v

SRO Exam McGuire Nuclear Station Plausible: this would be the correct answer if not required to trip at 0204.

Page 53 For Official Use Only Question #41 Ques_593

SRO Exam McGuire Nuclear Station Question #42 Bank Question: 594 Answer: C 1 Pt(s)

Which one of the following statements correctly describes the operation of the condenser dump valves during a loss of condenser vacuum? Condenser steam dump valves do not open because the C-7A arming signal is blocke Condenser steam dump valves isolate on a P4 signal when the reactor trip Condenser steam dump valves isolate upon a loss of C-9 signal when condenser pressure drops below 20 inches of vacuu Condenser steam dump valves continue to dump steam to the condenser until condenser reaches atmospheric pressur................................--------------------------------------------------....

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Distracter Analysis: Incorrect: C-7A will arm on a 10% step change in load Plausible: If the C-7A interlock did not pick up and arm the condenser dump valve, they would not open Incorrect: The P4 signal does not close the condenser dump valves Plausible: A P4 signal would block the atmospheric dump valves Correct answer Incorrect: The condenser dump valves would close on loss of C-9 Plausible: The condenser dump valves normally open for a reactor trip.

Page 54 For Official Use Only Ques_594

SRO Exam McGuire Nuclear Station Bank Question: 595 Answer: D I Pt(s)

During step 22 of ECA-0.0 (Loss of All AC Power), the operators are directed to depressurize intact S/Gs to 210 psig at the maximum rate if the standby makeup pump cannot be started. What is the basis for depressurizing at the maximum rate instead of a slower more controlled rate? To cooldown as quickly as possible to prevent the loss of pressurizer leve To reduce NC pressure as rapidly as possible to prevent voiding in the upper head regio To maximize natural circulation flow to prevent excessive thermal stratification in the NC loop To minimize the loss of NC system inventory through the NCP seal....

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Distracter Analysis: Incorrect: Depressurization at the maximum rate will actually increase the chances of losing pressurizer level as the cooldown causes NC system inventory contractio Plausible: ECA-0.0 has a specific note that the cooldown should be continued even of pressurizer level is lost. The loss of PZR level is often a termination criterion for depressurization in other procedure Incorrect: Depressurization at the maximum rate will actually increase the chances of voiding in the upper head region as the cooldown causes NC system inventory contractio Plausible: This is an EOP basis for maximizing the cooldown rate in other EOPs - not the depressurization rat Incorrect: Maximizing the cooldown rate will increase the thermal driving head but this prevent, not enhance thermal stratification in the NC loops. Thermal stratification in the NC loops occurs when natural circulation flow is lost due to heat losses to ambien Plausible: Maximizing the cooldown rate will increase the temperature difference across the core because the SG temperatures will decreas Correct answer. The standby makeup pump provides flow to the NCP seals. If the pump cannot be started, rapid seal failure will occur.

Page 55 For Official Use Only Question #43 Ques_5905

Question #44 SRO Exam McGuire Nuclear Station Bank Question: 596 Answer. C I Pt(s)

Units 1 and 2 were operating at 100% power when a fire broke out in the back of the control room. Given the following conditions:

"* The fire has not effected or degraded any control systems

"* Heavy black smoke is throughout the control room

"* The SRO implements AP/17 (Loss of Control Room)

Which one of the following statements correctly describes the operator response to this event? Immediately trip both unit turbines and reactors and evacuate the control room to the auxiliary shutdown panel Evacuate the control room; trip both unit turbines and reactors on the way to the auxiliary shutdown pane Evacuate the control room: proceed to the auxiliary shutdown panels and trip both unit turbines and reactors when directed by the SROs at the auxiliary shutdown panel Evacuate the control room, proceed to the safe shutdown facility and trip both unit turbines and reactors when directed by the SRO at the standby shutdown facilit...

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Distracter Analysis: Incorrect: There is a specific caution in AP-1 7 that warns against tripping the reactor until the SRO is stationed at the ASP and the SRO directs the reactor be trippe Plausible: This could be a conservative thing to do before evacuating. Many plants require the reactor to be tripped prior to evacuatio Incorrect: There is a specific caution in AP-17 that warns against tripping the reactor until the SRO is stationed at the ASP and the SRO directs the reactor be trippe Plausible: If the candidate does not recognize this caution. This would be a convenient and expeditious action to take. This was the old AP-17 response and is now the AP/24 respons _eI. flffmnIl IUse Onlv Page 56 Ques_59b I-i Vlý1 ~

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SRO Exam McGuire Nuclear Station Correct answer Incorrect: Evacuate to the ASP not the SSF Plausible: The operators would evacuate to the SSF if the fire degraded control systems Page 57 For Official Use Only Question #44 Ques_596

SRO Exam McGuire Nuclear Station Bank Question: 597 Answer: D 1 Pt(s)

Unit 2 was shutdown in Mode 4, cooling down to a refueling outage. The following annunciator lights are provided for identification purposes in answering the question below:

Annunciators on panel 2AD-10:

E-1 = Upper Cont Airlock Aux. Door Open F-I = Upper Cont. Airlock Rx. Door Open E-2 = Lower Cont. Airlock Aux. Door Open F-2 = Lower Cont. Airlock Rx. Door Open Annunciators on panel 2AD-13:

A-8 = VE Door Open Approval was given for normal passage into the containment to perform work - no approval has been given for any compensatory measure Which one of the following alarm conditions requires corrective action under MSD 585, (Reactor Building Personnel Access and Material Control for Modes 1, 2, 3 and 4)? AD-10 E-I and 2AD-10 E-2 actuated AD-10 F-i and 2AD-10 F-2 actuated AD-10 E-i and 2AD-10 F-2 actuated AD-10 E-2 and 2AD-13 A-8 actuated Distracter Analysis: Incorrect: Allowable to have one door open in each airlock Plausible: 2 airlock doors are open at the same time Incorrect: Allowable to have one door open in each airlock Plausible: 2 airlock doors are open at the same time Incorrect: Allowable to have one door open in each airlock Plausible: 2 airlock doors are open at the same time Correct answer although only one containment airlock door is open, the alarm on the VE annulus door - if left open for > 2minutes, requires compensatory security action because an ECCS phase B actuation will auto-start the VE system to establish a negative pressure in the annulus - which can't occur if the annulus door is open.

Page 58 For Official Use Only Question #45 Ques__597

SRO Exam McGuire Nuclear Station Bank Question: 598 Answer. C 1 Pt(s)

Unit I was responding to an internal flow blockage condition in the core that required a reactor trip and entry into FR-C.2 (Response to Degraded Core Cooling).

Step 15.e of FR-C.2 states:

Dump steam to condenser from intact S/Gs while maintaining cooldown rate in NC T-colds less than 100 'F in an hou Given the following times and temperatures during the event:

Time 0200 0210 0220 0230 0240 0250 0300 NC T-cold 'F 557 560 565 558 540 530 520 Time 0300 0310 0320 0330 0340 0350 0400 NC T-cold 'F 520 495 468 467 444 428 420 Time 0400 0410 0420 0430 0440 0450 0500 NC T-cold 'F 420 405 390 371 350 320 310 If the cooldown started at 0230, what time did the operators first exceed the cooldown limit of FR-C.2? Distracter Analysis: Incorrect: cooldown rate was 45 'F for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months - did not exceed 100

'F in one hour Plausible: 108 0F/hr instantaneous cooldown rate for the 10-minute interval exceeded 100 0F/hr. In addition, the applicant has to consider the NCS temperature prior to the tri Incorrect: cooldown rate was 97 0F for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months - did not exceed 100

'F in one hour but came very close.

Page 59 For Official Use Only Question #46 Ques_598

SRO Exam McGuire Nuclear Station Plausible: 162 0F/hr instantaneous cooldown rate for the I 0-minute interval exceeded 100 0F/h Correct Answer: cooldown rate was 102 OF in one hour - the instantaneous cooldown rate was only 96 0F/h Incorrect: although the cooldown rate was 108 OF for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the operators exceeded the limit at 0350 - not the first time Plausible: Exceed both the instantaneous rate and the I-hour rat If the applicant misses the correct calculation for answer C, this is the next time when the cooldown rate is exceeded.

Page 60 For Official Use Only Question #46 QuesL_598

SRO Exam McGuire Nuclear Station Bank Question: 599 Answer. D 1 Pt(s)

Unit I was conducting a plant startup at 5% power when a control rod in control bank "A" drops into the core. Given the following events and conditions:

"* The reactor remains critical during the recovery of the control rod

"* Tave is allowed to drop to 550°F Which one of the following statements correctly describes the adverse considerations? Reduced Tave could cause thermal shock on the pressurizer spray nozzl Thermal power best estimate would indicate higher than reactor power (by Power Range N/Is). Moderator temperature coefficient (MTC) could exceed the minimum safety analysis value (i.e. become too positive) late in core lif Moderator temperature coefficient (MTC) could exceed the minimum safety analysis value (i.e. become too positive) early in core lif...................................................------------------------------.........

Distracter Analysis: Incorrect: The spray nozzle can withstand much higher temperature differentials Plausible: One of the tech spec bases for the minimum temperature for critical operations is to ensure that adequate pressurizer spray capability is maintaine Incorrect: Thermal power would indicate lower than reactor power due to increased thermalization of the neutron Plausible: If the candidate reverses the logic and the effec Incorrect: MTC would become too positive EARLY in core lif Plausible: The effect on MTC is correct - to reduce the coefficient Correct answer Page 61 For Official Use Only Question #47 Ques_599

Question #48 McGuire Nuclear Station SRO Exam Bank Question: 601 Answer. A 1 Pt(s)

Which one of the following selections correctly describes reflux boiling flow path during a large break LOC Steam enters the _(1)

of S/G U-tubes where the steam condenses and re-enters the core area via the S/G (2)

. hot lg hot le hot g cold leg cold le ho ItJg cold le coldleg

Distracter Analysis: Correct answer Incorrect: steam returns via the hot leg Plausible: the first part of the answer is correct Incorrect: the steam enters the hot leg Plausible: the second part of the answer is correct Incorrect: cold legs are not affected during reflux boiling Plausible: psychometric balance Page 62 For Official Use Only Ques_601

SRO Exam McGuire Nuclear Station Bank Question: 602 Answer: A I Pt(s)

Unit 2 is responding to a small break LOCA in ES-1.1 (SI Termination).

Given the following plant conditions:

"* NCPs tripped

"* Pressurizer level is steady

"* Only one train of ECCS is injecting

"* Loop A temperatures are representative of all 4 loops

"* Steam generator pressures are the same as steam header pressure Which one of the following sets of plant parameters is indicative ofnatural circulation occurring in the steam generators per enclosure 2 of ES-1.1 ?

Time Steam Header Pressure (psig)

NC System Pressure (psig)

Loop A T-hot (°F)

Loop A T-cold (°F) Steam Header Pressure (psig)

NC System Pressure (psig)

Loop A T-hot (0F)

Loop A T-cold (0F) Steam Header Pressure (psig)

NC System Pressure (psig)

Loop A T-hot (0F)

Loop A T-cold (0F)

0200 0205 0210 0215 1042 1968 579 548 1042 1968 579 548 1042 1968 579 548 1009 1964 574 544 1009 1972 582 544 1047 1964 574 549 976 1960 569 540 976 1975 585 540 1050 1960 569 548 945 1958 564 536 945 1981 595 536 1052 1958 564 550 Steam Header Pressure (psig)

1042 1047 1050 1052 NC System Pressure (psig)

1968 1972 1975 1981 Loop A T-hot (0F)

579 582 585 595 Loop A T-cold (0F)

548 544 540 536

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Distracter Analysis: The following conditions support natural circulation:

"* S/G pressure stable of decreasing

"* T-hot stable or decreasing

"* T-cold stable or decreasing

"* NC subcooling > 0 - NC pressure may trend up or down.

Page 63 For Official Use Only Question #49 Quesý6O2

SRO Exam McGuire Nuclear Station Correct: This shows indication of natural circulation flow occurring

- decreasing S/G pressure, T-cold at S/G saturation conditions and decreasing, T-hot decreasin Incorrect: T-hot is increasing while steam pressure is decreasing Plausible: Steam pressure and T-cold are both decreasing Incorrect: Steam pressure is increasing and T-cold is tracking along with this trend. Temperature difference is decreasing indicating that heat removal rate is decreasing. This is a classic case of gas binding Plausible: T-hot is decreasin Incorrect: Steam pressure increasing and T-hot is increasin Plausible: T-cold is decreasing Page 64 For Official Use Only Question #49 Ques_602

Question #50 McGuire Nuclear Station SRO Exam Bank Question: 603 Answer. B 1 Pt(s)

Unit I was operating at 100% power when the 1 A NV pump failed. Given the following events and conditions:

S1B NV pump was tagged out of service for maintenance

The Positive Displacement NV pump was tagged out of service

"* The plant is at normal operating temperature, pressure and level

"* Normal letdown is in service on the 75 gpm orifice

"* Identified leakage is at the Tech Spec Limit

"* Unidentified leakage is 1 cc/hr If no operator actions are taken, how much time will elapse before the pressurizer level reaches the low level alarm and the heaters trip?

REFERENCES PROVIDED:

Curve Book Encl 7.38 (PZR Volume vs. Level)

ITS 3.4.13 page I Less than 45 minutes minutes minutes Longer than 65 minutes

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Distracter Analysis:

The NC coolant will let down to the VCT at a rate of 75 gpm until 17% is reached in the pressurizer. At 17%, the PZR low-level alarm will isolate letdow Pressurizer level 55% = 7800 gal Pressurizer at 17% = 2900 gal Letdown flow = 75 gpm until isolation at 17% PZR level Identified leakage = 10 gpm (includes NCP seal leakage to NCDT)

Unidentified leakage is negligible and may be ignored NCP seal leak off= 4x3 = 12 gpm into VCT Total flow rate out of the NC system = 97 gpm until letdown isolation Time to reach 17% PZR level = (7800-2900 gal) / (97 gpm) = 50.5 minutes C...

flClrIlI UICs flnlv Page 65 Quesbui WN

SRO Exam McGuire Nuclear Station Incorrect: too short Plausible: if the candidate adds total NCP #1 seal injection flow instead of seal leak off (8x4=32 gpm) to the letdown leak rate 75 gpm and Tech Spec leak rate 10 gpm Time = 41 minutes, misreads the pressurizer level tank curve or makes another mistak Correct answer: Time to reach 17% PZR level = (7800-2900) / (75

+ 10 +12 gpm) = 50.5 minutes Incorrect: time is too long Plausible: if candidate forgets to add in max allowable Tech Spec leakage or NCP seal leak off Time to 17% PZR level = (7800-2900) / (75+10gpm) = 57.6 min Time to 17% PZR level = (7800-2900)/(75+12 gpm) = 56.3 min Incorrect: too long Plausible: if the candidate does not consider the letdown rate of 75 gpm and only considers the Tech Spec leak rate 10 gpm and/or NCP seal leakoff, misreads the pressurizer level tank curve or makes another mistake.

Page 66 For Official Use Only Question #50 Ques_603

SRO Exam McGuire Nuclear Station Question #51 Bank Question: 604 Answer: D I Pt(s)

Unit I was operating at 25% power following a reactor startup when intermediate range channel N35 failed. Given the following conditions and events:

"* N35 repairs have been made and N35 is being returned to service

"* N36readsl.5x10 4 amps

"* The N35 "level trip" switch was returned to the "normal" position If all power range nuclear instruments and N36 have been properly adjusted, which of the following operator conditions (if any) would cause the reactor to trip? N35 "Operation Selector" switch was left in "IOV<" position after retesting N35 was significantly under-compensated N35 control power fuses were never reinstalled A reactor trip would not occur

Distracter Analysis: At 10%, the operators manually block the hi IR Rx trip by procedure after P-10 is enabled on 2 of 4 PR detectors > 10% Incorrect: - the operation selector switch is taken out of the circuit when the level trip switch is taken to normal - and all IR high flux Rx trips are blocked by P-1 Plausible: - if the candidate thinks that a test signal can be inserted with level trip switch in the normal position Incorrect: - The IR high flux trip is blocked by P-10 Plausible: - inserting a test signal can cause a trip signal to be generated from the IR drawer - but will not go to SSPS Incorrect: - the reactor is above P-10 and although under compensation of N35 could cause the high flux setpoint to be reached, the IR trips are disabled by P-10 Plausible: - If the candidate does not recognize that a N36 level of 1.5x10' amps is above P-10 Correct answer

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flfriml IUsen Onlv Page 67 Quesbu't W1 il~ldlv~

SRO Exam McGuire Nuclear Station Bank Question: 605 Answer. B 1 Pt(s)

Unit 2 was operating at 100% power when a reactor trip occurred. The reactor trip caused the initiation of a tube leak in the 2B S/G. The leak rate was 100 gpm. Given the following conditions:

2EMF-33 (Condenser Air Ejector Exhaust) alarms in trip 2 If all the automatic features operate as designed (without operator intervention), which one of the following indications will provide the best indication (most sensitive and timely) to confirm that a S/G tube leak has occurred? Comparing S/G feed flow to steam flow mismatch Observing 2EMF-10, 11, 12 and 13 (steamline hi rad) Observing 2EMF-34 (S/G sample line to range) Observing 2EMF-71, 72, 73, 74 (N16 leakage)

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Distracter Analysis: Incorrect: Not a sensitive method of comparison - requires large gpm leak rates before this is noticeabl Plausible: This method will show gross SGTRs Correct answer: normally, EMF-71-74 are the most sensitive monitors. But these monitors detect N' 6 y radiation that has a high energy (7 MeV) y that only is generated when the reactor is operating at power (requires a neutron flux). Incorrect: S/G sample line will isolate at EMF-33 trip 2 - the sample line can only be lined up to 1 S/G at a time. If the leak is not in that S/G, there will be no indication of anything after isolatio Prior to isolation, it may show an increasing trend due to a general build up of activity in the feedwate Plausible: This would be a good answer if the automatic isolation did not occur Incorrect: most sensitive method as it detects N'6 y radiation Plausible: This was the correct answer for the 1997 NRC exam when the premise of the question had the reactor was operating at 100% power. In this question, the reactor has tripped and neutron flux has decreased - causing the N'6 y to decay off (TA2 is 7 seconds)

Page 68 For Official Use Only Question #52 Ques_605

SRO Exam McGuire Nuclear Station so that by the time that the steam line monitors see the contents of the S/G, the N"6 y has decayed away.

Page 69 For Official Use Only Question #52 Ques_6O5

SRO Exam McGuire Nuclear Station Bank Question: 606 Answer. D 1 Pt(s)

Unit 1 was operating at 100% when a steam generator tube rupture occurred in the IB S/G. Given the following list of valves in the S/G sample and blowdown systems:

"* 1NM-267 S/G Sample HDR RAD Monitor Inlet Isolation

"* Blowdown Blowoff Automatic Isolation Valves

"o IBB-119=ffomthe1AS/G

"o IBB-120 = from the lB S/G

"o IBB-121 =fromthe IC S/G

"o IBB-122=fromthe 1DS/G

"* S/G Sample HDR to Conventional Sample System valves

"o 1NM-269 from the 1A S/G

"o 1NM-270 = fromthe lB S/G

"o INM-271= from the IC S/G

"o 1NM-272 = from the ID S/G Which one of the following statements correctly describes the complete set of valves that would automatically close? NM-267 NM-267, IBB-120, 1NM-270 INM-267 1NM-269, 1NM-270, INM-271, 1NM-272 NM-267 1BB-119, 1BB-120, 1BB-121, 1BB-122, 1NM-269, lNM-270, 1NM-271, 1NM-272

Distracter Analysis: Incorrect: insufficient Plausible: closes sample header Incorrect: insufficient - partial list Plausible: these valves would isolate sample flow from the I B S/G with the ruptured tube Incorrect: insufficient - partial list - does not include blowdown system valves Plausible: all sample system (NM) valves isolate Correct answer - complete list Page 70 For Official Use Only Question #53 Ques_606

SRO Exam McGuire Nuclear Station Bank Question: 608 Answer: D 1 Pt(s)

OP/0/A/6350/OOIC, (250 VDC Auxiliary Power System) contains the following precaution:

"The DC bus ties will normally remain open. They are only to be closed during equalization charges of batteries, or on a loss of a battery or battery charger. "

Which one of the following is the basis for this precaution? Prevents damage to the battery chargers resulting from both battery chargers simultaneously supplying the same bus at different voltage output Prevents overloading one battery if the battery terminal voltages are significantly different which would lead to excessive hydrogen evolution and a possible explosive hazar Ensures both battery chargers are operated in parallel to be able to reach the terminal voltage (-271 VDC) required for an equalization charg Ensure DC channels remain independent of each other and that a fault on one bus does not adversely affect the other bus

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Distracter Analysis: Incorrect: The reason is DC channel independence Plausible: If 2 battery chargers were run in parallel, they could fight each other if they had vastly different voltage output characteristic Incorrect: The reason is DC channel independence Plausible: If the candidate thinks that batteries with different charge states could fight each other if connected in paralle Incorrect: The reason is DC channel independence - during an equalization charge, the battery being charged is charged from the standby charger due to the high termination voltage required to finish the charg Plausible: DC ties are closed during an equalization charge to allow one battery to power both buses - to allow the charged battery to achieve termination voltage which is higher than normal voltage and may damage equipment if applied on the bu Correct answer Page 71 For Official Use Only Question #54 Ques_608

Question #55 McGuire Nuclear Station SRO Exam Bank Question: 609 Answer. C I Pt(s)

The Unit 1 SRO was monitoring a release from the waste monitor tan Which one of the following alarms would terminate this release automatically? IEMF-31 (Turbine Bid Sump Disch) trip 2 EMF-44(L) (Cont Vent Drn Tank Out) trip 2 EMF-49(L) (Liquid Waste Disch) trip 2 EMF-50(L) (Waste Gas Disch) trip 2

Distracter Analysis: Incorrect: does not monitor the WMT release path Plausible: would terminate a liquid release from the turbine building sump Incorrect: does not monitor the WMT release path Plausible: would terminate a liquid release from the VUCDT using the same automatic valves as the IEMF-49 (WP-35 and WP-46) Correct answer: closes WP-35 and WP-46 to stop the releas Incorrect: does not monitor the WMT release path Plausible: would terminate a release from the WGTD Page 72 For Official Use Only Ques_609

Question #56 SRO Exam McGuire Nuclear Station Bank Question: 610 Answer: C I Pt(s)

Unit 2 is conducting a core reload and one hundred thirty fuel assemblies have been loaded into the core. The following data has been recorded upon completion of each assembly reload sequence group:

Based on the given data, during which reload group (if any) would you predict that the reactor would reach criticality? The reactor will not reach criticalit Distracter Analysis: Using the thumb rule that if the count rate doubles, the reactor is '/z way to criticality, the following calculation shows:

300-600 counts - count rate doubles between reload sequences 10 and 15 For Official Use Only Page 73 Ques 610

SRO Exam McGuire Nuclear Station 2500 + 2000 + 3000 + 2500 + 2000 + 2500 = 14500 PCM added to cause count rate to doubl + 3000 + 4500 = 12000 PCM < 14500 - not critical on reloads 16-18 12000 + 4000 = 16000 PCM > 14500 - beyond critical on reload 19 Incorrect: The reactor will go critical on reload sequence 19 Plausible: based on misapplication of the thumb rule (count rate doubles at criticality). Incorrect: The reactor will go critical on reload sequence 19 Plausible: based on CR doubling rule using # of assemblies added rather than actual reactivity of the assemblie Correct: Incorrect: The reactor will go critical on reload sequence 19 Plausible: if the candidate adds the Ap from Group 9 to the reactivity from groups 10-15 to determine the amount of reactivity that it took to double count rate, this is the answer. In addition, this is the expected outcome when conducting a core reload.

Page 74 For Official Use Only Question #56 Ques_610

SRO Exam McGuire Nuclear Station Bank Question: 611 Answer: D 1 Pt(s)

Unit 2 has just completed a plant shutdown after a record run when a leak was suspected from the relief valve on the waste gas decay tank that had been placed in service at the start of the shutdown. The tank was empty prior to being placed in service for the shutdown. The SRO directs RP to confirm the existence and determine the location of the suspected lea Which one of the following statements would be an effective method of locating the leak in the waste gas system? Radiological Protection could monitor for alpha particle emission from the radioactive decay of entrained tritium ga Radiological Protection could monitor for flammable levels of Hydrogen gas that accumulate in the WGDTs from purging the VC Radiological Protection could monitor for ammonia (NH 4) gas from the breakdown of ammonium hydroxide (NH 4OH), which is added to the NC system for pH contro Radiological Protection could monitor for beta/gamma emission from the radioactive decay of particulate from long-lived fission product gaseous isotope...

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Distracter Analysis: Incorrect: Tritium gas does not emit alpha particles - will not detect Tritium by monitoring for alpha emissio Plausible: Tritium builds up in the waste gas system from reactor operations and has a relatively long half-lif Incorrect: waste gas system recombiners remove Hydrogen during shutdown prior to storage in a WGDT. The Hydrogen gas concentration is reduced below flammable levels prior to storage in the WGDT to assure that it is safe to release to the environmen Plausible: Hydrogen gas is removed from the VCT, PRT and NCDT by the waste gas system during shutdow Incorrect: Ammonium Hydroxide is not added to the NC system for chemistry control of pH. It is added to the condensate system Plausible: Ammonia gas is produced in the NC system by the breakdown of Hydrazine (N2H4) when temperature is raised above 250 'F during startup. Ammonia gas builds up in the pressurizer and Page 75 For Official Use Only Question #57 Ques_611

SRO Exam McGuire Nuclear Station enters the waste gas system during degas operations. It is not removed in the waste gas decay syste Correct answer: The waste gas system would contain fission product gasses.

Page 76 For Official Use Only Question #57 Ques_611

SRO Exam McGuire Nuclear Station Bank Question: 612 Answer: A 1 Pt(s)

A large break LOCA occurred on Unit 1. The operators entered ECA- (Loss of Emergency Coolant Recirculation) for a complete loss of emergency coolant recirculation due to a blockage in the containment sumps, causing large increase in containment temperatures and pressure Which one of the following parameter changes would indicate that significant core uncovery was occurring? Source range instruments show a rapid increase Power range instruments show a rapid increase tEMF-51/52 (Containment TRN A/B High Range) shows a rapid increase EMF-9 (Rx Bid Incore Inst Rm) radiation ARM shows a rapid increase

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Distracter Analysis: The loss of containment cooling will cause core uncovery. All of the answers will show indications of degrading conditions inside containmen Correct answer Incorrect: power range instruments are calibrated to detect high levels of neutron flux and to compensate for gamma flu Plausible: An increase in power range output would indicate recriticality, not uncover Incorrect: EMF-51/2 (Containment High Range Radiation Monitor) would rise as fission products are transported into the containment atmosphere - but would not show a rapid increase when core uncovery occurs. The loss of shielding effect (water shielding EMF-51/2) would be very small compared to the other effects primarily the amount and location of fission products in the containment atmospher Plausible: EMF-51/2 will increase throughout the accident Incorrect: The reactor building incore instrument room is essentially the same area as the Seal Table from the SAMGs. The radiation increase is indicative of a core melt and failure of an incore instrument tube Plausible: Used by SAMGs Page 77 For Official Use Only Question #58 Ques_612

SRO Exam McGuire Nuclear Station Bank Question: 613 Answer. B I Pt(s)

Unit 2 was operating at 100% when the following indications occurred:

Pressurizer level began decreasing

IA NV Pump ammeter showed running amps decreased

Normal letdown was in service If all automatic control system appeared to operate normally, which one of the following conditions would cause the IA NV pump running amps to decrease to the minimum value? NV-238 (Charging Line Flow Control) failed open NV-238 (Charging Line Flow Control) failed closed NI-241 (Seal Inj Flow Control) failed open NV-241 (Seal Inj Flow Control) failed closed

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Distracter Analysis: Incorrect: pump amps and pressurizer level would initially increase Plausible: If the candidate does not understand the charging flow path or does not understand the relationship between pump amps and flo Correct answer: This would block the charging flow path and pumps amps would reduce to minimum as all the charging flow was diverted through NV-i 50 and NV-IS51 miniflow valves Incorrect: this would increase charging flow, which would increase charging pump amps. It would also increase pressurizer level, which would cut back on charging pump speed to offset the flow increase and stabilize the pressurizer leve Plausible: If the candidate thought that this could divert charging flow or did not understand the relationship between pump amps and flo Incorrect: NV-241 closing would increase backpressure on the charging system, which would divert more charging flow through the NCP seals. However, the drop in pressurizer level would act to increase the running speed of the charging pump to compensate for the flow reduction. The overall effect would be to increase pump amps as flow would remain the same but at a higher backpressure.

Page 78 For Official Use Only Question #59 Ques_613

SRO Exam McGuire Nuclear Station This would also cause letdown isolation due to the loss of NV flow through the regenerative heat exchange Plausible: If the candidate does not consider the effect of the charging pump speed control circui For Official Use Only Page 79 Ques613 Question #59

SRO Exam McGuire Nuclear Station Bank Question: 616 Answer. A I Pt(s)

Unit I is operating at full power. Given the following events and conditions on the NCPs:

"* An OAC alarm indicates loss of KC flow to the to the NCP "* The KC supply outside containment isolation valve (IKC-338) is close * Seal injection flow rate to each NCP is 8 gp What are the likely consequences if the operators do not respond to this alarm? The NCPs should operate without KC indefinitel The NCP motor bearings will overheat causing motor damag The NCP stator windings will overheat causing motor damag The NCPs will experience seal failure within 3-5 minute...

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Distracter Analysis: This failure condition is not explicitly covered in the training materials however the candidates should be familiar with the component failure. When KC flow is stopped to the NCPs, this will stop cooling flow to the motor lube oil coolers for the bearings. This will cause lube oil temperatures to exceed allowable values leading to bearing failur Incorrect: Although the NCP thermal barriers can be operated indefinitely as long as seal injection flow is maintained, the loss of KC flow to the motor coolers will cause motor bearing temperatures to overhea Plausible: The NCP seals will operate indefinitely without KC Correct Answer : Motor bearings will overheat Incorrect: NCP Stator windings are cooled by air coolers Plausible: if the candidate forgets that the NCP stators are air cooled Incorrect: seal failure will not occur as long as seal injection is maintaine Plausible: If seal injection is lost to a pump along with KC flow to the thermal barrier, this would be true.

Page 80 For Official Use Only Question #60 QuesL616

SRO Exam McGuire Nuclear Station Bank Question: 617 Answer. C I Pt(s)

Unit 2 was operating at 90% power after a start-up from a refueling outag A PORV is found to be leaking and the associated PORV block valve was shut. The PRT was cooled down to the following PRT conditions:

"* PRT Level - 65%

"* PRT Pressure - 8 psig

"* PRT Temperature - 1 000F

"* Lower Containment Temperature - 118 'F What actions are required to restore and maintain normal operating conditions to the PRT for the long term? Vent/purge the PRT to containmen Continue to cool the PRT to 90" Vent/purge the PRT to the waste gas syste Lower the PRT level to 50%.

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Distracter Analysis: The priority of action to reduce pressure is:

1. Cool the PRT 2. Reduce level 3. Purge to waste gas Incorrect: cannot be performed at power as the vent valve is inside containment and is inaccessible at power Plausible: venting to containment would accomplish the required action Incorrect: Lower Containment temp is 118 'F. Cooling the tank further would only delay the time when it would heat back up again and require further coolin Plausible: Cooling the PRT is the 1V' priority of action to be taken to reduce pressure. Cooling will reduce pressure temporarily but will not allow the PRT to reheat to its normal limit of 114 0F without getting a high-pressure conditio Correct: Incorrect: but will lower level below its normal operating band between 64% and 88%.

Plausible: reducing level will reduce pressure ad is the 2"d priority of action to be taken.

Page 81 For Official Use Only Question #61 Ques_617

SRO Exam McGuire Nuclear Station Bank Question: 618 Answer. B I Pt(s)

Unit I is recovering from a loss of offsite power in ES-0.2 (Natural Circulation Cooldown). The operators reach step 17 which states:

IFA TANY TIME cooldown rate must be raised to greater than 50'F in an hour, THEN GO TO EP/1/A/5000/ES-O. 3 (Natural Cooldown with Steam Void in Vessel)

Given the following plant conditions:

T-hot= 560 OF

NC Pressure = 1225 psig

RVLIS = 100% upper range, 64% lower range

FWST level = 405 inches

All plant equipment is operating as designed

Cooldown rate is 47 0F/hr Which statement correctly describes the condition of the core and the proper procedure flow path?

REFERENCES PROVIDED:

Steam Tables Curve Book Curves 1.IOB, 1.IOC, 1.JOD The core is in a superheated condition - transition to ES-0.3 to continue the cooldown The core is in a superheated condition - remain in ES-The core is in a subcooled condition - transition to ES-0.3 to continue the cooldown The core is in a subcooled condition - remain in ES-Distracter Analysis: ES-0.2 does not provide specific guidance for this transitional step. The EOP bases for this step is to make the transition if:

"* There is limited condensate storage

"* No CRDM fans are operatin The entry conditions for ES-03 are:

"* If cooldown rate must be raised above setpoint

"* If reactor vessel indicates not full and it is determined that depressurization must occur Page 82 For Official Use Only Question #62 Ques_618

SRO Exam McGuire Nuclear Station When evaluating transitions, the operators are required to use the curves in the Data Book instead of steam tables. These curves include an instrument error offset of 20 0F. If the operators refer to steam tables to evaluate subcooling, the core conditions are subcooled (ignoring instrument error). If they refer to the curve, it will show that the core is in the saturation region which means it is below the saturation curve and below the subcooled region

- which means it is superheate Incorrect: Transition to ES-0.3 is not required - no indications of void in the core Plausible: The core is in the "saturation region" of curve 1.1 OB which means that it is below the saturation curve and in the superheat region of instrument error is considered. If the candidate thought that RVLIS readings showed that a void was forming in the core, transitioning would be correc Correct: Incorrect: The core is in the superheat condition and there is no requirement to transition to ES-Plausible: If the candidate used the steam tables to evaluate the core saturation conditions. If the candidate thought that RVLIS readings showed that a void was forming in the cor Incorrect: The core is in the superheat condition Plausible: Remaining in ES-0.2 is correct Page 83 For Official Use Only Question #62 Ques__618

SRO Exam McGuire Nuclear Station Bank Question: 620 Answer. C 1 Pt(s)

Page 84 For Official Use Only Unit I was operating at 100% power when a loss of VI system air pressure occurred. Which one of the following statements correctly describes the condition of the upper containment airlock seals? The seals will slowly depressurize. They can be manually re inflated using SA system air pressur The seals will slowly depressurize. They can be manually re inflated using the VB system The seals will remain pressurized by an air supply from local air tank The seals will remain pressurized by a backup line from the SA syste Distracter Analysis: Incorrect: The seals will not depressurize Plausible: VI provides air to the seal supply Incorrect: The seals will not depressurize Plausible: VI provides air to the seal supply Correct: Incorrect: There is no backup connection to the SA system Plausible: The seals will remain pressurized on a loss of VI.

Question #63 Ques__62o

SRO Exam McGuire Nuclear Station Bank Question: 621 Answer: B I Pt(s)

Unit 2 was responding to a large break LOCA in E-I (Loss of Reactor or Secondary Coolant). Given the following events and conditions:

"* The 4160/600 VAC supply transformer to load center 2ELXD failed

"* Motor control center 2EMXD was deenergized Which one of the following statements correctly describes the actions needed to start containment air return fan 2B? Transfer 2ELXD to transformer 2ELXB Transfer 2ELXD to transformer 2ELXF Transfer 2EMXD to transformer 2ELXF Manually start air return fan 2B

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Distracter Analysis: Containment air fan 2B is used to mix the containment atmosphere during a LOCA to enhance removal of Iodine and other fission products by containment spray. Fan 2B is powered from motor control center 2EMXD through load center 2ELXD. The fan is automatically started by a safety injection signa Incorrect: Prohibited by a Kirk Key interlock Plausible: This would physically repower the fan Correct: Incorrect: Not physically possible Plausible: If the transfer could be physically done, it would repower fan 2B Incorrect: Will not start - no power to the fan Plausible: If the candidate does not determine which motor control center powers containment air return fan 2B Page 85 For Official Use Only Question #64 Ques_621

SRO Exam McGuire Nuclear Station Bank Question: 622 Answer: D I Pt(s)

Unit 2 was operating at 5% power during a plant startup when a total loss of AC power (station blackout) occurred. Given the following events and conditions:

"* The plant was operating within normal limits and bands

"* All protection systems operated as designed

"* All emergency diesel generators failed to start

"* No safety injection occurred

"* No operator action was taken Which one of the following statements correctly describes the response of the reactor trip system? No automatic reactor trip would occur and the reactor would remain critica The shunt coils in the reactor trip and bypass breakers would energize and a reactor trip would occu The under-voltage and shunt coils in the reactor trip breakers would energize and a reactor trip would occu The CRDMs would deenergize and the rods would drop into the cor............

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Distracter Analysis: The loss of the safety bus would cause the rod drive MG sets to lose power and power would be lost to the CRDMs. Below 10%

power, the at-power reactor trips are bypassed so no automatic trip would occur. SSPS would remain energized from the 120 VAC Instrument bus but would not get a trip signa Incorrect: CRDM power would be lost from the rod drive MG sets Plausible: If the candidate thinks that the rod drive MG sets are powered from the 120 VAC instrument bus or a DC bu Incorrect: No trip signal would be generated below 10% power, P-Plausible: SSPS remains energized and could generate a trip signal Incorrect: No trip signal would be generated below 10% power, P-Plausible: SSPS remains energized and could generate a trip signa Correct:

Page 86 For Official Use Only Question #65 Ques_622

SRO Exam McGuire Nuclear Station Bank Question: 623 Answer. A 1 Pt(s)

Unit 2 was operating at 100% power when an alarm was received on the 2B NCP standpipe level. Which one of the following statements correctly describes the cause of the standpipe level alarm? A high standpipe level indicates excessive leakoff through the #2 sea A high standpipe level indicates reduced leakoff through the #3 sea A low standpipe level indicates excessive leakoff through the #2 sea A low standpipe level indicates reduced leakoff through the #3 sea...

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Distracter Analysis: Correct: Incorrect: the standpipe level maintains a backpressure on the #2 seal - reduced leakoff by the #3 seal has no effect on standpipe level as this leakoff goes directly to the RCOT Plausible: If the candidate thinks that the leakoff from the #3 seal effects standpipe leve Incorrect: excessive leakoff from the #2 seal would lead to a higher standpipe level as the standpipe maintains a backpressure on the #2 sea Plausible: if the candidate thinks that excessive leakoff would cause standpipe level to drop or thinks that the standpipe is on the #3 sea Incorrect: reduced leakoff from the #3 seal would not lead to a lower standpipe level as the standpipe maintains a backpressure on the #2 sea Plausible: If the candidate thinks that flow from the #3 seal goes to the standpipe Page 87 For Official Use Only Question #66 Ques__623

SRO Exam McGuire Nuclear Station Bank Question: 624 Answer. D 1 Pt(s)

Unit 2 was operating at 100% power with train B components in service. If a high strainer differential pressure alarm occurs on the 2B RN pump, what statement describes the RN system alignment upon completion of all automatic actions? A RN pump is running from the SNSWP 2A RN strainer is in service A RN pump is running from the low level intake 2B RN strainer is in backwashing B RN pump is running from the SNSWP 2B RN strainer is in backwashing B RN pump is running from the low level intake 2B RN strainer is backwashing

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Distracter Analysis: Incorrect: The low D/P alarm causes the 2B strainer to backwash Plausible: IF the candidate thinks that the low D/P alarm on the B strainer causes the running RN trains to shift to the A train Incorrect: the RN pumps do not shift for a high D/P alarm Plausible: the RN pumps would be manually shifted to the SNSWP for a low suction pressure alarm, not a high D/P alarm Incorrect: the suction does not shift to the SNSWP Plausible: the 2B RN strainer backwashes Correct:

Page 88 For Official Use Only Question #67 Ques_624

SRO Exam McGuire Nuclear Station Bank Question: 625 Answer. B I Pt(s)

Unit I was operating at 5% power following a reactor startup after a refueling outage. Given the following conditions and events:

"* A mixed bed demineralizer that had been isolated at the end of the last fuel cycle was placed in service

"* T-ave=5580 F

"* All systems are aligned normally for the existing plant conditions What will be the effect (if any) on T-ave? T-ave will increase due to the exchange of Lithium ions T-ave will increase due to the exchange of boric acid T-ave will decrease due to the exchange of boric acid T-ave will not change

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Distracter Analysis: Placing a mixed bed demineralizer in service will change reactivity by exchanging boric acid (borate ions) for OH ions. The mixed bed demineralizer was exposed to low boric acid concentrations at the end of core life prior to the shutdown so it reached equilibrium saturation conditions of very low concentrations of boric acid. When placed in service at beginning of core life, it will exchange boric acid for OH ions and reduce the boric acid concentration in the NC syste Incorrect: T-ave will increase due to exchange of boric acid Plausible: pH may increase but not temp due to the exchange of Li ions for hydroxyl ions depending on the saturation state of the resi Correct Answer: Incorrect: T-ave will increase due to exchange of boric acid Plausible: If the candidate reverses the effect of this action Incorrect: T-ave will increase due to exchange of boric acid Plausible: This is the correct answer for cation bed resin which exchanges Lithium for OH - but not boric acid Page 89 For Official Use Only Question #68 Ques_625

SRO Exam McGuire Nuclear Station Bank Question: 627 Answer. C I Pt(s)

Unit I was operating at 100% power when panel board I EKVB was unintentionally deenergized. Which one of the following lists of ESS loads was deenergize Process Protection Channel I Safeguards Test Cabinet Train A SSPS Channel I (Trains A&B)

SSPS Train A Output Cabinet Auxiliary Safeguards Cabinet Train A Process Protection Channel IV Safeguards Test Cabinet Train B SSPS Channel IV (Trains A&B)

SSPS Train B Output Cabinet Auxiliary Safeguards Cabinet Train Process Protection Channel II SSPS Channel II (Trains A & B) Process Protection Channel III SSPS Channel III (Trains A & B)

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Distracter Analysis: Incorrect: powered from EKVA Plausible: if the candidate does not know the power supply Incorrect: powered from EKVD Plausible: if the candidate does not know the power supply Correct: Incorrect: powered from EKVC Plausible: if the candidate does not know the power supply Page 90 For Official Use Only Question #69 Ques_627

SRO Exam McGuire Nuclear Station Bank Question: 628 Answer. C 1 Pt(s)

Unit 2 was responding to a small-break LOCA in E-I (Loss of Reactor or Secondary Coolant). Given the following conditions:

"* Containment pressure = 0.7 psig (at peak pressure for the event)

"* 2ETA was deenergized due to a bus fault

"* The VI header inside containment was depressurized and isolated due pipe rupture

"* The VI system outside containment remained pressurized Which one of the following statements correctly describes the positions of valves 2RV-79A and 2RV-80B?

REFERENCES PROVIDED Station Drawing MCFD-1604-03.00 Flow Diagram of RV RV-79A is open, 2RV-80B is open RV-79A is shut, 2RV-80B is open RV-79A is open 2RV-80B is shut RV-79A is shut 2RV-80B is shut

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Distracter Analysis: These valves are RV containment isolation valve They are air operated and will close on a high-high containment pressure signal (phase B isolation). They fail closed upon loss of operating air pressure. There are no backup nitrogen accumulators to provide operating pressure (as with the PORVs) even though they are safety-related valve In this question, the valves will not auto close because pressure remains below the phase B actuation point (3.0 psig) but they will close due to a loss of operating air pressur Incorrect: the valves are shut due to the loss of VI air pressure Plausible: if the candidate thinks that they fail open or that they are electrically operated Incorrect: the valves are shut due to the loss of VI air pressure Plausible: if the candidate thinks that they are electrically operated and fail shut Page 91 For Official Use Only Question #70 Ques_628

SRO Exam McGuire Nuclear Station Correct Answer: 2RV-79A remains open without a phase B signal and 2RV-80B fails closed due to the loss if VI pressure inside containmen Incorrect: 2RV-79A remains open because no phase B isolation was generated Plausible: if the candidate thinks that a phase B signal was generated or if he does not recognize that VI pressure is maintained outside of containment - or if he thinks 2RV-79A is powered from 1 ETA and fails closed.

Page 92 For Official Use Only Question #70 Ques_628

SRO Exam McGuire Nuclear Station Bank Question: 629 Answer. A 1 Pt(s)

With Unit I was operating at 75% power with rods in automatic control when turbine load drops 10%. Which of the following correctly indicates the change in plant parameters when the transient is complete? T-ave decreases approximately 2 to 30F due to decreased Tre Tave stays the same due to automatic rod motio Tave increases approximately 2 to 3 IF due to rod motio Tave stays the same due to decreased Tre.......................

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Distracter Analysis: T-ave is ramped from 557 TF at 0% power to 585 TF at 100% power. A 10% drop in load causes a 10% reduction in T-ave within the operating band - so T-ave would drop by 10% of 28 TF or - 3 Correct: Incorrect: Tave decreases 10%.

Plausible: If you think rod control compensates for load change by maintaining T-ave Incorrect: Tave decreases 10%.

Plausible. If you think Tref change reflects load increas Incorrect: Tave decreases 10%.

Plausible: If you think Tref decreases and compensates for load decrease.

Page 93 For Official Use Only Question #71 Ques_629

SRO Exam McGuire Nuclear Station Bank Question: 630 Answer. A 1 Pt(s)

Which of the following statements correctly describes the major effect of a failure of a large number of lower ice condenser doors to open for an unisolable main steam line break with an associated SGTR (tube rupture on the faulted S/G) accident inside containmen Containment peak pressure would be higher and would be achieved sooner in the even Containment peak pressure would be higher but would be achieved later in the even Containment sump water inventory would not be adequate after shift to recirculation mod Containment sump water inventory would not be adequate to maintain long-term subcriticality during the cooldow...

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Distracter Analysis: This question is a variation on a bank question which is normally asked from the perspective of a LBLOCA. However, the analysis is also true for a faulted ruptured S/G inside containment. The difference in these events is that the MSL break is a HELBIC and the SGTR is a small break LOCA - so this would cause the water in the containment sump to include the secondary water, which is unborated. In addition, the amount of water from the S/G would be less than the water from a LBLOC Correct: Incorrect: containment pressure would peak sooner Plausible: if the candidate thought that the ice condenser melt was delayed due to the door not opening - causing the release of the cold water to occur later Incorrect: sufficient water is added from the FWST during the injection phase Plausible: the ice water would not be immediately available as melting would be delayed. In addition, the amount of water that enters the sump from the S/G would be less than the water from a LBLOCA. A SGTR is essentially a SBLOCA and water would not be released from the core at a rapid rate to add to the sump before swap over was require Incorrect: the FWST would provide sufficient borated water to maintain shutdown throughout the process Page 94 For Official Use Only Question #72 Ques_630

McGuire Nuclear Station Plausible: release of the borated water from the ice melt would be delayed and thus would not be available until after melting. In addition, the water from the S/G is unborated.

Page 95 For Official Use Only SRO Exam Question #72 Ques_63o

SRO Exam McGuire Nuclear Station Bank Question: 631 Answer: D 1 Pt(s)

The Unit 2 NV system cold leg flow path balance test procedure throttles high pressure injection flow between a minimum value to limit (1)

and a maximum value to limit

_(2)

. Pump overheating pipe erosion Break flow pipe erosion Pump overheating pump runout Break flow pump runout

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Distracter Analysis: Incorrect: pump overheating is not a problem at the higher flow rates and pipe erosion is not a limiting problem for this system Plausible: low flow rates can cause pump overheating and high flow rates can cause pipe erosion Incorrect: safe-end erosion is not a limiting factor Plausible: break spillage is the basis for the minimum throttle limit Incorrect: pump overheating is not a problem at the higher flow rates and pipe erosion is not a limiting problem for this system Plausible: pump runout is the basis for the higher throttle setting Correct:

Page 96 For Official Use Only Question #73 Ques_631

McGuire Nuclear Station Bank Question: 632 Answer: D I Pt(s)

Unit I was responding to a station blackout in ECA-0.0 (Loss of all AC Power). What pressurizer heaters are available to control reactor pressure? Group A backup heaters can be controlled from the SS Group B backup heaters can be controlled from the AS Group C backup heaters can be controlled from the AS Group D backup heaters can be controlled from the SS......................................................-----------------------------.......

Distracter Analysis: Only about 10% of Group D heaters will be available under these conditions Incorrect: Group A has no power and is controlled from the AS Plausible: If confuses group A with D and keys on SSF as the power sourc Incorrect: Group B has no powe Plausible: If associates group B with its correct local control statio Incorrect: Group C has no power and is controlled from the MC Plausible: If doesn't know power supplies and chooses based on logical extension of normal pressure contro Correct:

Page 97 For Official Use Only SRO Exam Question #74 Ques_632

McGuire Nuclear Station Bank Question: 633 Answer. B I Pt(s) Incorrect:

Plausible: Correct: Incorrect:

Plausible: Incorrect:

Plausible:

systems.

the spray valves open and pressure decrease based on a misunderstanding of how the controller work Backup heaters come on but do not control pressur based on misconception of an IPE system "degas" mod There are no block valve based on possible confusion between spray and PORV Page 98 For Official Use Only Unit 2 was operating at 100% power when the pressurizer spray valve failed open. Given the following conditions:

"* PZR Channel Select is in 1-2 position

"* PZR Pressure Control in AUTO

"* NCS Pressure is 2245 psig Which one of the following describes the response of the PZR pressure control system to these conditions? PZR pressure does not decrease because the spray valves will not open below 2260 psig PZR pressure decreases to 1945 psig where the RPS reactor trip PZR pressure decreases to 2210 psig, where the backup heaters take control of pressur PZR pressure decreases to 2185 psig where the spray line block valves clos...................................--

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Distracter Analysis:

SRO Exam Question #75 Ques_633

McGuire Nuclear Station Bank Question: 634 Answer: A I Pt(s)

Which one of the following statements correctly describes the response of the turbine driven CA pump if turbine speed exceeds 4500 rpm? A mechanical flyweight assembly unlatches a trip hook on the turbine stop valve A mechanical flyweight assembly unlatches a trip hook on the turbine governor valve The Woodward governor will generate a signal that trips the turbine stop valve tripping latch assembly The Woodward governor will generate a trip signal that trips the turbine governor valve S--------------------------------------------------------------------------------------

Distracter Analysis: Correct Answer: Incorrect: turbine stop valve trips Plausible: the mechanical flyweight assembly causes the trip Incorrect: the Woodward governor does not cause the trip Plausible: the turbine stop valve is the valve that actually trips Incorrect: the Woodward governor does not cause the trip Plausible: this is the mechanism by which the turbine speed is limited Page 99 For Official Use Only SRO Exam Question #76 Ques_634

McGuire Nuclear Station Bank Question: 635 Answer: B I Pt(s)

Unit 1 was conducting refueling in Mode 6. RP requested the control room operator to independently verify the adjustment of the trip 2 setpoint for 1EMF-16 (Containment Refueling Bridge) area radiation monito If the trip 2 setpoint was required to be set at 1/2/2 decade above the background and background radiation levels were 3.0 mR/hr, what is the correct value for the trip 2 setpoint? mR/hr mR/hr mR/hr mR/hr

Distracter Analysis: V2 decade is 3.16x background above background that is commonly considered 3x background for setting trip 2 setpoint Incorrect: too low - the correct value is 3x3 = 9 mR/hr Plausible: if the candidate thinks that /2 decade = twice background levels Correct Answer: 3x3 mR/hr = 9 mRlhr Incorrect: too high - 9 mR/hr is the correct answer Plausible: if the candidate adds the background level to the calculation-i.e. 3x3+3 = 12 mR/hr Incorrect: too high - 9 mR/hr is correct Plausible: if the candidate thinks that A decade is 5 x background (as one decade is 72 of 10 X background) 5x3 = 15 Page 100 For Official Use Only SRO Exam Question #77 Ques_635

McGuire Nuclear Station Bank Question: 637 Answer. D I Pt(s)

Unit 1 was conducting a reactor startup. Given the following conditions:

"* All shutdown rod banks have been fully withdrawn at 222 steps

"* Control bank "A" rods are being withdrawn at 80 step "* The RPI Urgent Failure Annunciator alarm Which of the following conditions would cause this alarm? A control bank "A" rod is misaligned from its bank position by 8 step Data "A" failure has occurred on one or more rods in shutdown bank "A". The rod control bank overlap unit has detected an improper rod step sequenc A rod in shutdown bank "C" has dropped into the bottom of the reacto....................................................-----------------------------.........

Distracter Analysis: Incorrect: requires a position deviation > 12 steps - and will not withdraw control bank C at this point - overlap unit restricts Plausible: if candidate does not know the criteria for deviation alarm Incorrect: Will not cause an urgent failure alarm Plausible: Will cause a non-urgent failure alarm Incorrect: will not cause an urgent failure Plausible: could confuse with a Non-Urgent Failure alar Correct Answer:

For Official Use Only SRO Exam Question #78 Page 101 Ques_637

2 May 2800 2:B4PM Brian Hakgensen PSHA FAX:

868 739 0333 PACE

Question #78 I Pt(s)

McGuire Nuclear Station SRO Exam Unit I was operating at 60% power. Given the following events and conditions:

"* Pressurizer pressure decreased to 1940 psi "* The SSPS train A low PZR pressure trip logic relay failed to actuat What effect would this failure have on the function of the reactor protection system? The reactor would not trip because the Train A logic relay would not remove power from the UV coil for RT The reactor would not trip because the Train B logic relay would not remove power from the UV coil for RT The reactor would trip because the Train B logic relay would remove power from the UV coil for RT The reactor would trip because the Train B logic relay would remove power from the L'V coil for RTA.

For Official Use Onlv Ques_638.dloc P::nna 7Al

2 May 2H01 2:814PM Brian Haagensen PSHIA FAX:

860 739 8333 PAGE Question #78 McGuire Nuclear Station SRO Exam Bank Question: 638 Answer. C I Pt(s)

Unit I was operating at 60% power. Given the following events and conditions:

"* Pressurizer pressure decreased to 1940 psi "* The SSPS train A low PZR pressure trip logic relay failed to actuat What effect would this failure have on the function of the reactor protection system? The reactor would not trip because the Train A logic relay would not remove power from the IV coil for RT The reactor would not trip because the Train B logic relay would not remove power from the UV coil for RT The reactor would trip because the Train B logic relay would remove power from the UV coil for RT The reactor would trip because the Train B logic relay would remove power from the UV coil for RT Distracter Analysis: Incorrect: The reactor will tri Plausible: based on misunderstanding of RPS redundancy. SSPS Train A had the failure so it makes sense that this could potentially fail to cause the trip Incorrect: The reactor will tri Plausible: based on misunderstanding of RPS redundanc Correct: Incorrect: The Train B logic does not affect RT Plausible: based on misunderstanding of RPS redundancy. If the candidate thinks that SSPS Train B opens RTA.

For Official Use Only Ques_638.doc Page 98

SRO Exam McGuire Nuclear Station Bank Question: 638 Answer. C 1 Pt(s)

Unit 1 was operating at 60% power. If the SSPS train A low PZR pressure trip logic relay fails to respond to a valid sensor trip signal, what effect would this failure have on the function of the reactor protection system? The reactor would not trip because the Train A logic relay would not remove power from the UV coil for RT The reactor would not trip because the Train B logic relay would not remove power from the UV coil for RT The reactor would trip because the Train B logic relay would remove power from the UV coil for RT The reactor would trip because the Train B logic relay would remove power from the UV coil for RT S--------------------------------------------------------------------------------------

Distracter Analysis: Incorrect: The reactor will tri Plausible: based on misunderstanding of RPS redundancy. SSPS Train A had the failure so it makes sense that this could potentially fail to cause the trip Incorrect: The reactor will tri Plausible: based on misunderstanding of RPS redundanc Correct: Incorrect: The Train B logic does not affect RT Plausible: based on misunderstanding of RPS redundancy. If the candidate thinks that SSPS Train B opens RTA.

Page 102 For Official Use Only Question #79 Ques_638

SRO Exam McGuire Nuclear Station Bank Question: 639 Answer: D 1 Pt(s)

Unit 1 is operating at 35% power with all systems in normal alignment for plant conditions. Given the following plant conditions:

The steam pressure input for Steam Flow for "B" Steam Generator fails HIGH (1300psig).

If the operators take no action, which one of the following describes the steam generator level control system initial response to the steam pressure failure until the level error signal counteracts the SM/CF flow mismatch? Indicated steam flow decreases, due to the decrease in density compensation, and S/G Level will decreas Indicated steam flow decreases, due to the increase in density compensation, and S/G Level will increas Indicated steam flow increases, due to the decrease in density compensation, and S/G Level will decreas Indicated steam flow increases, due to the increase in density compensation, and S/G Level will increas...................................--

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Distracter Analysis: Incorrect: indicated steam flow increases and S/G level increases Plausible: If the candidate reverses the effect of steam pressure on density compensatio Incorrect: indicated steam flow increases and S/G level increases Plausible: apparent steam density does correctly increase if steam pressure increases - if the candidate reverses the effect of density compensation on indicated steam flow Incorrect: density compensation increases and S/G level increases Plausible: if the candidate reverses the effect of steam pressure on apparent steam density - and reverses the effect of a density decrease on steam flow - then the answer follows as steam flow increase does cause S/G level to increase Correct:

Page 103 For Official Use Only Question #80 Ques_639

McGuire Nuclear Station Bank Question: 640 Answer. A 1 Pt(s)

Unit 2 is responding in E-I (Loss of Reactor or Secondary Coolant) to a LOCA inside containment. Given the following conditions:

"* Phase B containment isolation actuated

"* Containment pressure remained above 3 psig

"* The FWST level decreased to 20 inche Which of the following best describes the steps necessary to prevent damaging the NS pumps? Reset NS, stop the NS pump Reset CPCS, stop the NS pump Reset containment phase B isolation, stop the NS pump Override CPCS, stop the NS pump Distracter Analysis: Correct: Incorrect: There is no CPCS rese Plausible: based on confusion between NS and CPCS actuation logi Incorrect: phase B will not reset - > 3 psig Plausible: based on confusion between phase B and NS actuation logic - can reset N Incorrect: Overriding CPCS will not reset phase B or N Plausible: based on confusion with CPCS failure actions.

Page 104 For Official Use Only SRO Exam Question #81 Ques_640

SRO Exam McGuire Nuclear Station Bank Question: 642 Answer. D I Pt(s)

Unit I was operating at 100% power when a design basis earthquake caused a loss of all AC power (station blackout). Given the following events and conditions:

"* The suction line to the IA KF pump sheared during the earthquake

"* Spent fuel pool (SFP) makeup was aligned from the FWST to compensate for any loss of SFP level as require "* FWST level was at 300 inche "* The operators entered ECA-0.0 (Loss of All AC Power)

Which of the following events would cause spent fuel pool level to continue to decrease? The ruptured suction line on the 1A KF pump caused the water to be siphoned out of the spent fuel poo The loss of containment and spent fuel pool ventilation fans caused a change in the differential pressure between the spent fuel pool and the reactor cavit The FWST gravity makeup line to the spent fuel pool was not properly isolated and water has been siphoned back into the FWS The standby makeup pump was in operatio...

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Distracter Analysis: Incorrect: The suction line inlet is very close to the surface of the spent fuel pool and a rupture in the line would not cause an appreciable drop in level Plausible: If the suction line inlet was not designed to prevent this event, it would drain the pool Incorrect: The reactor cavity isolation devices are close Plausible: based on misunderstanding plant-operating condition The spent fuel pool level will change if there is a change in differential pressure between containment and the refueling building

- has occurred in the pas Incorrect: FWST makeup does not reverse siphon until below level is below 100 inches.

Page 105 For Official Use Only Question #82 Ques_642

SRO Exam McGuire Nuclear Station Plausible: based on misunderstanding the SFP/FWST design criteri Correct: The standby makeup pump takes suction from the spent fuel pool Page 106 For Official Use Only Question #82 Ques_642

McGuire Nuclear Station Bank Question: 643 Answer. D I Pt(s)

Unit 1 was operating at 100% power when a steam generator tube rupture occurred in the IB S/G. If the operators respond properly in E-3 (Steam Generator Tube Rupture) and isolate the I B S/G, which of the following conditions are indicative of successful isolation prior to commencing the initial cooldown of the NC system? S/G level decreases as S/G water flows back through the break into the NC syste S/G pressure decreases as steam generator pressure equalizes with NC system pressur S/G level decreases as S/G water flows back through the break into the NC syste S/G pressure increases as steam generator pressure equalizes with NC system pressur S/G level increases as NC system coolant water flows through the break into the S/ S/G pressure decreases as steam generator pressure equalizes with NC system pressur S/G level increases as NC system coolant water flows through the break into the S/ S/G pressure increases as steam generator pressure equalizes with NC system pressur..............................................--------------------------------------.....

Distracter Analysis: Incorrect: S/G level increases and S/G pressure increases Plausible: provided for psychometric balance Incorrect: S/G level increases Plausible: S/G pressure increases correctly Incorrect: S/G pressure increases Plausible: S/G level increases correctly Correct Answer:

For Official Use Only SRO Exam Question #83 Page 107 Ques_643

McGuire Nuclear Station Bank Question: 644 Answer. A 1 Pt(s)

Unit 2 has tripped due to instrument technician error during a surveillance test. The moisture separator reheaters (MSRs) did not reset. Assuming no operator action, what effect would this failure have on the plant response to this transient? The NCS would be overcooled because the main steam supply to the MSRs would not isolat Safety injection will actuate on low SG pressure because MSR steam supply valve 2SM-15 fails to clos Safety injection will actuate on low SG pressure because the main steam supply to the MSRs would not isolat The NCS would be overcooled because MSR steam supply valve 2SM-15 fails to clos...

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

Distracter Analysis: Correct: Incorrect: Unit 2 does not have low SG pressure safety injectio Plausible: The cooldown will continue until MSIV closure on low SG pressure. A plausible answer for Unit Incorrect: Unit 2 does not have low SG pressure safety injectio Plausible: The cooldown will continue until MSIV closure on low SG pressure. A plausible answer for Unit 1 which does have low SG pressure safety injectio Incorrect: 2SM-1 5 closure is a manual action if the MSRs do not rese Plausible: based on a misunderstanding of the MSR system configuration.

For Official Use Only SRO Exam Question #84 Page 108 Ques_644

McGuire Nuclear Station Bank Question: 648 Answer: B 1 Pt(s)

Which one of the following correctly describes the normal loading of the 125VDC vital battery chargers? (1) Battery on "charge", (1) 125VDC DC distribution center, (1) 125VDC DC panel board, (1) 120VAC AC static inverte (1) Battery on "float", (1) 125VDC DC distribution center, (2) 125VDC DC panel boards, (2) 120VAC AC static inverter (1) Battery on "float", (1) 125VDC DC distribution center, (1) 125VDC DC panel board, (1) 120VAC AC static inverte (1) Battery on "charge", (2) 125VDC DC distribution centers, (2) 125VDC DC panel boards, (2) 120VAC AC static inverter Distracter Analysis: Incorrect: There are 2 panel boards and inverters, one per uni Plausible: based on misunderstanding of float, and a one unit only perspectiv Correct: Incorrect: There are 2 panel boards and inverters, one per uni Plausible: based on one unit only perspectiv Incorrect: There is only one distribution center normally aligne Plausible: based on a misunderstanding of float, and battery capacity versus normal alignment.

For Official Use Only SRO Exam Question #85 Page 109 Ques_648

McGuire Nuclear Station Bank Question: 651 Answer. D Which of the following actions occur when an emergency diesel generator reaches 95% of rated speed during an emergency start? Generator field flashed and voltage and frequency automatically controlle Low lube oil pressure trip reinstated and starting air secure Generator field flashed, and starting air secure Low lube oil pressure trip reinstated and voltage and frequency automatically controlled.

Distracter Analysis: Incorrect:

Plausible: Incorrect:

Plausible: Correct:

Field flash occurs at 40%.

based on a logical progression for auto start logi Starting air is secured at 40%.

based on misunderstanding of starting air reset setpoin Field flash and starting air secured occur at 40%.

based on misunderstanding of correct setpoint.

For Official Use Only 1 Pt(s)

SRO Exam Question #86 Page 110 Ques_651

McGuire Nuclear Station Bank Question: 653 Answer. C 1 Pt(s)

During preparation for a waste gas release, the pre-release surveillance test of I EMF-50(L) (Waste Gas Disch (Lo Range)) revealed that there was no response to the source check. IAE reported that the scintillation detector had failed. What action is necessary to begin releasing the waste gas decay tank? Take manual grab samples of the waste gas decay tank prior to any gaseous waste releas Source check the GM detector for tEMF-50(H) (Waste Gas Disch (Hi Range)) prior to any gaseous waste releas Verify 1EMF-36(L) (Unit Vent Gas (Lo Range)) is in service prior to any gaseous waste releas Repair 1EMF-50(L) prior to any gaseous waste releas......

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Distracter Analysis: Incorrect: Either EMF-50 or 36 must be in service or unit vent samples take Plausible: based on misunderstanding of the sample locatio Incorrect: IEMF-50(H) does not provide automatic protection for the waste gas release path - the correct EMF is I EMF-50(L) - which is NOT a GM detector Plausible: based on 1 of 2 detectors faile Correct: Incorrect: EMF-36 can substitute for EMF-5 Plausible: based on not knowing any others are correct.

Page 111 For Official Use Only SRO Exam Question #87 Ques_653

McGuire Nuclear Station Bank Question: 654 Answer. A I Pt(s)

Which one of the following interlocks is designed to prevent a water hammer in the RC piping if the RC pumps trip? The vacuum breaker valves automatically open if all RC pumps tri There is a 45 second time delay before pump discharge valves close to allow coast dow The pump discharge valves remain open until flow decreases below a preset valu The pump discharge valves close over 120 seconds while the pump is coasting dow......

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Distracter Analysis: Correct: Incorrect: There is no time dela Plausible: based on a logical alternativ Incorrect: The valves close regardless of RC flow rat Plausible: based on a logical alternativ Incorrect: Not designed to preclude water hamme Plausible: based on a logical alternative.

For Official Use Only SRO Exam Question #88 Page 112 Ques_654

McGuire Nuclear Station Bank Question: 655 Answer. B I Pt(s)

On May 19 th, the NLO was directed by the unit supervisor to perform a sequence of steps using a working copy of a procedure in progress that had previously been correctly validated against the controlled copy on May I" Which one of the following statements correctly describes the required actions of the NLO? Perform just the designated steps as directed using the existing working cop Re-validate the working copy of the procedure and perform just the designated steps from the existing working cop Obtain a new working copy of the procedure and perform just the designated steps from the new working cop Obtain a new working copy of the procedure and inform the shift supervisor that all procedure steps must be performed or validated from the first step in the procedur......

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Distracter Analysis: Incorrect: the working copy cannot be used unless it has been validated every 14 days Plausible: if the candidate thinks that the validation requirement exceeds 14 days Correct Answer: Incorrect: There is no requirement to obtain a new working copy if the validation has exceeded 14 days - in addition this would now have working 2 copies of the procedure with completed steps initialed on each - would be hard to keep track of the configuration control Plausible: if the candidate thinks that once the working copy has exceeded its validation requirement, it must be replaced Incorrect: There is no requirement to obtain a new working copy if the validation has exceeded 14 days - or to revalidate all steps in the procedure Plausible: This answer is overly conservative - but some candidates might select the most conservative answer if they do not know the requirement.

For Official Use Only SRO Exam Question #89 Page 113 Ques_655

SRO Exam McGuire Nuclear Station Bank Question: 657 Answer: D 1 Pt(s)

During a surveillance test of NI system valves, 2NI-9A (NC COLD LEG INJ FROM NV) did not respond to a safety injection signal. The SOM requests that the OSM issue a special order pre-assigning a dedicated operator by name on each shift to open 2NI-9A should a safety injection signal occur before the cause of this condition is correcte Which one of the following statements best describes the requirements for issuance of this special order? The OSM is not authorized to issue special orders, only the SOM can issue a special orde The special order cannot be authorized for any situation when a procedure change is require The special order cannot be issued until after an operability evaluation has been complete The special order cannot be issued until after the procedure change has been issue..........................................------------------------------------------------

Distracter Analysis: Incorrect: OSMs are authorized to issue special orders at McGuir Plausible: based on the standard industry practice having Ops Mgr issue all special order Incorrect: Special orders can be issued after a procedure chang Plausible: based on guidance prohibiting issuance until a procedure change is implemente Incorrect: Special orders can be issued after an operability evaluatio Plausible: based on operability evaluations coming from NSD-20 Correct:

Page 114 For Official Use Only Question #90 Ques_657

McGuire Nuclear Station Bank Question: 659 Answer. B I Pt(s)

Unit 1 is at 1% power, starting up from a plant trip due to multiple power range nuclear instrument failures. Unit 2 is shutting down (30% power) to Mode 3, to investigate the potential common mode failure mechanism. The Unit 2 power range nuclear instrument channel N41 has been tagged out in preparation for the investigatio Which of the following best describes the TSAIL entry for power range nuclear instrument inoperability during this maintenance for Unit 2? No TSAIL entry is required because N41 will not be required to be operable in Mode A TSAIL entry is required because N41 is inoperable in Mode No TSAIL entry is required because N41 will be within the action statement time limit A TSAIL entry is required for tracking only

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Distracter Analysis: Incorrect: A TSAIL entry is require Plausible: based on knowledge that no entry would be required in Mode Correct: tagging N41 out of service makes N41 inoperable regardless of the outcome of the investigation into the common mode failur Incorrect: A TSAIL entry is require Plausible: based on misunderstanding of TSAIL entry requirement Incorrect: A TSAIL entry is required due to N41 inoperability in Mode Plausible: based on the requirement for a "tracking only" entry.

For Official Use Only SRO Exam Question #91 Page 115 Ques_659

McGuire Nuclear Station Bank Question: 660 Answer: D 1 Pt(s)

Unit 2 was operating at 100% power. There is a packing leak on a VI system containment isolation valve inside lower containment. You are reviewing an RWP that controls the work permit to inspect and repair the VI valv What are the minimum dosimetry requirements for this job? Thermoluminescent and neutron dosimeters Electronic alarming and neutron dosimeters Electronic alarming and pocket ion chamber dosimeters Thermoluminescent and electronic alarming dosimeters

Distracter Analysis: All personnel entering an RCA must have a real time reading dosimeter to alert them of dangerous radiation levels and a permanent record dosimeter to record their exposure for their legal recor In addition, neutron exposure is measured by either measuring neutron radiation levels in the general area and multiplying by actual stay time, or by measuring the ratio of neutron dose to gamma dose and then ratioing the TLD dose measuremen Incorrect: Must have a PIC or EA Plausible: if the candidate focuses on the requirement to monitor neutron exposure - inside containment of an operating reacto Incorrect: Must have a TLD - no legal record of gamma dos Plausible: if the candidates think that EADs or PICs are permanent record dosimeter Incorrect: No neutron dosimetry and no legal record of dose receive Plausible: if the candidate is not aware of requirement to have a TLD Correct: Neutron dose can be measured by ratioing the gamma dose to the neutron dose and using the TLD readout.

For Official Use Only SRO Exam Question #92 Page 116 Ques_660

McGuire Nuclear Station Bank Question: 661 Answer. A I Pt(s)

Units 1 and 2 are at 100% power. Given the following conditions:

Unit 2 has experienced 2 fuel pin failure *

The mechanical seal has failed on NI pump 2 *

The NI-2B pump room general area is 200 mrem/h *

In order to reach the NI-2B pump room the workers must transit through 6 rem/hr high radiation area for 1 minute and retur *

Worker A has an accumulated annual dose of 400 mrem, respectivel How long can worker A participate in the seal repair on NI Pump 2B without exceeding the alert flag exposure limit for external exposure? No longer than 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months No longer than 5.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months No longer than 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months No longer than 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months

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Distracter Analysis:

The candidate should determine that the alert flag exposure limit is 80% of 2000 mrem admin limit = 1600 mrem Transient exposure is 200 mrem (6000mrem/hr x 2/60hr). (During transit to and from job).

400 mrem + 200 mrem = 600 mrem 1600 mrem - 600 mrem = 1000 mrem allowable before reaching alert flag exposure admin limit 1000 mrem / 200 mrem/hr = 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Correct: Incorrect: The answer is 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Plausible: based on calculating a one-way transit dos Incorrect: The answer is 5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Plausible: based on no transit dos Incorrect: The answer is 5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Plausible: based on using admin limit (2000) versus alert flag.

Page 117 For Official Use Only SRO Exam Question #93 Ques_661

McGuire Nuclear Station Bank Question: 662 Answer. A I Pt(s)

Which one of the following statements correctly describes the use of Severe Accident Management Guides (SAMGs) by the control room operators? Control room operators are the implementers for the SAMGs, which are used in place of ERG Control room operators are the implementers and SAMGs are used simultaneously with ERG Control room operators are evaluators for the SAMGs, which are used in place of ERG Control room operators are evaluators and SAMGs are used simultaneously with ERG...................................-----------------------------------------------------...

Distracter Analysis: Correct Answer: Incorrect: SAMGs replace ERGs when implemented Plausible: partially correct - the control room operators are the implementers Incorrect: control room operators are implementers Plausible: partially correct - the SAMGs are used in place of the ERGs Incorrect: control room operators are implementers - SAMGs replace the ERGs Plausible: psychometric balance Page 118 For Official Use Only SRO Exam Question #94 Ques_662

McGuire Nuclear Station Bank Question: 663 Answer. C 1 Pt(s) Incorrect: NS train was inoperable Plausible: based on failure to know the effect of manual closure of the valv Incorrect: NS train was inoperable Plausible: based on failure to isolate the valve during maintenanc Correct: Incorrect: Not operable until NS-32A is cycle Plausible: based on failure to know the effect of manual closure of the valve.

For Official Use Only With the plant at 10% power, an Instrument Technician was allowed to adjust the limit switches on NS32A (containment spray pump A discharge containment isolation valve) without a tag-out. He cycled the valve using the manual hand wheel to set up the limit switches. Upon completion of the work, NS32A was manually closed using the manual hand whee Which one of the following statements is correct concerning this maintenance activity? "A" NS train was operable during the maintenance but NS32A must be cycled electricall "A" NS train was operable during the maintenance because NS-32A would have opened on deman "A" NS train was inoperable during the maintenance and NS-32A must be cycled electricall "A" NS train was inoperable during the maintenance, but is now operabl..........................................------------------------------------------------

Distracter Analysis:

SRO Exam Question #95 Page 119 Ques_663

McGuire Nuclear Station Bank Question: 664 Answer: D 1 Pt(s)

Unit 1 was conducting control rod drop tests during a plant startup at 4%

reactor power when a complete loss of RN occurred. Given the following events and conditions:

"* Control room operators enter AP/20 (Loss of RN)

"* Several NCP motor stator windings exceed 311 'F

"* The operators manually trip the reactor but the trip breakers fail to open

"* Reactor power is 5%

"* Pressurizer pressure = 1930 psig Which one of the following statements correctly describes the proper procedural flow path for these conditions? Remain in AP/20, trip all NCPs and commence a reactor shutdow Implement FR-S.I (Response to Nuclear Power Generation/ATWS) concurrently with AP/2 Terminate AP/20, enter E-0 (Reactor Trip or Safety Injection)

and immediately transition to FR-Enter E-0 and immediately transition to FR-S.t while continuing on in AP/20 as time and conditions permi...........................

...

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Distracter Analysis: Incorrect: AP/20 directs the operator to enter E-0 in step 2 Plausible: E-0 entry is not required if< P-II and performing control rod drop tests below 5% power per OMP 4-3 Incorrect: Do not enter FRPs unless ERGs have been implemented Plausible: Meet the criteria for RED PATH on subcriticality CSF Incorrect: should not terminate AP/20 as still have a loss of R Plausible: the priority is to make the reactor subcritical Correct Answer:

Page 120 For Official Use Only SRO Exam Question #96 Ques_664

McGuire Nuclear Station Bank Question: 665 Answer: A Unit 1 was shutdown in Mode 4 preparing to enter Mode 5 during a forced outage to repair a problem on INS-I8A (A NS Pump Suct From Cont Sump). Given the following events and conditions:

IA and lB ND pumps were in operation INS-I8A was tagged shut The following indications existed on the iNI-185A A ND & NS) controls (RB Sump to Train Derse Which one of the following actions could cause a loss of shutdown cooling by draining the ND system to the containment sump? Manually depress the "OPEN" pushbutton for 1NI-185A Manually push the "REL" pushbutton for 1NI-185A Manually push the "SS RESET" pushbutton on 1NI-185A The "S LATCHED" light illuminates for 1NI-185A

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Distracter Analysis: The normal ND valve interlock between IND-19A and INI-185A prevents opening INI-185A when IND-19A is ope This interlock can be bypassed when the BYPASS pushbutton is depressed as long as INS-I 8A is close Correct Answer: Allows flow path between ND system and containment sump - ND drains to containment Incorrect: This will only release the bypass pushbutton For Official Use Only 1 Pt(s)

S SRO Exam Question #97 Page 121 Ques_665

McGuire Nuclear Station Plausible: if the candidate does not know the function of this control Incorrect: This action restores the INI-185A interlock with IND 19A and resets the S latch Plausible: if the candidate does not know the ND valve interlock controls Incorrect: S latch allows auto opening of 1NI-185A on FWST low level and allows manual opening of INI-1 85A - but the valve is not manually opened in this distracter Plausible: If a safety injection signal were to occur, INI-1 85A would auto open when FWST level drops to the low level alarm but 1ND-19A would then auto close For Official Use Only SRO Exam Question #97 Page 122 Ques_665

McGuire Nuclear Station Bank Question: 666 Answer: D 1 Pt(s)

Unit 1 was operating at 100% power when a total loss of offsite power occurred. Given the following events and conditions:

"* The diesel generators started and loaded as designed

"* The operators completed E-0 (Reactor Trip Response)

"* The operators reached step 1 I of ES-0. I (Natural Circulation Cooldown)

which requires the cooldown of the NC system Which one of the following components are necessary to prevent the formation of a void in the reactor vessel while cooling down the plant? NI pumps Head vent VL/VU fans CRDM fans

......

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

Distracter Analysis: Incorrect: NI pumps are not required - SI is blocked Plausible: If the candidate thinks that SI is required to prevent void in reactor vessel Incorrect: not required for cooldown Plausible: head vent would relieve a void in the reactor vessel but would not prevent a void Incorrect: not required for cooldown Plausible: these are another set of fans in containment that provide cooling to various components Correct answer For Official Use Only SRO Exam Question #98 Page 123 Ques-666

McGuire Nuclear Station Bank Question: 672 Answer. C I Pt(s)

Which one of the following conditions would cause IEMF-51A (Containment TRN A (Hi Range)) to increas An increase in alpha radiation from a tritium leak A cloud of radioactive gas that emits beta radiation An increase in gamma flux from a failed fuel event An increase in neutron radiation from a criticality event S--------------------------------------------------------------------------------------

Distracter Analysis: Incorrect: does not respond to alpha radiation - nor does tritium emit and alpha particle Plausible: a type of radiological hazard - provided for psychometric balance Incorrect: does not respond to beta radiation Plausible: some detectors respond to beta such as scintillation detectors Correct answer Incorrect: does not respond to neutron radiation Plausible: would seem appropriate to measure neutron radiation for criticality events For Official Use Only SRO Exam Question #99 Page 124 Ques_672

McGuire Nuclear Station Bank Question: 676 Answer. B I Pt(s)

Unit 1 was operating in Mode 4, shutdown cooling. Given the following conditions and events:

"* A mixed bed demineralizer was being pre-treated in preparation for placing it in servic "* The operators had initiated automatic makeup to compensate for a reduction in pressurizer leve "* Both trains of ND were in servic Which one of the following statements correctly describes a condition which would cause an inadvertent dilution of the NCS? NCS Boron depletion had occurred over the life of the core in the NC system and initiation of automatic makeup caused a dilution even INVSS5450 (BA FLOW CNTRL) for INV-267A had been incorrectly set to 5.0 when it should have been set at INSS5460 (BA BLEND DISCH CNTRL) for INV-252A had been incorrectly set to 4.0 when it was required to be set at A cation bed demineralizer that had previously been in service at the beginning of the fuel cycle and was not pre-treated was inadvertently placed in service in place of the pre-treated mixed bed demineralize...................................----------------------------------------------------...

Distracter Analysis: Incorrect: Boron depletion causes the amount of B' in the NC system to be reduced due to neutron flux interactions. This reduces the effective cross-section compared to the same concentration of Boric acid that has not been exposed to neutron flux in the NC system. When blended makeup into the NCS occurs, the boric acid contains a higher proportion of B'0 for the same concentration of boric acid. This effectively results in adding makeup water that has higher effective concentration of B'0.

Plausible: If the candidate reverses the effect of boron depletion thinks it causes positive reactivity to be added - not negative reactivity. This phenomenon was described in PIP I-M99-2394 (included in lesson plan PS-NV).

For Official Use Only SRO Exam Question #100 Page 125 Ques_676

SRO Exam McGuire Nuclear Station Correct Answer: This problem would cause a decrease in the addition rate of boric acid to the blender which would decrease the concentration of the boric acid that was added to the NC syste This would cause a dilution of the NC system boric acid concentratio Incorrect: A reduction in the set point for the controller for total blended flow would cause 1NV-252A to throttle makeup flow to the blender which would effectively increase the concentration of boric acid added to the NC system - which would not cause the opposite of a dilution even Plausible: if the candidate reverses the effect of the setpoint of 1NV-267 Incorrect: adding an untreated cation demineralizer that had last been in service at the start of the cycle would not cause a dilution event. Cation resin does not achieve equilibrium with boric acid (as does anion resin in mixed bed demineralizers which exchanges borate ions for OH ions) and no change would occur to boric acid concentration. In addition, the cation demineralizer was in service at the beginning of the fuel cycle when NCS boric acid concentration was highest - so any residual water in the demineralizer would be at a higher concentration of boric acid - not lower - causing a boration not dilutio Plausible: if the candidate reverses the effect of adding the demineralizer.

Page 126 For Official Use Only Question #1o00 Ques_676

INITIAL SUBMITTAL MCGUIRE EXAM 2000-301 MAY 8-MAY 22 50-370/2000-301 12, MAY-25, 19, 2000 INITIAL SUBMITTAL WRITTEN EXAMINATION

[lE / s /

A)

50-369/2000-301 AND SRO Aul".

ex Ye

McGuire Sample Plan PWR SRO Examination Outline ES-401-4 Facility: McGuire Date of Exam: 5119/00 Exam Level: SRO K/A Category Points Point Tier Group K

K K

K A

A A

A

2

4

6

2

4 G

Total T

1

5

4

2 Emergency &

1

3

4

16

Abnormal

1

0

1

3

Plant Tier Evolutions Totals

7

8

4

2

0

1

3

2

1

2

3

1

2

1

17

Plant

0

1

0

4 Systems Tier Totals

3

5

3

6

3

40

Generic Knowledge and Cat I Cat 2 Cat3ICat4 Abilities

4

17

arget

6

9

0

3/21/00 For Official Use Only Summary

McGuire Sample plan

[ES-A01 offrmF..i IES.1403 Bank r

EnoeiencneandAbno.PIMsa Thntos~o a-pr s

ýzsu

/APIE S/ Nary I Safety Function K/

K3AI A Topic(s) Pit Queswo 0026 Loss of Condensert CooI1 Wool I VVu M0029 Anticipated Transient -I-Soim 11I

,acOl0 Steam Lif Ruptur. -Enom.N.llHeast Transfer IV MOBE0 RCS 0yarcotling - pTS I IV MOB0 Ing mT' lV 000055 Station Blackout I VI 000057 Lose of Vital Alc Ete. Inst. Rue fVI 000059 Accidental Liquid Radwaste R.I. I IX 000002 Loss of Nuclear Service Wear I IV 000087 Plant FIre On-itle IX fiCsO$ Control Room EVrc. l VIi 000089 (WE'14) Loss of CTMT Integrlty I V 000074 (W1M0 E07) lnod. Core Coolg I IV 76 igh Reactor Coolant Activity f IX WM02 St lenosinestion f III K/WACa a Totala:

100001 Continuous Rod Wiithdrawald I 1 000003 Droppedl Control Rod 11

)0cat I~pr:a;-bleffituck Control Rod 11 0011 Large BreMk LOCA ll WIE04 LOCA Outside Conainirent I III WIE02 SI Te, lnat.os. I II1 000015417 RCP Malfunction I IV F

o0o Natural Circ. AV 00024 EmerlgenCy Bontion 0 I

I 20W 261 2.01 2.03 F307 3.01 302 304 3,07

'V

'.3

1.05 1Of 1,02 J~I I

2,2 209 I

I 4,41 4 24

.----

5-

.s,5nInsuAlr.has no4gof threasonsfort o

e chp iitforeTa 101 422

_j4,1 A 344.2 41/4 A 3.84W Ability to Interpret co nmroom indications to verity the statuls and operation of system and how operator actions and directives affect plant system condions nowledge of tie operational implications of tie follOWing concepts. naIural drecliation and cOoling including mrieis boling Mbileiy Wo operate andlor nmonitor tMe following. desired operating results during aorarl and emergency situatons Mible, to delemnine and interpret tie following, adherence to appropnate procedures and operation vitisn liltationsl In tie fatliies license and amendmerlts Ability to detennine and interpret, when to secure RCPs on high stator tenperaltres Knowledge of tie operatonal implicalions of tMe folowing concepts as they apply to te.. coiponanls. capacity and funcionl of emergency systems Knowledge of the operational implications of tie followng concepts..relationship between boron ad/o and reactor power Knowledge of loss of cooding water procedures nowledge of the inteelationships between.breaken, relays and disconnects Ability to operate and/or monitor,,manual and automatic RPS tip initiation

"Knoledgoe of the int*,srsationships between

,components and furncons of control and safety systems Including instromentation, signals, interlocks failure modes, and aUtometic end manual feat ures

"Knotsedge of tie masons for mse following responses loss of steam dump capability upon toes of condenser vacuum

"Knoedge of the reasons for tMe followng responses. actions contained in EOP for toes of onsite end

ctite power Ability to operate and/or montor manuel control of ompornent forvwhich automatic conbol is lost

"Knoold oftihe nttemnialiorships between radioacbve liquid monitors Knowtedge of bases for piodrzing safety functions dunng abnormal nemergency operations Knowedge of tie masons for tMe following responses.s actiOs contained in EOP for plant I. onsite

"Knowledge of ie rmasons for tie folloWing responses maintenanco Of SGo levels Using AIW Conbol valves Knowledg, e ofe ointerelltionships between, personnel access hatch and emerogency

.cces hatcht Ability to operate and/o monitr.RCS cooldown rate 1, Knosedge of ibnormal condition procedures

,nnidedg aof -

on*errelaioships beIwdeen. facilitys heat removal system inc1uding primary coofant. emergency coolant, decay heat removal systems, and Aol 601 507 602 242 666 592 241 593 372 594 595

609 604 596 501 597 598 307 i= I~l I

-

-~

.-..

J

-.

.I I

4.

Ii -

For Official Use Only EAPEs TI G1 3/21/00

. 1 map P.... ý.os 1.1

0 jily W discrete ancor morc t

t, reftchts 4 W42

I I d

5

41 i

4 3/4.;

401 3 5.4/'.0/ /.3/3 7 29"/31V 3.4/ n13 4 3/ /3,5 2 7/2 3,0/ /.0/43 2 B-/2,1 394 /38 348356

I

1

SO U -3dý Alýo m Pý I-J WtM Md qd..S.1-0-W

McoGuie Sample plan System I / N K 002 Realcto Coolant N Emergency Core Cooling 10 preaeurzr Pre r

osurf Control oil presurIlzer Level Control 012 Reactor Prove,,ten 011 Noon uclearlnetrufenttlofl 027 Containment Ionn Removal 028 Hydronen Recom.bine' and Purge Control o0n Contalnment Purge 033 SOpnt Fuel Pool CoOling 634 Fuel Handling Equitpentio 035 Stast GneraMrtor 039 lain and Reheat Stail us5 Condarenr Aulr Reoval 162 AC Electrical Distribution N05 artagency Diesel GeNIerato 073 Proce.. Radiation Monitoring 078 Circleating Water 079 Station Air ON. Fia protection 103 Con"inmernt WVA Catol Totins:

1K 2K 2.01 I12 201 1.05

22 pi m

I~deke de kfifrefpoor, ES.A 4-31K 41K5KGAILI K/A Topclus l I uesrtiaon 1.00 Ablity to predict end/or monitor changes in paMora ters (O prevent exceeding design limits) associated with operating the RCS fT-ve 37/38 Sya.....

..

I I

A 3A 2.0M.1/4 3 37131 Knovweoge of desin featareis) and/or Interock(s) which provide for HPI flow throttling

KnOWledge of bius power supplies to.. PzR heatder

Ability to prdiact the impacts or the foloving malfunction or operalloit and based on those predictions. use pnocures to correct, control or mitigate the

orssqencet o n..Inadvertent praessurizer spray edumellon

Knoledge of the effect that a loss or malfundion Will have on.hlp Iogi circults

Koteledge o phoyslcal connections adrn cause and e-eci relationshlpe S/.

Knorledge Of hus power supplies to fanos Ahitty o predict and/or monitor chanrges In peramtee (to prevent enceeding design linitsi assocated with operating the saent fuel pool w.ter level Ability to InthepreetMontmrl roon InducatIons to verify the status and operation of sysiem, and underthand how operator acrlono and directives affect plant end system conditions Abilit y to manually operate and/osor monitor it the control roof, 5/C isolationl on Mean leak or tlte npoture/leak Knlvoiedge of the Wedfet hat a Imross nu alfurmnco, will have on RICS Knoweadifd of the effied~ that a hasS or Malfunction Wrll hae. on, mupr Knowledge of design leature(s) and/or interlockls) which provide for-nrooMplate-sarln rlay Kbnllny ad predige of th eafolohn rolloonm omalftnhazon ar operasion and headed on Mm desamqerlos, mse pproedures to orrent, contrOl or Milionath the mannaeqenods of., madetcoo failure K*n ýowld of d asign feature(s} ando interloc (s) whih povie fo a d canot Man 1AS K on ý W 9a o f t h e f allO ed o p e ml io fl l im pi icl l to n h a z od$ to I n o nnlo'n at u

result of fire type or nethods o1 suppreslon Knowledge of physical onnections and/or cause and effec relationspltlp personnel accs hahandMaroci emeny aosccess hatch 631 632 SYSTEMS T2 G2 838 639 021 642 810 843 844 243 851 853 451

-

=8k Jl Trhh r2Q a u

.1. 1-M/,4" 275/3 3.5.5/ /327 2./3 2 207.2 5 I

on9

McdumUi.Smpi. Pla F., Offdal ti 0*t SYSTEM. W2 G3

=~I'm

McGulre Sample Plan ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-5 Exam Facility: McGuire Date of Exam: 6119100 Level: SRO Bank Category K/A #

Topic im Points Question 2.1.21 Ability to obtain and verify controlled procedure copy 3.1/.1.27 Knowledge of system purpose or function 2.8/ Conduct of 2.1.32 Ability to explain and apply all system limits and precautions 3.4/ Operations 2.1.15 Ability to manage short term information such as night and standing orders 2s3/30

657 Total

4 2.2.23 Ability to track limiting conditions for operations 2.61.28 Knowledge of new and spent fuel movement procedures 2.6/35

216 Equipment 2.2.22 Knowledge of limiting conditions for operations and safety limits 3.4/ Ability to manipulate the console controls as required to operate the racility between Control 2. shutdown and designated power levels 4.0/.221 Knowledge of pre and post maintenance operability requirements 2.3/ Total 5 6 2. Knowledge of 10 CFR 20 and related facility radiation control requirements 2.6/ Knowledge of radiation exposure limits and contamination control, including 2. permissible levels in excess of those authorized 2.5/ Ability to perform procedures to reduce excessive levels ot radiation and guard Radiation 2.3.10 against personnel exposure 2.9/ Control 2. Knowledge of use and function of personnel monitoring equipment 2.3/ Total

41 Knowledge of how the event-based emergencylabnomal operating procedures are 2. used in conjunction with symptom-based EOPs 3.0/.4.39 Knowledge of the RO's responsibilities in emergency plan implementation 3.3/ Ability to diagnose and recognize trends in an accurate and timely manner utilizing Emergency 2.4.47 the appropriate control room reference material 3.41 Procedures 2.4.13 Knowledge of crew roles and responsibilities during EOP... use 3.3/ and Plan Total

4 Tier 3 Point Total

17

3/21/00 For Official Use Only Generics

IR 05000369/2000301

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2000 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(1) 10(0) Security Operator Exams 301(0) Emergency Planning Other

IR 05000369/2000301

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