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2015 Davis-Besse Nuclear Power Station Initial License Examination Proposed RO-SRO Written Examination
	ML15217A607
Person / Time
Site:	Davis Besse
Issue date:	04/09/2015
From:	; Division of Nuclear Materials Safety III
To:	;
	Reeser, D.
Shared Package
ML15217A093	List: ML15217A607; ML15245A236; ML15245A261; ML15245A264; ML15245A266;
References
	Download: ML15217A607 (103)
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Davis-Besse 1LOT15 NRC Written Exam r1

1. Which of the following describes the purpose and function of the Anticipatory Reactor Trip System (ARTS)?

The purpose of ARTS is to initiate a reactor trip to minimize the __(1)__.

ARTS initiates a reactor trip by opening contacts in series with the Control Rod Drive Trip Breaker

__(2)__.

A. (1) severity of a Main Feed Water line break accident (2) Shunt Trip Coils B. (1) severity of a Main Feed Water line break accident (2) Undervoltage Coils C. (1) probability of actuation of the Power Operated Relief Valve (PORV)

(2) Shunt Trip Coils D. (1) probability of actuation of the Power Operated Relief Valve (PORV)

(2) Undervoltage Coils Answer: D Explanation/Justification:

A. Incorrect - minimize PORV lifting is ARTS purpose per Tech Spec Bases 3.3.16. ARTS trips are in series with UV coils - See Tech Spec Bases 3.3.4. MFW break is plausible because it is one of the design bases events for SFRCS and an SFRCS trip causes an ARTS trip. See Tech Spec Bases 3.3.11. Shunt Trip Coils is plausible because the shunt trip is actuated by UV sensing relay in parallel with CRD breaker UV coil, so Shunt Trip will trip when ARTS trips.

B. Incorrect - minimize PORV lifting is ARTS purpose per Tech Spec Bases 3.3.16. MFW break is plausible because it is one of the design bases events for SFRCS and an SFRCS trip causes an ARTS trip. See Tech Spec Bases 3.3.11.

C. Incorrect - ARTS trips are in series with UV coils - See Tech Spec Bases 3.3.4. Plausible because shunt trip is actuated by UV sensing relay in parallel with CRD breaker UV coil. Part 1 is correct.

D. Correct - PORV - see Tech Spec Bases 3.3.16; UV Trip - See Tech Spec Bases 3.3.4. See also DB-OP-06403 R20 RPS and NI Operating Procedure Attachment 4.

Sys # System Category KA Statement 000007 Reactor Trip Generic Knowledge of the purpose and function of major system components and controls K/A# 2.1.28 K/A Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

Tech Spec Bases 3.3.16 and 3.3.4; DB-OP-06403 R20 Attachment 4 Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41.7)

Objective: OPS-GOP-303-05K

Davis-Besse 1LOT15 NRC Written Exam r1

2. The plant is operating at 100% power.

The following conditions are noted:

4-1-D PZR RLF VLV OPEN alarm

Containment Air Cooler 1 Suction Temperature TI1356 is 160 ºF

Containment Air Cooler 2 Suction Temperature TI1357 is 155 ºF

Computer alarm T770 RC PRZR PRESS RLF OUT TMP, RC12-2 high Which of the following describes:

(1) the event that has occurred?

(2) its effect, if any, on indicated Pressurizer (PZR) level?

A. (1) Leaking PZR Power Operated Relief Valve (2) No effect on PZR level B. (1) Leaking PZR Power Operated Relief Valve (2) PZR level indicates higher than actual C. (1) Leaking PZR Code Safety Relief Valve (2) No effect on PZR level D. (1) Leaking PZR Code Safety Relief Valve (2) PZR level indicates higher than actual Answer: D Explanation/Justification:

A. Incorrect - Safety valve leak indicated by computer alarm. See DB-OP-02513 R11 step 2.5.2. PORV leak would have T773 computer alarm.

See DB-OP-02513 R11 step 2.2.4. PZR level reads high due to reference leg heat up. Elevated CAC suction temperatures indicate reference leg heat up. See DB-OP-06003 R30 PZR Operating Procedure step 2.2.9. Plausible because of similarities in computer alarm nomenclature; PORV leak symptoms step 2.4.3 lists no change in PZR level. See DB-OP-02513 R11 step 2.4.3.

B. Incorrect - Safety valve leak indicated by computer alarm. See DB-OP-02513 R11 step 2.5.2. PORV leak would have T773 computer alarm.

See DB-OP-02513 R11 step 2.2.4. Plausible because of similarities in computer alarm nomenclature. Part 2 is correct.

C. Incorrect - PZR level reads high due to reference leg heat up. Elevated CAC suction temperatures indicate reference leg heat up. See DB-OP-06003 R30 PZR Operating Procedure step 2.2.9. Plausible because DB-OP-02513 R11 section 2.5 symptoms for leaking safety is silent on PZR level. Part 1 is correct.

D. Correct - Safety valve leak indicated by computer alarm. See DB-OP-02513 R11 step 2.5.2. PZR level reads high due to reference leg heat up.

See DB-OP-06003 R30 PZR Operating Procedure step 2.2.9.

Sys # System Category KA Statement 000008 Pressurizer AK2 Knowledge of the interrelations between the Pressurizer Sensors and detectors (PZR) Vapor Space Accident and the following:

Vapor Space Accident K/A# AK2.02 K/A Importance 2.7* Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02513 R11 step 2.5.2; DB-OP-06003 R30 step 2.2.9 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.7 / 45.7)

Objective: OPS-GOP-113-01K

Davis-Besse 1LOT15 NRC Written Exam r1

3. The plant was tripped from 100% power due to a Reactor Coolant System (RCS) leak.

The operators have performed all required actions prior to implementing the applicable Symptom Mitigation Section of the governing procedure.

The plant has been stabilized. Current conditions:

The RCS leak has been isolated.

RCS pressure is 1275 psig.

RCS Thot is 540 ºF.

Incore temperatures are 545 ºF.

Borated Water Storage Tank (BWST) level is 38 feet.

Which of the following Technical Specifications ACTIONS are required to be met for current conditions?

A. Restore RCS pressure to 2064.8 psig within 30 minutes per LCO 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits.

B. Restore RCS cooldown rate to within limits within 30 minutes per LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits.

C. Immediately initiate action to lower incore temperature to 535 ºF per LCO 3.4.6 RCS Loops - MODE 4.

D. Restore BWST level to > 38.6 feet within one hour per LCO 3.5.4 Borated Water Storage Tank.

Answer: D Explanation/Justification:

A. Incorrect - LCO 3.4.1 is not applicable. LCO 3.4.1 applies in MODE 1 only and the plant is in MODE 3 due to the reactor trip. Plausible because LCO 3.4.1 pressure limit is not met due to low RCS pressure from SBLOCA and Action would be correct.

B. Incorrect - LCO 3.4.3 is met. RCS P/T is within the limits of Figure 1 of the PTLR and RCS did not cool down 100 ºF. Plausible because LCO 3.4.3 is applicable at all times and required Action and Completion Time would be correct.

C. Incorrect - LCO 3.4.6 is not applicable. LCO 3.4.6 applies in MODE 4 and the plant is in MODE 3 due to the reactor trip. Plausible for determining that the required RCS loop is not in operation because all RCPs were stopped. Action would place plant in compliance with LCO 3.4.6 NOTE b and Completion Time is consistent with Condition A.

D. Correct - LCO 3.5.4 is applicable because the plant is in MODE 3 due to the reactor trip. BWST is inoperable due to low water volume, so Condition B applies which has one hour completion time. See also DB-OP-02003 R16 Window 3-1-C step 3.3 Sys # System Category KA Statement 000009 Small Break Generic Knowledge of less than or equal to one hour LOCA Technical Specification action statements for systems K/A# 2.2.39 K/A Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

LCO 3.5.4; DB-OP-02003 R16 Window 3-1-C Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 / 41.10 / 43.2

/ 45.13)

Objective: OPS-GOP-343-01K

Davis-Besse 1LOT15 NRC Written Exam r1

4. The plant is operating at 100% power.

A Design Basis Loss of Coolant Accident (DBLOCA) occurs.

Which of the following describes the bases for the Borated Water Storage Tank (BWST) level at which the operators transfer Low Pressure Injection (LPI) Suction to the Emergency Sump?

The specified BWST level for transfer of LPI suction to the Emergency Sump is designed to __(1)__ and

__(2)__.

A. (1) maximize Core cooling during the DBLOCA Injection Phase (2) minimize Containment pressure during the DBLOCA Injection Phase B. (1) maximize Core cooling during the DBLOCA Injection Phase (2) ensure sufficient LPI Pump NPSH prior to the completion of the transfer of LPI Suction to the Emergency Sump C. (1) ensure sufficient LPI Pump NPSH during the DBLOCA Recirculation Phase (2) ensure sufficient LPI Pump NPSH prior to the completion of the transfer of LPI Suction to the Emergency Sump D. (1) ensure sufficient LPI Pump NPSH during the DBLOCA Recirculation Phase (2) minimize Containment pressure during the DBLOCA Injection Phase Answer: C Explanation/Justification:

A. Incorrect - Plausible misconception because Injection Phase water from the BWST is colder than Recirculation Phase water from the Containment Sump. Both items occur with larger Injection Phase volumes, but are not the bases of the transfer setpoint.

B. Incorrect - Plausible because it contains one of the correct items.

C. Correct - See System Description for Decay Heat Removal System SD-042 R6 item 2.1.2.3 page 2-6 D. Incorrect - Plausible because it contains one of the correct items.

Sys # System Category KA Statement 000011 Large Break EK3 Knowledge of the reasons for the following responses as they Criteria for shifting to recirculation mode LOCA apply to the Large Break LOCA: Criteria for shifting to recirculation mode K/A# EK3.15 K/A Importance 4.3 Exam Level RO References provided to Candidate None Technical

References:

SD-042 R6 item 2.1.2.3 page 2-6 Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR 41.5 / 41.10 / 45.6 /

45.13)

Objective: OPS-GOP-309-04K

Davis-Besse 1LOT15 NRC Written Exam r1

5. The plant is operating at 70% power.

RCS Loop 1 flow is 74 mpph

RCS Loop 2 flow is 75 mpph RCP 2-2 trips (1) Which of the following is the signal the ICS will receive for Tave input?

(2) How will the trip of RCP 2-2 impact SG levels?

A. (1) Loop 2 Tave (2) SG 1 level will be higher than SG 2 Level B. (1) Loop 1 Tave (2) SG 1 level will be higher than SG 2 Level C. (1) Loop 1 Tave (2) SG 2 level will be higher than SG 1 Level D. (1) Loop 2 Tave (2) SG 2 level will be higher than SG 1 Level Answer: B Explanation/Justification:

A. Incorrect. Plausible since Loop 2 Tave is the normal controlling Tave Loop. Since a Loop 2 RCP trips, Loop 1 will have the highest flow FW flow and therefore SG level will be higher in SG 1 which is correct.

B. Correct. The Smart Analog Selector Switch (SASS) for Tave automatically selects the Loop with the Highest RCS Flow when a RCP is stopped.

Since a Loop 2 RCS trips, Loop 1 will have the highest flow and Loop 1 Tave will be selected. ICS will ratio FW flow to the Steam Generators based on RCS flow or about 2.5 to 1 with the 2 RCP loop SG receiving the higher Feedwater Flow and will operate at a higher Steam Generator Level.

C. Incorrect. Plausible Since a Loop 2 RCP trips, Loop 1 will have the highest flow FW flow and therefore SG level will be higher in SG 1.

D. Incorrect. Plausible since Loop 2 Tave is the normal controlling Tave Loop.

Sys # System Category KA Statement 000015/0 Reactor AK1 Knowledge of the operational implications of the following Consequences of an RCPS failure 00017 Coolant concepts as they apply to Reactor Coolant Pump Malfunctions (Loss Pump (RCP) of RC Flow):

Malfunctions K/A# AK1.02 K/A Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02515 R12 RCP and Motor Abnormal Operation Attachment 1 RCP Shutdown Question Source: Bank - #172683 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.8 / 41.10 / 45.3)

Objective: OPS-GOP-115 05K

Davis-Besse 1LOT15 NRC Written Exam r1

6. The plant is operating at 100% power.

Component Cooling Water (CCW) Pump 1 is operating.

A CCW System leak occurs.

CCW Surge Tank Level side 1 lowers to 33 inches and stabilizes

CCW Surge Tank Level side 2 lowers to 30 inches and continues lowering slowly NO operator actions have been taken.

Which of the following is correct regarding the status of CCW loads?

A. Neither MU Pump has CCW cooling because CC1460 CCW to MU Pump Coolers has closed.

B. The Pressurizer Quench Tank Cooler has no CCW cooling because CC1411B CCW to Containment has closed.

C. The Control Rod Drive Mechanisms have no CCW cooling because CC5097 Non-Essential CCW Containment Building Return Line 1 Isolation has closed.

D. Reactor Coolant Pump 1-1 Seal Cooler has no CCW cooling because CC1495 CCW to Aux Building Non-Essential Header has closed.

Answer: C Explanation/Justification:

A. Incorrect - flow through MU pump #1 oil coolers still exists via the #1 Essential Header supply and operating CCW Pump 1. See OS-0021 sheet 2 R31, F-27. Plausible because CC1460 is closed at 35 inches to isolate the non-essential supply to both MU Pumps. See OS-0021 sheet 1 R37, CL-3.

B. Incorrect - Quench Tank Cooler is isolated by CC1495. See OS-0021 sheet 2 R31, C-19 and OS-0021 sheet 1 R37, B-10. Plausible for misconception that the Quench Tank Cooler is in Containment with the Quench Tank instead of outside Containment.

C. Correct - See OS-0021 sheet 2 R31, C-28 and OS-0021 sheet 1 R37 J-11, CL-7.

D. Incorrect - RCP 1-1 seal cooler is isolated by CC1411B. See OS-0001B sheet 1 R26, J-3 and OS-0021 sheet 2 R30, D-29. Plausible because RCP Seal Return Coolers are supplied by CC1495. See OS-0021 sheet 2 R30, D-20.

Sys # System Category KA Statement 000026 Loss of AA1 Ability to operate and / or monitor the following as they apply to Loads on the CCWS in the control room Component the Loss of Component Cooling Water:

Cooling Water (CCW)

K/A# AA1.02 K/A Importance 3.2 Exam Level RO References provided to Candidate None Technical

References:

OS-0021 sheet 2 R31, C-28; OS-0021 sheet 1 R37 J-11, CL-7.

Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.7 / 45.5 / 45.6)

Objective: OPS-GOP-123-04K

Davis-Besse 1LOT15 NRC Written Exam r1

7. The plant is operating at 100% power.

All attempts to trip the Reactor from the Control Room have failed.

An operator is dispatched to the Low Voltage Switchgear Rooms to open Reactor Trip Breakers.

Which pair of Reactor Trip Breakers located in the Low Voltage Switchgear Rooms will trip the Reactor when opened?

A. A and B B. C and D C. A and C D. B and D Answer: A Explanation/Justification:

A. Correct - DB-OP-02000 directs opening of Trip Breakers A, B, and C in the Low Voltage Switchgear Rooms. See DB-OP-02000 R27 step 3.3 RNO. A and B open will cause a reactor trip. See DB-OP-06402 R25 CRD Operating Procedure Attachment 4 CRD System Power Diagram (page 162)

B. Incorrect - Trip Breaker D is in the CRD Cabinet Room. Plausible because C and D open would trip the reactor.

C. Incorrect - CRDMs still energized via B and D. Plausible because both breakers are in the Low Voltage Switchgear Rooms.

D. Incorrect - CRDMs still energized via A and C. Plausible for faulty recall of CRD power supply diagram.

Sys # System Category KA Statement 000029 Anticipated EK2 Knowledge of the interrelations between the following and an Breakers, relays, and disconnects Transient ATWS:

Without Scram (ATWS)

K/A# EK2.06 K/A Importance 2.9* Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02000 R27 step 3.3; DB-OP-06402 R25 Attachment 4 (page 162)

Question Source: Bank - 165796 Modified Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR 41.7 / 45.7)

Objective: OPS-GOP-302-05K

Davis-Besse 1LOT15 NRC Written Exam r1

8. The plant WAS operating at 100% power.

A Steam Generator Tube Rupture (SGTR) occurred on Steam Generator (SG) 2.

All operator actions were completed in accordance with the governing procedures.

The Reactor Coolant System (RCS) is now at 500 °F and 1000 psig with SG 2 isolated.

All four Reactor Coolant Pumps (RCPs) are running.

The operators initiate an RCS Cooldown in accordance with the governing procedures.

Which of the following describes the criteria for securing an RCP during an SGTR?

The first RCP stopped during the RCS Cooldown should be in _____.

A. RCS Loop 1 to maximize cooling of SG 2 from reverse heat transfer B. RCS Loop 1 to maximize RCS pressure control from Pressurizer Spray C. RCS Loop 2 to minimize heat addition to the RCS from RCP operation D. RCS Loop 2 to minimize contamination of SG 2 from flow through ruptured tube Answer: B Explanation/Justification:

A. Incorrect -Plausible because stopping a Loop 1 RCP puts 70% of total RCS flow through Loop 2, which would tend to raise reverse heat transfer.

Additional cooling of an isolated SG is desirable to raise maximum RCS Cooldown with an isolated SG. See DB-OP-06903 R47 Plant Cooldown section 5.0 Cooldown with one SG step 5.4. Reverse heat transfer is described in DB-OP-02000 NOTE 8.44 (page 120)

B. Correct - All RCPs left running per DB-OP-02000 Section 8. At step 8.52, exit to DB-OP-06903 Plant Cooldown with REFER TO DB-OP-02531 SGTL. DB-OP-02531 step 4.13 for RCS Cooldown is REFER TO DB-OP-02543 Rapid Cooldown. DB-OP-02543 step 4.17 directs stop of Loop 1 RCP to maximize PZR spray capability.

C. Incorrect - Loop 1 preferred for first RCP stopped. Plausible misconception because stopping RCPs will lower RCS heat input. RCPs are stopped to minimize RCS heat input during Lack of Heat Transfer event - see Bases and Deviation document for DB-OP-02000 R20 Step 6.3 D. Incorrect - Loop 1 preferred for first RCP stopped. Plausible for misconception that pressure control is not the highest priority.

Sys # System Category KA Statement 000038 Steam EK3 Knowledge of the reasons for the following responses as they Criteria for securing RCP Generator apply to the SGTR:

Tube Rupture (SGTR)

K/A# EK3.08 K/A Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02000 R27 step 8.52; DB-OP-02531 R20 SGTL step 4.13; DB-OP-02543 R9 step 4.17 Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR 41.5 / 41.10 / 45.6 /

45.13)

Objective:

Davis-Besse 1LOT15 NRC Written Exam r1

9. The plant is operating at 100% power.

Auxiliary Feed Water (AFW) Pump 1 is out of service.

Both Main Feed Water Pumps trip.

AFW Pump 2 trips when it starts.

Both Steam Generators (SGs) are at 1000 psig and 24 inches Startup Level When the RO enables the Motor Driven Feedwater Pump (MDFP) Discharge Valves, the:

The CONTROL VALVE OFF light for FW 6459 SG 1 level control valve goes OFF

The CONTROL VALVE OFF light for FW 6460 SG 2 level control valve stays LIT The RO verifies LIC 6459 and LIC 6460 are set to minimum output.

Which of the following describes the sequence for restoring level in both SGs that establishes feedwater flow THE FASTEST without running out the MDFP?

(1) Start the MDFP (2) Raise SG 1 level to 49 inches at full flow using LIC 6459 as required (3) Raise SG 2 level to 49 inches at full flow using LIC 6460 as required (4) Direct local operator to throttle closed FW6398 MDFP TO AUXILIARY FEED LINE 1 ISOLATION (5) Direct local operator to throttle closed FW6397 MDFP TO AUXILIARY FEED LINE 2 ISOLATION A. 1, 2, 4, 3 B. 1, 3, 5, 2 C. 4, 1, 2, 3 D. 5, 1, 3, 2 Answer: B Explanation/Justification:

A. Incorrect - This method results in runout of the MDFP from full flow through both feed lines. FW6460 is failed open and not yet isolated in this sequence. Plausible for candidate reverse interpretation of the Control Valve Off lights.

B. Correct - FW6460 is failed open (see DB-OP-06225 R21 step 5.1.4 and OS-0012A sheet 1 R26, B-16) and FW6459 is closed. When the MDFP is started, MDFP flow will be limited to 800 gpm by the Cavitating Venturi. See OS-0017A sheet 1 R31, G-10. When proper level is reached, FW6460 is isolated by local operator closing FW6397. See OS-0012A sheet 1 R26, B-17. Full flow can then be established to SG 1 using LIC 6459 without running out the MDFP. See DB-OP-06225 R21 CAUTION 5.1.10. Faster to start MDFP first, then throttle closed manual valve.

C. Incorrect - This method throttles the wrong manual valve. Plausible for candidate reverse interpretation of the Control Valve Off lights and misconception that MDFP connects to AFW lines downstream of Cavitating Venturis.

D. Incorrect - Faster to start MDFP first, then throttle closed manual valve. Plausible for misconception that MDFP connects to AFW lines downstream of Cavitating Venturis.

Sys # System Category KA Statement 000054 Loss of Main AA1 Ability to operate and / or monitor the following as they apply to AFW controls, including the use of alternate AFW Feedwater the Loss of Main Feedwater (MFW): sources (MFW)

K/A# AA1.01 K/A Importance 4.5 Exam Level RO References provided to Candidate None Technical

References:

OS-0012A sheet 1 R26; OS-0017A sheet 1 R31; DB-OP-06225 R21 step 5.1.4 and CAUTION 5.1.10 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.7 / 45.5 / 45.6)

Objective: OPS-GOP-303-05K

Davis-Besse 1LOT15 NRC Written Exam r1

10. The Plant has experienced a complete loss of AC Power.

Performance of DB-OP-02521 Loss of AC Bus Power Sources is in progress.

At 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months following the beginning of the event AC power is still lost.

Battery 1P is in service supplying Panel D1P only.

Batteries 1N, 2P and 2N are isolated from all loads.

How long will it be before DC power is no longer available?

A. less than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months B. approximately 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C. approximately 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months D. greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Answer: D Explanation/Justification: DC Bus Load shedding is performed to reduce Discharge Rate and therefore extend battery life.

A. Incorrect - Plausible because the batteries are designated as having a 1500 amp-hour rating based on an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months discharge rate.

B. Incorrect - Plausible if it is assumed there are 250V loads required to remain energized following load shedding C. Incorrect - Plausible since one battery (1P) will remain in service and 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months is a multiple of 8.hours D. Correct - Approximately 39 hours4.513889e-4 days 0.0108 hours 6.448413e-5 weeks 1.48395e-5 months for D1P followed by D2P. See DB-OP-02521 R23 Attachment 17 (page 129) last paragraph.

Sys # System Category KA Statement 000055 Loss of EK1 Knowledge of the operational implications of the following Effect of battery discharge rates on capacity Offsite and concepts as they apply to the Station Blackout :

Onsite Power (Station Blackout)

K/A# EK1.01 K/A Importance 3.3 Exam Level RO References provided to Candidate Technical

References:

DB-OP-02521 R23 Attachment 17 page 129 last paragraph Question Source: Bank - DB 2013 NRC Exam #48 Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR 41.8 / 41.10 / 45.3)

Objective: OPS-SYS-121-11K

Davis-Besse 1LOT15 NRC Written Exam r1

11. The plant has been operating at 100% power for one year.

A Loss of Offsite Power occurs.

NO operator actions are taken.

Reactor Coolant System (RCS) temperatures stabilize post-trip.

Which of the following describes the values of RCS Thot and Tcold?

(1) RCS Thot will be _____.

(2) RCS Tcold will be _____.

A. (1) 596 to 600 °F (2) 550 to 554 °F B. (1) 596 to 600 °F (2) 530 to 534 °F C. (1) 550 to 554 °F (2) 550 to 554 °F D. (1) 550 to 554 °F (2) 530 to 534 °F Answer: A Explanation/Justification:

A. Correct - LOP causes loss of RCPs and SFRCS Isolation Trip on reverse FW dP due to loss of MFW pumps on loss of AC oil pumps. Lowest SG safety valve lift pressure is 1050 psig (1065 psia). Tsat SG 552.3 °F. Full power core T 46 °F, so Thot about 598 °F B. Incorrect - T > 50 °F. See DB-OP-06903 R47 Plant Cooldown Section 6.0 Cooldown on Natural Circulation step 6.3. Part 1 is correct. 530 to 534 °F Tcold plausible because this is normal MODE 3 Tave for a reactor startup.

C. Incorrect - LOP causes loss of RCPs so RCS T should be near the full power value of 46 °F, not zero. Part 2 is correct. Plausible because post-trip forced flow T approaches zero.

D. Incorrect - Plausible T for normal MODE 3 Tave for a reactor startup following sustained operation at lower power (40-45%).

Sys # System Category KA Statement 000056 Loss of AA2 Ability to determine and interpret the following as they apply to RCS hot-leg and cold-leg temperatures Offsite Power the Loss of Offsite Power:

K/A# AA2.57 K/A Importance 3.9 Exam Level RO References provided to Candidate Steam Tables Technical

References:

TS Table 3.7.1-1; DB-OP-06903 R47 Plant Cooldown Section 6.0 Cooldown on Natural Circulation step 6.3 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective: OPS-GOP-303-04K

Davis-Besse 1LOT15 NRC Written Exam r1

12. The plant experienced a loss of 120V AC Essential Panel Y3.

The problem with Y3 has been corrected.

Y3 has been re-energized from Transformer XY3.

Y3 will be transferred from Transformer XY3 to Inverter YV3 as part of the recovery process.

Which of the following describes the correct sequence of steps to swap Y3 from XY3 to YV3?

A. 1. Depress the ALTERNATE SOURCE TO LOAD pushbutton on YV3.

2. Place Inverter YV3 MANUAL BYPASS SWITCH in the ALTERNATE position

3. Check YV3 ALTERNATE SOURCE SUPPLYING LOAD RED light ON.

B. 1. Depress the ALTERNATE SOURCE TO LOAD pushbutton on YV3.

2. Place the MANUAL BYPASS SWITCH on YV3 in the NORMAL position.

3. Depress the INVERTER TO LOAD pushbutton on YV3.

C. 1. Place the MANUAL BYPASS SWITCH on YV3 in the NORMAL position.

2. Depress the INVERTER TO LOAD pushbutton on YV3.

3. Depress the ALTERNATE SOURCE TO LOAD pushbutton on YV3.

D. 1. Place the MANUAL BYPASS SWITCH on YV3 in the NORMAL position.

2. Depress the ALTERNATE SOURCE TO LOAD pushbutton on YV3.

3. Check YV3 INVERTER SUPPLYING LOAD yellow light ON.

Answer: B Explanation/Justification:

A. Incorrect - this is sequence for transfer from YV3 to XY3. Plausible for inversion of normal and alternate sources.

B. Correct - see DB-OP-06319 R29 Instrument AC System Procedure section 3.44 (page 87)

C. Incorrect - plausible candidate inversion of switch functions.

D. Incorrect - plausible because it is 2 of the 3 required actions in proper order.

Sys # System Category KA Statement 000057 Loss of Vital AA1 Ability to operate and / or monitor the following as they apply to Manual inverter swapping AC Electrical the Loss of Vital AC Instrument Bus:

Instrument Bus K/A# AA1.01 K/A Importance 3.7* Exam Level RO References provided to Candidate None Technical

References:

DB-OP-06319 R29 Instrument AC System Procedure section 3.44 (page 87)

Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR 41.7 / 45.5 / 45.6)

Objective: OPS-SYS-408-05K

Davis-Besse 1LOT15 NRC Written Exam r1

13. The plant is operating at 100% power.

Charger DBC2P is aligned to Battery 2P

Charger DBC2N is aligned to Battery 2N The following conditions are observed:

Annunciator 1-6-G DC BUS 2 TRBL alarms

CHARGER DBC2N indicator II 6284 reads zero amps

BATTERY 2N indicator II 6290 reads 100 amps DISCHARGE Which of the following will eventually occur if NO operator actions are taken?

A. Power Operated Relief Valve (PORV) RC2A wont open if required B. Battery Charger DBC2PN automatically charges Battery 2N C. Reactor Protection System Channel 3 de-energizes D. Main Feed Pump 1 Emergency Bearing Oil Pump wont start if required Answer: A Explanation/Justification:

A. Correct - With no operator action, battery 2N will continue to discharge and voltage will continue to lower on 125V DC Panel D2N until the RC2A solenoid coils will no longer function. RC2A is a D2N load. See DB-OP-02540 R08 Loss of D2N and DBN Attachment 1 (page 13)

B. Incorrect - Swing charger must be manually aligned. Plausible because this is a procedure-driven manual action. See DB-OP-02001 R30 Window 1-6-G step 3.7.3 C. Incorrect - Rectifier YRF4 will continue to supply 120V AC panel Y4 via Inverter YV4. See UFSAR R30 8.3.2.1.5 (page 8.3-46). Plausible because Y4 would be supplied from battery 2N during a concurrent loss of AC input. RPS Channel 3 supplied from Y4.

D. Incorrect - MFP 1 EBOP is DC MCC 1 load. Plausible for loss of either DBC1P or DBC1N. See OS-0060 sheet 1 R29 Sys # System Category KA Statement 000058 Loss of DC AK1 Knowledge of the operational implications of the following Battery charger equipment and instrumentation Power concepts as they apply to Loss of DC Power:

K/A# AK1.01 K/A Importance 2.8 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02540 R8 Attachment 1 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.8 / 41.10 / 45.3)

Objective: OPS-GOP-137-03K

Davis-Besse 1LOT15 NRC Written Exam r1

14. The plant is operating at 100% power.

Component Cooling Water (CCW) Pump 1 is operating.

The Control Room receives a report of a Service Water (SW) rupture on SW Pump 1 strainer.

Bus C1 locks out concurrent with the rupture report.

Which of the following describes:

(1) the steps to allow restoration of normal SW operating parameters in the affected loop?

(2) the status of SW Loop 1 when flow is restored?

(1) Close SW Pump 1 discharge valve SW20 and align the Backup SW Pump to SW Loop 1 in _____

speed.

(2) SW Loop 1 is _____.

A. (1) FAST (2) OPERABLE B. (1) FAST (2) INOPERABLE C. (1) SLOW (2) OPERABLE D. (1) SLOW (2) INOPERABLE Answer: B Explanation/Justification:

A. Incorrect - BUSW Pump use makes the SW loop INOPERABLE per DB-OP-06261 R63 SW System Operating Procedure NOTE 4.1.3 (page 86).

Part 1 is correct. Plausible because BUSW Pump in FAST speed restores normal SW Loop operating parameters.

B. Correct - Closing SW20 isolates the leak. Aligning the BUSW Pump to Loop 1 in FAST speed bypasses SW Pump 1 & 3 which cannot be used due to the C1 lockout. See OS-0020 sheet 1 R95 and DB-OP-02511 R16 Loss of SW Pumps/Systems Attachment 5. BUSW Pump use makes the SW loop INOPERABLE per DB-OP-06261 R63 SW System Operating Procedure NOTE 4.1.3 (page 86).

C. Incorrect - SLOW speed does not restore normal flow. FAST speed is required per DB-OP-02511 R16 Loss of SW Pumps/Systems Attachment 5 step 10. BUSW Pump use makes the SW loop INOPERABLE per DB-OP-06261 R63 SW System Operating Procedure NOTE 4.1.3 (page 86).

Plausible because SLOW is the normal speed for BUSW Pump operation as the Dilution Pump and flow is restored in SW Loop 1.

D. Incorrect - BU SW Pump in FAST speed per Attachment 5 step 10. Part 2 is correct. Plausible because SLOW is the normal speed for BUSW Pump operation as the Dilution Pump and flow is restored in SW Loop 1.

Sys # System Category KA Statement 000062 Loss of AA2 Ability to determine and interpret the following as they apply to The valve lineups necessary to restart the SWS Nuclear the Loss of Nuclear Service Water: while bypassing the portion of the system causing Service the abnormal condition Water K/A# AA2.03 K/A Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

OS-0020 sheet 1 R95; DB-OP-02511 R16 step 4.1.7 and Attachment 5 step 10; DB-OP-06261, Note 4.1.3 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective: OPS-GOP-111-02K

Davis-Besse 1LOT15 NRC Written Exam r1

15. The plant is operating at 100% power.

PI810 INSTRUMENT AIR HEADER PRESS lowers to 50 psig and stabilizes.

The operators perform the required abnormal procedure actions then implement DB-OP-02000 RPS, SFAS, SFRCS Trip or SG Tube Rupture.

During the performance of Attachment 1 Primary Inventory Control Actions the following indications are noted:

Letdown Flow FI MU7 45 gpm

RCP 1-1 Seal Injection Flow FI MU30C 15 gpm

RCP 1-2 Seal Injection Flow FI MU30D zero gpm

RCP 2-1 Seal Injection Flow FI MU30A 15 gpm

RCP 2-2 Seal Injection Flow FI MU30B 15 gpm Which of the following additional failures is consistent with these indications?

A. RCP tripped during the transient B. Essential DC Distribution Panel D2P de-energized C. Inadvertent SFAS Level 3 actuation of Channel 2 only D. Seal Injection Isolation valve Air Volume Tank leak Answer: D Explanation/Justification:

A. Incorrect - Plausible because seal injection flow lowers when an RCP is stopped, but only to around 3 gpm. See Makeup & Purification System Description SD-048 R04 page 2-9 B. Incorrect - D2P loss would also close MU66A and MU3, resulting in zero flow on FI MU7 and FI MU30A See DB-OP-06405 R13 SFAS Procedure step 3.2.4 (page 11) and Attachment 4 Page 4 of 4 (page 79)

C. Incorrect -SFAS Level 3 on Actuation Channel 2 would also have closed MU66A, resulting in zero gpm on FI MU30A, too. See DB-OP-02000 Table 4 (page 427)

D. Correct - Question is written for Air Volume Tanks which serve the same purpose as backup Nitrogen supply at D-B. The Air Volume Tanks are supplied from the Instrument Air System via check valves which prevent back-leakage from the tanks to the depressurized Instrument Air supply.

MU3 letdown isolation and MU66 valves are equipped with Air Volume Tank to maintain them open. Air Volume Tank leak for MU66D would result in its closure and zero flow on FI MU30D. See DB-OP-02528 R22 Instrument Air System Malfunctions Attachment 18 Failure Position of Pneumatic Valves (page 110), Makeup & Purification System Description SD-048 R04 2.4.8 (pages 2-24 and 2-25) and OS-0002 sheet 2 Sys # System Category KA Statement 000065 Loss of AA2 Ability to determine and interpret the following as they apply to Whether backup nitrogen supply is controlling Instrument the Loss of Instrument Air: valve position Air K/A# AA2.07 K/A Importance 2.8* Exam Level RO References provided to Candidate None Technical

References:

OS-0002 sheet 2 R21; DB-OP-02528 R22 Attachment 18 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective: OPS-GOP-128-08K

Davis-Besse 1LOT15 NRC Written Exam r1

16. The plant is operating at 25% power.

An event occurs.

The following conditions are noted:

8-5-A SWYD ACB 34560 TRIP alarm

8-5-B SWYD ACB 34561 TRIP alarm

14-6-D ICS IN TRACK alarm

Turbine Bypass Valves and Atmospheric Vent Valves are open

JI 6003 MEGAWATTS indicates 43 MWe Which of the following describes the response of the following controls during the event?

(1) Main Generator Voltage Regulator (2) Main Turbine DEHC Load Control A. (1) remains in AUTO (2) transfers to MANUAL B. (1) transfers to MANUAL (2) remains in AUTO C. (1) remains in AUTO (2) remains in AUTO D. (1) transfers to MANUAL (2) transfers to MANUAL Answer: A Explanation/Justification:

A. Correct - Load rejection has occurred. See DB-OP-02520 R 7 Load Rejection 2.1 Symptoms. Part 1 - Automatic Voltage Regulator trips to manual on loss of potential transformer signals or Generator Field Breaker trip. See DB-OP-02016 R25 Window 16-4-B. Field breaker stays closed on Load Rejection because the main generator transformer lockout relays dont actuate.

B. Incorrect - Part 1 incorrect. Part 2 incorrect. Part 1 plausible for generator trip. Part 2 plausible because Power Load Unbalance circuit actuates to place turbine in MANUAL and this signal does not interface with the ICS transfer to MANUAL logic. See M-00175 R4 Logic String 1.

C. Incorrect - Part 2 is incorrect. Part 1 correct. Plausible because Power Load Unbalance circuit actuates to place turbine in MANUAL and this signal does not interface with the ICS transfer to MANUAL logic.

D. Incorrect - Part 1 incorrect. Part 2 correct. Plausible for generator trip.

Sys # System Category KA Statement 000077 Generator AK2 Knowledge of the interrelations between Generator Voltage and Turbine / generator control Voltage and Electric Grid Disturbances and the following:

Electric Grid Disturbances K/A# AK2.07 K/A Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02520 R 7 Load Rejection 2.1 Symptoms; DB-OP-02016 R25 Window 16-4-B Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)

Objective: OPS-GOP-120-01K

Davis-Besse 1LOT15 NRC Written Exam r1

17. The plant has experienced a Loss of ALL Feedwater.

The following conditions exist:

Incore temperatures 620 ºF stable

Reactor Coolant System (RCS) pressure 1770 psig stable

Steam Generator (SG) levels 10 inches stable

SG pressures 700 psig slowly lowering The local operator reports Auxiliary Feedwater Pump Turbine (AFPT) 1 Trip Throttle Valve (TTV) is reset. The operator is standing by to open ICS38C AFPT1 TTV to restore feedwater.

Which of the following describes establishing Auxiliary Feed Water (AFW) flow to SG 1?

As ICS38C is opened, the desired initial flow to SG 1 on AUX FW FLOW FI6426 is __(1)__.

An indication of SG heat transfer being established is RCS pressure __(2)__ and SG 1 PRESS PI SP12B __(3)__.

A. (1) 100 gpm (2) lowering (3) rising B. (1) 100 gpm (2) stable (3) lowering C. (1) full flow (2) lowering (3) rising D. (1) full flow (2) stable (3) lowering Answer: C Explanation/Justification:

A. Incorrect - No AFW flow limit for dry SG during Lack of Heat Transfer (LOHT). See DB-OP-02000 R27 Attachment 5 Section B NOTE 4 (page 285). Plausible because items 2 & 3 are correct (see Correct Answer explanation), item 1 is the correct flow limit if not in LOHT because all RCPs would be stopped for lack of adequate subcooling margin.

B. Incorrect - No AFW flow limit for dry SG during Lack of Heat Transfer (LOHT). See DB-OP-02000 R27 Attachment 5 Section B NOTE 4 (page 285). Plausible because item 1 is the correct flow limit if not in LOHT, items 2 & 3 plausible because they are the initial response to the initiation of AFW to SG 1 before heat transfer is established. See Bases and Deviation Document for DB-OP-02000 R20 step 6.11 (page 94).

C. Correct - Specific Rule 4 requires full continuous AFW flow until SG reaches setpoint. AFW flow is limited to about 800 gpm by the Cavitating Venturi. See DB-OP-02000 R27 Attachment 5 Section B step 5 (page 285) and Specific Rule 4.3.1 (page 246). RCS is saturated, so RCS pressure will lower as voids condense due to primary to secondary heat transfer. SG pressure will rise. See Areva Technical Document 74-1152414-10 Part II Section 3.3 Indication of Primary to Secondary Coupling page Vol.3, II.B-10 D. Incorrect - items 2 & 3 are the initial response to the initiation of AFW to SG 1 before heat transfer is established. See Bases and Deviation Document for DB-OP-02000 R20 step 6.11 (page 94). Plausible for initial response and item 1 being correct

Davis-Besse 1LOT15 NRC Written Exam r1 Sys # System Category KA Statement BW/E04 Inadequate Generic Ability to interpret control room indications to Heat Transfer verify the status and operation of a system, and

- Loss Of understand how operator actions and directives Secondary affect plant and system conditions Heat Sink K/A# 2.2.44 K/A Importance 4.2 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02000 R27 Attachment 5 Section B step 5 (page 285) and Specific Rule 4.3.1 (page 246);

Areva Technical Document 74-1152414-10 Part II Section 3.3 page Vol.3, II.B-10 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.12)

Objective: OPS-GOP-305-02K

Davis-Besse 1LOT15 NRC Written Exam r1

18. The plant is experiencing an unisolable steam leak in Containment.

Which of the following describes an action required of the Reactor Operator and the reason for the action?

A. Open the Atmospheric Vent Valve on the affected Steam Generator to blow it down to atmosphere to ensure compliance with TNC 8.7.1 Steam Generator Pressure/Temperature Limitation.

B. Open the Atmospheric Vent Valve on the affected Steam Generator to blow it down to atmosphere to ensure compliance with LCO 3.6.1 Containment.

C. After blowing down the affected Steam Generator, close its Atmospheric Vent Valve to ensure compliance with LCO 3.6.1 Containment.

D. After blowing down the affected Steam Generator, close its Atmospheric Vent Valve to ensure compliance with TNC 8.7.1 Steam Generator Pressure/Temperature Limitation.

Answer: C Explanation/Justification:

A. Incorrect - Plausible because AVV is opened for steam leak in Containment and opening AVV would reduce SG pressure if TNC 8.7.1 was applicable. .

B. Incorrect - Plausible because AVV is opened for steam leak in Containment and opening #2 AVV limits containment pressure rise.

C. Correct - 2 AVV must be closed following SG blowdown to isolate direct path from containment atmosphere through steam rupture to outside atmosphere via AVV. See DBOPBASES R20 step 7.26 (page 138).

D. Incorrect - Plausible because closing AVV after SG blowdown is correct action.

Sys # System Category KA Statement BW/E05 Steam Line EK3 Knowledge of the reasons for the following responses as they RO or SRO function within the control room team Rupture - apply to the (Excessive Heat Transfer): as appropriate to the assigned position, in such a Excessive way that procedures are adhered to and the Heat Transfer limitations in the Facilities license and amendments are not violated.

K/A# EK3.4 K/A Importance 3.8 Exam Level RO References provided to Candidate None Technical

References:

Bases and Deviation Document for DB-OP-02000 R20 Step 7.26 page 138 Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41.5 / 41.10, 45.6, 45.13)

Objective: OPS-GOP-306-03K

Davis-Besse 1LOT15 NRC Written Exam r1

19. The plant is operating at 100% power.

A power reduction to 70% is started.

The Integrated Control System (ICS) is in Full Automatic

The Unit Load Demand (ULD) is in Automatic with its Rate of Change set to 0.5%/min.

At 85% power, the following conditions are noted:

Annunciator 5-2-E CRD ASYMMETRIC ROD alarms

ASYMMETRY FAULT light on the Rod Control Panel is lit.

Control Rod 7-1 indicates 7% higher than Group 7 average Which of the following identifies:

(1) the power level limit (power < value) for the initial attempt to realign Control Rod 7-1?

(2) the control station to use for the power change to the power level limit?

A. (1) 60%

(2) ULD B. (1) 42%

(2) ULD C. (1) 60%

(2) Rod Control Panel in MANUAL D. (1) 42%

(2) Rod Control Panel in MANUAL Answer: A Explanation/Justification:

A. Correct - event is misaligned rod. See DB-OP-02516 R14 CRD Malfunctions step 2.2.1; 60% correct per step 4.2.2. ULD is preferred station per DB-OP-02504 R20 Rapid Shutdown step 4.1.

B. Incorrect - 42% is 3-RCP limit for recovering a misaligned rod. Part 2 is correct. Plausible for misapplication of power limit.

C. Incorrect - ULD is preferred station per DB-OP-02504 R20 Rapid Shutdown step 4.1. Part 2 is correct. Plausible because control rod will be recovered with the Rod Control Panel in MANUAL.

D. Incorrect - 60% is 4-RCP limit for recovering a misaligned rod. ULD is the preferred station for the power change per DB-OP-02504 R20 Rapid Shutdown step 4.1. Plausible for misapplication of power limit and because control rod will be recovered with the Rod Control Panel in MANUAL.

Sys # System Category KA Statement 000005 Inoperable/St AA1 Ability to operate and / or monitor the following as they apply to Reactor and turbine power uck Control the Inoperable / Stuck Control Rod:

Rod K/A# AA1.04 K/A Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02516 R14 CRD Malfunctions steps 2.2.1 and 4.2.2; DB-OP-02504 R20 Rapid Shutdown step 4.1.

Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.7 / 45.5 / 45.6)

Objective: OPS-GOP-116-03K

Davis-Besse 1LOT15 NRC Written Exam r1

20. The plant is operating at 100% power.

Assuming NO operator actions have been taken, which of the following describes the plant response to a leak on the reference leg of the selected Pressurizer Level Transmitter?

Makeup Tank level __(1)__.

Pressurizer Heaters __(2)__.

A. (1) lowers (2) de-energize B. (1) lowers (2) remain energized C. (1) rises (2) de-energize D. (1) rises (2) remain energized Answer: D Explanation/Justification:

A. Incorrect - MUT rises and PZR heaters stay energized. Plausible because this is plant response to variable leg leak (level input failing low). See DB-OP-02513 R11 Pressurizer System Abnormal Operation steps 2.6.4 and 2.6.5.

B. Incorrect - response describes letdown leak. Plausible because item 2 is correct. See DB-OP-02522 R13 Small RCS Leaks Attachment 13 Background Information Letdown System Leaks (page 50).

C. Incorrect -response describes significant RCS leak after automatic transfer of MU Pump suction to the BWST at 17 inch MU Tank level. See OS-0002 sheet 2 R21 DUN 13-0024-001-001 CL-8. Plausible for misconception of significant potential RCS mass loss from reference leg leak. See DB-OP-02522 R13 Small RCS Leaks Attachment 12 Align MU Pump Recirc to the BWST (page 47).

D. Correct - PZR level indication uses a wet reference leg dP transmitter - see RCS System Description SD-039A R06 section 2.5.1.10 (page 2-55)

Reference leg leak causes level input to indicate higher than actual level. High level causes PZR Level Control Valve MU32 to throttle closed to lower MU flow. Lower MU flow with constant letdown flow causes MU Tank level to rise. See DB-OP-02513 R11 Pressurizer System Abnormal Operation step 2.6.3. PZR heaters are affected by low level, not high level, so they remain energized. DB-OP-02513 R11 step 2.6.5 Sys # System Category KA Statement 000028 Pressurizer AK1 Knowledge of the operational implications of the following PZR reference leak abnormalities (PZR) Level concepts as they apply to Pressurizer Level Control Malfunctions:

Control Malfunction K/A# AK1.01 K/A Importance 2.8* Exam Level RO References provided to Candidate None Technical

References:

SD-039A R06 section 2.5.1.10 (page 2-55); DB-OP-02513 R11 steps 2.6.3 and 2.6.5.

Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.8 / 41.10 / 45.3)

Objective: OPS-GOP-113-05K

Davis-Besse 1LOT15 NRC Written Exam r1

21. The plant is in MODE 3 with Tave at 532 °F.

Control Rod Safety Group 1 is fully withdrawn.

Which of the following will cause a loss of Source Range Nuclear Instrument (NI) 1?

A. 120V AC Distribution Panel Y4 breaker Y408 RPS Channel 4 in OFF B. 120V AC Distribution Panel Y3 breaker Y308 RPS Channel 3 in OFF C. Reactor Protection System Channel 2 SYSTEM AC POWER breaker in OFF D. Reactor Protection System Channel 1 SYSTEM AC POWER breaker in OFF Answer: C Explanation/Justification:

A. Incorrect - Y408 powers RPS Channel 4 which powers NI-3. See DB-OP-06403 R20 RPS and NI Operating Procedure steps 3.1.4 and 3.1.17.

Plausible for incorrect concept of RPS Channel number to NI.

B. Incorrect - Y308 powers RPS Channel 3 which powers NI-4. See DB-OP-06403 R20 RPS and NI Operating Procedure steps 3.1.4 and 3.1.18.

Plausible for incorrect concept of RPS Channel number to NI.

C. Correct - SYSTEM AC POWER breaker open de-energizes RPS cabinet 2 which de-energizes NI-1. See DB-OP-06403 R20 RPS and NI Operating Procedure steps 3.1.6 and 3.1.17.

D. Incorrect - RPS Channel 1 powers NI-2. See DB-OP-06403 R20 RPS and NI Operating Procedure step 3.1.17. Plausible for misconception that RPS Channel number equals NI Channel number.

Sys # System Category KA Statement 000032 Loss of AK2 Knowledge of the interrelations between the Loss of Source Power supplies, including proper switch positions Source Range Nuclear Instrumentation and the following:

Range Nuclear Instrumentati on K/A# AK2.01 K/A Importance 2.7* Exam Level RO References provided to Candidate None Technical

References:

DB-OP-06403 R20 RPS and NI Operating Procedure steps 3.1.3, 3.1.4, 3.1.6 and 3.1.17 Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR 41.7 / 45.7)

Objective: OPS-SYS-502-03K

Davis-Besse 1LOT15 NRC Written Exam r1

22. The plant experienced a Loss of Coolant Accident (LOCA) inside Containment.

A 10 gpm non-isolable leak from the Containment Sump to the Auxiliary Building is discovered.

Containment pressure is 35 psia.

What will be the approximate leak rate when Containment pressure lowers to 20 psia?

A. 8.4 gpm B. 7.6 gpm C. 5.0 gpm D. 2.5 gpm Answer: C Explanation/Justification:

A. Incorrect - see explanation of correct answer. For this distracter 50 and 35 were used for the dP values. F2 = (10 x 5.916) / 7.071= 8.37.

Plausible for gauge to absolute pressure relationship inversion.

B. Incorrect - This distracter based on using 35 and 20 for dP values. F2 = (10 x 4.472) / 5.916 = 7.56. Plausible for candidate using values given as gauge pressures (containment pressure - zero).

C. Correct - dP for calculation is containment pressure - atmospheric pressure. dP1 = 35 psia - 15 psia = 20 psi; dP2 = 20 psia - 15 psia = 5 psi.

Relationship is (F1 / dP1 ) = (F2 / dP2 ). F2 = (F1 x dP2 ) / dP1 . F2 = (10 x 2.236) / 4.472 = 5.0 D. Incorrect - see explanation of correct answer. This distracter based on linear ratio of dPs (20 and 5) to leak rates. F2 = (10 x 5) / 20 = 2.5.

Plausible for candidate forgetting the square root in relationship.

Sys # System Category KA Statement 000069 Loss of AK1 Knowledge of the operational implications of the following Effect of pressure on leak rate Containment concepts as they apply to Loss of Containment Integrity:

Integrity K/A# AK1.01 K/A Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

Question Source: Bank - ANO 2011 #24 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.8 / 41.10 / 45.3)

Objective: OPS-GOP-311-03K

Davis-Besse 1LOT15 NRC Written Exam r1

23. A power change from 60% power to 100% power is in progress.

The Integrated Control System (ICS) is in Automatic

The Load Control Panel is in Automatic with its Rate of Change set to 0.5%/min.

Annunciator 2-1-A LETDOWN RAD HI alarms.

Which of the following would meet the REQUIRED operator actions for these conditions?

(1) Check RI 1998 FAILED FUEL IN LETDOWN INDICATOR to determine if the radiation monitor is operating properly.

(2) Press OPEN on the standby Mixed Bed Demineralizer inlet valve switch HISMU10A or HISMU10B and observe Letdown Flow FI MU7 rises.

(3) Divert Letdown to the Clean Waste Receiver Tank and batch to the Reactor Coolant System at the present Boron concentration.

(4) Press MAN on the ICS Load Control Panel and observe the ULD SETPOINT changes to the current ULD OUTPUT.

A. 1 and 2 B. 2 and 3 C. 1 and 4 D. 3 and 4 Answer: C Explanation/Justification:

A. Incorrect - item 2 placing the standby Mixed Bed Demineralizer in service not a requirement and would not, by itself, raise letdown flow. See DB-OP-02535 R09 High Activity in the RCS step 4.10 which lists it as a potential action to evaluate. See also DB-OP-06006 R35 step 3.20.2. Item 1 is correct - see DB-OP-02535 R09 High Activity in the RCS step 4.5. Plausible because additional RCS cleanup may be desirable.

B. Incorrect - item 2 incorrect (see above); item 3 is also not a requirement. Plausible because additional RCS cleanup may be desirable and feed and bleed of RCS would provide some activity reduction.

C. Correct - See DB-OP-02535 R09 High Activity in the RCS steps 4.1 and 4.5 D. Incorrect - item 3 incorrect (see above). Plausible because feed and bleed of RCS would provide some activity reduction.

Sys # System Category KA Statement 000076 High Reactor AA1 Ability to operate and / or monitor the following as they apply to Failed fuel-monitoring equipment Coolant the High Reactor Coolant Activity:

Activity K/A# AA1.04 K/A Importance 3.2 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02535 R09 High Activity in the RCS steps 4.1 and 4.5 Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR 41.7 / 45.5 / 45.6)

Objective: OPS-GOP-135-02K

Davis-Besse 1LOT15 NRC Written Exam r1

24. The plant is operating at 90% power.

Which of the following conditions will cause an Integrated Control System (ICS) Runback?

A. Reactor Coolant Pump 1-1 current 300 amps B. Main Feed Pump 1 Lube Oil pressure 10 psig C. Main Feed Pump 2 discharge pressure 1470 psig D. Deaerator Storage Tank 1 level 6.0 feet Answer: C Explanation/Justification:

A. Incorrect - Plausible because RCP trip causes a runback and current is 40 amps high. See DB-OP-02014 R14 Window 14-3-C B. Incorrect - MFP trips at 4 psig lube oil pressure. Plausible because MFPT trip causes a runback and lube oil pressure is low.

C. Correct - MFP high discharge pressure runback actuates at 1433 psig. See DB-OP-02014 R14 Window 14-3-D.

D. Incorrect - Low DAST level runback occurs at 4.0 feet. See DB-OP-02014 R14 Window 14-3-D. Plausible because 6.0 feet is below the low level alarm.

Sys # System Category KA Statement BW/A01 Plant AK2 Knowledge of the interrelations between the (Plant Runback) Components, and functions of control and safety Runback and the following: systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

K/A# AK2.1 K/A Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02014 R14 Window 14-3-D Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41.7 / 45.7)

Objective: OPS-SYS-514-03K

Davis-Besse 1LOT15 NRC Written Exam r1

25. The plant is operating at 100% power.

Auxiliary Feed Water (AFW) Pump 1 is out of service.

The following alarms actuate:

11-1-E CLNG TWR BASIN LVL LO

15-1-F HP CNDSR PRESS HI annunciator

15-2-F LP CNDSR PRESS HI annunciator

15-3-F CNDSR PIT FLOODED annunciator After the control room operators take the prescribed actions to stabilize the plant, which of the following is correct?

Feed Water is being supplied by __(1)__.

Equipment issues due to local water level are being addressed per __(2)__.

A. (1) AFW Pump 2 only (2) DB-OP-06272 Station Drainage and Discharge System B. (1) AFW Pump 2 only (2) DB-OP-02517 Circulating Water System Malfunctions C. (1) AFW Pump 2 and the Motor Driven Feed Pump (2) DB-OP-06272 Station Drainage and Discharge System D. (1) AFW Pump 2 and the Motor Driven Feed Pump (2) DB-OP-02517 Circulating Water System Malfunctions Answer: B Explanation/Justification:

A. Incorrect -Flooding mitigation would be addressed using DB-OP-02517. DB-OP-06272 is plausible since it provides guidance for normal station drains operation.

B. Correct - Flooding is in progress in the Condenser Pit. MDFP would not be running because of flooding. See DB-OP-02517 Attachment 3 step 4.0. Leak isolation and flooding issues are addressed using DB-OP-02517 Attachment 3. .

C. Incorrect - MDFP would not have been started. DB-OP-02517 used for flooding issues. Plausible because MDFP would be started by DB-OP-02000 Specific Rule 4 step 4.1 if not for the flooding. DB-OP-06272 is plausible since it provides guidance for normal station drains operation.

D. Incorrect - MDFP would not have been started. Part 2 is correct. Plausible because MDFP would be started by DB-OP-02000 Specific Rule 4 step 4.1 if not for the flooding.

Sys # System Category KA Statement BW/A07 Flooding AA2 Ability to determine and interpret the following as they apply to Facility conditions and selection of appropriate the (Flooding): procedures during abnormal and emergency operations K/A# AA2.1 K/A Importance 3.0 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02517 R06 Circulating Water System Malfunctions steps 2.3, 4.3.1, and Attachment 3.

Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective: OPS-GOP-117-04K

Davis-Besse 1LOT15 NRC Written Exam r1

26. Plant conditions:

The plant tripped from 100% power due to a loss of offsite power.

A Reactor Coolant System (RCS) Cooldown has been initiated to comply with Technical Specification requirements.

RCS Cooldown rate is 10 °F/hr.

The following conditions are observed:

RCS pressure stable

Makeup Tank level sudden rise

Pressurizer (PZR) level sudden rise

Both RCS Tsat meters indicate 30 °F subcooling margin (SCM)

Reactor Vessel Head Vent temperature T012 indicates 19 °F subcooled Which of the following describes the REQUIRED operator action(s) for these conditions?

A. Turn on PZR heaters to compress the RCS steam bubble that is NOT in the PZR.

B. Initiate full MU/HPI flow to compress the RCS steam bubble that is NOT in the PZR.

C. Open the RCS Loop 1 and Loop 2 High Point Vents to vent off the steam bubble(s) in the Hot Leg(s).

D. Throttle open the AVVs for RCS Cooldown rate of 100 °F/hr to condense the steam bubble(s) in the Hot Leg(s).

Answer: A Explanation/Justification:

A. Correct - steam bubble exists in a location other than the PZR. See DB-OP-06903 R47 Plant Cooldown steps 6.4 and 6.5 (page 80).

B. Incorrect - full MU/HPI is NOT REQUIRED because SCM 20 °F per Tsat meters. See DB-OP-02000 step 4.1 (page 18) and Specific Rule 3.2.1 (page 241). Plausible because full MU/HPI flow would compress the non-PZR steam bubble.

C. Incorrect - not required by procedure. Plausible because opening the Loop High Point Vents is an action for Lack of Heat Transfer. See DB-OP-02000 step 6.14 RNO (page 62). Inadequate local heat transfer led to the Hot Leg steam bubble formation. .

D. Incorrect - not required by procedure. Plausible because raising the steaming rate of the SGs promotes natural circulation flow which would help condense the steam bubble(s).

Sys # System Category KA Statement BW/E09 Natural EK3 Knowledge of the reasons for the following responses as they Manipulation of controls required to obtain desired Circulation apply to the (Natural Circulation Cooldown): operating results during abnormal, and Cooldown emergency situations K/A# EK3.3 K/A Importance 3.8 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-06903 R47 Plant Cooldown steps 6.4 and 6.5 (page 80).

Question Source: Bank -#178901 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 / 41.10, 45.6, 45.13)

Objective: OPS-GOP-206-03K

Davis-Besse 1LOT15 NRC Written Exam r1

27. The plant is operating at 100% power.

A Loss of Offsite Power occurs.

NO operator actions have been taken.

Which of the following annunciator alarms has the highest priority for operator response under these conditions?

A. 1-1-A EDG 1 TRBL B. 9-1-F INST AIR HDR PRESS LO C. 10-2-G AFPT 1 OVRSPD TRIP D. 14-2-D ICS/NNI 118 VAC PWR TRBL Answer: C Explanation/Justification:

A. Incorrect - Plausible because 1-1-A indicates potential EDG problem. EDG trip would require Specific Rule 6 implementation for loss of power to Bus C1, but lower priority than Specific Rule 4. See DB-OP-02000 R27 Specific Rule 6.1. (page 250).

B. Incorrect - Plausible because action is required per DB-OP-02000 R27 step 4.7. Specific Rule 4 has higher priority.

C. Correct - operators perform Attachments 5 and 6 to start the MDFP per Specific Rule 4.1. See DB-OP-02000 R27 page 245. Specific Rule 4 is the highest priority condition. See Bases and Deviation Document R20 Specific Rule Prioritization (page 8).

D. Incorrect - Plausible because action is required per DB-OP-02000 R27 step 4.6. Specific Rule 4 has higher priority.

Sys # System Category KA Statement BW/E13 EOP Rules Generic Ability to prioritize and interpret the significance of each annunciator or alarm K/A# 2.4.45 K/A Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02000 R27 Specific Rule 4.1 (page 245);

Bases and Deviation Document R20 Specific Rule Prioritization (page 8)

Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.3 /

45.12)

Objective: OPS-GOP-300-05K

Davis-Besse 1LOT15 NRC Written Exam r1

28. The plant is operating at 35% power.

RCP 2-1 Upthrust Bearing temperature is 220 °F.

What operator actions are REQUIRED?

A. Trip the Reactor and stop RCP 2-1.

B. Perform a Rapid Shutdown and stop RCP 2-1.

C. Stop RCP 2-1 and notify I & C to reduce the RPS High Flux Trip setpoints.

D. Lower CCW temperature to 85 °F and start RCP 2-1 AC Lift Oil Pump.

Answer: C Explanation/Justification:

A. Incorrect - Reactor trip not required. Plausible because these are the actions for reactor critical with 3 RCPs operating. See DB-OP-02515 R12 RCP and Motor Abnormal Operation step 4.6.1 RNO.

B. Incorrect - Rapid Shutdown not required. Power is 35%, Rapid Shutdown required if power is > 72%. See DB-OP-02515 R12 RCP and Motor Abnormal Operation Attachment 1 RCP Shutdown step 1. Plausible because these would be the correct actions at full power.

C. Correct - See DB-OP-02515 R12 RCP and Motor Abnormal Operation step 4.6.1 RNO and Attachment 1 RCP Shutdown steps 3 and 7.

D. Incorrect - RCP stop required for bearing temperature 190 ºF per DB-OP-02515 R12 RCP and Motor Abnormal Operation step 4.6.1 RNO.

Plausible because these are the actions for bearing temperature above 185 ºF but less than 190 ºF. See DB-OP-06005 R31 RCP Operation steps 4.2.3 and 4.2.5.

Sys # System Category KA Statement 003 Reactor Generic Ability to evaluate plant performance and make Coolant operational judgments based on operating Pump characteristics, reactor behavior, and instrument System interpretation (RCPS)

K/A# 2.1.7 K/A Importance 4.4 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02515 R12 RCP and Motor Abnormal Operation step 4.6.1 RNO and Attachment 1 RCP Shutdown steps 3 and 7.

Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.12 /

45.13)

Objective: OPS-SYS-105-08K

Davis-Besse 1LOT15 NRC Written Exam r1

29. The plant is operating at 70% power.

Core T (Reactor Coolant System Thot - Tcold) is 33 ºF Reactor Coolant Pump (RCP) 1-2 breaker opens spuriously.

NO operator actions are taken.

Which of the following describes the change in Core T when the plant stabilizes?

Core T will _____.

A. go to approximately 1 ºF following the automatic reactor trip B. lower to approximately 25 ºF due to the reduction in RCS flow C. remain at 33 ºF since reactor power does not change D. rise to approximately 44 ºF due to the reduction in RCS flow Answer: D Explanation/Justification:

A. Incorrect - reactor does not trip from stop of RCP at 70% power. Plausible for RCP trip at higher power.

B. Incorrect - Plausible for misapplication of Q=mT. This value is 75% of the given value pf 33 ºF.

C. Incorrect - Plausible for misapplication of Q=mT.

D. Correct - for power constant at 70% with RCS flow reduction to 75%, T = 33 ÷ 0.75 = 44 ºF. See DB-PF-06703 R22 Miscellaneous Operation Curves CC2.2 and CC2.3 Sys # System Category KA Statement 003 Reactor K3 Knowledge of the effect that a loss or malfunction of the RCPS RCS Coolant will have on the following:

Pump System (RCPS)

K/A# K3.01 K/A Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

DB-PF-06703 R22 CC2.2 and CC2.3 Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41.7 / 45.6)

Objective: OPS-SYS-105-03K

Davis-Besse 1LOT15 NRC Written Exam r1

30. A Lack of Heat Transfer event has occurred.

The operators are performing the actions for Recovery from MU/HPI PORV Cooling.

Reactor Coolant System (RCS) Subcooling Margin (SCM) has been restored.

Letdown flow has been established through Orifice Block MU4 and Letdown Flow Control MU6.

The normal Makeup System alignment has been established with flow through Pressurizer Level Control Valve MU32 only.

The operators close the Power Operated Relief Valve (PORV) RC2A.

MU6 fails closed.

Which of the following describes the operator action required to offset the MU6 failure when controlling RCS pressure?

A. Raise flow through MU4.

B. Raise flow through MU32.

C. Lower flow through MU4.

D. Lower flow through MU32.

Answer: D Explanation/Justification:

A. Incorrect - MU4 block orifice valve is already full open, so flow cannot be raised. Plausible for candidate inversion of MU4 operation and MU6 operation. MU6 is a throttle valve. Raising letdown flow would restore the letdown - makeup flow balance for RCS pressure control.

B. Incorrect - Plausible if candidate misses that SG heat transfer must be established to recover from MU/HPI PORV Cooling and focuses on the cooling effect of raising MU flow. Lowering RCS temperature lowers RCS pressure.

C. Incorrect - closing MU4 orifice block valve makes the letdown - makeup flow imbalance worse, causing a higher rise in RCS pressure. Plausible for candidate inversion of the effect of letdown flow on solid RCS pressure control.

D. Correct - MU6 closure lowers letdown flow which raises RCS pressure. MU32 must be throttled closed to compensate. DB-OP-02000 R27 CAUTION 6.9 and step 7 (page 402)

Sys # System Category KA Statement 004 Chemical and K6 Knowledge of the effect of a loss or malfunction on the following Methods of pressure control of solid plant (PZR Volume CVCS components: relief and water inventory)

Control System K/A# K6.26 K/A Importance 3.8 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02000 R27 CAUTION 6.9 and step 7 (page 402)

Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.7 / 45.7)

Objective: OPS-GOP-305-05K

Davis-Besse 1LOT15 NRC Written Exam r1

31. The Decay Heat Removal System design feature which can provide flow to the Makeup (MU) and High Pressure Injection (HPI) Systems is called Piggyback operation.

Which of the following describes the Design Basis of Piggyback operation?

Piggyback operation was designed to ensure _______.

A. maximum HPI flow following a loss of all MU flow capability B. maximum MU/HPI flow during a Lack of Heat Transfer event C. adequate NPSH for the MU Pumps during Injection Phase of a Loss of Coolant Accident D. adequate NPSH for the HPI Pumps during Recirculation Phase of a Loss of Coolant Accident Answer: D Explanation/Justification:

A. Incorrect - not the design basis. Plausible because HPI Piggyback operation is initiated for this event. See DB-OP-02512 R14 Makeup and Purification System Malfunctions step 4.1.11 RNO B. Incorrect - not the design basis. Plausible because MU/HPI Piggyback operation is initiated for this event. See DB-OP-02000 R27 step 6.3.3 (page 56) and Attachment 8 step 2.b (page 314)

C. Incorrect - not the design basis. Plausible because there is no MU flow limit based on NPSH when Piggybacked. See DB-OP-02000 R27 Specific Rule 3.2.4 (page 241)

D. Correct - see UFSAR Section 6.3.2.11 page 6.3-6 Sys # System Category KA Statement 005 Residual K4 Knowledge of RHRS design feature(s) and/or interlock(s) which Lineup for "piggy-back" mode with high-pressure Heat provide or the following: injection Removal System (RHRS)

K/A# K4.08 K/A Importance 3.1* Exam Level RO References provided to Candidate None Technical

References:

UFSAR Section 6.3.2.11 page 6.3-6 Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.7)

Objective: OPS-SYS-303-06K

Davis-Besse 1LOT15 NRC Written Exam r1

32. The plant is operating at 100% power.

Which of the following will prevent High Pressure Injection Pump 1 from starting?

_____ Lockout Relays actuated A. Emergency Diesel Generator 1 B. Emergency Diesel Generator 2 C. 4160V AC Bus C1 D. 4160V AC Bus D1 Answer: C Explanation/Justification:

A. Incorrect - C1 remains energized from its normal source (Bus C2) via AC110 and nothing interferes with an automatic start of HPI Pump 1. The EDG lockouts do not affect AC110. See OS-0041A CD-2 (sheet 1). Plausible because HPI Pump 1 wont automatically start during a LOCA concurrent with loss of offsite power, which is its design function B. Incorrect - EDG 2 supports HPI Pump 2. Plausible for inversion of HPI pump power supplies.

C. Correct - See OS-0003 R36 CL-2.

D. Incorrect - plausible for inversion of HPI pump power supplies.

Sys # System Category KA Statement 006 Emergency K2 Knowledge of bus power supplies to the following: ECCS pumps Core Cooling System (ECCS)

K/A# K2.01 K/A Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

OS-0003 R36 CL-2 Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.7)

Objective: OPS-SYS-302-03K

Davis-Besse 1LOT15 NRC Written Exam r1

33. The plant is operating at 100% power.

Which of the following describes:

(1) An inadvertent SFAS Incident Level trip that requires entry into DB-OP-02000 RPS, SFAS, SFRCS Trip or SG Tube Rupture?

(2) The DB-OP-02000 Response Not Obtained action for the Verify Turbine Valves closed Immediate Action?

A. (1) 3 (2) Initiate and Isolate SFRCS B. (1) 2 (2) Initiate and Isolate SFRCS C. (1) 3 (2) Stop BOTH EHC Fluid Pumps D. (1) 2 (2) Stop BOTH EHC Fluid Pumps Answer: A Explanation/Justification:

A. Correct -SFAS Level 3 requires DB-OP-2000 entry. See DB-OP-02000 R27 step 1.2.2. Initiate & Isolate SFRCS per step 3.5 RNO.

B. Incorrect - Inadvertent Level 2 does NOT require entry into DB-OP-02000. Part 2 is correct. Plausible because Level 2 isolates letdown which requires a plant shutdown if not corrected. See DB-OP-02512 R14 Makeup and Purification System Malfunctions step 4.3.8 RNO.

C. Incorrect - Initiate & Isolate SFRCS per step 3.5 RNO. Part 1 correct. Stop both EHC pumps plausible because that is the RNO action for tripping the turbine in the turbine trip procedure. See DB-OP-02500 R13 Turbine Trip step 4.1 RNO.

D. Incorrect - Inadvertent Level 2 does NOT require entry into DB-OP-02000. See DB-OP-02000 R27 step 1.2.2. Initiate & Isolate SFRCS per step 3.5 RNO. Plausible because Level 2 isolates letdown which requires a plant shutdown if not corrected. See DB-OP-02512 R14 Makeup and Purification System Malfunctions step 4.3.8 RNO. Stop both EHC pumps plausible because that is the RNO action for tripping the turbine in the turbine trip procedure. See DB-OP-02500 R13 Turbine Trip step 4.1 RNO.

Sys # System Category KA Statement 006 Emergency Generic Knowledge of EOP entry conditions and Core Cooling immediate action steps System (ECCS)

K/A# 2.4.1 K/A Importance 4.6 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02000 R27 steps 1.2.2 and 3.5 Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.13)

Objective: OPS-GOP-304-03K

Davis-Besse 1LOT15 NRC Written Exam r1

34. The plant is operating at 100% power.

The following conditions are noted in the control room:

LI225 Quench Tank Level 9 feet and slowly lowering

LI1721 Reactor Coolant Drain Tank Level 20 inches and slowly rising

Quench Tank Circ Pump GREEN light is LIT Which of the following is occurring?

A. Pressurizer Code Safety Valve leakage B. Pressurizer Power Operated Relief Valve leakage C. Pressurizer High Point Vent line valves leaking by D. Quench Tank Demineralized Water makeup valves leaking by Answer: D Explanation/Justification:

A. Incorrect - safety valve leakage would raise Quench Tank temperature which would start the Circ Pump. See OS-0001A sheet 3 CL-9 (sheet 4).

Plausible because the Circ Pump is not required to drain the Quench Tank. See DB-OP-06004 R10 Quench Tank NOTE 4.2.4.

B. Incorrect - same as Safety Valve leakage.

C. Incorrect - same as Safety Valve leakage.

D. Correct - Demin Water in-leakage causes Quench Tank level rise. At 9.5 ft, RC225A opens to start water transfer to the RCDT See OS 0001A sheet 3 R26 C-43 and sheet 4 R24 CL-10. Demin Water in-leakage would not cause a rise in Quench Tank temperature, so Circ Pump would not start. See OS-0001A Sheet 3 R26 CL-9.

Sys # System Category KA Statement 007 Pressurizer A3 Ability to monitor automatic operation of the PRTS, including: Components which discharge to the PRT Relief Tank

/Quench Tank System (PRTS)

K/A# A3.01 K/A Importance 2.7* Exam Level RO References provided to Candidate None Technical

References:

OS-0001A sheets 3 & 4 CL-9 and CL-10 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 / 45.5)

Objective: OPS-SYS-104-04K

Davis-Besse 1LOT15 NRC Written Exam r1

35. Decay Heat (DH) Cooler 1 Component Cooling Water (CCW) side will be vented to remove a void discovered during ultrasonic testing.

Fill water will be provided from the CCW Surge Tank volume.

Which of the following describes:

(1) the preferred status of CCW Pump 1 during DH Cooler 1 filling?

(2) the method to maintain CCW Surge Tank level?

(1) CCW Pump 1 should be _____.

(2) Maintain CCW Surge Tank level 51 to 53 inches by _____ as required.

A. (1) stopped to minimize potential air entrainment (2) opening DW2643 using HIS 2643 DEMIN WTR MAKEUP B. (1) stopped to minimize potential air entrainment (2) opening SW234 and SW233 SW HEADER 1 TIE TO CCW SYSTEM ISOLATION valves C. (1) operating to ensure complete filling of the DH Cooler (2) opening DW2643 using HIS 2643 DEMIN WTR MAKEUP D. (1) operating to ensure complete filling of the DH Cooler (2) opening SW234 and SW233 SW HEADER 1 TIE TO CCW SYSTEM ISOLATION valves Answer: A Explanation/Justification:

A. Correct - CCW pump OFF to minimize air entrainment per DB-OP-06262 R36 CCW System Procedure step 2.2.17. Demin Water makeup per section 3.23.

B. Incorrect - Demin Water makeup per section 3.23. Part 1 is correct. Plausible because SW is emergency backup to demin water. See DB-OP-06262 R36 section 5.1.

C. Incorrect - CCW pump OFF to minimize air entrainment. Part 2 is correct. Plausible for higher pressure = better fill.

D. Incorrect - CCW pump OFF to minimize air entrainment. Demin Water makeup per section 3.23. Plausible for higher pressure = better fill and because SW is backup to demin water.

Sys # System Category KA Statement 008 Component A4.02 Ability to manually operate and/or monitor in the control room: Filling and draining operations of the CCWS Cooling including the proper venting of the components Water System (CCWS)

K/A# A4.02 K/A Importance 2.5* Exam Level RO References provided to Candidate Technical

References:

DB-OP-06262 R36 CCW System Procedure step 2.2.17 and Section 3.23.

Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.7 / 45.5)

Objective: OPS-SYS-304-07K

Davis-Besse 1LOT15 NRC Written Exam r1

36. The plant is operating at 100% power.

Component Cooling Water (CCW) Pump 1 is operating.

PI810 INSTRUMENT AIR HEADER PRESS lowers to 50 psig and stabilizes.

The operators perform the required abnormal procedure actions then implement DB-OP-02000 RPS, SFAS, SFRCS Trip or SG tube Rupture.

During the performance of the applicable abnormal procedure in parallel with DB-OP-02000, the operators take actions to mitigate the effects of this failure.

Which of the following describes an impact of the low Instrument Air pressure and the operator action to mitigate its effects?

A. CCW Pump 1 experiences run-out when Decay Heat Cooler 1 outlet valve CC1467 fails open. The operators manually isolate CCW flow to Decay Heat Cooler 1 to restore normal CCW Pump flow rate.

B. The CCW Containment Header experiences flow starvation when Decay Heat Cooler 2 outlet valve CC1469 fails open. The operators manually isolate CCW flow to Decay Heat Cooler 2 to restore normal CCW Containment Header flow rate.

C. The Control Rod Drive (CRD) Booster Pump experiences flow starvation when the Spent Fuel Pool (SFP) Heat Exchanger Outlet Valves CC1454 and CC1457 fail open. The operators manually isolate CCW flow to the SFP Heat Exchangers to restore normal CRD Booster Pump flow rate D. Reactor Coolant Pumps lose CCW flow through their seal coolers when CCW to Aux Building Non-essential Header isolation valve CC1495 fails closed. The operators open the manual bypass valve CC43 to restore Reactor Coolant Pump Seal Cooling.

Answer: A Explanation/Justification:

A. Correct - CC1467 fails open See DB-OP-02528 R22 Instrument Air System Malfunctions Attachment 18 Failure Position of Pneumatic Valves (page 106). DHR HX is isolated per Attachment 8 CCW System Actions step 3. Attachment 8 CAUTION 1 second bullet describes run-out damage concern.

B. Incorrect - CCW Pump 1 is supplying the Containment Header, so flow starvation affecting letdown cooling on CCW Pump 2 wont occur.

Plausible because CC1469 fails open (DB-OP-02528 R22 Attachment 18 page 106) and its isolation is directed by DB-OP-02528 R22 Attachment 8 step 3.

C. Incorrect - CCW System flow is not affected when CC1454 and CC1457 fail open because their supply is isolated when CC1495 fails closed.

See OS-0021 sheet 2 and sheet 1. Plausible because CC1454 and CC1457 fail open.

D. Incorrect - RCP cooling is supplied by the Containment Header. See OS-0021 sheet 2. Plausible because this header supplies the RCP Seal Return Coolers, CC1495 fails closed (DB-OP-02528 R22 Attachment 18 page 106), and CC43 is directed to be opened per DB-OP-02528 R22 Attachment 8 CCW System Actions step 4.

Sys # System Category KA Statement 008 Component A2 Ability to (a) predict the impacts of the following malfunctions or Effect of loss of instrument and control air on the Cooling operations on the CCWS, and (b) based on those predictions, use position of the CCW valves that are air operated Water procedures to correct, control, or mitigate the consequences of those System malfunctions or operations:

(CCWS)

K/A# A2.05 K/A Importance 3.3* Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02528 R22 Attachment 18 page 106, Attachment 8 page 62 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.3 /

45.13)

Objective: OPS-SYS-304-07K

Davis-Besse 1LOT15 NRC Written Exam r1

37. The following conditions exist:

Reactor Coolant System pressure is 485 psig.

Quench tank pressure is 20 psig.

The Pressurizer Power Operated Relief Valve (PORV) RC2A lifts.

What is the PORV tailpipe downstream temperature?

A. 235 to 245 ºF B. 260 to 270 ºF C. 330 to 340 ºF D. 460 to 470 ºF Answer: C Explanation/Justification:

A. Incorrect - plausible because it is saturation temperature for 20 psia (common error)

B. Incorrect - plausible because it is 1205 BTU/lb expanded to 35 psia, but at constant entropy C. Correct - enthalpy at 500 psia 100% quality is 1205 BTU/lb. expand to 35 psia = 335 ºF D. Incorrect - plausible because it is saturation temperature for 500 psia (common error).

Sys # System Category KA Statement 010 Pressurizer K5 Knowledge of the operational implications of the following Constant enthalpy expansion through a valve Pressure concepts as they apply to the PZR PCS:

Control System (PZR PCS)

K/A# K5.02 K/A Importance 2.6 Exam Level RO References provided to Candidate Steam Tables with Mollier Technical

References:

Diagram Question Source: Bank # 167005 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 / 45.7)

Objective: OPS-SYS-104-03K

Davis-Besse 1LOT15 NRC Written Exam r1

38. The plant was operating at 100% power.

A manual Reactor trip was initiated, but the Control Rod Drive (CRD) breakers did NOT open.

Reactor was shut down from the Control Room by alternate means.

NO Trip Confirm signal was generated

The CRD breakers are still closed Which of the following describes Steam Generator (SG) pressure control for these conditions?

SG pressures are being maintained at about __(1)__ by the __(2)__.

A. (1) 880 psig (2) Turbine Bypass Valves B. (1) 880 psig (2) Atmospheric Vent Valves C. (1) 995 psig (2) Turbine Bypass Valves D. (1) 995 psig (2) Atmospheric Vent Valves Answer: A Explanation/Justification:

A. Correct - The turbine was manually tripped per Immediate Action 3.4. Reactor Shutdown by de-energizing E2 and F2 leaves the CRD trip breakers closed. The reactor tripped status input to the ICS turbine header pressure control logic is the Trip Confirm signal. See DB-OP-06402 R25 CRD Operating Procedure Attachment 2 Rod Control Panel Indicating Lights item 1 (page 147). Since there is no Trip Confirm signal and the turbine is tripped, there is no bias added to the Turbine Header Pressure set point of 880 psig. See DB-OP-06401 R23 ICS Operating Procedure Attachment 9 (page 103).

B. Incorrect - SG pressure control transfers from the TBVs to the AVVs on low vacuum of closure of either MSIV. See DB-OP-06401 R23 Attachment 10 item 5 (page 106). Neither of these conditions exists. Plausible because part 1 is correct SG pressure control is based on individual SG pressure signals when the turbine stop valves are closed. See DB-OP-06401 R23 Attachment 10 item 2. Individual SG pressure signal control is often confused with AVV control.

C. Incorrect - no 115 psi bias signal because there is no Trip Confirm signal. Part 2 is correct. Plausible because this is the normal response to a reactor trip.

D. Incorrect - both parts are incorrect. Plausible because 995 psig is normal pressure control set point post-trip and misapplication of pressure signal transfer to AVVs.

Sys # System Category KA Statement 012 Reactor K3 Knowledge of the effect that a loss or malfunction of the RPS will Steam Dump System Protection have on the following:

System (RPS)

K/A# K3.03 K/A Importance 3.1* Exam Level RO References provided to Candidate None Technical

References:

DB-OP-06402 R25 (page 147); DB-OP-06401 R23 (page 103)

Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 / 45.6)

Objective: OPS-SYS-501-06K

Davis-Besse 1LOT15 NRC Written Exam r1

39. The plant is operating at 100% power.

Two of the Reactor Protection System (RPS) trip functions utilize variable setpoints that are calculated by Trip Setpoint Calculators.

Which of the following sets of Trip Setpoint Calculator malfunctions will cause BOTH RPS Channels to trip?

(1) RPS Channel 1 Flux/Delta Flux/Flow Trip setpoint fails __(1)__.

(2) RPS Channel 2 RCS Pressure/Temperature Trip setpoint fails __(2)__.

A. (1) low (2) low B. (1) low (2) high C. (1) high (2) low D. (1) high (2) high Answer: B Explanation/Justification:

A. Incorrect - P-T bistable trips when RCS pressure is lower than the calculated setpoint. The P-T setpoint calculator output failing low drives the variable low pressure trip setpoint even farther below actual RCS pressure. Setpoint pressure calculation is 16.25 Thot - 7885.5 psig, so normal full power setpoint is about 1945 psig. Plausible because item 1 is correct and candidate can easily invert parameter - setpoint relationship when answering part 2.

B. Correct - Flux/Delta Flux Flow bistable trips when reactor power signal is higher than the calculated setpoint. Setpoint calculator output low = trip because actual NI power is above the failed low setpoint. See UFSAR 7.2.1.2.2 item 7 (page 7.2-6). Pressure/Temperature (aka Variable Low RC Pressure) bistable trips when RCS pressure signal is lower than the calculated setpoint. Setpoint calculator output high = trip because actual RCS pressure is below the failed high setpoint. See UFSAR 7.2.1.2.2 item 4 (page 7.2-5) and TRM Table 8.3.1-2 C. Incorrect - Flux/Delta Flux Flow bistable trips when reactor power signal is higher than the calculated setpoint. The setpoint calculator failing high drives the Flux/Delta Flux/Flow high power trip setpoint even higher above actual power. P-T bistable trips when RCS pressure is lower than the calculated setpoint. The P-T setpoint calculator output failing low drives the variable low pressure trip setpoint even farther below actual RCS pressure. Both plausible because candidate can easily invert parameter - setpoint relationship when answering. Item 1 also plausible if candidate equates setpoint calculator high failure with high Delta Flux input which would cause a trip.

D. Incorrect - Flux/Delta Flux Flow bistable trips when reactor power signal is higher than the calculated setpoint. The setpoint calculator failing high drives the Flux/Delta Flux/Flow high power trip setpoint even higher above actual power. Part 2 is correct. Plausible because candidate can easily invert parameter - setpoint relationship when answering Sys # System Category KA Statement 012 Reactor K6 Knowledge of the effect of a loss or malfunction of the following Trip setpoint calculators Protection will have on the RPS:

System (RPS)

K/A# K6.11 K/A Importance 2.9* Exam Level RO References provided to Candidate None Technical

References:

UFSAR 7.2.1.2.2; TRM Table 8.3.1-2 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 / 45/7)

Objective: OPS-SYS-504-05K

Davis-Besse 1LOT15 NRC Written Exam r1

40. The plant is operating at 100% power.

A Design Bases Loss of Coolant Accident (DBLOCA) occurs.

Which of the following describes:

(1) the definition of a Safety Features Actuation System (SFAS) safety train?

(2) the operational implication of the failure of one SFAS safety train on the DBLOCA Analyses assumptions?

(1) SFAS Channels __(1)__ Output Modules comprise Safety Actuation Train 1.

(2) DBLOCA Analyses assumptions __(2)__ met.

A. (1) 1 and 4 (2) are B. (1) 1 and 3 (2) are C. (1) 1 and 4 (2) are NOT D. (1) 1 and 3 (2) are NOT Answer: B Explanation/Justification:

A. Incorrect - SFAS Channels 1 and 3 Output Modules comprise Actuation Channel 1. See Bases 3.3.7 2nd paragraph. Part 2 is correct. Plausible for Channels 1 & 4 Output Modules = Safety Train 1.

B. Correct - SFAS Channels 1 and 3 Output Modules comprise Actuation Channel 1. See Bases 3.3.7 2nd paragraph. DBLOCA Analyses assumptions are met. See UFSAR R30 6.3.2.11.

C. Incorrect - SFAS Channels 1 and 3 Output Modules comprise Actuation Channel 1. See Bases 3.3.7 2nd paragraph. DBLOCA Analyses assumptions are met. See UFSAR R30 6.3.2.11. Plausible for Channels 1 & 4 Output Modules = Safety Train 1 and single Safety Train actuation provides insufficient ECCS.

D. Incorrect - DBLOCA Analyses assumptions are met. See UFSAR R30 6.3.2.11. Part 1 is correct. Plausible for single Safety Train actuation provides insufficient ECCS.

Sys # System Category KA Statement 013 Engineered K5 Knowledge of the operational implications of the following Definitions of safety train and ESF channel Safety concepts as they apply to the ESFAS:

Features Actuation System (ESFAS)

K/A# K5.01 K/A Importance 2.8 Exam Level RO References provided to Candidate None Technical

References:

Bases 3.3.7 2nd paragraph; UFSAR R30 6.3.2.11 page 6.3-8 Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41.5 / 45.7)

Objective: OPS-SYS-506-05K

Davis-Besse 1LOT15 NRC Written Exam r1

41. The plant is operating at 100% power.

Containment Air Coolers 1 and 2 are operating.

The following events occur:

Large Break Loss of Coolant Accident

Loss of Offsite Power

Bus D1 Lockout NO operator actions are taken.

Five minutes after the above events, what is the status of the Containment Air Cooler (CAC) Fans and their associated Service Water (SW) valves?

CAC Fan 1 speed is __(1)__; its inlet SW1366 is __(2)__.

CAC Fan 2 speed is __(3)__; its inlet SW1367 is __(4)__.

A. (1) slow (2) open (3) zero (4) open B. (1) zero (2) open (3) slow (4) open C. (1) zero (2) closed (3) zero (4) closed D. (1) slow (2) open (3) slow (4) open Answer: A Explanation/Justification:

A. Correct - CACs start in SLOW from SFAS Level 2 signal. See DB-OP-02000 R27 page 418. CAC SW inlet valves initially close on restoration of power, then go through a refill sequence back to fully open. Refill sequence takes inlet valve back to fully open over about 4 minutes. See OS-0020 sheet 2 CL-11. CAC 2 is at zero speed because it has no power due to D1 lockout; its inlet valve is open because it had no power to initiate the refill sequence (alos D1 lockout).

B. Incorrect - backwards. Plausible for misconception of D1 power to CAC 1 (#1 bus to #1 component).

C. Incorrect - Plausible because this would be the status of both CACs following LOP only. See OS-0020 sheet 2 CL-11. Fans are normally in FAST speed. See OS-0033A Note 13.

D. Incorrect - Plausible because this would be the status without the D1 lockout.

Sys # System Category KA Statement 022 Containment K2 Knowledge of power supplies to the following: Containment cooling fans Cooling System (CCS)

K/A# K2.01 K/A Importance 3.0* Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02000 R27 page 418; OS-0020 sheet 2 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7)

Objective: OPS-SYS-305-05K

Davis-Besse 1LOT15 NRC Written Exam r1

42. The plant is operating at 100% power.

The following events occur:

Double-ended guillotine Hot Leg break at the inlet of a Steam Generator

Offsite Power Loss

Bus D1 Lockout The locked out bus can NOT be restored.

At the specified level in the Borated Water Storage Tank (BWST), the operators transfer Low Pressure Injection (LPI) Suction to the Emergency Sump.

Which of the following describes the INITIAL response of Containment pressure after the transfer is complete?

Containment Pressure __(1)__ due to __(2)__.

A. (1) rises (2) elevated Containment Sump temperature B. (1) rises (2) throttling of Containment Spray Discharge Valve C. (1) lowers at a slower rate (2) elevated Containment Sump temperature D. (1) lowers at a slower rate (2) throttling of Containment Spray Discharge Valve Answer: A Explanation/Justification:

A. Correct - See UFSAR R30 Section 6.2.1.3.2 page 6.2-11 Long-term Containment Analysis and Figure 6.2-10. Containment pressure rises for the first 2000 seconds (half hour) after swap to sump.

B. Incorrect - per UFSAR Section 6.2.1.3.2 page 6.2-11 Long-term Containment Analysis, the majority of heat removal from Containment during recirculation is performed by the CAC and the Decay Heat Removal Cooler, so throttling of spray flow has a minor effect. Plausible because Containment pressure trend is correct and spray flow is lowered by throttling.

C. Incorrect - Plausible because elevated sump temperature is correct. Slower lowering might be seen for the misconception of a smaller effect on pressure or for both trains of Containment Cooling operating.

D. Incorrect - Plausible for the misconception of slower lowering due to a smaller effect of sump temperature on pressure or for both trains of Containment Cooling operating. Spray flow is lowered by throttling.

Sys # System Category KA Statement 026 Containment A1 Ability to predict and/or monitor changes in parameters (to Containment pressure Spray System prevent exceeding design limits) associated with operating the CSS (CSS) controls including:

K/A# A1.01 K/A Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

UFSAR R30 Section 6.2.1.3.2 page 6.2-11, Figure 6.2-10 Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.5 / 45.5)

Objective: OPS-SYS-306-02K

Davis-Besse 1LOT15 NRC Written Exam r1

43. The plant is operating at 100% power.

A large break Loss of Coolant Accident occurs.

The operators take all required procedure actions and are preparing to transfer Low Pressure Injection (LPI) Suction to the Emergency Sump.

At this point it is observed that the Main Steam Isolation Valves (MSIVs) MS100 and MS101 are closed.

Which of the following describes why the MSIVs are closed?

A. Manual closure required by procedure B. Manual trip of all Reactor Coolant Pumps C. Automatic Steam Feed Rupture Control System Actuation D. Automatic Safety Features Actuation System Level 4 Actuation Answer: C Explanation/Justification:

A. Incorrect - no manual SFRCS trip is directed in the routing for a Large LOCA. Plausible because Containment isolation is desirable.

B. Incorrect - RCP trip causes Actuation Only SFRCS trip. Plausible because manual trip of all RCPs will be performed and candidate may have misconception that this causes an SFRCS Isolation trip.

C. Correct - trip of all RCPs is required for loss of Subcooling Margin per Specific Rule 2. Trip of all RCPs causes Steam & Feed Rupture Control System (SFRCS) Actuation Only Trip which starts Auxiliary Feed Water (does NOT close MSIVs). SFAS Level 2 actuation raises the SG level control setpoint from 49 inches to 124 inches, so full AFW flow is supplied to both SGs. Since the RCS and the SGs are no longer hydraulically coupled due to the LOCA, SG pressures lower rapidly. An SFRCS Isolation trip occurs at 630 psig and closes the MSIVs. See DB-OP-02000 R27 Specific Rule 2 (page 240), step 5.7 (page 42), Table 2 (page 419), and Table 1 (page 415). MSIVs are Containment Isolation Valves D. Incorrect - SA Level 4 actuates Level 3 Containment Isolation (the highest), but it does not close MSIVs. See DB-OP-02000 R27 Table 2 (page 421) and UFSAR 6.2.4.2.1 page 6.2-57. Plausible because Containment Isolation is desirable.

Sys # System Category KA Statement 039 Main and K4 Knowledge of MRSS design feature(s) and/or interlock(s) which Reactor building isolation Reheat provide for the following:

Steam System (MRSS)

K/A# K4.07 K/A Importance 3.4 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02000 R27 Specific Rule 2, step 5.7, Table 2, and Table 1 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7)

Objective: OPS-SYS-202-06K

Davis-Besse 1LOT15 NRC Written Exam r1

44. The plant is at 45% power.

Reactor Coolant Pump 1-1 is NOT operating.

The reactor trips.

Which of the following describes the response of Integrated Control System (ICS) Rapid Feedwater Reduction (RFR)?

RFR will _____.

A. NOT actuate since one Steam Generator is on Low Level Limits B. actuate causing Main Feedwater Pump speed to go to approximately 4600 rpm C. NOT actuate because one Main Feedwater Pump is tripped D. actuate causing the Main Feedwater Control Valves to go to 15% open Answer: B Explanation/Justification:

A. Incorrect - RFR actuates. SG Low Level Limits is not an input to RFR actuation logic. See M-533 00178 R13 Logic String 11. Plausible for confusion of single SG release from RFR at Low Level Limits or after 2.5 minutes. See M-533-00178 R13 Logic String 15 and DB-OP-02000 R27 step 2.1.4.

B. Correct - RFR actuates. See M-533 00178 R13 Logic String 11. MFPT to target speed per Logic string 11 FWD27.1 and FWD27.2 and M-533-00176-2 R FW21.9. Target speed 4600 rpm per DB-OP-02000 R27 step 2.1.4.

C. Incorrect - RFR actuates. See M-533 00178 R13 Logic String 11. Plausible because both MFPTs tripped would prevent RFR actuation.

D. Incorrect - MFW Control valves close. Plausible for inversion with SUFW valves.

Sys # System Category KA Statement 059 Main K4.18 Knowledge of MFW design feature(s) and/or interlock(s) which Automatic feedwater reduction on plant trip Feedwater provide for the following:

(MFW)

System K/A# K4.18 K/A Importance 2.8* Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02000 R27 step 2.1.4; M-533-00178 R13 Logic String 11 Question Source: Bank #168736 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7)

Objective: OPS-SYS-207-06K

Davis-Besse 1LOT15 NRC Written Exam r1

45. The plant is operating at 100% power.

Computer point T879 SG 1 AFW NOZZLE TEMP alarms.

T879 indicates 210 ºF and rising.

The high temperature condition is confirmed locally at Auxiliary Feedwater (AFW) to Steam Generator (SG) 1 Line Stop valve AF608.

Which of the following describes the operational implications of this condition?

A. AFW flow to SG 1 may be limited by line voiding. Only one Emergency Feedwater Train is INOPERABLE during mitigation of the condition.

B. AFW Train 1 is steam bound and may NOT produce sufficient flow. Mitigation of the condition requires all three Emergency Feedwater Trains to be made INOPERABLE for a short period of time.

C. AF608 may not close if needed due to being outside of its Environmental Qualification temperature. NO Emergency Feedwater Trains are INOPERABLE during mitigation of the condition.

D. Water hammer could induce a steam break on SG 1 if AFW flow is initiated. Mitigation of the condition requires two Emergency Feedwater Trains to be made INOPERABLE for a short period of time Answer: B Explanation/Justification:

A. Incorrect - DB-OP-06233 R37 NOTE 4.9.5 references step 2.1.6 which states all three EFW Trains are inoperable while AF608 is closed for venting. Plausible because flow could be limited and AF608 closure effect on the other two trains could be overlooked.

B. Correct - See DB-OP-06233 R37 Section 4.9 Discovery and Resolution of Steam Binding in AFW Train 1 Components.

C. Incorrect - Plausible because AF608 is an EQ valve, no lineup changes would be required for cooling it down.

D. Incorrect - All three EFW Trains inoperable while AF608 is closed for venting - See DB-OP-06233 R37 step 2.1.6. Plausible because water hammer could induce an AFW line break; MDFP and AFW Train 1 both discharge through AF608.

Sys # System Category KA Statement 061 Auxiliary / K5 Knowledge of the operational implications of the following Feed line voiding and water hammer Emergency concepts as they apply to the AFW:

Feedwater (AFW)

System K/A# K5.05 K/A Importance 2.7 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-06233 R37 Section 4.9 and step 2.1.6 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 / 45.7)

Objective: OPS-SYS-213-13K

Davis-Besse 1LOT15 NRC Written Exam r1

46. If the Auxiliary Feedwater (AFW) Pumps start automatically, the Control Room contacts an operator to locally shift the AFW Pump recirculation flow path.

Performing this action maintains which of the following?

A. AFW Pump seal and bearing temperatures within limits B. Condensate Storage Tank chemistry parameters within specification C. Offsite radioactive material releases As Low As Reasonably Achievable D. The margin assumed in the Condensate Storage Tank Capacity analysis Answer: D Explanation/Justification:

A. Incorrect - local operator opens AF50 & AF51, then closes AF59. See DB-OP-02000 R27 step 4.18 (page 34) and Bases and Deviation Document for DB-OP-02000 R20 step 4.18 (page 43). Plausible because pumps are self-cooled. Without the installed restriction orifices RO 501 and RO 555, candidate could determine that pump flows go up because two valves are opened and one is closed on the recirc line. See OS-0010 R23 sheet 1.

B. Incorrect - Plausible because this is the reason for the normal lineup having AF59 closed. See Bases and Deviation Document for DB-OP-02000 R20 step 4.18 (page 43)

C. Incorrect - Plausible for secondary side radioactive contamination because closing AF50 isolates the flow path to the CST overflow and ultimately the environment. See OS-0010 R23 sheet 1 D. Correct - local operator opens AF50 & AF51, then closes AF59. See DB-OP-02000 R27 step 4.18 (page 34) and Bases and Deviation Document for DB-OP-02000 R20 step 4.18 (page 43)

Sys # System Category KA Statement 061 Auxiliary / Generic Knowledge of local auxiliary operator tasks during Emergency an emergency and the resultant operational Feedwater effects (AFW)

System K/A# 2.4.35 K/A Importance 3.8 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02000 R27 step 4.18 (page 34) and Bases and Deviation Document for DB-OP-02000 R20 step 4.18 (page 43)

Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.13)

Objective: OPS-SYS-213-11K

Davis-Besse 1LOT15 NRC Written Exam r1

47. The plant is at 100% power.

120V AC Panel YAR will be transferred to its alternate supply from 120V AC Panel YBR.

The governing procedure requires the following before YAR is transferred:

Diverse Scram System (DSS) Channel 1 is de-energized

Caldon Cabinet C5757E is de-energized These actions are performed in accordance with the applicable procedures.

Which of the following describes the effect on:

(1) DSS and (2) Integrated Control System (ICS) Unit Load Demand (ULD)?

A. (1) DSS will trip the Reactor if required.

(2) ICS ULD can be operated in AUTO.

B. (1) DSS will trip the Reactor if required.

(2) ICS ULD can NOT be operated in AUTO.

C. (1) DSS will NOT trip the Reactor if required.

(2) ICS ULD can be operated in AUTO.

D. (1) DSS will NOT trip the Reactor if required.

(2) ICS ULD can NOT be operated in AUTO.

Answer: C Explanation/Justification:

A. Incorrect - part 2 is correct. Part 1 is incorrect because DSS is a 2/2 coincidence energize to trip system. Plausible for examinee misconception of DSS as 1/2 de-energize to trip system.

B. Incorrect - Part 1 is incorrect because DSS is a 2/2 coincidence energize to trip system. Plausible for examinee misconception of DSS as 1/2 de-energize to trip system. Part 2 is incorrect because ICS ULD can be operated in AUTO using MFW Flow Venturis for thermal power calculation.

Plausible because ULD is placed in MANUAL prior to taking LEFM signal to bypass (DB-OP-06407 R15 step 4.20.3.c.3) or for examinee misconception that Venturis cant provide input to ULD heat balance.

C. Correct - DSS is a 2/2 coincidence energize to trip system. See DB-OP-06402 R25 CRD Operating Procedure NOTE 4.18 (page113). Leading Edge Flow Meter (LEFM) signal is bypassed in ULD when de-energizing Caldon cabinet. See DB-OP-06407 R15 NNI Operating Procedure step 4.20.3.c.4 (page 40). ICS ULD can be operated in AUTO using MFW Flow Venturis for thermal power calculation. See NOTE and step 4.20.3.c.5 D. Incorrect - Part 1 is correct. . Part 2 is incorrect because ICS ULD can be operated in AUTO using MFW Flow Venturis for thermal power calculation. Plausible because ULD is placed in MANUAL prior to taking LEFM signal to bypass (DB-OP-06407 R15 step 4.20.3.c.3) or for examinee misconception that Venturis cant provide input to ULD heat balance.

Sys # System Category KA Statement 062 AC Electrical A1 Ability to predict and/or monitor changes in parameters (to Effect on instrumentation and controls of switching Distribution prevent exceeding design limits) associated with operating the ac power supplies System distribution system controls including:

K/A# A1.03 K/A Importance 2.5 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-06402 R25 CRD Operating Procedure NOTE 4.18; DB-OP-06407 R15 NNI Operating Procedure step 4.20.3.c.4 and 4.20.3.c.5)

Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.5 / 45.5)

Objective: OPS-SYS-408-02K

Davis-Besse 1LOT15 NRC Written Exam r1

48. The plant is at 100% power.

Which of the following will cause a rise in Charger DBC1N amps?

A. Start of Main Turbine Generator Emergency Bearing Oil Pump B. Start of Low Pressure Injection Pump 1 C. Inverter YVA transfer to Alternate Source D. Inverter YVB transfer to Alternate Source Answer: A Explanation/Justification:

A. Correct -EBOP powered from DC MCC 1 so its start raises load on Charger DBC1N. See OS-0060 sheet 1 R25.

B. Incorrect - No effect on Charger DBC 1N. Plausible for inversion with HPI Pump1 which has a DC lube oil pump.

C. Incorrect - YVA normal source is DC MCC 1, so Charger DBC1N output would lower when YVA transfers to alternate, not rise. Plausible for confusion with Essential Inverters (YV1, YV2, YV3, & YV4) which are normally powered by their respective AC rectifiers (YRF1, YRF2, YRF3, &

YRF4). The Essential Inverters switch to DC power from chargers when their respective rectifiers fail or lose power. See UFSAR Rev 30 page 8.3-19 D. Incorrect - YVBA normal source is DC MCC 2, so Charger DBC1N output would not be affected. Plausible for confusion with Essential Inverters.

See OS-0060 sheet 1 R25.

Sys # System Category KA Statement 063 DC Electrical K1 Knowledge of the physical connections and/or cause-effect Battery charger and battery Distribution relationships between the DC electrical system and the following System systems:

K/A# K1.03 K/A Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

OS-0060 sheet 1 R25 Question Source: Bank # 167376 modified Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Objective: OPS-SYS-409-10K

Davis-Besse 1LOT15 NRC Written Exam r1

49. The plant is at 100% power.

Component Cooling Water (CCW) Pump 1 is operating.

A Loss of Offsite Power occurs.

Five minutes later, EDG 1 trips.

NO operator actions have been taken.

Which of the following describes the status of CCW Pump 1 and Service Water (SW) Pump 1 breakers?

(1) CCW Pump 1 breaker is _____.

(2) SW Pump 1 breaker is _____.

A. (1) open (2) open B. (1) open (2) closed C. (1) closed (2) closed D. (1) closed (2) open Answer: D Explanation/Justification:

A. Incorrect - CCW Pump 1 breaker remains closed. Bus UV does not affect CCW Pump 1 breaker. See OS-0021 sheet 1 R37 CL-2. Part 2 is correct. Plausible for misconception that Bus UV opens CCW Pump breaker.

B. Incorrect - CCW Pump 1 breaker remains closed. Bus UV does not affect CCW Pump 1 breaker. See OS-0021 sheet 1 R37 CL-2. SW Pump 1 breaker opens on Bus UV. See OS-0020 sheet 2 R51 CL-3. Plausible for candidate inversion of pump responses to Bus UV.

C. Incorrect - SW Pump 1 breaker opens on Bus UV. See OS-0020 sheet 2 R51 CL-3. Part 1 is correct. Plausible for misconception that SW Pump responds the same as CCW pump.

D. Correct - CCW and SW Pumps are the major loads that sequence on to the EDG for a non-SFAS condition. CCW Pump 1 breaker remains closed. Bus UV does not affect CCW Pump 1 breaker. See OS-0021 sheet 1 R37 CL-2. SW Pump 1 breaker opens on Bus UV. See OS-0020 sheet 2 R51 CL-3.

Sys # System Category KA Statement 064 Emergency K3 Knowledge of the effect that a loss or malfunction of the ED/G Systems controlled by automatic loader Diesel system will have on the following:

Generator (ED/G)

System K/A# K3.01 K/A Importance 3.8* Exam Level RO References provided to Candidate None Technical

References:

OS-0021 sheet 3 R12 CL-13 and sheet 1 CL-2; OS-0020 sheet 2 R51 CL-3 Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41.7 / 45.6)

Objective: OPS-SYS-405-19K

Davis-Besse 1LOT15 NRC Written Exam r1

50. The plant is at 100% power.

Emergency Diesel Generator (EDG) 1 is started and loaded for a normal surveillance test.

EDG 1 Jacket Water (JW) Thermostatic Control Valve JW112 is stuck closed.

NO operator actions are taken.

Which ONE of the following describes the effect on EDG 1?

EDG 1 __(1)__ temperature(s) rise(s) until EDG 1 stops from __(2)__.

A. (1) JW only (2) high JW temperature trip B. (1) JW and Lube Oil (2) high JW temperature trip C. (1) JW only (2) the engine seizing up D. (1) JW and Lube Oil (2) the engine seizing up Answer: B Explanation/Justification:

A. Incorrect - JW112 stuck closed prevents cooling of the JW and lube oil. See OS-0041A sheet 1 R32. Part 2 is correct. Plausible for misconception of lube oil cooling directly by CCW.

B. Correct - JW112 stuck closed prevents cooling of the JW and lube oil. See OS-0041A sheet 1 R32. High JW temperature trips EDG because EDG start was manual, not Emergency. See DB-OP-02001 R30 Window 1-1-B.

C. Incorrect - JW112 stuck closed prevents cooling of the JW and lube oil. See OS-0041A sheet 1 R32. High JW temperature trips EDG because EDG start was manual, not Emergency. See DB-OP-02001 R30 Window 1-1-B. Plausible for misconception of lube oil cooling directly by CCW and misapplication of Emergency start bypass of JW temperature trip. See DB-OP-02000 Bases and Deviation Document R20 page 455.

D. Incorrect - High JW temperature trips EDG because EDG start was manual, not Emergency. See DB-OP-02001 R30 Window 1-1-B. Part 1 is correct. Plausible for misapplication of Emergency start bypass of JW temperature trip. See DB-OP-02000 Bases and Deviation Document R20 page 455.

Sys # System Category KA Statement 064 Emergency K1 Knowledge of the physical connections and/or cause-effect D/G cooling water system Diesel relationships between the ED/G system and the following systems:

Generator (ED/G)

System K/A# K1.02 K/A Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

OS-0041A sheet 1 R32; DB-OP-02001 R30 Window 1-1-B Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Objective: OPS-SYS-406-06K

Davis-Besse 1LOT15 NRC Written Exam r1

51. A release of Clean Waste Monitor Tank (CWMT) 1 is in progress.

The release is using CWMT Transfer Pump 1.

Both Clean Liquid Waste System Discharge Monitors RE1770A and RE1770B are functional.

Which of the following describes how, if at all, that the release can be stopped from the Control Room?

A. Release can NOT be stopped from the Control Room. Local valve operation is required.

B. Push HIS1708 CWMT Transfer Pump 1 STOP button.

C. Press CLOSE on HIS1771 Clean Waste System Outlet Flow Valve.

D. Press the TEST button on RE1770A module until the HIGH alarm comes in.

Answer: D Explanation/Justification:

A. Incorrect - Plausible because all actions to stop release are local per DB-OP-03011 R23 Radioactive Liquid Batch Release. See step 4.9.25.e (page 74), 4.9.32 (page 78), and 4.9.33 (page 79). See OS-0028A sheet 4 R14 B. Incorrect - CWMT Transfer Pump control is local only. See OS-0028A sheet 4 R14. Plausible because this is how the release is normally terminated. See DB-OP-03011 R23 Radioactive Liquid Batch Release step 4.9.33 (page 79).

C. Incorrect - WC1771 control is local only. See OS-0028A sheet 4 R14. Plausible because WC1771 is closed to stop release. See DB-OP-03011 R23 Radioactive Liquid Batch Release steps 4.9.25.e (page 74) and 4.9.32 (page 78).

D. Correct -Trip check of monitor prior to release is performed in this manner. See DB-OP-03011 R23 Radioactive Liquid Batch Release step 4.9.18 (page 71).

Sys # System Category KA Statement 073 Process A4 Ability to manually operate and/or monitor in the control room: Effluent release Radiation Monitoring (PRM)

System K/A# A4.01 K/A Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-03011 R23 Radioactive Liquid Batch Release step 4.9.18 (page 71)

Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.7 / 45.5 to 45.8)

Objective: OPS-SYS-115-06K

Davis-Besse 1LOT15 NRC Written Exam r1

52. The plant is operating at 100% power.

Component Cooling Water (CCW) Pump 1 is operating.

Which of the following describes the impact of extended cold weather operation on the Service Water (SW) System and the actions required to mitigate the potential consequences?

Extended cold weather operation challenges SW Relief Valve __(1)__ lift setting.

To mitigate the potential consequences, the operators establish flow through the spare __(2)__.

A. (1) Loop 1 SW3962 (2) CCW Heat Exchanger B. (1) Loop 2 SW3963 (2) CCW Heat Exchanger C. (1) Loop 1 SW3962 (2) Containment Air Cooler D. (1) Loop 2 SW3963 (2) Containment Air Cooler Answer: A Explanation/Justification:

A. Correct - Challenges to SW relief valves With CCW Pump 1 operating, SW Loop 1 is the primary loop. Standby CCW HX is used for pressure control of primary loop. See DB-OP-06261 R63 SW System Operating Procedure step 2.2.9.

B. Incorrect - With CCW Pump 1 operating, SW Loop 2 is the secondary loop. Standby TPCW HX is used for pressure control of secondary loop.

See DB-OP-06261 R63 SW System Operating Procedure Section 3.13. Plausible because pressure control would be established for high Loop 2 pressure.

C. Incorrect - CCW HX would be used for Loop 1 pressure control. See DB-OP-06261 R63 SW System Operating Procedure step 2.2.9. Plausible because DB-OP-06016 R33 CAC System Procedure section 4.6 describes CACs being used for SW header pressure control, but only in MODE 5 or lower.

D. Incorrect - Standby TPCW HX would be used for pressure control of SW loop 2. See DB-OP-06261 R63 SW System Operating Procedure Section 3.13. Plausible because pressure control would be established for high Loop 2 pressure. CAC plausible because DB-OP-06016 R33 CAC System Procedure section 4.6 describes CACs being used for SW header pressure control, but only in MODE 5 or lower.

Sys # System Category KA Statement 076 Service A2 Ability to (a) predict the impacts of the following malfunctions or Service water header pressure Water operations on the SWS; and (b) based on those predictions, use System procedures to correct, control, or mitigate the consequences of those (SWS) malfunctions or operations:

K/A# A2.02 K/A Importance 2.7 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-06261 R63 SW System Operating Procedure step 2.2.9 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45/3 /

45/13)

Objective: OPS-SYS-305-03K

Davis-Besse 1LOT15 NRC Written Exam r1

53. The plant is operating at 100% power.

Which of the following will result in the stable readings on the gauges below?

A. Instrument Air Dryer Switching Failure B. Leak on the air header to the Atmospheric Vent Valves C. Leak on the air header to the Turbine Plant Cooling Water Heat Exchanger temperature control valves D. Leak on the process air header to the Moisture Separator Reheater Demineralizer skid resin transfer system Answer: D Explanation/Justification:

A. Incorrect - IA and SA pressure would be about equal and lower for this failure. See DB-OP-02528 R19 Attachment 24 (page 124). Plausible for misdiagnosis since both headers have low pressure. See DB-OP-02528 R22 step 2.2.2.

B. Incorrect - IA pressure would still lower. This is an IA leak which is downstream of IA450. See OS-0019A sheet 2 R19 H-20. Plausible for misconception of header supplying valves.

C. Incorrect - IA pressure would still lower. This is an IA leak which is downstream of IA72. See OS-0019A sheet 2 R19 D-22 and DB-OP-02528 R22 Attachment 17 (page 101). Plausible for misconception of header supplying valves.

D. Correct - Leak is on station air header See DB-OP-02528 R22 Instrument Air System Malfunctions step 4.1.6 and Attachment 24 Background Information page 124 2nd paragraph. MSRD skid resin transfer air is on SA header. See OS-0019B sheet 2 R21 D-30.

Sys # System Category KA Statement 078 Instrument A4 Ability to manually operate and/or monitor in the control room: Pressure gauges Air System (IAS)

K/A# A4.01 K/A Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02528 R22 step 4.1.6 and Attachment 24 (page 124); OS-0019B sheet 2 R21 Question Source: Oconee 2010 #53 modified Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 / 45.5 to 45.8)

Objective: OPS-GOP-128-01K

Davis-Besse 1LOT15 NRC Written Exam r1

54. The plant is operating at 100% power.

All Control Room Instrument Air (IA) and Station Air (SA) System pressure indicators lower to 80 psig and stabilize.

Which of the following lists the correct order of automatic actions that occurred?

(1) Station Air Compressor (SAC) 1 started (2) IA Dryer Bypass Valves IA932 and IA962 opened (3) Emergency Instrument Air Compressor (EIAC) started A. 1, 2, 3 B. 1, 3, 2 C. 3, 1, 2 D. 3, 2, 1 Answer: B Explanation/Justification:

A. Incorrect - See DB-OP-02528 R22 IA Malfunctions page 123. Plausible for misconception that EIAC is last resort action.

B. Correct - See DB-OP-02528 R22 IA Malfunctions page 123.

C. Incorrect - See DB-OP-02528 R22 IA Malfunctions page 123. Plausible for off-normal lineup of EIAC operating in LEAD status.

D. Incorrect - See DB-OP-02528 R22 IA Malfunctions page 123. Plausible for off-normal lineup of EIAC operating in LEAD status and dryers bypassed earlier to support that status.

Sys # System Category KA Statement 078 Instrument A3 Ability to monitor automatic operation of the IAS, including: Air pressure Air System (IAS)

K/A# A3.01 K/A Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02528 R22 IA Malfunctions page 123 Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.7 / 45.5)

Objective: OPS-SYS-602-08K

Davis-Besse 1LOT15 NRC Written Exam r1

55. The plant is operating at 100% power.

HIS 2022A SFAS CHANNEL 1 ACTUATE switch is located on the MANUAL ACTUATION section of Control Room Safety Features Actuation Panel C5717.

HIS 2022A arcs across its TRIP position contacts.

This causes a Channel 1 Manual SFAS Actuation

The circuit problem that caused the inadvertent Channel 1 Manual SFAS Actuation clears a few moments after the actuation.

No other adjacent switches or circuits are affected.

Which of the following describes the effect of the inadvertent Channel 1 Manual SFAS Actuation on Containment and the actions required to mitigate the potential consequences?

A. Containment air temperature lower limit is challenged. Reset SFAS and return Containment Air Cooler Fan 1 to FAST speed per DB-OP-06910 Trip Recovery.

B. Containment Equipment Qualification is challenged. Block SFAS and stop Containment Spray Pump 1 per DB-OP-02000 Attachment 9 Miscellaneous Post Accident Actions.

C. The Containment Vessel Negative Pressure Analysis is challenged. Reset SFAS and reopen the Containment Vacuum Relief Isolation Valves per DB-OP-06910 Trip Recovery.

D. The Containment Fire Hazards Analysis is challenged. Block SFAS and reopen the Component Cooling Water (CCW) Containment Isolation Valves per DB-OP-02000 Attachment 9 Miscellaneous Post Accident Actions.

Answer: C Explanation/Justification:

A. Incorrect - CAC is less efficient in SLOW speed, so upper temperature limit is challenged. See DB-OP-02525 R10 Steam Leaks page 59 third paragraph. Plausible because recovery actions are correct (see DB-OP-06910 steps 6.3.2 and 6.3.3.e) and for misconception of CACs being more efficient in SLOW (hence the discussion in DB-OP-02525 as to why manual shift to SLOW is no longer performed for a steam leak).

B. Incorrect - Manual Containment Spray actuation is a separate circuit. See DB-OP-06405 R13 SFAS Operating Procedure step 5.2.2.b (page 65).

Plausible because Manual SFAS Actuation actuates all of the Level 4 Containment Isolation functions, but does not start the Spray Pump. See DB-OP-06405 R13 SFAS Operating Procedure step 5.2.2.a (page 65). Stop of a Containment Spray Pump that was inadvertently started could be performed per DB-OP-02000 Attachment 9.

C. Correct - SFAS Manual Actuation Channel initiates SA Levels 1-4 of Containment Isolation, but does NOT start the Containment Spray Pump.

See DB-OP-06405 R13 SFAS Operating Procedure step 5.2.2.a (page 65). Channel 1 SA Level 2 isolates 5 of the 10 Containment Vacuum Breakers. See DB-OP-02000 R27 page 418. At least 6 vacuum breakers are required per the Inadvertent Containment Spray analysis. See UFSAR R30 page 6.2-14. Trip Recovery provides guidance for SFAS reset and reopening the Vacuum breaker isolation valves. See DB-OP-06910 R27 steps 6.3.2 (page 47) and 6.3.3.g (page 49). Attachment 9 does NOT provide guidance for reopening the vacuum breaker isolations.

See DB-OP-02000 R27 page 322.

D. Incorrect - Major fire hazard in Containment is the RCP lube oil system per FHAR R26 9.1.3.1 (page 9-12). Operators stop all RCPs when CCW and seal injection are both lost per DB-OP-02515 R12 RCP and Motor Operation step 4.4.1, which mitigates the challenge. Plausible because CCW supplies cooling for the RCP oil coolers. CCW valves could be reopened per Attachment 9 step 2.6.

Sys # System Category KA Statement 103 Containment A2 Ability to (a) predict the impacts of the following malfunctions or Phase A and B isolation System operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

K/A# A2.03 K/A Importance 3.5* Exam Level RO References provided to Candidate None Technical

References:

DB-OP-06405 R13 SFAS Operating Procedure step 5.2.2.a (page 65), DB-OP-02000 R27 page 418, UFSAR R30 page 6.2-14, DB-OP-06910 R27 steps 6.3.2 (page 47) and 6.3.3.g (page 49)

Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5/43.5/45.3/45.13)

Objective: OPS-SYS-108-02K

Davis-Besse 1LOT15 NRC Written Exam r1

56. The following plant conditions exist:

The plant is in Mode 3.

Control Rod Drive (CRD) Trip Breakers C and D have been closed locally.

Which condition listed below will prevent the closing of CRD Trip Breakers A and B when the TRIP RESET pushbutton on the Rod Control Panel is depressed?

A. Annunciator 5-1-G, RPS CH 1 TRIP, is lit.

B. The operating CRD Booster Pump flow is 116 gpm.

C. The ELECTRONIC TRIP D light on the Rod Control Panel is lit.

D. The APSR GROUP IN LIMIT light on the Rod Control Panel is off.

Answer: B Explanation/Justification:

A. Incorrect - 2 channels of RPS tripped would be required to prevent breaker closure. See DB-OP-02005 R18 Window 5-1-G NOTE 2.1. Plausible because CRD Operating Procedure resets all RPS trips prior to breaker closure. See DB-OP-06402 R25 step 3.7.5.

B. Correct - Minimum CRD flow for breaker closure is 146 gpm. See DB-OP-06402 R25 CRD Operating Procedure step 3.7.7.

C. Incorrect - this light is expected to be lit and goes off when the Trip Reset button is pressed. See DB-OP-06402 R25 NOTE 3.7.15.c. and step 3.7.15.e. Plausible for misconception on Source Interruption Device actuated by UV on one Power Supply Train, not two. See DB-OP-06402 R25 NOTE 3.7.16.a. One of the other items that will light the ELECTRONIC TRIP D light is CRD Trip Breaker D open. See DB-OP-06402 R25 page 150.

D. Incorrect - APSR (Group 8) IN LIMIT is not required to close a CRD breaker. Plausible because Group 1-7 IN LIMITS are required. See DB-OP-06402 R25 NOTE 3.7.11-14 and step 3.7.14.

Sys # System Category KA Statement 001 Control Rod K4 Knowledge of CRDS design feature(s) and/or interlock(s) which Resetting of CRDM circuit breakers Drive provide for the following:

K/A# K4.11 K/A Importance 2.7 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-06402 R25 CRD Operating Procedure step 3.7.7 Question Source: Bank - #167286 Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41.7)

Objective: OPS-SYS-501-07K

Davis-Besse 1LOT15 NRC Written Exam r1

57. The Data Acquisition and Display System/Safety Parameter Display System (DADS/SPDS) computer is being used to perform a Reactor Coolant System Water Inventory Balance calculation.

DADS/SPDS automatic data retrieval is being utilized.

NO computer points are out of service.

Which of the following describes the requirements for using the computer for this calculation?

A. NO parameter values must be manually entered. There are NO restrictions on maintaining steady state plant conditions after the final values are recorded.

B. Reactor Coolant Pump Seal Leakage Indicator values must be manually entered. Steady state plant conditions must be maintained for at least 15 minutes after the final values are recorded.

C. Reactor Coolant Pump Seal Leakage Indicator values must be manually entered. There are NO restrictions on maintaining steady state plant conditions after the final values are recorded.

D. Reactor Coolant Pump Seal Leakage Indicator and Quench Tank parameter values must be manually entered. Steady state plant conditions must be maintained for at least 15 minutes after the final values are recorded.

Answer: B Explanation/Justification:

A. Incorrect - Plausible because this is how an automated data retrieval calculation would work if RCP Seal Leakage totalizers had computer points.

B. Correct - manual entry of RCP Seal Leakage readings required per DB-SP-03357 R19 RCS Water Inventory Balance step 4.1.12.b. 15 minute wait period required per step 4.1.8.

C. Incorrect - Plausible for misconception that final data time = end of calculation.

D. Incorrect - Per step 4.1.12, Quench Tank level is NOT manually entered for computer calculation. Plausible because RCP Seal Leakage Indicator values are recorded on an attachment entitled Attachment 1 RCP Seals Leak Rate and Quench Tank In-Leakage Calculation Sheet. 15 minute wait period is correct.

Sys # System Category KA Statement 002 Reactor A4 Ability to manually operate and/or monitor in the control room: RCS leakage calculation program using the Coolant computer System (RCS)

K/A# A4.01 K/A Importance 3.5* Exam Level RO References provided to Candidate None Technical

References:

DB-SP-03357 R19 RCS Water Inventory Balance steps 4.1.12.b and 4.1.8.

Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41.7 / 45.5 to 45.8)

Objective: OPS-SYS-525-01S

Davis-Besse 1LOT15 NRC Written Exam r1

58. The plant is operating at 100% power.

Component Cooling Water (CCW) Pump 1 is operating.

A Loss of Offsite Power occurs.

NO operator actions have been taken.

Which of the following additional malfunctions, if any, will cause ZERO Makeup Pumps to be operating one minute after the Loss of Offsite Power?

A. Bus C1 locks out.

B. Containment Pressure rises to 18.0 psia.

C. Emergency Diesel Generator 2 does NOT start.

D. None. Zero Makeup Pumps will be running with NO additional malfunction.

Answer: C Explanation/Justification:

A. Incorrect. Since MU Pump 2 was previously running and is not affected by C1 lockout, it restarts. Plausible because bus lockout trips and locks out its associated MU Pump. See OS-00002 sheet 3 R33 CL-10.

B. Incorrect. Plausible for Containment pressure above 18.4 psia which would cause SFAS Level 3 start of LPI Pump 2 which would trip MU Pump 2 after auto-restart. See OS-00002 sheet 3 R33 CL-10.

C. Correct - MU Pump 2 was running prior to the LOP per normal alignment. Previously running MU Pump load sheds on bus UV, then restarts 2.5 seconds after its associated EDG breaker closes. Since EDG doesnt start, zero MU Pumps will be running. See OS-0002 sheet 4 R24 CL-15.

D. Incorrect. Plausible for prior design which had no auto-start feature for MU Pumps.

Sys # System Category KA Statement 011 Pressurizer K2 Knowledge of bus power supplies to the following: Charging pumps Level Control System K/A# K2.01 K/A Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

OS-0002 sheet 4 R24 CL-15 Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.7)

Objective: OPS-SYS-106-14K

Davis-Besse 1LOT15 NRC Written Exam r1

59. The plant is at 100% power.

Power Range Nuclear Instrument (NI) 5 Power indicates 100%

NI 5 Imbalance indicates -10%.

The power supply cable to the upper detector of Power Range NI 5 becomes disconnected.

Which of the following describes the effect of the loss of NI 5 upper detector power supply?

(1) NI 5 Power indicates _____.

(2) NI 5 Imbalance indicates _____.

A. (1) 55%

(2) -55%

B. (1) 55%

(2) +45%

C. (1) 45%

(2) -55%

D. (1) 45%

(2) +45%

Answer: A Explanation/Justification:

A. Correct - Imbalance = Power upper - Power lower, so prior to the failure, Power upper = 45% and Power lower = 55%. When the upper detector power supply cable becomes disconnected, Power upper = 0, so total power = 55% and imbalance = -55%.

B. Incorrect - Imbalance = -55%. Part 1 is correct. Plausible for inversion of Imbalance relationship (Lower - Upper).

C. Incorrect - Power = 55%. Part 2 is correct. Plausible for inversion of power values.

D. Incorrect - both parts wrong. Plausible for inversion of Imbalance relationship (Lower - Upper).

Sys # System Category KA Statement 015 Nuclear K6 Knowledge of the effect of a loss or malfunction on the following Component interconnections Instrumentati will have on the NIS:

on System K/A# K6.03 K/A Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

UFSAR R30 pages 7.2-2 and 7.8-2 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 / 45.7)

Objective: OPS-SYS-502-03K

Davis-Besse 1LOT15 NRC Written Exam r1

60. The plant is at 100% power.

The Reactor Coolant System (RCS) Loop 1 Flow signal shorts to ground in the Non-Nuclear Instrumentation (NNI) cabinet.

Which of the following describes the effect, if any, on plant protection systems?

A. No effect on plant protection systems.

B. One Reactor Protection System (RPS) channel trips.

C. One Steam Feed Rupture Control System (SFRCS) channel trips.

D. One RPS channel and one SFRCS channel trip.

Answer: A Explanation/Justification:

A. Correct - RPS provides RCS Flow signal to NNI. See DB-OP-06403 R20 RPS and NI Operating Procedure step 4.3.3. RPS isolation amplifier prevents fault from feeding back into protection system, so RPS does not trip on Flux/Flux/Flow. See UFSAR R30 section 7.1.2.3.

B. Incorrect - RPS isolation amplifier prevents fault from feeding back into protection system. See UFSAR R30 section 7.1.2.3. Plausible because RPS provides RCS Flow signal to NNI and channel would trip on Flux/Flux/Flow if isolation amplifier didnt prevent fault from feeding back into RPS cabinet. See DB-OP-06403 R20 step 4.3.3.

C. Incorrect - SFRCS monitors RCP motor current for pump status input, so failure doesnt affect SFRCS. See UFSAR R30 7.4.1.3.10.4 (page 7.4-8). Plausible for misconception that SFRCS monitors flow for RCP status.

D. Incorrect - RPS isolation amplifier prevents fault from feeding back into protection system, so RPS does not trip on Flux/Flux/Flow. SFRCS monitors RCP motor current for pump status input, so failure doesnt affect SFRCS. Plausible because RPS provides RCS Flow signal to NNI and RPS channel would trip on Flux/Flux/Flow if isolation amplifier didnt prevent fault from feeding back into RPS cabinet. Plausible for misconception that SFRCS monitors flow for RCP status.

Sys # System Category KA Statement 016 Non-nuclear K5 Knowledge of the operational implication of the following Separation of control and protection circuits Instrumentati concepts as they apply to the NNIS:

on K/A# K5.01 K/A Importance 2.7* Exam Level RO References provided to Candidate None Technical

References:

UFSAR R30 section 7.1.2.3 Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.5 / 45.7)

Objective: OPS-SYS-507-12K

Davis-Besse 1LOT15 NRC Written Exam r1

61. The plant is operating at 100% power.

The operating Spent Fuel Pool (SFP) Pump 1 trips and can NOT be restarted.

Which ONE of the following describes the preferred order of the listed options for SFP cooling?

(1) Use SFP Pump 2 (2) Use Decay Heat Train 2 (3) Makeup from Demineralized Water A. 1, 2, 3 B. 1, 3, 2 C. 3, 1, 2 D. 3, 2, 1 Answer: A Explanation/Justification:

A. Correct - See DB-OP-02547 R4 SFP Cooling Malfunctions step 4.1.12 and DB-OP-06021 R26 SFP Operating Procedure section 3.19 B. Incorrect - Demin water 3rd option per DB-OP-02547 R4 SFP Cooling Malfunctions step 4.1.12 RNO. Plausible because Demin Water makeup is normal source to make up for evaporation.

C. Incorrect - Plausible because Demin Water makeup is normal source to make up for evaporation.

D. Incorrect - Plausible because Demin Water makeup is a normal evolution while DHR requires entry into an LCO.

Sys # System Category KA Statement 033 Spent Fuel Generic Knowledge of abnormal condition procedures Pool Cooling K/A# 2.4.11 K/A Importance 4.0 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02547 R4 SFP Cooling Malfunctions step 4.1.12, DB-OP-06021 page 2 Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.13)

Objective: OPS-GOP-147-01K

Davis-Besse 1LOT15 NRC Written Exam r1

62. The plant is in MODE 3 following a transient.

Reactor Coolant System (RCS) Tave is 545 ºF

All Reactor Coolant Pumps (RCPs) are operating.

The operators will perform an RCS cool down to 532 ºF using the Turbine Bypass Valves.

Which of the following Steam Generator (SG) pressure changes produces an RCS cool down to 532 ºF at the maximum allowable rate?

A. Lower SG pressure by 50 psia over 4 minutes.

B. Lower SG pressure by 50 psia over 8 minutes.

C. Lower SG pressure by 100 psia over 8 minutes.

D. Lower SG pressure by 100 psia over 16 minutes.

Answer: C Explanation/Justification:

A. Incorrect - 50 psi lowering only cools down to about 538 ºF. Plausible because rate is correct.

B. Incorrect - 50 psi lowering only cools down to about 538 ºF. Plausible because rate would be correct for heatup or natural circulation cooldown.

See DB-OP-06903 R47 Plant Cooldown step 6.2 (page 80).

C. Correct - Maximum cooldown rate for forced circulation is 100 ºF/hr or 1.67 ºF/min. See DB-OP-06910 Trip Recovery R27 step 2.2.1.a. SG pressure from 1000 psia to 900 psia equals cooldown from 545 ºF to 532 ºF. 13 ºF ÷ 1.67 ºF/min = 8 min.

D. Incorrect - Cooldown rate is only 50 ºF/hr. Plausible because final temperature is correct and rate would be correct for heatup natural circulation cooldown.

Sys # System Category KA Statement 035 Steam A1 Ability to predict and/or monitor changes in parameters (to S/G pressure Generator prevent exceeding design limits) associated with operating the S/GS controls including:

K/A# A1.02 K/A Importance 3.5 Exam Level RO References provided to Candidate Steam Tables Technical

References:

DB-OP-06910 Trip Recovery R27 step 2.2.1.a Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 / 45.5)

Objective: OPS-SYS-201-08K

Davis-Besse 1LOT15 NRC Written Exam r1

63. The plant is at 8% power following a mid-cycle outage.

The Rod Control Panel is in AUTO.

The Reactor Demand Station is in HAND.

Main Steam Isolation Valve (MSIV) MS101 10% Closed limit switch spuriously actuates.

MS101 remains full open.

Without operator actions, what effect, if any, will this have on Reactor Power and Reactor Coolant System (RCS) Tave?

A. Reactor Power lowers. RCS Tave rises.

B. No effect on Reactor power. RCS Tave rises.

C. Reactor Power lowers. No effect on RCS Tave.

D. No effect on Reactor Power. No effect on RCS Tave.

Answer: D Explanation/Justification:

A. Incorrect - Reactor Demand Station in HAND with Rod Control Panel in AUTO maintains Reactor power constant. Tave doesnt change.

Plausible for misapplication of AVV steam flow of 5% per valve and the natural reactivity feedback that would make power lower if not for the status of rod control.

B. Incorrect - Tave doesnt change. Part 1 is correct. Plausible for misapplication of AVV steam flow of 5% per valve.

C. Incorrect - Reactor Demand Station in HAND with Rod Control Panel in AUTO maintains Reactor power constant. Part 2 is correct.

D. Correct - When in HAND, Reactor Demand maintains constant neutron power, so Reactor power is not affected. See DB-OP-06401 R23 ICS Operating Procedure page 94 and M-533-00179 R4. When the MSIV 10% Closed switch actuates, the Turbine Bypass Valves (TBVs) close and the Steam Generator (SG) pressure control signals are transferred to the Atmospheric Vent Valves (AVVs). See DB-OP-06401 R23 page 106.

AVVs can pass 10% steam flow so Tave doesnt change. See UFSAR 10.4.4.3 (page 10.4.7).

Sys # System Category KA Statement 041 Steam K3 Knowledge of the effect that a loss or malfunction of the SDS will RCS Dump/Turbin have on the following:

e Bypass Control K/A# K3.02 K/A Importance 3.8 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-06401 Att. 2, step 1, M-533-00179, DB-OP-06401 R23 ICS Operating Procedure page 106, UFSAR 10.4.4.3 (page 10.4.7)

Question Source: Bank - #172527 2008 NRC modified Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 / 45.6)

Objective: OPS-SYS-201-08K

Davis-Besse 1LOT15 NRC Written Exam r1

64. The plant is operating at 100% power.

The Main Turbine trips.

Which of the following describes the response of the Main Generator?

(1) The Main Generator Output Breakers ACB34560 and ACB34561 trip __(1)__.

(2) The Generator Field Breaker __(2)__.

A. (1) after a timer times out (2) trips when the ACBs trip B. (1) after a timer times out (2) stays closed C. (1) immediately (2) stays closed D. (1) immediately (2) trips when the ACBs trip Answer: A Explanation/Justification:

A. Correct -See DB-OP-02000 R27 step 2.1.5.b and DB-OP-02016 R25 Window 16-6-C. Also System Description SD-005 R4 Main Generator and Auxiliaries pages 2-27 and 2-28.

B. Incorrect - Field breaker opens when ACBs open. Part 1 is correct. Plausible for misinterpretation of DB-OP-02500 Turbine Trip Attachment 2 which implies manual opening of field breaker is required.

C. Incorrect - ACBs open based on timer and field breaker opens at the same time. Part 1 plausible for misdiagnosis as generator trip. Part 2 field breaker staying closed is plausible for misinterpretation of DB-OP-02500 Turbine Trip Attachment 2 which implies manual opening required.

D. Incorrect - ACBs open based on timer. Part 2 is correct. 16-1-C actuated by 81U2 and 81U1 which also actuate generator lockout. See DB-OP-02016 R Window 16-1-C and OS-0055 sheet 2 R38 CD-1. Plausible for misdiagnosis as generator trip.

Sys # System Category KA Statement 045 Main Turbine A3 Ability to monitor automatic operation of the MT/G system, Generator trip Generator including:

K/A# A3.11 K/A Importance 2.6* Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02000 R27 step 2.1.5.b and DB-OP-02016 R25 Window 16-6-C Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41/7 / 45.5)

Objective: OPS-SYS-401-02K

Davis-Besse 1LOT15 NRC Written Exam r1

65. The plant is operating at 100% power.

A Loss of Offsite Power occurs.

The following annunciators alarm:

1-3-D BUS C1 LOCKOUT

1-3-H BUS D1 LOCKOUT

10-5-G AFP 1 SUCT PRESS LO

10-5-H AFP 2 SUCT PRESS LO

13-1-B CNDS STRG TK LVL Condensate Storage Tank 1 Level LI 512 indicates ZERO feet.

Condensate Storage Tank 2 Level LI 516 indicates ZERO feet.

Which of the following describes a method, if any, to supply feedwater to the Steam Generators (SGs) under these conditions that does NOT require the installation of temporary piping or hoses?

A. Startup Feed Pump from the Deaerator Storage Tanks B. Auxiliary Feedwater Pump 1 from the Diesel Fire Pump C. Motor Driven Feedwater Pump from the Backup Service Water Pump D. NO feedwater is available for the SGs Answer: B Explanation/Justification:

A. Incorrect - The only available onsite AC power source is the Station Blackout Diesel Generator (SBODG) which powers Bus D2. SUFP is powered from Bus C2 which has no power and cant be aligned to Bus D2 due to the lockouts of C1 and D1 Buses. Plausible because this would be a feedwater source if C2 power could be restored. See DB-OP-02000 R27 Attachment 5 Section C (page 287) and DB-OP-06226 R15 Startup Feed Pump Operating Procedure NOTE 5.1 (page 14).

B. Correct - See DB-OP-02600 R13 Operational Contingency Response Action Plan Attachment 12 AFW Emergency Fire Protection Water to AFW Pump Suction (page 73). Diesel Fire Pump is available to provide suction head for AFW Pump operation.

C. Incorrect - Bus C1 lockout has stopped Service Water Pump 1 which is the emergency backup suction supply for the MDFP. See OS-0012A sheet 1 R26. BUSW Pump cant be used in place of SW Pump 1 because Bus C2 cant be powered form the SBODG due to the C1 and D1 lockouts. Plausible because the MDFP can be powered from the SBODG.

D. Incorrect - Plausible for misconception that Diesel Fire Pump is not available. See OS-0047A sheet 1 R25.

Sys # System Category KA Statement 086 Fire K1 Knowledge of the physical connections and/or cause-effect AFW system Protection relationships between the Fire Protection System and the following System systems:

(FPS)

K/A# K1.03 K/A Importance 3.4* Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02600 R13 Operational Contingency Response Action Plan Attachment 12 AFW Emergency Fire Protection Water to AFW Pump Suction (page 73)

Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Objective: OPS-SYS-601-02K

Davis-Besse 1LOT15 NRC Written Exam r1

66. The plant is in MODE 6 with Fuel Movement in progress.

Refueling Canal Water Level LI 1627 is stable at 23.5 ft.

No water additions or drain operations are planned or will be allowed this shift.

Decay Heat (DH) Loop 1 has been operating continuously for the past 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to provide core cooling.

The Fuel Handling Director requests DH Pump 1 be stopped to aid in Fuel Movement.

In accordance with Technical Specifications, what is the maximum time DH Pump 1 may be stopped, if at all?

A. DH Pump 1 may NOT be stopped.

B. 15 minutes C. 30 minutes D. 60 minutes Answer: D Explanation/Justification:

A. Incorrect - maximum allowable 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months stopped per eight hour period per LCO 3.9.4 NOTE. Plausible because it would be correct without NOTE.

B. Incorrect - Plausible because this is the allowable stopped time for LCO 3.9.5 which would apply for LI 1627 at 22.5 feet.

C. Incorrect - Plausible for multiple of LCO 3.9.5 allowable stop time.

D. Correct - 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months stopped per eight hour period allowed per DB-NE-06101 R25 Fuel/Control Component Shuffle step 2.2.2 and LCO 3.9.4.

Sys # System Category KA Statement N/A N/A Generic Knowledge of the Refueling process K/A# 2.1.41 K/A Importance 2.8 Exam Level RO References provided to Candidate None Technical

References:

LCO 3.9.4 Question Source: Bank - Oconee 2010 #95 Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.2 / 41.10 / 43.6/

45.13)

Objective: OPS-GOP-439-01K

Davis-Besse 1LOT15 NRC Written Exam r1

67. The plant is at 100% power.

Which of the following describes the system(s)/component(s) used to control dissolved Oxygen?

__(1)__ is used to control Dissolved Oxygen in the __(2)__.

A. (1) Hydrogen Cylinders and Makeup Tank (2) Pressurizer B. (1) Hydrogen Cylinders and Makeup Tank (2) Reactor Coolant System C. (1) Chemical Addition System and Makeup and Purification System (2) Reactor Coolant System D. (1) Chemical Addition System and Makeup and Purification System (2) Pressurizer Answer: B Explanation/Justification:

A. Incorrect - PZR Oxygen control by venting to RC Drain Tank. See DB-OP-06033 R13 Hydrogen Addition and Degasification Section 3.3.

Plausible because PZR is connected to RCS and part 1 is correct.

B. Correct - Lower RCS Oxygen by adding Hydrogen to MU Tank. See UFSAR R30 pages 9.3-13 and 9.3-17.

C. Incorrect - Lower RCS Oxygen by adding Hydrogen to MU Tank per DB-OP-06033 R13 Hydrogen Addition and Degasification Section 3.1.

Plausible because Chemical Addition System is used to add Hydrazine to RCS at lower temperatures. See UFSAR 9.3.6.2.2 page 9.3-30 D. Incorrect - PZR Oxygen control by venting to RC Drain Tank. See DB-OP-06033 R13 Hydrogen Addition and Degasification Section 3.3.

Plausible because hydrazine is added to PZR for Oxygen control prior to drawing a bubble. See DB-OP-06900 R61 Plant Heatup step 3.50.

Sys # System Category KA Statement N/A N/A Generic Knowledge of system purpose and/or function K/A# 2.1.27 K/A Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

UFSAR R30 pages 9.3-13 and 9.3-17 Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.7)

Objective: OPS-SYS-106-01K

Davis-Besse 1LOT15 NRC Written Exam r1

68. During a plant shutdown, Heater Drain Tank 1 is placed on recirculation per DB-OP-06227 Low Pressure Feedwater Heaters.

Which ONE of the following valves on DB-OP-06227 Attachment 5 Page 1 of 2 is in its normal full power position when the attachment is completed?

References provided A. HD 5 B. HD 49 C. HD 27 D. HD 35 Answer: B Explanation/Justification:

A. Incorrect - Heater Drain Pumps are operating at full power with HD 5 open. Attachment 5 has HD 5 closed. Plausible because HD 5 is closed when Heater Drain Pump 1 is stopped. See DB-OP-06227 step 3.5.4.

B. Correct - See OS-0013 sheet 1 R15. Operations Schematics show 100% Power lineups C. Incorrect - HD 27 is open at full power. Attachment 5 has HD 27 closed. Plausible for misconception of valve name drain vs process line.

D. Incorrect - HD 35 is open at full power. Attachment 5 has HD 27 closed. Plausible because HD 35 is closed during loop seal restoration. See DB-OP-06227 step 4.5.5.

Sys # System Category KA Statement N/A N/A Generic Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.

K/A# 2.2.15 K/A Importance 3.9 Exam Level RO References provided to Candidate DB-OP-06227 Attachment Technical

References:

DB-OP-06227 R14 Attachment 5 Page 1 of 2 and 5 Page 1 of 2 and OS-0013 OS-0013 sheet 1 R15.

sheet 1.

Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.10 / 43.3 / 45.13)

Objective: OPS-GOP-505-02K

Davis-Besse 1LOT15 NRC Written Exam r1

69. The plant is at 100% power.

Emergency Diesel Generator 1 is out of service for a planned 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months maintenance outage.

Which of the following additional equipment problems, if not corrected, would require action to be taken within one hour?

A. Emergency Diesel Generator 2 is discovered to be INOPERABLE B. Containment Spray Pump 2 automatic start circuit is discovered to be INOPERABLE C. Control Room Emergency Vent Fan 2 is discovered to be INOPERABLE D. Station Emergency Ventilation System Fan 2 is discovered to be INOPERABLE Answer: A Explanation/Justification:

A. Correct - 3.8.1 Condition B applies when EDG 2 becomes inoperable which requires breaker alignment surveillance within one hour.

B. Incorrect - LCO 3.8.1 Action B.2 gives 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from redundant feature inoperable to declare supported feature (Spray Pump 1) inoperable.

Plausible because both trains of containment spray will become inoperable.

C. Incorrect - LCO 3.8.1 Action B.2 gives 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from redundant feature inoperable to declare supported feature (CREV Fan 1) inoperable.

Plausible because inoperability of CRE requires immediate suspension of fuel movement regardless of power supply status; however, CRE operability is not affected by fan status.

D. Incorrect - LCO 3.8.1 Action B.2 gives 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from redundant feature inoperable to declare supported feature (SFP EVS) inoperable. Plausible because both trains of SFP EVS inoperable requires immediate suspension of fuel movement in the SFP.

Sys # System Category KA Statement N/A N/A Generic Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations K/A# 2.2.36 K/A Importance 3.1 Exam Level References provided to Candidate None Technical

References:

LCO 3.8.1 Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (41.7 / 41.10 / 43.2 / 45.13)

Objective: OPS-GOP-438-02A

Davis-Besse 1LOT15 NRC Written Exam r1

70. The plant is at 5% power during a startup.

13.8 kV Bus A locks out.

Which of the following describes required operator action?

A. Initiate Reactor shutdown due to the delay in plant startup.

B. Adjust Turbine Bypass Valve controls following the AUTOMATIC Reactor trip.

C. Adjust Turbine Bypass Valve controls in response to the lowered Reactor Coolant flow.

D. Initiate Reactor shutdown because less than three Reactor Coolant Pumps are operating.

Answer: D Explanation/Justification:

A. Incorrect - wrong reason. Plausible because this would be the action and reason for a different delay if the reactor was still subcritical. See DB-OP-06912 R17 Approach to Criticality step 4.23.

B. Incorrect - A Bus lockout trips RCPs 1-2 and 2-1. This leaves one RCP running in each loop, a combination that does not result in an automatic flux to pumps RPS trip. See Tech Spec Table 3.3.1-1. No automatic Flux-Delta Flux Flow trip. Trip setpoint lowers to around 60% power as RCS flow lowers when the RCPs stop. Power Range channels are all at zero, so no trip occurs. Plausible because automatic trip would occur at higher power level or for misconception of RCP/loop/13.8 kV bus relationship. TBVs would close when the 115 psi bias was applied by the reactor trip. DB-OP-06401 R23 ICS Operating Procedure Attachment 9. TBVs would have to be placed in HAND or the header pressure setpoint lowered to maintain Tave constant. See DB-OP-02000 Attachment 2 step 2 RNO.

C. Incorrect - No TBV adjustment is required. Header pressure setpoint does not change and TBVs are controlling SG pressures in auto. RCS T is < 1 ºF at 5% power, so Tavg change due to loss of 25% of RCS flow is negligible. Plausible for TBVs in HAND.

D. Correct - A Bus lockout trips RCPs 1-2 and 2-1. This leaves one RCP running in each loop, a combination that does not result in an automatic RPS trip. See Tech Spec Table 3.3.1-1. Operating License 2.C(3)(a) states FENOC shall not operate the reactor in MODES 1 and 2 with < 3 RCPs in operation. Inserting rods places the unit in MODE 3 where the license condition does not apply.

Sys # System Category KA Statement N/A N/A Generic Knowledge of conditions and limitations in the facility license K/A# 2.2.38 K/A Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

Operating License 2.C(3)(a)

Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 / 41.10 / 43.1 /

45.13)

Objective: OPS-GOP-115-06K

Davis-Besse 1LOT15 NRC Written Exam r1

71. Today is February 2. An operator has accumulated 200 mRem Total Effective Dose Equivalent (TEDE) radiation exposure so far this year, all of it at Davis-Besse.

Assuming the operator is permitted to continue work until BOTH limits are reached, which of the following describes the cumulative time the operator could perform normal work in a 50 mRem/hr radiation field before the:

(1) Site Administrative Control Limit dose is reached?

(2) 10CFR20 dose limit is reached?

A. (1) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (2) 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months B. (1) 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months (2) 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months C. (1) 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months (2) 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months D. (1) 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months (2) 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months Answer: C Explanation/Justification:

A. Incorrect. Plausible for quarterly limitations on Site ACL and on 10CFR20 annual limit values (like in the old days).

B. Incorrect - Plausible for quarterly limitation on 10CFR20 annual limit value (like in the old days). 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months is correct for Site ACL.

C. Correct - Site ACL is 1000 mRem/yr. 800 Mrem remaining ÷ 50 mRem/hr = 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months . See NOP-OP-4201 R2 Routine External Exposure Monitoring NOTE 6.5.1 (page14). 10CFR20 dose limit is 5.0 Rem/yr. 4800 MRem remaining ÷ 50 mRem/hr = 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months . See 10CFR20 D. Incorrect - Plausible for Site ACL = 10CFR20 limit. 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months is correct 10CFR20 limit.

Sys # System Category KA Statement N/A N/A Generic Knowledge of radiation exposure limits under normal or emergency conditions K/A# 2.3.4 K/A Importance 3.2 Exam Level RO References provided to Candidate None Technical

References:

NOP-OP-4201 R2 Routine External Exposure Monitoring NOTE 6.5.1 (page14); 10CFR20 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.12 / 43.4 / 45.10)

Objective: OPS-GOP-511-02K

Davis-Besse 1LOT15 NRC Written Exam r1

72. Which of the following are actions an operator is REQUIRED to PERFORM prior to EACH USE of a portable radiation survey instrument per DBBP-RP-1007 Meter Source and Response Testing?

(1) Perform an instrument _____.

(2) Make an entry in the _____.

A. (1) Calibration (2) Use/Response Log B. (1) Calibration (2) Daily Source Check Log C. (1) Response Check (2) Use/Response Log D. (1) Response Check (2) Daily Source Check Log Answer: C Explanation/Justification:

A. Incorrect - Calibration not performed by operator, just checked current per sticker. See DBBP-RP-1007 R32 Meter Source and Response Testing step 3.2.1.1. Part 2 is correct. Plausible because instrument calibration must be current.

B. Incorrect - Calibration not performed by operator, just checked current per sticker. Plausible because daily source check log entry is required for daily source check.

C. Correct - Response check required per DBBP-RP-1007 R32 Meter Source and Response Testing step 3.2.2.1. Use/Response Log entry required per step 3.2.2.1.H D. Incorrect - Source Check Log entry not made because operator does not perform source check. Part 1 is correct.

Sys # System Category KA Statement N/A N/A Generic Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

K/A# 2.3.5 K/A Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

DBBP-RP-1007 R32 Meter Source and Response Testing steps 3.2.2. and 3.2.2.1.H Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.11 / 41.12 / 43.4 /

45.9)

Objective: OPS-GOP-511-02K

Davis-Besse 1LOT15 NRC Written Exam r1

73. The plant is at 75% power.

An operator will be making a Containment entry to manually throttle RC2 Pressurizer Spray valve per the governing abnormal procedure.

Which of the following describes requirements for this entry per DB-OP-01101 Containment Entry?

Continuous Radiation Protection coverage __(1)__ required.

The Containment Elevator __(2)__ be used during this entry.

A. (1) is (2) should NOT B. (1) is (2) should C. (1) is NOT (2) should NOT D. (1) is NOT (2) should Answer: A Explanation/Justification:

A. Correct - See NOP-OP-4104 R6 Job Coverage step 4.4.1 and DB-OP-01101 R13 Containment Entry CAUTION 5.2.9.

B. Incorrect - Containment Elevator NOT used for personnel use during power entries. See DB-OP-01101 R13 Containment Entry CAUTION 5.2.9.

Part 1 is correct. Plausible because elevator is operational for Containment entries during shutdown.

C. Incorrect - Continuous RP coverage is required. See DB-OP-01101 R13 Containment Entry step 6.1.2. Part 2 is correct. Plausible because continuous RP coverage is not required for Containment entries during shutdown.

D. Incorrect - Containment Elevator NOT used for personnel use during power entries. See DB-OP-01101 R13 Containment Entry CAUTION 5.2.9.

Continuous RP coverage is required. See DB-OP-01101 R13 Containment Entry step 6.1.2. Plausible because elevator is operational and continuous RP coverage is not required for Containment entries during shutdown.

Sys # System Category KA Statement N/A N/A Generic Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

K/A# 2.3.12 K/A Importance 3.2 Exam Level RO References provided to Candidate None Technical

References:

NOP-OP-4104 R6 Job Coverage step 4.4.1 and DB-OP-01101 R13 Containment Entry CAUTION 5.2.9.

Question Source: New Question Cognitive Level: Low - Recall 10 CFR Part 55 Content: (CFR: 41.12 / 45.9 / 45.10)

Objective: OPS-GOP-511-03K

Davis-Besse 1LOT15 NRC Written Exam r1

74. The plant is at 100% power.

Which of the following REQUIRES FES Unit System Dispatch notification per NOBP-OP-1015 Event Notifications?

A. Transferring 13.8 KV Bus A to Startup Transformer 01 B. Idle Starting Emergency Diesel Generator 1 C. Transferring Main Generator Voltage Regulator from AUTOMATIC to MANUAL D. Adjusting Main Generator output by 10 MEGAVARS OUT to maintain the Voltage Schedule Answer: C Explanation/Justification:

A. Incorrect - no notification requirement. See DB-OP-06314 R13 13.8 KV Buses Switching Procedure section 3.8 and NOBP-OP-1015 R3 Event Notifications Attachment 66 (page 221). Plausible because transfer is to offsite power source.

B. Incorrect - no notification requirement. See DB-OP-06316 R57 Diesel Generator Operating Procedure section 4.30 and NOBP-OP-1015 R3 Event Notifications Attachment 66 (page 221). Plausible because it would add generation to the grid if it were loaded.

C. Correct - See DB-OP-06301 R27 Generator and Exciter Operating Procedure step 3.4.1 and NOBP-OP-1015 R3 Event Notifications Attachment 66 (page 221).

D. Incorrect - no notification requirement for small MVAR changes. See DB-OP-06301 R27 Generator and Exciter Operating Procedure section 3.5.

Threshold for MVAR reporting is >100 per NOBP-OP-1015 R3 Event Notifications Attachment 66 (page 221). Plausible because it does affect the grid conditions.

Sys # System Category KA Statement N/A N/A Generic Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator K/A# 2.4.30 K/A Importance 2.7 Exam Level RO References provided to Candidate None Technical

References:

NOBP-OP-1015 R3 Event Notifications Attachment 66 (page 221)

Question Source: New Question Cognitive Level: Low - Memory 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.11)

Objective: OPS-GOP-510-02K

Davis-Besse 1LOT15 NRC Written Exam r1

75. The plant is operating at 100% power.

A non-fire incident requires evacuation of the Control Room.

The required Immediate Actions of the governing procedure are performed.

NO Supplemental Actions are performed prior to Control Room Evacuation.

Which of the following describes actions that are REQUIRED to be performed outside the Control Room by the Balance of Plant (BOP) Reactor Operator (RO) under these conditions?

A. Tripping the Main Turbine to stop an Overcooling event B. Opening Reactor Trip Breakers to shut down the Reactor C. Tripping both Main Feed Pumps to initiate the Steam Feed Rupture Control System D. Isolating Instrument Air to the Atmospheric Vent Valves to allow manual operation.

Answer: D Explanation/Justification:

A. Incorrect - Overcooling event is not in progress because SFRCS Isolation Trip was manually actuated in Immediate Action 3.2. SFRCS Isolation Trip closes the MSIVs MS101 and MS100 which prevents an Overcooling event due to Main Turbine failure to trip. See DB-OP-02508 R16 Control Room Evacuation step 3.2 and DB-OP-02000 R27 Table 1. Immediate Actions were performed per the stem. Plausible because BOP RO trips turbine per step 2 of Attachment 4 since Supplementary Actions were not completed prior to evacuation.

B. Incorrect - Reactor is shut down in Immediate Action 3.1. Plausible because the BOP RO opens CRD breakers if the Immediate Actions were NOT performed. See DB-OP-02508 R16 Control Room Evacuation Attachment 4 step 1.1.

C. Incorrect - SFRCS initiation is an Immediate Action. See DB-OP-02508 R16 Control Room Evacuation step 3.2. Plausible because BOP RO would trip both feedpumps if Immediate Actions had NOT been performed.

D. Correct - See DB-OP-02508 R16 Control Room Evacuation Attachment 4 step 3.0.

Sys # System Category KA Statement N/A N/A Generic Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects K/A# 2.4.34 K/A Importance 4.2 Exam Level RO References provided to Candidate None Technical

References:

DB-OP-02508 R16 Control Room Evacuation Attachment 4 step 3.0.

Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 /

45.13)

Objective: OPS-GOP-108-03K

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

76. The plant is operating at 100% power.

Makeup Pump 2 is operating.

The following annunciators are received:

(2-2-C) MU TK LVL LO

(4-2-E) PZR LVL LO The following conditions are noted:

MU Pump 2 Discharge Pressure PI MU25B is 1700 psig.

Total Seal Injection flow FI MU19 is 30 gpm and lowering.

Seal Injection Flow Control Valve MU19 demand is 80% and rising.

Makeup Flow Control Valve MU32 demand is 100%.

CTMT Normal Sump level is constant.

ECCS ROOM 2 SUMP PUMP RUNNING lights IL4621A and IL4621B are lit.

Based on these indications, which DB-OP-02522, Small RCS Leaks attachment and action requires implementation NEXT to mitigate this event?

A. Perform Attachment 6, Isolation of Leaks in the Makeup System and stop Makeup Pump 2 B. Perform Attachment 6, Isolation of Leaks in the Makeup System and place the Alternate Injection line in service C. Perform Attachment 8, Isolation of Leaks in the Seal Injection Header and close the seal injection isolation to the affected Reactor Coolant Pump D. Perform Attachment 8, Isolation of Leaks in the Seal Injection Header and close MU19, Seal Injection Flow Control Valve Answer: A Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 first bullet. SRO is required to assess plant conditions and select the appropriate attachment to address mitigating the leak. The SRO must be familiar with the procedure actions and implementation priority to select the attachment to perform and have an understanding of the actions contained in the procedure.

A. Correct - Per DB-OP-02522, Small RCS Leaks, Attachment 4, the key indication is the lower than normal Makeup Pump Discharge Pressure which is less than 2200 psig which directs performance of Attachment 6. The first action in Attachment 6 is to stop the makeup pumps to attempt to stop the leak supply B. Incorrect - Plausible since isolation of MU32 may stop the leak and placing the alternate injection line in service may provide makeup but the first action is to stop the leak by stopping makeup flow C. Incorrect - Plausible since DB-OP-02522, Small RCS Leaks, Attachment 4 identifies high or low seal injection flow as possible leaks in the injection line and seal injection leaks in containment would be from an individual line. Attachment 8 directs isolation of the leak in an individual RCP injection line if diagnosed.

D. Incorrect - Plausible since a leak in the Seal injection system would cause abnormal flow and demand indications and closing MU19 isolates seal injection to all RCPs Sys # System Category KA Statement 000022 Loss of Reactor AA2 Ability to determine and interpret the following as they Whether charging line leak exists Coolant Makeup apply to the Loss of Reactor Coolant Makeup:

K/A# AA2.01 K/A Importance 3.8 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02522 R13 Small RCS Leaks Attachments 4 and 6; OS-0002 sheet 3 R33.

Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR 43.5/ 45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

77. The Plant is in Mode 6 with a Refueling Outage is in progress. RCS Level is at 80 inches, following Reactor Head Removal. Decay Heat Removal Train 2 is in service. Both SG primary side manways are removed to vent the reactor coolant system.

The following events occur:

All Offsite Power is Lost

EDG 1 does not start and cannot be manually started.

EDG 2 starts.

The SBODG cannot be started.

The following annunciator alarm is received:

(1-3-H) BUS D1 LOCKOUT Which of the following DB-OP-02527, Loss of Decay Heat Removal Action and Attachment must be performed to mitigate this event?

A. Start #1 DHR Pump to provide core cooling per Attachment 1, Starting Decay Heat Pump 1.

B. Restart #2 DHR Pump to provide core cooling per Attachment 2, Starting Decay Heat Pump 2.

C. Start either train of Auxiliary Feedwater and establish Steam Generator Heat Transfer per Attachment 3, Establish Steam Generator Heat Transfer.

D. Align the BWST to provide injection flow to the RCS to establish Feed and Bleed cooling per Attachment 10: Using Gravity Drain of the BWST to the RCS.

Answer: D Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 first bullet. SRO is required to have knowledge of the content of the procedure Attachments and be able to diagnose the plant configuration based upon the equipment status stated in the stem. The SRO must analyze the status of the electrical power supplies select the available strategy to mitigate the loss of RHR.

A. Incorrect - Plausible because at reduce inventory, DHR Pump 1 is maintained in standby to provide core cooling. In this scenario, C1 bus will not be energized and placing #1 DHR Pump in service is not possible based on no power available.

B. Incorrect - Plausible because on a loss of off-site power, #2 EDG Auto starts to restore power to D1. As a result, the operator only has to restart DHR Train 2 to provide Core Cooling. In this scenario, D1 is locked out and cannot be repowered.

C. Incorrect - Plausible because with different operating conditions, establishing SG heat transfer is the preferred heat removal mode during loss of Decay Heat Removal.. In this condition with the SG Manways removed, SG Heat Transfer is not possible.

D. Correct - Given the Plant Conditions, no electrical power will be available to provide inventory to the Reactor Coolant System. Feed and Bleed will be established using gravity drain of the BWST and venting the steam from the RCS with the SG Manways.

Sys # System Category KA Statement 000025 Loss of RHR Generic Knowledge of low power/shutdown implications in System accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies K/A# 2.4.9 K/A Importance 4.2 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02527 R19 Loss of Decay Heat Removal step 4.1.7 RNO and Attachment 3 step 3.

Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

78. The plant is operating at 100 percent power.

Component Cooling Water (CCW) Pump 1 is operating.

The following conditions are noted:

PRSRC2B Reactor Coolant (RC) Pressure Loop 1 is 2090 psig and slowly lowering.

PRSRC2A2 RC Pressure Loop 2 is 2090 psig and slowly lowering.

All Pressurizer Heater Banks are ON.

FI MU34 Makeup (MU) Flow Train 2 indicates 25 gpm.

Which of the following describes:

(1) the correct section of DB-OP-02513 Pressurizer System Abnormal Operation to implement? and (2) the procedure to implement if the initial mitigation actions are NOT successful?

A. (1) Section 3.2 Pressurizer Spray Valve RC 2 Failed Open (2) NOP-OP-1010 Operational Decision Making B. (1) Section 3.2 Pressurizer Spray Valve RC 2 Failed Open (2) DB-OP-02504 Rapid Shutdown C. (1) Section 3.7 Pressurizer Vapor Space Leak (2) NOP-OP-1010 Operational Decision Making D. (1) Section 3.7 Pressurizer Vapor Space Leak (2) DB-OP-02504 Rapid Shutdown Answer: B Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 second bullet. SRO is required to have knowledge of the content of the procedure action and mitigation strategies. Requires selection of appropriate abnormal procedure and knowledge of decision point in the body of the procedure. The SRO must diagnose the plant response to the failed equipment and select the correct procedural actions. The SRO is required to understand the actions of the procedure and alternate actions if they are not successful. The plant power will have to be reduced to secure the RCP with the affected valve.

A. Incorrect - Decision point at step 4.2.1.b requires power reduction and RCP stop. Part 1 is correct. NOP-OP-1010 plausible for evaluation of continued operation with the spray block valve closed if RCS pressure was stable, but power reduction for RCP stop required because RCS pressure is still lowering.

B. Correct - 25 gpm is normal MU flow value. The spray valve failure does not affect MU flow. See DB-OP-02513 R11 PZR System Abnormal Operation step 2.2.3. Decision point at step 4.2.1 requires power reduction and RCP stop - all PZR heaters are already ON with RCS pressure lowering per the stem and isolation attempts have failed.

C. Incorrect - PZR level rises per DB-OP-02513 step 2.7.1, so MU flow would lower to 12 gpm which is MU32 bypass value when MU32 is closed (minimum 10 gpm per DB-OP-06006 R35 step 2.2.40). Plausible because RCS pressure lowering and all heaters ON are consistent with vapor space leak. NOP-OP-1010 plausible for continued operation with leak to containment per step 4.7.5.

D. Incorrect - PZR level rises per DB-OP-02513 step 2.7.1, so MU flow would lower to 12 gpm MU32 bypass value when MU32 closed. Plausible because RCS pressure lowering and all heaters ON are consistent with vapor space leak. Part 2 correct since RCS pressure is slowly lowering in the stem, rapid shutdown per step 4.7.1.

Sys # System Category KA Statement 000027 Pressurizer AA2 Ability to determine and interpret the following as they apply to Makeup flow indication Pressure the Pressurizer Pressure Control Malfunctions:

Control System (PZR PCS)

Malfunction K/A# AA2.07 K/A Importance 3.1 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02513 R11 PZR System Abnormal Operation steps 2.2.3, 2.7.1, and 4.2.1 RNO Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

79. The plant is operating at 100% power.

During Main Turbine Stop Valve (TSV) testing, TSV 3 displays sluggish response due to valve stem binding.

(1) Which of the following describes the Technical Specifications Condition with the shortest Completion Time that applies?

A rupture downstream of Main Steam Non-Return Valve (NRV) Line 1 MS210 and upstream of the TSVs occurs.

Steam Generator 2 Main Steam Isolation Valve MS100 fails open.

TSV 3 fails open.

(2) Which Steam Feedwater Rupture Control System (SFRCS) actuated components prevent depressurization of SG 2 through the rupture?

A. (1) LCO 3.3.11 Condition B for all Loss of RCPs SFRCS Instrumentation Channels INOPERABLE. Be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(2) Main Steam NRVs Line 1 and 2 MS210 and MS209.

B. (1) LCO 3.7.4 Condition A for one TSV INOPERABLE. Close the inoperable TSV within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(2) TSVs 1 and 2 C. (1) LCO 3.3.11 Condition B for all Loss of RCPs SFRCS Instrumentation Channels INOPERABLE. Be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(2) TSVs 1 and 2 D. (1) LCO 3.7.4 Condition A for one TSV INOPERABLE. Close the inoperable TSV within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(2) Main Steam NRVs Line 1 and 2 MS210 and MS209.

Answer: B Explanation/Justification: Meets the requirements of the SRO only white paper Section II .B page 7 first and third bullet. SRO is required to have knowledge of Required Actions and Surveillance Requirements for Tech Specs and TRM. The SRO must determine which TS condition is required to be entered and understand the function of SFRCS system.

A. Incorrect - LCO 3.7.4 Condition A applies. TSVs 1 and 2 isolate SG 2 from the rupture. LCO 3.3.11 Condition B plausible because Loss of RCPs SFRCS trip generates a turbine trip (see DB-OP-02000 Table 1). Loss of RCPs SFRCS trip actuates the AFW Initiation Function, but not the Main Steam Isolation Function. NRVs are plausible for isolation, but not SFRCS-actuated and do not stop forward flow.

B. Correct - Binding TSV wont meet isolation time of 1 second required by SR 3.7.4.1. See TNC 8.3.11 Table 8.3.11-1 SFRCS Response Time.

Failure to meet Surveillance means LCO 3.7.4 is not met. See Generic SR 3.0.1. TSVs 1 and 2 isolate SG 2 from the rupture. See OS-0008 sheet 2 R21 and OS-0008 sheet 5 R29. SRO because it requires application of Required Actions and Surveillance Requirements for Tech Specs and TRM.

C. Incorrect - LCO 3.7.4 Condition A applies. Part 2 correct. LCO 3.3.11 Condition B plausible because Loss of RCPs SFRCS trip generates a turbine trip, but not the Main Steam Isolation Function. See B3.3.11.

D. Incorrect - TSVs 1 and 2 isolate SG 2 from the rupture. Part 1 correct per SR 3.7.4.1. NRVs are plausible for isolation, but not SFRCS-actuated and do not stop forward flow.

Sys # System Category KA Statement 000040 Steam Line Generic Knowledge of system purpose and/or function Rupture K/A# 2.1.27 K/A Importance 4.0 Exam Level SRO References provided to Candidate None Technical

References:

LCO 3.7.4 page 3.7.4-1 and Bases 3.7.4 page B3.7.4-1 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

80. The plant is operating in Mode 1 at 100% power.

Engineering has completed review of the recently completed Battery 1P Performance Discharge Test, conducted per SR 3.8.6.6

The analysis shows that Battery 1P is able to produce a maximum capacity of 78% of the battery nameplate rating Which of the following actions, if any, are required to comply with Technical Specifications requirements?

A. No action is required. Battery 1P is sized to be able to provide all loads with sufficient capacity at end of battery life which is defined as 75% of nameplate rating.

B. Declare DC Train 1 inoperable and enter TS 3.8.4 DC Sources Operating ONLY.

C. Declare Battery 1 P inoperable and enter TS 3.8.6 Battery Parameters ONLY.

D. Declare Battery 1P AND DC Train 1 inoperable. Enter TS 3.8.4 DC Sources Operating AND TS 3.8.6 Battery Parameters.

Answer: D Explanation/Justification: Meets the requirements of the SRO only white paper Section II .B page 3 third bullet. SRO is required to know the battery nameplate rating from the TS bases and that both TS for the Battery and DC sources are applicable based upon information from the TS bases.

A. Incorrect - Plausible if the candidate does not recognize that 78% is less than the minimum percent of nameplate capacity. This minimum name plate capacity is provided in the Bases for TS 3.8.4.

B. Incorrect - Plausible because TS 3.8.6 SR 3.8.6.6 requires entry into TS 3.8.4 (Battery Inoperable) if capacity test is less than 80%.

C. Incorrect - Plausible because the surveillance requirement not met is SR 3.8.6.6 associated with TS 3.8.6.6 but TS 3.8.4 entry is also required D. Correct - Failing SR 3.8.6.6 (less than 80% capacity) requires entering TS 3.8.6 and declaring the Battery Inoperable. TS 3.8.4 Bases specifies An OPERABLE DC electrical power source requires two batteries and one charger per battery to be operating and connected to the associated DC bus. TS 3.8.4 requires The Train 1 and Train 2 DC electrical power sources shall be OPERABLE therefore, TS 3.8.4 is also entered.

Sys # System Category KA Statement 000058 Loss of DC Generic Ability to apply Technical Specifications for a Power system K/A# 2.2.40 K/A Importance 4.7 Exam Level SRO References provided to Candidate None Technical

References:

Bases TS 3.8.4 and 3.8.6 Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.10 / 43.2 / 43.5 /

45.3)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

81. While operating at 100% power, a reactor trip occurs due to a loss of all Main Feedwater. The post trip review performed in accordance with DB-OP-06910, Trip Recovery determined the peak RCS Pressure reached during the event was 2775 psig.

The plant is currently in Mode 3, at Normal operating temperature and pressure. The Motor Driven Feedwater Pump is in service in the Main Feedwater Mode.

Which of the following requirements, if any, must be met prior to restarting the reactor?

A. No action is required. The RCS did not exceed the hydrostatic test pressure of 125% of design pressure.

B. No action is required. The Pressurizer Safety Valves prevented exceeding RCS design pressure by more than 10% in accordance with Section III of the ASME code for Nuclear Plant Components.

C. Request a fracture mechanics evaluation of the reactor vessel material AND evaluate compliance with TS 3.4.3, RCS Pressure and Temperature (P/T) Limits D. The Corrective Action Review Board (CARB) must review the DB-OP-06910, Trip Recovery, Attachment 6, Post Trip Review prior to restart.

Answer: C Explanation/Justification: Meets the requirements of the SRO only white paper Section II B page 3 first bullet. SRO is required to know the design pressure and administrative limitations of the RCS and actions required if the limits are exceeded. The administrative requirements are SRO only knowledge.

A. Incorrect - Plausible - while the RCS did not exceed the hydrostatic test pressure (design time 1.25 = 3125 psig), this does not preclude performing the DB-OP-06910, Trip Recovery required actions for exceeding a safety limit of 2750 psig.

B. Incorrect - Plausible because the TS Bases for the safety limits describes the code safety valve size as preventing exceeding the design pressure by more than 10% by the ASME Code,Section III.

C. Correct per Step 4.3 of Attachment 6 of DB-OP-06910, Trip Recovery.

D. Incorrect - Plausible because this event would lead to a Root Cause investigation and investigation at that level are reviewed by the CARB, but this review is not a specific requirement for restart following exceeding a Safety Limit and because a review is required by the Plant Operations Review Committee (PORC)

Sys # System Category KA Statement BW/E10 Post-Trip EA2 Ability to determine and interpret the following as they apply to Adherence to appropriate procedures and Stabilization the (Post-Trip Stabilization): operation within the limitations in the facilitys license and amendments K/A# EA2.2 K/A Importance 4.0 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-06910, Trip Recovery Attachment 6, DB-PF-06703 CC1.3 (pages 13-14)

Question Source: New Question Cognitive Level: Low 10 CFR Part 55 Content: (CFR: 43.5, 45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

82. The Miscellaneous Waste Monitor Tank (MWMT) has been aligned to transfer the contents of the MWMT to the Miscellaneous Waste Drain Tank (MWDT) for further processing to reduce radioactivity levels in accordance with DB-OP-06111, Miscellaneous Waste Liquid Waste System Section 4.5, Transferring the Contents of the MWMT to the MWDT.

Once the transfer is started, the following conditions are noted:

Annunciator 9-1-G FIRE OR RADIATION TROUBLE alarms

Miscellaneous Radwaste System Outlet RI 1878A1 and B1 High Alarm lights are lit

Station Effluent Indicator RI 8433 indications have not changed

Radiation Control Monitor (RCM) RE 8433 Low Flow light is lit.

Computer alarm Z670 MISC WST SYS OUT VLVS NC Based on these indications, which of the following actions, if any, are required?

A. Continue the transfer. These indications are consistent with the proper transfer of high activity liquid from the MWMT to the MWDT.

B. Continue the transfer. Refer to RA-EP-02861, Radiological Incidents to have Radiation Protection take additional surveys as required for radiation monitors in alarm.

C. Stop the transfer per DB-OP-03011 Radioactive Liquid Batch Release Attachment 22 Response to a RE Warn or High Alarm. An accidental release was NOT in progress.

D. Stop the transfer and restore the valve lineup per DB-OP-06111 Section 4.5. An accidental release was in progress.

Answer: D Explanation/Justification: Meets the requirements of the SRO only white paper Section II .D page 6 second bullet. The SRO must diagnose the event in progress based upon indications, then select the appropriate procedure to mitigate the event. They must determine that an accidental release was in progress based upon indications provided.

A. Incorrect - Plausible because Radiation Element RE1778A and B are in the recirc flowpath for the MWMT Pump used for this evolution, but not in the flowpath for transferring MWMT contents to the MWDT.

B. Incorrect - Plausible because Radiation Element RE1778A and B are in the recirc flowpath for the MWMT Pump used for this evolution, but not in the flowpath for transferring MWMT contents to the MWDT. RA-EP-02861, is used to respond to high radiation levels.

C. Incorrect - Attachment 22 of DB-OP-03011 only adjusts RE setpoints or release flow rate in response to an alarm. It does not stop the MWMT Pump or reposition any valves. Plausible because RE1778A&B are not in the transfer flowpath so transfer must be stopped and Attachment 22 title looks like it would do that. No release plausible for candidate missing the low sample flow condition on RE8443.

D. Correct - The flowpath for the transfer does not use RE1878A/B, so valves are misaligned and the transfer must be stopped. DB-OP-06111 Section 4.5 stops the MWMT Pump and closes the MWMT outlet valve WM1855 which stops the release. The computer alarm indicates a flowpath to the collection box exists. The misleading stable indication on RE 8433 is due to loss of sample flow.

Sys # System Category KA Statement 000059 Accidental AA2 Ability to determine and interpret the following as they apply to Failure modes, their symptoms, and the causes of Liquid the Accidental Liquid Radwaste Release: misleading indications on a radioactive-liquid Radwaste monitor Release K/A# AA2.03 K/A Importance 3.6 Exam Level SRO References provided to Candidate Technical

References:

DB-OP-06111, OS -29 Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

83. The plant is operating at 100 percent power.

Component Cooling Water (CCW) Pump 1 is operating.

A Loss of Offsite Power occurs.

The following annunciators are received:

1-4-H BUS D1 VOLTAGE

9-1-F INST AIR HDR PRESS LO

11-1-B CCW HX 1 OUTLET TEMP HI Both Emergency Diesel Generators (EDGs) are supplying their respective emergency buses.

Which ONE of the following requires implementation to correct the highest priority condition?

A. DB-OP-02511 Loss of Service Water Pumps/Systems B. DB-OP-02523 Component Cooling Water System Malfunctions C. DB-OP-02000 Attachment 6 Reenergization of Buses D2, F7, and MCC F71 D. DB-OP-02000 Attachment 28 Restore Power to C1 or D1 Bus from the SBODG Answer: A Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 first bullet. SRO is required to have knowledge of the content of the alarm response procedures. The SRO must diagnose the plant response to the failed equipment and select the correct procedural actions. Specific knowledge of the annunciator input is required to differentiate the procedure selection and priority.

A. Correct - Highest priority condition is EDG 1 operating with degraded CCW cooling because it is a precursor to required EDG shutdown. 11-1-B alarms at 115 ºF; CCW temperature limit for EDG operation is 130 ºF per Specific Rule 6.1.4. DB-OP-02511 Attachment 1 provides guidance to correct. SRO because 11-1-B is a common symptom to both DB-OP-02511 and DB-OP-02523 and knowledge of procedure content is required for proper selection.

B. Incorrect - DB-OP-02523 will not correct the high temperature condition. DB-OP-02523 section 4.4 for high CCW temperature takes manual control of the SW outlet valve to restore temperature. SW outlet is failed open because Instrument Air pressure is lost with the LOP, so DB-OP-02523 actions will not be effective. Plausible because the high CCW temperature is the highest priority condition.

C. Incorrect - Attachment 6 addresses restoration of instrument air pressure, which is a lower priority (DB-OP-02000 step 4.7 RNO) than maintaining Bus C1 powered by EDG 1 (Specific Rule 6). Plausible because restoration of instrument air pressure makes plant control from the control room much easier.

D. Incorrect - Attachment 28 is directed to be performed if both C1 and D1 remain de-energized. See Specific Rule 6.2. Attachment 28 is written for both EDG breakers open, so it will not correct the problem with EDG 1 since both buses are already energized by the EDGs. Plausible for misconception of proper Attachment 28 application.

Sys # System Category KA Statement BW/A05 Emergency Generic Ability to prioritize and interpret the significance of Diesel each annunciator or alarm Actuation K/A# 2.4.45 K/A Importance 4.3 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02000 R27 Specific Rule 6.1.4, DB-OP-02511 R16 Attachment 1 Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.3 /

45.12)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

84. Plant conditions:

Refueling operations in progress The following event occurs:

A fuel assembly has just been transferred from the Spent Fuel Pool

Main FH Bridge Operator reports assembly at full up in mast over basket

Refueling Canal Transfer Mechanism Operator reports the empty basket is malfunctioning and will not lower

3-1-A, REFUELING CANAL LVL alarms in the Control Room

4-3-A CTMT NORM SUMP LVL HI alarms in the Control Room

Spent Fuel Pool Bridge Operator reports Spent Fuel Pool Level approximately 15 feet above assemblies and lowering Which of the following identifies the actions that are required to be initiated FIRST based on the above conditions AND the procedure that will be transitioned to from DB-OP-02003, ECCS Alarm Panel Annunciator to direct completion of these actions?

A. Lower assembly into the core and evacuate Containment in accordance with DB-NE-06101, Fuel/Control Component Shuffle B. Lower assembly back into the basket and manually lower the basket in accordance with DB-NE-06306, Fuel Transfer System Operating Procedure C. Leave assembly in mast and abandon the bridges in accordance with DB-OP-00030, Fuel Handling Operations D. Leave assembly in mast and evacuate Containment in accordance with DB-OP-02530, Fuel Handling Accident Answer: C Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 third bullet. The SRO is required to know the specific procedure details as to the action to take when level is lower than to 19 feet. The procedure action is different depending upon the level of water above the assemblies.

A. Incorrect - plausible because lowering the assembly into the core would be an option except below 19 feet the direction is to abandon the bridges. DB-NE-06101, Fuel/Component Shuffle procedure provides directions for moving fuel B. Incorrect - plausible because lower into the basket and lowering the basket is an option except below 19 feet the direction is to abandon the bridges. DB-NE-06306 provides directions for manual operations C. Correct - DB-NE-02003 for refueling Canal level alarm will direct GO TO DB-OP-00030 which will direct abandoning the bridges if unable to lower loaded masts before canal level falls to 19 feet.

D. Incorrect - Plausible because Fuel Handling Accident directs evacuating containment and may be selected due to abnormal condition related to Fuel Handling but there is no accident without fuel damage, so entry conditions of DB-OP-02530 are not met.

Sys # System Category KA Statement BW/A08 Refueling AA2 Ability to determine and interpret the following as they apply to Facility conditions and selection of appropriate Canal Level the (Refueling Canal Level Decrease): procedures during abnormal and emergency Decrease operations K/A# AA2.1 K/A Importance 4.0 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02003, 3-1-A and DB-OP-00030 Question Source: Modified from TMI 2011 Question Cognitive Level: Low-Memory 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0 85 The plant is operating at 100% power with all components in normal alignment.

A large break Loss of Coolant Accident (LOCA) occurs.

10 minutes after the start of the LOCA the following indications are observed:

Reactor Coolant System (RCS) pressure is 170 psig

Low Pressure Injection (LPI) flow is 1500 gpm in each injection line

Borated Water Storage Tank (BWST) level is 37 feet Low Pressure Injection (LPI) Pump 1 trips.

Which of the following DB-OP-02000 attachments requires implementation at this time?

A. Attachment 11 HPI Flow Balancing B. Attachment 12 Establishing Long Term Boron Dilution using the Alternate Method C. Attachment 14 Establishing HPI Alternate Minimum Recirc Flowpath D. Attachment 22 Cross Connect LPI Pump Discharge Answer: D Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 first bullet. SRO is required to have knowledge of the content of the post EOP attachments. The SRO must determine RCS drain rate and status of the available equipment, then select the correct attachment to implement.

A. Incorrect - Plausible for misconception of piggybacked HPI Pump 1 lost along with LPI Pump 1.

B. Incorrect - Attachment 12 is performed after swap to sump. See DB-OP-02000 step 10.17. Plausible because this would be the correct method for establishing long term boron dilution.

C. Incorrect - Attachment 14 would be performed if BWST level was lowering at < 2 ft/hr. See DB-OP-02000 step 10.12 RNO 3. BWST has lowered from 40 feet to 37 feet over 10 minutes for a rate of 18 ft/hr. Loss of LPI Pump 1 still puts rate > 9 ft/hr. Plausible because HPI would have to remain in service after the swap to sump unless Attachment 22 is performed and 1350 gpm flow is maintained in both LPI lines for 20 minutes or more. See DB-OP-02000 step 10.12 and Specific Rule 3.5.1.

D. Correct - See DB-OP-02000 R27 step 10.7 RNO.

Sys # System Category KA Statement BW/E14 EOP Generic Ability to evaluate plant performance and make Enclosures operational judgments based on operating characteristics, reactor behavior, and instrument interpretation K/A# 2.1.7 K/A Importance 4.7 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02000 R27 step 10.7 Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.12 /

45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

86. The plant is at 100% power with all systems in a normal alignment with the exception of #1 MU Pump which is out of service for planned maintenance A Reactor Trip occurs. Subcooling Margin is lost.

The ATC Reactor Operator reports DB-OP-02000, RPS, SFAS, SFRCS Trip or SG Tube Rupture, Attachment 8, Place HPI, LPI, MU in Service has been completed with no deficiencies. The Reactor Operator later reports Makeup Tank level is 86 inches and slowly rising.

Based on these indications, which Section of DB-OP-02000 Attachment 13 requires implementation to mitigate the rising MU Tank Level?

Attachment 13, Controlling the MU System, Section ___________________.

A. Section 1, B, Transferring MU Pump Recirculation to the BWST.

B. Section 4, Diverting Letdown to the Clean Waste Receiver Tank.

C. Section 5, Transferring MU Pump Suctions to the BWST.

D. Section 7, Placing the MU Alternate Injection Line in Service.

Answer: A Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 second bullet. SRO is required to have knowledge of the content of the EOP Attachment procedures. The SRO must determine the plant configuration based procedure guidance with the loss of subcooling. Then the SRO must have detailed procedure knowledge of Attachment 13 subsections.

A. Correct - With a loss of Subcooled Margin, MU Pump Suctions are locked on the BWST per Attachment 8, with the recirculation flowpath still aligned to the MU Tank. This caused MU tank level to rise. Transferring recirculation flow to the BWST will terminate the increase.

B. Incorrect - Plausible because in a normal alignment, diverting Letdown to the CWRT will reduce MU Tank Level, however Letdown is isolation per SFAS actuation and Attachment 8. Transferring Letdown to the CWRT will not affect MU Tank Level.

C. Incorrect - Plausible because in a normal alignment, transferring MU suctions to the BWST will lead to an auto transfer back to the MU tank at 86 inches, however, per Attachment 8 with a loss of SCM, MU Pump Suctions are Locked on the BWST preventing this auto transfer.

D. Incorrect - Plausible because in a normal alignment, placing the Alternate Injection Line in service would increase MU Flow and cause MU Tank Level to lower, however Attachment 8 places the Alternate Injection Line in service only when 2 MU Pumps are available. Performing this action would result in two injection lines on a single MU Pump which is not allowed.

Sys # System Category KA Statement 004 Chemical and Generic Ability to interpret control room indications to Volume verify the status and operation of a system, and Control understand how operator actions and directives System affect plant and system conditions.

(CVCS)

K/A# 2.2.44 K/A Importance 4.4 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02000 Attachment 8 (pages 318 and 320)

Attachment 13 (pages 341 and 342)

Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.12)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

87. Plant conditions:

Plant is in Mode 5 with Reactor Coolant System Loops filled

Decay Heat Pump 2 is in service

The Motor Driven Feedpump and Startup Feed Pump are available

SG secondary side fill to wet layup is in progress

SG 1 Full Range level is 645 inches

SG 2 Startup Range level indicates 10 inches

Emergency Diesel Generator 1 is INOPERABLE for testing The following event occurs:

Decay Heat Pump 2 trips on overcurrent

Decay Heat Pump 1 is placed in service on the RCS (1) Are Two Loops OPERABLE to comply with Tech Spec 3.4.7 RCS Loops - Mode 5, Loops Filled and, (2) if not, which action would ensure compliance?

A. (1) Two Loops are NOT OPERABLE.

(2) Drain SG1 to 620 inches Full Range in accordance with DB-OP-06230, Steam Generator Secondary Side Fill, Drain and Layup Procedure B. (1) Two Loops are NOT OPERABLE.

(2) Fill SG 2 to 16 inches on the Startup Range in accordance with DB-OP-06226, Startup Feed Pump Operating Procedure C. (1) Two Loops are NOT OPERABLE.

(2) Place EDG 1 in standby in accordance with DB-OP-06316, Diesel Generator Operating Procedure D. (1) Two RCS Loops are OPERABLE.

(2) No action is required.

Answer: A Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 first bullet and II.B Page 3 third bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant status and determine which procedure to implement to meet the TS Operability requirements for RCS loops. Detailed knowledge of the bases information is required to select the correct procedure actions.

A. Correct -. This is correct Per Tech Spec Bases 3.4.7 the steam generator maximum level must be maintained low enough such that the steam generator remains capable of heat removal by maintaining a steam flow path (i.e., 625 inches full range level). DHR 1 does not require an emergency power source to be considered operable. Per Tech Spec Bases 3.4.7, DHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required.

B. Incorrect - Plausible because 16 inches is the cutoff for indication of a dry SG per DB-OP-02000. This would not make RCS Loop 1 OPERABLE:

Per Tech Spec Bases 3.4.7, An OPERABLE SG requires 35 inches of secondary side water level above the lower tube sheet. DB-OP-06226 provides no direction for filling SGs.

C. Incorrect - Plausible because DHR Loop1 may be considered INOPERABLE due to its emergency power supply (see explanation in correct answer). RCS loops may be determined to be operable because an electric feed pump is available and water exists in the SGs. Per TS Bases 3.4.7 to ensure that the SGs can be used as a heat sink, an electrically driven feed pump is needed, because it is independent of steam D. Incorrect - Plausible since DHR 1 is Operable and an electric feed pump is available and water exists in the SGs. Per TS Bases 3.4.7 However, to ensure that the SGs can be used as a heat sink, an electrically driven feed pump is needed, because it is independent of steam

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0 Sys # System Category KA Statement 005 Residual A2 Ability to (a) predict the impacts of the following malfunctions or RHR pump/motor malfunction Heat operations on the RHRS, and (b) based on those predictions, use Removal procedures to correct, control, or mitigate the consequences of those System malfunctions or operations:

(RHRS)

K/A# A2.03 K/A Importance 3.1 Exam Level SRO References provided to Candidate None Technical

References:

TS 3.4.7 Bases page 3.4.7-3, DB-OP-06230, pg 2 Question Source: New Question Cognitive Level: High-Comprehension 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.3 /

45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

88. SR 3.3.5.2, SFAS Channel 1 Monthly Functional is scheduled to be performed.

Various SFAS Channel 1 parameters will be INOPERABLE during this test.

The Test is scheduled for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Which of the following statements describes how Technical Specification 3.3.5 will be applied during this test?

Entry into associated Conditions and Required Actions_____________

A. will be required during the performance of this test unless compliance will cause undesired actuation of safety system components B. may be delayed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , provided two other channels of the same SFAS instrumentation Parameter are OPERABLE C. will not be required since SFAS Channel 1 is required to be restored to OPERABLE status at least once per hour during the performance of the test D. may be delayed indefinitely provided all three of the remaining channels of the same SFAS instrumentation Parameter are OPERABLE Answer: B Explanation/Justification: Meets the requirements of the SRO only white paper Section II .B page 3 first bullet. SRO is required to have knowledge of the system limits. The SRO is required to know that the entry into the TS Conditions and Actions is modified by a note that allows for a time limit to perform the surveillance test.

A. Incorrect - Plausible since undesired actuation of SFAS equipment could be an unintended consequence B. Correct - This is as stated in the Note to SR 3.3.5.2 and DB-SC-03110 R20 SFAS Channel 1 Functional Test L&P step 2.1.2.a.4 C. Incorrect - Plausible since no action would be required if inoperability time was less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D. Incorrect - Plausible since one RPS Channel can be bypassed indefinitely implying 2 out of 3 logic is sufficient Sys # System Category KA Statement 013 Engineered Generic Ability to explain and apply system limits and Safety precautions Features Actuation System (ESFAS)

K/A# 2.1.32 K/A Importance 4.0 Exam Level SRO References provided to Candidate None Technical

References:

SR 3.3.5.2 page 3.3.5-3 of Tech Specs Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.10 / 43.2 / 45.12)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

89. A Waste Gas Decay Tank release is in progress when the following occurs:

Annunciator 9-1-G FIRE OR RADIATION TROUBLE alarms

Alarm 7-1-C, WST GAS SYS OUT RAD HI alarms.

RE1822A Detector is observed to be failed high

RE1822B is indicating normal.

(1) What, if any, is the effect on the release of this failed detector?

(2) Where are the operability requirements addressed for the failed detector?

A. (1) Release still in progress due to the redundant instrument operating correctly (2) Technical Specifications B. (1) Release still in progress due to the redundant instrument operating correctly (2) Off-Site Dose Calculation Manual C. (1) The release is terminated by the failed detector (2) Technical Specifications D. (1) The release is terminated by the failed detector (2) Off-Site Dose Calculation Manual Answer: D Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 third bullet. The operator is required to diagnose and predict the impact of the monitor failure on the release pathway. The SRO is also required to know that the entry into the ODCM for corrective actions is required for the detector failure.

A. Incorrect - (1) Plausible because most safety and protective systems have redundant trip functions with coincidence logic (2) is plausible because Technical Specifications is a source for addressing other Operability requirements B. Incorrect - (1) Plausible because most safety and protective systems usually redundant trip functions with coincidence logic (2) is correct C. Incorrect - (1) is correct (2) ) is plausible because Technical Specifications is a source for addressing other Operability requirements D. Correct - Either detector in alarm will trip close the gaseous release outlet valves per OS-0030 SH2 R20 CL-1 & DB-OP-03012 steps 4.3.10 and 4.3.11 and operability for RE1822A/B in ODCM Section 3 Sys # System Category KA Statement 073 Process A2 Ability to (a) predict the impacts of the following malfunctions or Detector failure Radiation operations on the PRM system; and (b) based on those predictions, Monitoring use procedures to correct, control, or mitigate the consequences of (PRM) those malfunctions or operations:

System K/A# A2.02 K/A Importance 3.2 Exam Level SRO References provided to Candidate Technical

References:

ODCM Table 3-1 page 56, OS-0030 SH2 R20 CL-1 Question Source: New Question Cognitive Level: Low-Memory 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.3 /

45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

90. The plant is operating at 100% power with Service Water Returns aligned to the Cooling Tower.

A seismic event occurs that significantly damages piping in the non-seismic portion of the Service Water System.

Which of the following actions are required to respond to this event?

A. Align Circ Water to supply Service Water Essential Header per DB-OP-02511, Loss of Service Water Pumps/System, Attachment 9, Lineup Circulating Water to Supply SW Primary Loads.

B. Align Circ Water to supply Service Water Secondary Loads per DB-OP-02511, Loss of Service Water Pumps/System, Attachment 10, Lineup Circulating Water to Supply SW Secondary Loads.

C. Align Service Water Returns to the Collection Box per DB-OP-02511, Loss of Service Water Pumps/System, Attachment 14, Align SW Returns to a Safety Grade Flowpath.

D. Align Service Water Returns to the Intake Forebay per DB-OP-02511, Loss of Service Water Pumps/System, Attachment 14, Align SW Returns to a Safety Grade Flowpath.

Answer: D Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 first bullet. SRO is required to have knowledge of the content of the procedure content. The SRO must determine the plant configuration based event in progress and select the appropriate action to mitigate the loss of SWS event, which renders the system incapable of performing the design function. The SRO must have detailed knowledge of the procedure content to select the Attachment that restores the system flowpath.

A. Incorrect - Plausible because the essential loads are those that must be cooled to maintain safety functions. The Cooling Tower Return flowpath is non-seismic piping which may be pinched by the seismic event.

B. Incorrect - Plausible because the SW Piping to the Secondary loads is non-seismic piping. There are important loads such as the MDFP the candidate would prefer to have available to respond to plant events..

C. Incorrect - Plausible because the Cooling Tower Line is not seismic. If the line collapses, a loss of flowpath could exist. This alignment could restore a flowpath, but the inventory from the Ultimate Heat Sink would be lost.

D. Correct - The initial conditions has SW aligned to Cooling Tower Makeup. Following a seismic event, this alignment could lead to depletion of the Ultimate Heat sink. Action is required within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months to protect the ultimate heat sink inventory.

Sys # System Category KA Statement 076 Service A2 Ability to (a) predict the impacts of the following malfunctions or Loss of SWS Water operations on the SWS; and (b) based on those predictions, use System procedures to correct, control, or mitigate the consequences of those (SWS) malfunctions or operations:

K/A# A2.01 K/A Importance 3.7* Exam Level SRO References provided to Candidate Technical

References:

DB-OP-02511 R16, Loss of Service Water Pumps/System Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45/3 /

45/13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

91. The plant is in Mode 1 at 100% power following a refueling outage.

A Condition Report is received that identifies an error occurred when the Containment Purge Exhaust Valve (Inside CTMT Valve) local leak rate testing was conducted prior to entry into Mode 4. The leak rate was incorrectly calculated. The actual leakage exceeded the allowed leakage for that penetration but the leak rate does not exceed the overall containment leakage rate acceptance criteria.

Which of the following Technical Specification actions are required?

A. No action required because the isolation valve leakage does not exceed the overall containment leakage rate acceptance criteria.

B. No action is required because the penetration flow path remains isolated by the Purge Exhaust Valve (Outside CTMT Valve) which is already de-activated in the closed position.

C. Enter TS 3.6.1 Containment and be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Mode 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to allow repair/testing of the affected valve.

D. Enter TS 3.6.3 Containment Isolation Valves and verify the affected penetration is isolated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and once per 31 days thereafter.

Answer: D Explanation/Justification: Meets the requirements of the SRO only white paper Section II B page 3 first bullet. The SRO must evaluate the TS impact of the valve that exceeds the allowable leak rate and the impact on overall CNMT operability. The TS action per the surveillance is SRO knowledge. The SRO must know that the TS is applicable in Mode 4 and the actions required.

A. Incorrect - Plausible because the overall CTMT leak rate is still met, therefore the expected leakage following a design bases event would be less than the assumed leakage in the calculations that estimate off-site dose impact of an event.

B. Incorrect - Plausible because the conditions stated are true, an operable closed valve remains in the flowpath.

C. Incorrect - Plausible because per TS 3.6.3, the use of administrative controls to unisolate the penetration for testing is not permitted requiring a return to Mode 5 for repair/testing.

D. Correct - Leakage in excess of allowed requires entry into 3.6.3 Condition D, but with overall CTMT leakage less than acceptance criteria entry into TS 3.6.1 is not required.

Sys # System Category KA Statement 029 Containment Generic Knowledge of the bases in Technical Purge Specifications for limiting conditions for operations and safety limits K/A# 2.2.25 K/A Importance 4.2 Exam Level SRO References provided to Candidate None Technical

References:

TS 3.6.3 Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.5 / 41.7 / 43.2)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

92. The plant is in Mode 1 at 100% power.

The following annunciators alarm:

13-2-B CNDS PMP DISCH HDR PRESS

13-4-B HP CNDSR HOTWELL LVL LO

13-4-C DEAR STRG TK 1 LVL

13-4-D DEAR STRG TK 2 LVL

15-3-F CNDSR PIT FLOODED Based on these indications, which of the following describes the effect on the Condensate System and the procedures to implement for this failure?

The Condensate Pump motors __(1)__. Implement DB-OP-02000 RPS, SFAS, SFRCS Trip or SG Tube Rupture and transition to __(2)__.

A. (1) trip on low hotwell level (2) DB-OP-06910 Trip Recovery Section 4.0 Recovery from Reactor Trip and SFRCS Actuation B. (1) trip on low hotwell level (2) DB-OP-06903 Plant Cooldown Section 3.0 Cooldown of the NSSS from HOT STANDBY (MODE 3)

Condition.

C. (1) become submerged and fault (2) DB-OP-06910 Trip Recovery Section 4.0 Recovery from Reactor Trip and SFRCS Actuation D. (1) become submerged and fault (2) DB-OP-06903 Plant Cooldown Section 3.0 Cooldown of the NSSS from HOT STANDBY (MODE 3)

Condition Answer: A Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 second bullet. SRO is required to have knowledge of the content of the procedures and actions taken based upon conditions of the plant. The SRO must evaluate the potential water level that could result from the flooding of the condensate pumps and the interlocks associated with the pumps. Requires detailed knowledge of the procedure routing following the reactor trip.

A. Correct - Malfunction is condensate header rupture. Condensate Pumps trip on low hotwell level of 24 inches - see OS-0010 sheet 3 R15 CL-5.

Spilled liquid in Turbine Building ends up in condenser pit and actuates automatic trip of Circ Pumps at 2.5 feet - see DB-OP-06272 R24 Station Drainage and Discharge System Attachment 3 pages 70 & 74 and OS-0016A R36. Loss of Circ Pumps results in loss of vacuum and automatic trip of Main Feedwater Pumps - see DB-OP-02518 R6 High Condenser Pressure page 22 last paragraph. MFW Pump trips causes SFRCS Isolation Trip on reverse Feedwater dP - see OS-0012A sheet 2 R32 CL12 and DB-OP-02000 Table 1. DB-OP-02000 R27 step 4.23 provides routing to DB-OP-06910. DB-OP-06910 R26 step 3.1.1 provides routing to Section 4.0.

B. Incorrect - DB-OP-02000 R27 step 4.23 provides routing to DB-OP-06910. Part 1 is correct. Plausible because a plant cooldown would be required to fix the condensate header rupture.

C. Incorrect - Condensate Pumps trip on low hotwell level of 24 inches. Part 2 is correct. Plausible because this is the result for condenser pit flooding from Circ Water without low hotwell level - see DB-OP-02517 R6 Circulating Water System Malfunctions Background Information for Large Leak/Rupture page 61. Even if condensate pump motors did become submerged, they were de-energized when low hotwell level opened their breakers.

D. Incorrect - Condensate Pumps trip on low hotwell level of 24 inches; DB-OP-02000 R27 step 4.23 provides routing to DB-OP-06910. Plausible for condenser pit flooding from Circ Water without low hotwell level and because a plant cooldown would be required to fix the condensate header rupture.

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0 Sys # System Category KA Statement 056 Condensate A2 Ability to (a) predict the impacts of the following malfunctions or Loss of condensate pumps operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

K/A# A2.04 K/A Importance 2.8* Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02518 page 22, OS-0010 CL-5 and DB-OP-02000 supplemental step 4.23 Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.3 /

45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

93. The plant is operating at 100% power The following event occurs:

SG1 Startup Level selected to ICS indicates 0 inches The following annunciators alarm:

12-3-A SG 1 OPERATE LVL HI

12-4-A SG 1 LVL LO

14-4-E ICS INPUT MISMATCH

14-5-E ICS SG 1 ON LO LVL LIMIT (1) Select the correct PROCEDURE to be implemented to mitigate this event?

(2) What is Technical Specification BASIS for Technical Specification LCO challenged by this event?

A. (1) DB-OP-02526 Primary to Secondary Heat Transfer Upset (2) To ensure Steam generator water inventory is maintained to provide adequate primary to secondary heat transfer B. (1) DB-OP-02526 Primary to Secondary Heat Transfer Upset (2) To preserve the initial condition assumptions for the steam generator inventory used in the main steam line break (MSLB) accident analysis C. (1) DB-OP-02000 RPS, SFAS, SFRCS TRIP OR SG TUBE RUTURE (2) To ensure Steam generator water inventory is maintained to provide adequate primary to secondary heat transfer D. (1) DB-OP-02000 RPS, SFAS, SFRCS TRIP OR SG TUBE RUTURE (2) To preserve the initial condition assumptions for the steam generator inventory used in the main steam line break (MSLB) accident analysis Answer: B Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 first bullet and II.B Page 3 third bullet. The SRO must diagnose the primary to secondary plant heat transfer event and cause and then select the appropriate mitigation procedure. The SRO is required to understand the TS bases for the LCO that is challenged by the failure.

A. Incorrect - Part 1 is correct. Part 2 is plausible because this is the TS basis for the low level limit B. Correct -Malfunction is failure of controlling Startup (SU) level low which results in SG overfeed. See M-533-00171 R10. DB-OP-02526 R4 is correct procedure. See step 2.1.4. TS Bases as listed in B 3.7.18 page B 3.7.18-1 C. Incorrect - Plausible if it is determined the Reactor has or should be tripped on low or high level. Part 2 is plausible because this is the TS basis for the low level limit D. Incorrect - Plausible if it is determined the Reactor has or should be tripped on low or high level. Part 2 is plausible because this is the TS basis for the low level limit. Part 2 is correct Sys # System Category KA Statement 035 Steam A2 Ability to (a) predict the impacts of the following malfunctions or Pressure/level transmitter failure Generator operations on the Steam Generator System; and (b) based on those System predictions, use procedures to correct, control, or mitigate the (SG/S) consequences of those malfunctions or operations:

K/A# A2.03 K/A Importance 3.6 Exam Level SRO References provided to Candidate None Technical

References:

DWG M-533-171-10, DB-OP-02526 page 4 and Bases B3.7.18 page B3.7.18-1 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.3 / 45.5)

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

94. The plant is at 100% power at minimum staffing levels.

At 0200, the following events occur.

The At the Controls Reactor Operator (ATC RO) falls in the Control Room and is unconscious.

The First Aid Team requests an ambulance to transport the ATC RO to the hospital.

Which of the following actions is required in response to this event?

A. Unit Supervisor must accompany ATC RO to the hospital in the ambulance.

B. Maintain the plant in a stable condition until the next shift of operators arrives for day shift.

C. Immediately callout a Reactor Operator to return to a minimum functional shift complement.

D. Have the Safe Shutdown Equipment Operator assume the RO position to comply with the Technical Requirements Manual.

Answer: C Explanation/Justification: Meets the requirements of the SRO only white paper Section II .A page 3 third bullet. The SRO is required to know the content of the administrative procedures related to shift staffing and the Technical Specification requirements. The actions to restore shift staffing are a SRO responsibility.

A. Incorrect - Loss of an SRO would make shift manning level worse. Plausible because supervisor is notified; however, RA-EP-02000 R5 Medical Emergencies step 6.2.9 states that when on-duty manning is minimal, a Management Representative shall be called to meet the patient at the treatment facility.

B. Incorrect - Plausible because NOP-OP-1002 step 4.1.13.3 does direct maintaining stable conditions, but allowing 3-4 hours to elapse is not consistent with taking action immediately.

C. Correct - per NOP-OP-1002 (R09), Conduction of Operations Step 4.1.13.3.

D. Incorrect - Even if the SSEO was licensed, minimum manning is not met per NOP-OP-1002 R9 Conduct of Operations Attachment 4. Plausible because the TRM does not require any non-licensed operators (see TRM 10.2.1) and per NOP-OP-1002, Conduct of Operations step 4.1.13.3 if the Shift Manager becomes incapacitated, the senior on shift licensed operator assumes the Shift Manager position; however, no such provision exists for other positions.

Sys # System Category KA Statement N/A N/A Generic Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.

K/A# 2.1.4 K/A Importance 3.8 Exam Level SRO References provided to Candidate None Technical

References:

NOP-OP-1002 (R09) Step 4.1.13.3 Question Source: New Question Cognitive Level: Low 10 CFR Part 55 Content: (CFR: 41.10 / 43.2)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

95. The plant is at 100% power at 200 EFPD. Rod Height is 290 Rod Index.

Reactor Coolant Pump (RCP) 1-1 develops an oil leak and must be shutdown.

Once stable, the following conditions are noted:

Reactor Power 72%

RCP 1-1 stopped

Axial Power Imbalance is -10%.

Rod Height is 260 Rod Index.

Which of the following actions, if any, are the FIRST required to comply with Technical Specifications requirements?

References provided A. No Action is required B. Verify FQ and FNH are within limits by using the Incore Detector System to obtain a power distribution map within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months C. Reduce THERMAL POWER to 40% RTP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months D. Reduce THERMAL POWER to less than or equal to the THERMAL POWER allowed by the regulating rod group insertion limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Answer: B Explanation/Justification: Meets the requirements of the SRO only white paper Section II B page 3 first bullet. SRO must interpret the TS 3.2.1.

Surveillance and graph, recognizing that the plant is in the restricted region and then select the appropriate TS action for this condition.

A. Incorrect - Plausible if the candidate uses the more typical COLR Figure 2a curve, 0 to 300 +10 EFPD, Four RC Pumps--2817 MWt RTP Davis-Besse 1, Cycle 19, instead of the correct three pump curve Figure 2c.

B. Correct - The plant is in the restricted region for 3 RCPs of Figure 2c. TS 3.2.1, Regulating Rod Insertion Limits Condition A requires performance of SR 3.2.5.1 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

C. Incorrect - This is the required action if TS 3.2.3, Axial Power Imbalance is not met which is possible for a rapid power reduction. In this case, Axial power imbalance is within the limits of TS 3.2.3 and therefore, not applicable D. Incorrect - This is the required action if TS 3.2.1 Condition A is not met which would be required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from the initiating event .

Sys # System Category KA Statement N/A N/A Generic Ability to interpret reference materials, such as graphs, curves, tables, etc.

K/A# 2.1.25 K/A Importance 4.2 Exam Level SRO References provided to Candidate Core Operating Limit Technical

References:

LCO 3.2.1; COLR Figure 2c, LCO 3.2.1 Action A and SR 3.2.5.1 Report Figures 2a, 2b, 2c, and 2d.

Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.12)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

96. An overhead Annunciator Alarm in the Control Room is not operating properly.

To avoid nuisance alarms, the Operations Manager has determined that the Annunciator will be disabled by removing the affected Annunciator Point Card.

Which of the following documents must be completed to remove this point card to disable the affected annunciator alarm?

1. Annunciator System Operating Procedure

2. Work Order for point card removal

3. 50.59 Regulatory Applicability Determination (RAD) and/or Screen

4. Engineering Change Package

5. Temporary Modification Tags

6. Clearance and Tags A. 1 and 3 B. 2, 4, and 5 C. 1 and 6 D. 2, 3 and 6.

Answer: A Explanation/Justification: Meets the requirements of the SRO only white paper Section II .C page 6 third bullet. The SRO is required to know the administrative requirements for disabling annunciators. Additionally the SRO must be knowledgeable of the requirements for implementing other types of work as well to correctly identify the required documents to disable the alarm.

A. Correct. Disabling an Annunciator Window is directed using DB-OP-06411, Station Annunciator Procedure Section 4.5 which also requires a 50.59 RAD and/or Screen.

B. Incorrect - Disabling an Annunciator Window is directed using DB-OP-06411, Station Annunciator Procedure Section 4.5 which requires a 50.59 RAD and/or Screen. Plausible for another craft such as I&C or IS to perform card removal under a Work Order, pulled circuit cards may be considered Temporary Modifications per NOP-CC-2003 R19 Engineering Changes step 2.1.3, TM Tags described in NOP-CC-2003 Attachment 7.

C. Incorrect - Disabling an Annunciator Window is directed using DB-OP-06411, Station Annunciator Procedure Section 4.5 which requires a 50.59 RAD and/or Screen. A Clearance is not necessary or directed to perform this activity. Plausible to use OPS Only Clearance for equipment control per NOP-OP-1001 R21 Clearance and Tagging Program Section 4.10.

D. Incorrect - Disabling an Annunciator Window is directed using DB-OP-06411, Station Annunciator Procedure Section 4.5 which requires a 50.59 RAD and/or Screen. A Clearance is not necessary or directed to perform this activity. Plausible for another craft such as I&C or IS to perform card removal under a Work Order and Clearance.

Sys # System Category KA Statement N/A N/A Generic Knowledge of the process for making design or operating changes to the facility K/A# 2.2.5 K/A Importance 3.2 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-06411 Section 4.5.

Question Source: New Question Cognitive Level: Low 10 CFR Part 55 Content: (CFR: 41.10 / 43.3 / 45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

97. The Plant is in Mode 5.

Based on planned maintenance, the Key Shutdown Defense in Depth for Electrical Power Availability meets the minimum number of points to be rated Yellow.

The following event occurs:

A Severe Thunderstorm Watch that includes Davis-Besse is issued by the National Weather Service.

Which of the following describes the impact on the Shutdown Defense In Depth indicator for the change in weather status and the maintenance controls that must be invoked?

Key Shutdown Defense in Depth for Electrical Power Availability ________________.

References provided A. remains Yellow. This indicator is not affected by the weather forecast. Continue to comply with Yellow Risk Requirements of NOP-OP-1007, Risk Management.

B. remains Yellow but would require transition to Orange if a Severe Thunderstorm Warning is issued.

Continue to comply with Yellow Risk Requirements of NOP-OP-1007, Risk Management.

C. would transition to Orange Risk. Comply with the Orange Risk Requirements of NOP-OP-1007, Risk Management.

D. would transition to Orange Risk, but require transition to Red if a Severe Thunderstorm Warning is issued. Comply with the Orange Risk Requirements of NOP-OP-1007, Risk Management.

Answer: C Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 first bullet. SRO is required to have knowledge of the content of the administrative procedures and actions taken based upon conditions of the plant. Requires detailed knowledge of the procedure and evaluation of the impact on th risk level based upon changes in the weather conditions.

A. Incorrect - Per NOP-OP-1005 Checklist, issuing a Severe Watch or Warning requires a reduction of one point which would cause the indicator to go to Orange.

B. Incorrect - Per NOP-OP-1005 Checklist, issuing a Severe Watch or Warning requires a reduction of one point which would cause the indicator to go to Orange. No further upgrade would be required if a warning is later issued.

C. Correct. Per NOP-OP-1005 Checklist, issuing a Severe Watch or Warning requires a reduction of one point which would cause the indicator to go to Orange. No further upgrade would be required if a warning is later issued.

D. Incorrect Per NOP-OP-1005 Checklist, issuing a Severe Watch or Warning requires a reduction of one point which would cause the indicator to go to Orange. Reduction of another point would drive the indicator to Red. No further upgrade would be required if a warning is later issued.

Sys # System Category KA Statement N/A N/A Generic Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

K/A# 2.2.18 K/A Importance 3.9 Exam Level SRO References provided to Candidate NG-DB-00117 and Form Technical

References:

NOP-OP-1005 Checklist and NOP-OP-1005 step NOP-OP-1005-02 4.3, NG-DB-00117 attachment 2 Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

98. A seismic event has occurred.

High Radiation Alarms are received on the following:

RE 8426 SFP Area

RE 8427 SFP Area

RE 8417 Fuel Handling Area

RE 8418 Fuel Handling Area

RE 8425 Equipment Hatch Area Spent Fuel Pool (SFP) Level LI1600 is 9 feet and stable.

Which one of the following actions and procedures require implementation?

A. Align a Decay Heat Removal Train to provide SFP Cooling per RA-EP-02820, Earthquake.

B. Evacuate the Spent Fuel Pool Area per RA-EP-02861, Radiological Incidents.

C. Perform off-site Dose Assessment per RA-EP-02240, Off-Site Dose Assessment.

D. Implement the Severe Accident Management Guidelines for a Severe Accident in the Spent Fuel Pool.

Answer: B Explanation/Justification: Meets the requirements of the SRO only white paper Section II .D page 6 second bullet. SRO must analyze the Radiation levels based upon the alarms received and then select the appropriate procedure to implement. The SRO is required to have knowledge of the procedure content which includes evacuating the area based upon the rise in activity and the current level in the SFP.

A. Incorrect - Plausible because RA-EP-02820 R9 would apply and step 6.2.2.h suggests this action for a loss of SFP cooling; however, this action is incorrect for a large leak. Minimum level to operate DHR Pump on SFP is 12 feet per DB-OP-02547 R4 SFP Cooling Malfunctions step 4.2.8 B. Correct - A minimum of 9.5 feet of level in the SFP is required to provide adequate biological shielding. With level below 9.5 feet and multiple high radiation alarms, RA-EP-02861 should be implemented and the area should be evacuated. RA-EP-02861 entry is also directed by DB-OP-02547 R4 SFP Cooling Malfunctions step 4.2.9.

C. Incorrect - Airborne release not in progress or imminent. See RA-EP-02240 R8 Offsite Dose Assessment step 5.0 Initiating Conditions.

Plausible because the inventory lost from the SFP has gone somewhere, however, Spent Fuel remains covered and even if the SFP contents are outside the SFP area, the inventory would not leave the site without specific action to pump the marsh area.

D. Incorrect - Plausible because a Spent Fuel Pool level of 1 foot requires entry into the Severe Accident Management Guidelines. See DB-OP-02547 R4 SFP Cooling Malfunctions step 4.2.17 RNO.

Sys # System Category KA Statement N/A N/A Generic Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities K/A# 2.3.14 K/A Importance 3.8 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02547 step 4.2.9 and USAR page 9.1-9.

Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.12 / 43.4 / 45.10)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0

99. The Plant is operating at 100% power when the following occurs:

Multiple Fire Alarms are received from Room 603, CONTROL ROOM AC EQUIPMENT ROOM, Fire Area HH

The Fire Brigade is dispatched in accordance with DB-OP-02529, Fire Procedure. The Fire Brigade Captain reports a significant fire is in progress and requests off-site assistance

The ATC Reactor Operator reports High Pressure Injection Pump 2 and Containment Spray Pump 2 have spuriously started

The Reactor trips

No other effects of the fire are indicated at this time Which of the following procedures should be transitioned to NEXT?

A. DB-OP-02000, RPS, SFAS, SFRCS Trip or SG Tube Rupture B. DB-OP-02501, Serious Station Fire C. DB-OP-02508, Control Room Evacuation D. DB-OP-02519, Serious Control Room Fire Answer: B Explanation/Justification: Meets the requirements of the SRO only white paper Section II .E page 7 first bullet. SRO is required to know the actions contained in the abnormal procedures. The SRO must diagnose that equipment has spuriously actuated and that the procedure rules of use govern implementation of the serious station fire procedure.

A. Incorrect. This answer is plausible because in general, the correct procedure to implement following a Reactor Trip is DB-OP-02000.

B. Correct -Spurious operation of safety related equipment requires implementation of DB-OP-02501which takes priority over DB-OP-02000. See DB-OP-01003 R14 Operations Procedure Use Instructions step 6.5.2.a.

C. Incorrect. This answer is plausible because DB-OP-02519, Serious Station Fire Attachment 20 for Fire Area HH directs use of DB-OP-02508, Control Room Evacuation if the fire in area HH affects Control Room Habitability. In addition, a fire in the Control Room AC area could introduce smoke into the Control Room.

D. Incorrect. This answer is plausible because a fire in the Control Room AC area could introduce smoke into the Control Room, however the Control Room circuits would not be involved in the fire which would require use of DB-OP-02519, Serious Control Room Fire.

Sys # System Category KA Statement N/A N/A Generic Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions K/A# 2.4.5 K/A Importance 4.3 Exam Level SRO References provided to Candidate Technical

References:

DB-OP-01003 step 6.5.2, DB-OP-02501, step 2.1 page 9 and Attachment 20 page 110.

Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.13)

Objective:

(SRO ONLY)

Davis-Besse 1LOT15 NRC Written Exam r0 100 A large break Loss of Coolant Accident has occurred. An Equipment Operator is directed to perform DB-OP-02000, Attachment 7, Section 1, Action to Close Breakers for DH7A, DH7B, DH9A, DH9B, and HP31. The assigned operator reports that RE8426 and 8427, Spent Fuel Pool Area Radiation Monitors located near MCC F11B read 30 REM/hr. The operator must access F11B to complete Attachment 7.

Which one of the following Emergency Operating Procedure DB-OP-02000 Attachments should be directed to be performed based on the indicated dose rate?

A. Continue with performance of DB-OP-02000, Attachment 7, (Section 1), Transferring LPI Suctions to the Emergency Sump. Worst case conditions for the route provided have been assumed in the development of this attachment.

B. Stop DB-OP-02000 Attachment 7 and perform Attachment 11, HPI Flow Balancing instead. The indicated dose rate prevents access to MCC F11B. As a result, train 2 of HPI will be lost requiring HPI Flow Balancing.

C. Stop DB-OP-02000 Attachment 7 and perform DB-OP-02000 Attachment 14, Establishing HPI Alternate Minimum Recirc instead. The indicated dose rate prevents access to MCC F11B. As a result, the normal recirc flowpath for train 2 of HPI will be lost. The Alternate HPI recirc flowpath must be placed in service D. Stop DB-OP-02000 Attachment 7 and perform Attachment 22, Cross Connect LPI Pump Discharge instead. The indicated dose rate prevents access to MCC F11B. As a result, Train 2 of LPI will lost. LPI must be cross connected to mitigate a possible LPI line break.

Answer: A Explanation/Justification: Meets the requirements of the SRO only white paper Section II .D page 6 second bullet. SRO must evaluate plant conditions based upon the dose rates and select a procedurally driven course of action. Requires detailed knowledge of the procedure actions and basis for assumed dose levels of transit paths contained in the procedure.

A. Correct - DB-OP-02000 Attachment 7 Warning provides information that the assumed worst case dose rate for performance of this action is 34 REM/hr. Continuing with Attachment 7 will allow transfer of the ECCS Pump Suctions to the Emergency Sump without exceeding the projected 2 REM total dose for this activity.

B. Incorrect - Plausible because if Attachment 7 is not performed, the High Pressure Injection System would lose suction once the BWST is depleted. The actions to close the breakers are required because the supply breakers are open to prevent spurious Mispositioning during a fire.

Normally, only a single train is protected for each serious station fire area, so it is plausible that only a single train of HPI would be lost and therefore Flow Balancing would be required.

C. Incorrect - Plausible because if Attachment 7 is not performed, the High Pressure Injection System Train 1 Recirc Flowpath to the BWST via HP31 would not have power. As a result, the candidate could assume the alternate recirc flowpath must be used..

D. Incorrect - Plausible because if Attachment 7 is not performed, the Low Pressure Injection System would lose suction once the BWST is depleted. The actions to close the breakers are required because the supply breakers are open to prevent spurious Mispositioning during a fire.

Normally, only a single train is protected for each serious station fire area, so it is plausible that only a single train of LPI would be lost and therefore cross connecting HPI would be required.

Sys # System Category KA Statement N/A N/A Generic Knowledge of the operational implications of EOP warnings, cautions, and notes K/A# 2.4.20 K/A Importance 4.3 Exam Level SRO References provided to Candidate None Technical

References:

DB-OP-02000 Attachment 7 Warning Question Source: New Question Cognitive Level: High 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.13)

Objective:

ML15217A607: Difference between revisions

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