ML20054A571

Proposed Written Exam with Answer Key

ML20054A571
Person / Time
Site:	Byron
Issue date:	12/16/2019
From:	Joseph Demarshall Exelon Generation Co
To:	NRC/RGN-III/DRS/OLB
Demarshall J
Shared Package
ML17214A839	List: ML20054A024 ML20054A142 ML20054A158 ML20054A485 ML20054A536 ML20054A571 ML20054A653 ML20054A682 ML20054A700
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Download: ML20054A571 (166)

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EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 1 of 166 25 July 2019 1

ID: 2019 NRC-2 Q 01 Points: 1.00 A manual reactor trip and SI have been actuated on Unit 1:

A PZR safety valve has been stuck slightly open for ten minutes.

RCS pressure is 1900 psig and slowly lowering.

PZR vapor space temperature is 630 °F and stable.

PRT level is 80% and slowly rising.

PRT pressure is 35 psig and slowly rising.

Containment pressure is 4 psig and slowly rising.

Currently, safety valve tailpipe temperature indication should be reading approximately:

A.

200°F B.

230°F C.

265°F D.

282°F Answer:

D Answer Explanation Explanation: All values are based on using a constant enthalpy value of 1145 for 1900 psig.

A. Incorrect. Plausible if the applicant misapplies the Mollier diagram, throttling to the saturation curve.

B. Incorrect. Plausible if the applicant misapplies the Mollier diagram, throttling to containment pressure (~19 psia). The given conditions support the PRT rupture disc remaining intact, therefore, this is incorrect.

C. Incorrect. Plausible if the applicant misapplies the Mollier diagram, throttling to 35 psia and not 50 psia (35 psig).

D. Correct. Correctly applying the Mollier diagram using the 1900 psig (1915 psia) and applying the throttling process down to 35 psig (50 psia), 282°F is determined.

Technical Reference(s):

Steam Tables Fundamentals Lesson Plan 193004 RO level question because:

10 CFR Part 55 Content:

55.41 3; 55.43

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 2 of 166 25 July 2019 2

ID: 2019 NRC-2 Q 02 Points: 1.00 A Small Break Loss of Coolant Accident (LOCA) has just occurred and Safety Injection (SI) was actuated.

All systems have responded as designed.

Reactor Coolant Pump (RCP) Trip Criteria have NOT been met.

Without any additional operator action, which of the following RCP temperatures will rise during the next 10 minutes of RCP operation.

A.

Thrust Bearing B.

Motor Lower Radial Bearing C.

Motor Stator Winding D.

Pump Lower Radial Bearing Answer:

C Answer Explanation Explanation:

C. Correct: SI and Phase A actuations cause RCFCs to run in Low Speed without Chilled Water for cooling (SX only) and Post-LOCA environment will cause higher temperatures in Containment. The Motor Windings are air cooled from the Containment atmosphere.

A. and B. are incorrect because CCW flow increases post-SI (with 2 CC pumps running),

providing additional cooling for the oil coolers and causing bearing temperatures to be stable or drop.

D. is incorrect since the additional seal injection flow rate post-SI aids in cooling and temperature will be stable or drop.

Meets the K/A because it requires the examinee to know how the RCS is affected by a loss of a RCP.

RO level question because:

10 CFR Part 55 Content:

55.41

__3__

References:

Steam Tables, Fundamentals Lesson Plan 193004 Lesson Plan N-BY-TQ-ILT-RC-13, Reactor Coolant Pump

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 3 of 166 25 July 2019 3

ID: 2019 NRC-2 Q 03 Points: 1.00 A Loss of Coolant Accident (LOCA) occurred.

The Reactor TRIPPED.

An automatic Safety Injection Signal occurred.

High Head SI Flow is 500 gpm.

Reactor Coolant System pressure is 150 psia.

RWST level is 58% and LOWERING.

Pressurizer level is 0%.

The crew has transitioned to 1BEP-1, LOSS OF REACTOR OR SECONDARY COOLANT, and is currently at step 11, Initiate Evaluation of Plant Status.

1. Reactor Coolant Pumps (RCP) will be ___(1)___.

2. Residual Heat Removal Pumps (RH) will be ___(2)___.

A.

1. running

2. running B.

1. running

2. stopped C.

1. stopped

2. running D.

1. stopped

2. stopped Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 4 of 166 25 July 2019 Explanation:

The question places the examinee in a situation where a Large Break LOCA has occurred.

During the execution of BEP-1, Step 1 directs the Operator to review RCP Trip Criteria. In this case, a trip criterion is met based on RCS pressure less than 1425 psig with High Head Injection flow greater than 100 gpm. The RH pumps will continue to run because the transition criteria to 1BEP ES-1.1 SI Termination (step 6 of 1BEP-1) has not been met based on RCS inventory, specifically PZR level. Another "gate" for stopping RH pumps is at step 8, Check if RH pumps should be stopped. That gate is not met because RCS pressure is less than 325 psig.

A. Incorrect: The first part is incorrect but plausible because in BFR-S.1 (another Emergency Operating Procedure) there is a caution to NOT Trip RCPs even if trip criteria is met. The second part is correct as stated above.

B. Incorrect: The first part is incorrect but plausible because in BFR-S.1 (another Emergency Operating Procedure) there is a caution to NOT Trip RCPs even if trip criteria is met. The second part is incorrect but plausible as there are procedural steps contained in 1BEP-1 for tripping the RH pumps.

C. Correct: See above.

D. Incorrect: The first part is correct as stated above. The second part is incorrect but plausible as there are procedural steps contained in 1BEP-1 for tripping the RH pumps.

Meets the K/A because it requires the examinee to know the status of the RCP and RH pumps under the stated conditions following a LOCA.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

1BEP-1, Operator Action Summary (RCP trip criteria), steps 6 and 8 1BFR-S.1, first caution of procedure, page 2

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 5 of 166 25 July 2019 4

ID: 2019 NRC-2 Q 04 Points: 1.00 Unit 1 was operating at 100% power when a transient occurred:

Pressurizer level is 57% and slowly lowering.

VCT level is 39% and slowly lowering.

Charging flow indicated on 1FI-121A is offscale low with 1CV121 in AUTO.

Annunciator 1-7-B2, RCP SEAL WTR INJ FLOW LOW, is LIT.

Annunciator 1-7-C3, RCP SEAL LEAKOFF FLOW LOW, is LIT.

Annunciator 1-9-A1, REGEN HX LTDWN TEMP HIGH, is LIT.

Annunciator 1-9-D3, CHG LINE FLOW HIGH LOW, is LIT.

The crew has entered 1BOA PRI-1, Excessive Primary Plant Leakage Unit 1.

Which of the following best describes the event that has occurred?

A.

Letdown line isolation.

B.

Pressurizer PORV failed open.

C.

Charging line break upstream of 1CV121.

D.

Charging line break downstream of 1CV182.

Answer:

C Answer Explanation Explanation:

A. Incorrect. Letdown line isolation would cause PZR level to rise. While charging flow may lower to near minimum, regen Hx temperature alarms should not come in and RCP seal injection and seal leakoff flow alarms should not come in.

B. Incorrect. An open PZR PORV would cause a lowering PZR level accompanied by an increase in charging flow. This would not cause RCP seal injection alarms or regen Hx high temp alarms.

C. Correct. A charging line break upstream of 1CV121 would allow minimal flow past 1CV121, reducing normal makeup to the RCS, reducing RCP seal injection flow, and prevent charging from cooling letdown water through the regen Hx. Letdown flow would remain online while PZR and VCT levels are dropping, until manual or automatic isolation occurs.

D. Incorrect. A charging line break downstream of 1CV182 would not cause RCP seal injection flow to be low or seal leakoff flow to be low.

Technical Reference(s):

1BOA PRI-1 1BOA PRI-15 RO level question because:

10 CFR Part 55 Content:

55.41 6

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 6 of 166 25 July 2019 5

ID: 2019 NRC-2 Q 05 Points: 1.00 Unit 2 is in Mode 4:

2A Train RHR is providing shutdown cooling RCS pressure is 350 psig RCS temperature is 300°F RCS heatup rate is 5°F/hr RHR total flow is 3300 gpm 2RH606 is throttled to 52% OPEN (1500 gpm)

Flow transmitter 2FT618, 2A RH Discharge Flow, fails LOW with the flow controller for 2RH618, 2A RH HX Bypass Valve, in automatic.

Assuming NO operator actions are taken, which of the following describes the expected plant response, if any, due to this failure?

A.

RCS heatup rate will RISE.

B.

RCS heatup rate will LOWER.

C.

RCS pressure will LOWER.

D.

RCS temperature will REMAIN THE SAME.

Answer:

A Answer Explanation With the 2RH618 valve (flow control) in automatic, the response to a sensed low flow condition would be for the valve to open. When 2RH618 opens, this means that more flow bypasses the heat exchanger (lower resistance to flow through valve that is more open) and the rate of RCS heatup will be higher. The distractors are plausible misconceptions of system response and alignment during shutdown cooling.

A. Correct. See explanation above.

B. Incorrect. Plausible as this is the exact opposite of the actual plant response and may be selected by an applicant who misunderstands the system response to the failure direction of the instrument. This would be the case if 2FT618 had failed high.

C. Incorrect. Plausible because during solid plant operations, RHR letdown via HCV128 to the CVCS system would be used to maintain RCS pressure, thus allowing for a potential misconception regarding the nature of the associated system response. However, even if solid plant conditions were assumed, the effect of RCS heatup would be for pressure to begin to rise before being stabilized by the operation of components in the letdown pathway.

D. Incorrect. Plausible because 2RH618 (RHR H/X bypass valve) is in parallel with 2RH606 (RHR H/X outlet valve) and it may be believed that 2RH606 would operate to offset the operation of 2RH618. However, 2RH606 would require manual adjustment in order for this to occur.

Meets the K/A because it requires the examinee to describe the expected plant response for a Loss of Residual Heat Removal System heat exchangers.

RO level question because:

10 CFR Part 55 Content: 55.41 (b) (7)

References:

Lesson Plan N-BY-TQ-ILT-RH-18, Residual Heat Removal System, Rev. 10

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 7 of 166 25 July 2019 6

ID: 2019 NRC-2 Q 06 Points: 1.00 Unit 1 is at 100%.

Pressure Control Channel Select Switch is in the 455/456 position.

1PT455 fails low.

Which of the following would be expected for this malfunction?

1. P-11 LIT on the Bypass Permissive Panel

2. Annunciator 1-14-B1, OTT HIGH RX TRIP ALERT

3. Annunciator 1-12-B2, PZR PORV OR SAF VLV OPEN

4. Annunciator 1-12-A2, PZR PRESS HIGH RX TRIP STPT ALERT A.

1 and 2 ONLY B.

3 and 4 ONLY C.

2 and 3 ONLY D.

1, 2, and 4 ONLY Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 8 of 166 25 July 2019 Explanation:

A) Incorrect. Plausible because a controlling PZR Pressure instrument failing low will bring in annunciator 2, but P-11 requires 2 of 3 logic on channels 455, 456, and 457 which will not be met as 456 and 457 will actually see rising pressure once all pressurizer heaters turn on as a result of the malfunction.

B) Incorrect. Plausible as a controlling PZR Pressure instrument failing low will energize the backup heaters, raising system pressure until PORV 1RY456 opens, resulting in annunciator 3.

1RY456 cycling will control and maintain pressure below the pressurizer pressure high reactor trip setpoint (annunciator 4).

C) Correct. The controlling PZR Pressure instrument failing low (single pressure channel) will result in annunciators 2 and 3. All heaters will energize and raise pressure, resulting in the opening of PORV 1RY456. This PORV will control pressure below the reactor trip setpoint.

D) Incorrect. Plausible as annunciator 2 will be received. The controlling channel failing low will cause back up heaters to energize raising actual system pressure. 1RY456 cycling will control and maintain pressure below the pressurizer pressure high reactor trip setpoint (annunciator 4).

In addition, P-11 requires 2 of 3 logic on channels 455, 456, and 457 to actuate.

Meets the K/A because it requires the examinee to prioritize and interpret the significance of each annunciator or alarm for the Pressurizer malfunction.

RO level question because:

10 CFR Part 55 Content: 55.41 (b) (10) & 55.43 (b) (5)

References:

1-14-B1, OTT HIGH RX TRIP ALERT 1-12-B2, PZR PORV OR SAF VLV OPEN

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 9 of 166 25 July 2019 7

ID: 2019 NRC-2 Q 07 Points: 1.00 BFR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS, Step 10 checks for either reactor coolant system temperature or steam generator pressure dropping in an uncontrolled manner.

Which of the following describes the basis for performing these checks under Anticipated Transient without Scram (ATWS) conditions?

A.

Preventing a challenge to RCS integrity.

B.

Avoiding a significant addition of positive reactivity.

C.

Conserving inventories in steam generators.

D.

Assessing isolation of potential containment release pathways.

Answer:

B Answer Explanation Explanation: From BD-FR-S.1 step 10: An uncontrolled cooldown of the RCS is indicated by either an uncontrolled RCS temperature reduction or an uncontrolled SG pressure reduction.

Such an RCS cooldown could add a significant amount of positive reactivity to the core, depending of the current value of the moderator temperature coefficient.

A. Incorrect. Plausible because RCS integrity can be challenged by excessive cooldown conditions, as could be the case in a BFR-P.1 (Response to Imminent Pressurized Thermal Shock) scenario.

B. Correct. See explanation above.

C. Incorrect. Plausible because while maintaining steam generator inventories is a concern in BFR-S.1, this concern is addressed via the tripping of the main turbine in step 2 of the procedure.

D. Incorrect. Plausible because while the prevention of potential releases of radioactive materials from containment is a concern in BFR-S.1, it is addressed via the isolation of non-essential containment ventilation penetrations in step 5 of the procedure.

Meets the K/A because it requires the examinee to understand negative temperature coefficient and how temperature change effects the plant during an ATWS event.

RO level question because:

10 CFR Part 55 Content: 55.41 (b) (8) & (10)

References:

2BFR-S.1, Response to Nuclear Power Generation/ATWS Unit 2 BD-FR-S.1, Response to Nuclear Power Generation/ATWS

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 10 of 166 25 July 2019 8

ID: 2019 NRC-2 Q 08 Points: 1.00 Unit 2 has tripped from full power.

2A AF pump is OOS for maintenance.

All S/G pressures are 380 psig slowly lowering.

All S/G narrow levels are approximately 8% and slowly lowering.

The crew has entered 2BCA-2.1, Uncontrolled Depressurization of All S/Gs.

Which of the following AFW flow condition(s) to the 2C S/G meet or exceed the minimum AFW flow requirements required in 2BCA-2.1?

2B AF Train Condition 1) 90 gpm Condition 2) 45 gpm Condition 3) 30 gpm Condition 4) 15 gpm A.

Conditions 1, 2, 3 and 4.

B.

Conditions 1, 2 and 3 only.

C.

Conditions 1 and 2 only.

D.

Condition 1 only.

Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 11 of 166 25 July 2019 Explanation:

Per 2BCA-2.1 caution prior to step 2, A minimum feed flow of 45 gpm must be maintained to each SG with a narrow range level less than 14% (34% adverse cnmt). A minimum verifiable feed flow to the SG allows components to remain in a wet condition, minimizing any thermal shock effects if feed flow is raised.

Condition 1 is greater than the minimum flow and it is also the green banded indication on the MCB.

Condition 2 is equal to the minimum flow, per the caution.

Condition 3 is NOT greater than the minimum flow, but it is plausible because it is greater than the minimum required flowrate identified in the bases of 25 gpm, but the procedure adds a margin of error to ensure flow is greater than this value.

Condition 4 is NOT greater than great than the minimum flow, but it is plausible because minimum recovery flowrate for re-establishing AF flow per 2BOA SEC-6 Loss of Feedwater Tempering Line Subcooling.

A. Incorrect. See above. Plausible because all conditions are plausible as described above.

B. Incorrect. See above. Plausible because conditions 1, 2 and 3 are plausible as described.

C. Correct. See above.

D. Incorrect. See above. Plausible if the candidate fails to identify the minimum value of 45 gpm Meets the K/A because it requires the examinee to understand the flow limits to a S/G during an Uncontrolled Depressurization of all Steam Generators.

RO level question because:

10 CFR Part 55 Content: 55.41 (b) (7)

References:

2BCA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS UNIT 2 BD-CA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS N-BY-TQ-ILT-BCA-2.1 (BCA-2.1 Lesson Plan) 2BOA SEC-6, LOSS OF FEEDWATER TEMPERING LINE SUBCOOLING UNIT 2

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 12 of 166 25 July 2019 9

ID: 2019 NRC-2 Q 09 Points: 1.00 1BFR-H-1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, has been implemented following a loss of all feedwater event.

AFW flow could not initially be established Bleed and feed has been initiated CETs are rising ALL SG wide range levels are currently 20% and lowering Containment pressure is 6 psig The 1B AF pump has just been started,

1) How should AFW flow INITIALLY be controlled in accordance with 1BFR-H-1, AND

2) What is the basis for that action?

A.

1.

Feed only one SG at the maximum rate.

2. SGs are considered to be dry under the given conditions.

B.

1. Feed only one SG at the maximum rate.

2. SGs are NOT considered to be dry under the given conditions.

C.

1.

Feed all SGs at between 60-80 gpm.

2. SGs are considered to be dry under the given conditions.

D.

1. Feed all SGs at between 60-80 gpm.

2. SGs are NOT considered to be dry under the given conditions.

Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 13 of 166 25 July 2019 Explanation: Per a BFR-H-1, Attachment B, note a dry SG is any SG with WIDE RANGE level less than 10% (27% ADVERSE CNMT), and no feed flow established. Attachment B, Step 4 (which applies after the initiation of bleed and feed with rising core exit temperatures present) directs that IF all SGs are dry, THEN perform the following Feed ONE SG at maximum rate until core exit temperatures drop. Containment is considered Adverse at 5 psig. With Containment pressure indicating 6 psig, the Adverse values are required to be used.

A) Correct. See explanation above.

B) First part correct / second part incorrect. Plausible because, with a WR level of 20%, per the Attachment B note, SGs would not be considered to be dry in the absence of adverse containment conditions.

C) First part incorrect / second part correct. Plausible because, in the absence of rising CETCs, Attachment B, Step 3 would direct feeding at 60-80 gpm.

D) First part incorrect / second part incorrect. Plausible because, in the absence of rising CETCs, Attachment B, Step 3 would direct feeding at 60-80 gpm. Also, per the Attachment B note, with a WR level of 20%, SGs would not be considered to be dry in the absence of adverse containment conditions.

Meets the K/A becasue requires knowledge to determine the required flow and why that flow is necessary.

Technical Reference(s):

1BFR-H.1, Response to Loss of Secondary Heat Sink Unit 1, Rev. 302 BD-FR-H.1, Response to Loss of Secondary Heat, Rev. 301 RO Question 10 CFR Part 55 Content:

55.41 5, 10; 55.43

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 14 of 166 25 July 2019 10 ID: 2019 NRC-2 Q 10 Points: 1.00 Byron Station Units 1 and 2 were both operating at 100% when a tornado struck the site.

Units 1 and 2 have both tripped A Loss of All AC power has occurred NO Diesel Generators were able to be started Under these conditions, the primary load shed of DC buses must be completed with 65 minutes in order to ensure that the station batteries will last at least _________.

A.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> D.

16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> Answer:

C Answer Explanation Explanation: 1BCA-0.0 Attachment B (Extended Loss of All AC Power Response) is implemented if no 4 KV ESF bus is available on the opposite unit. If no diesel generator can be started, then this attachment directs that 1BFSG-4, ELAP DC Bus Load Shed/Management, be performed. A note at the beginning of 1BFSG-4 states that the primary load shed should be initiated within 35 minutes and should be completed within 65 minutes of the event to ensure the batteries will last at least 8 HOURS.

A) Incorrect. Plausible because Technical Specification 3.8.4, Condition D, contains a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> required action when one DC electrical power subsystem is inoperable for reasons other than Condition A, B, or C.

B) Incorrect. Plausible because 1BCA-0.0 contains a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> time criteria, however, it is associated with the initiation of RCS cooldown.

C) Correct. See explanation above.

D) Incorrect. Plausible because 1BCA-0.0 contains a 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> time criteria, however, it is associated with the initiation of RCS boration.

Technical Reference(s):

1BFSG-4, ELAP DC Bus Load Shed/Management, Unit 1, Rev. 2 1BCA-0.0, Loss of All AC Power, Unit 1, Rev. 302 Technical Specification 3.8.4, DC Sources-Operating, Amendment 198 RO level question because:

10 CFR Part 55 Content:

55.41 8, 10; 55.43

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 15 of 166 25 July 2019 11 ID: 2019 NRC-2 Q 11 Points: 1.00 Unit 1 is at 100%.

Unit 1 tripped when a Loss of Offsite Power occurred.

Both 4kV ESF busses were re-energized by their respective Diesel Generators within the Tech Spec allowable time limits.

The 1A AFW pump failed to automatically start.

The 1B AFW pump automatically started.

The crew has entered 1BEP-0, Reactor Trip or Safety Injection Unit 1, and 1BOA ELEC-4, Loss of Offsite Power Unit 1.

Which of the following describes the sequences for manually starting the 1A AFW Pump from 1PM06J?

A.

C/S is placed to START and released; the pump and the aux oil pump start at the same time.

B.

C/S is placed to START and released; the aux oil pump starts, the 1A AFW pump starts when oil pressure exceeds 8 psig.

C.

C/S is placed to START and held; the pump and the aux oil pump start at the same time; then the C/S can be released.

D.

C/S is placed to START and held; the aux oil pump starts, the 1A AFW pump starts when oil pressure exceeds 8 psig; then the C/S can be released.

Answer:

D Answer Explanation Explanation:

A. Incorrect. The C/S must be held in START until the AOP develops > 8psig oil pressure, then the 1A AFW pump will start.

B. Incorrect. The C/S must be held in START until oil pressure > 8 psig.

C. Incorrect. The 1A AFW pump will only start with oil pressure > 8 psig.

D. Correct. The C/S must be taken and held in the START position, allowing the AOP to start and build oil pressure. Once oil pressure > 8 psig, the 1A AFW pump starts.

Meets the K/A because it addresses the manual actuation of the 1A AF pump.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

Lesson Plan N-BY-TQ-ILT-AF-26

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 16 of 166 25 July 2019 12 ID: 2019 NRC-2 Q 12 Points: 1.00 At 1000, Unit 1 is in MODE 3 PZR pressure was 2235 psig PZR PORV 1RY455A is in AUTO PZR PORV 1RY456 is isolated due to excessive seat leakage.

At 1005:

Power is lost to Instrument Bus 111 A RCS pressure transient occurs PZR pressure is currently 2360 psig What is the current state of PZR PORV 1RY455A?

A.

Closed and can be manually opened.

B.

Closed and CANNOT be manually opened.

C.

Open and will close automatically when PZR pressure reaches 2185 psig.

D.

Open and will close automatically when PZR pressure reaches 2315 psig.

Answer:

A Answer Explanation Explanation: Per 1BOA ELEC-2, Table A, Loss of Inst Bus 111 Effects, PORV 1RY455A will not open in AUTO or ARM LOW TEMP under these conditions. Although the 1RY455A open setpoint of 2345 psig has been exceeded, it will not open in AUTO but can be manually opened.

A) Correct.

B) Incorrect. Plausible because it may be incorrectly believed that manual opening capability is also affected by the loss of Instrument Bus 111.

C) Incorrect. Plausible because it may be incorrectly believed that 1RY455A responds to the failure by opening and will not close until the 2185 psig PORV closure interlock is met.

D) Incorrect. Plausible because it may be incorrectly believed that 1RY455A responds to the failure by opening and will function as designed for closure (e.g. resets at 2315 psig).

Meets the K/A because it requires the examinee to know how the loss of AC instrument bus impacts the operation of the PZR PORVs RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(3) Mechanical components and design features of reactor primary system.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

1BOA ELEC-2, Loss of Instrument Bus Unit 1, Rev. 109 N-BY-TQ-ILT-RY-14, Pressurizer (RY), Rev. 11

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 17 of 166 25 July 2019 13 ID: 2019 NRC-2 Q 13 Points: 1.00 Unit 1 at 100% power.

The Unit 0 CC pump is mechanically and electrically aligned to Unit 2.

1A DG is OOS due to a fuel injection pump failure.

ALL other systems are in their normal alignment.

The following sequence of events occurs:

1B CC pump trips due to a fire in the motor.

SAT 142-2 trips due to a sudden pressure relay actuation.

ALL other equipment functions as designed.

What actions will the U1 Operators will take?

A.

Trip Unit 1 and enter 1BCA-0.0 LOSS OF ALL AC POWER UNIT 1.

B.

Trip Unit 1 and enter 1BEP-0 REACTOR TRIP OR SAFETY INJECTION.

C.

Continue operation at power and enter 1BOA ELEC-3 LOSS OF 4KV ESF BUS.

D.

Continue operation at power and enter 1BOA ELEC-4 LOSS OF OFFSITE POWER.

Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 18 of 166 25 July 2019 Explanation: The SAT trip will de-energize both SATs. This will cause a Loss Of Offsite power to U1. With 1A DG unavailable, Bus 141 will remain de-energized. Bus 142 will have power restored when 1B DG starts and syncs to the bus. With the loss of 1B CC pump, 1A CC pump de-energized, and U0 CC pump aligned to the opposite unit, U1 will have NO CC pumps. This will require a U1 RX trip per 1BOA PRI-6 COMPONENT COOLING MALFUNCTION UNIT 1 and continued actions in both simultaneously.

A. Incorrect. This is credible if the candidate assumes the SAT fault extends to the ESF buses, as a RX trip with a Loss Of Offsite will cause a Loss of All AC after the Generator trips.

B. Correct. See explanation above.

C. Incorrect. This is credible as it would be correct without the loss of all CC pumps. 1BOA ELEC-3 LOSS OF 4KV ESF BUS would be used to address the loss of Bus 141 since the 1A DG would be incapable of starting.

D. Incorrect. This is credible since the candidate could identify the SAT fault would de-energize both ESF buses.

Meets the K/A because it requires the examinee to know which procedure is the correct one to utilize during a loss of ALL CC pumps.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

1BOA PRI-6 1BEP-0 1BOA ELEC-3 1BCA-0.0

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 19 of 166 25 July 2019 14 ID: 2019 NRC-2 Q 14 Points: 1.00 Unit 1 is at 100% power.

Instrument Air is lost to 1CV182, Charging Header Pressure Control Valve.

The valve has failed ___(1)___ and seal injection flow has ___(2)___?

A.

1. open

2. risen B.

1. open

2. lowered C.

1. closed

2. risen D.

1. closed

2. lowered Answer:

B Answer Explanation Explanation:

A. Incorrect: 1CV182 does fail open but causes seal injection flow to decrease. If the applicant believes the valve opens for more seal injection flow, this is a plausible answer.

B. Correct: 1CV182 is in the charging line, downstream of the tap-off for seal injection. It is normally throttled partially open. When instrument air is lost, it fails open, allowing more flow down the charging line, and less seal injection flow.

C. Incorrect: If 1CV182 failed closed, it would cause seal injection flow to increase, so this is a plausible answer D. Incorrect: If the applicant thinks the valve fails closed on loss of IA, and that the valve is throttled closed for less seal injection flow, this is a plausible answer.

Meets the K/A because it requires the examinee to know how the fail position of the 1CV182 and the impact it has on the CVCS system.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety sytems, including insturmentation, signals, interlocks, failure modes, and automatic and manual features.

References:

1BOA SEC-4, Loss of Insturment Air Unit 1, Rev 108 Chapter 15a; I1-CV-XL-01, Chemiical and Volume Control System, Rev 13.

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 20 of 166 25 July 2019 15 ID: 2019 NRC-2 Q 15 Points: 1.00 Byron Station Unit 1 and 2 are at 100%.

Unit 1 reactive load is 250 MVAR.

Unit 2 reactive load is 250 MVAR.

Unit 2 trips.

Unit 1 Main Generator Reactive (MVAR) loading will (1)

and Field Current will (2)

A.

1. rise 250 MVAR

2. lower B.

1. rise less than 250 MVAR

2. lower C.

1. rise 250 MVAR

2. rise D.

1. rise less than 250 MVAR

2. rise Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 21 of 166 25 July 2019 Explanation:

All the answer choices are plausible since reactive load and field current will be impacted by the transient.

A. Incorrect; The reactive load that was being supplied by the tripped unit will be shared amongst the generators connected to the grid; therefore the reactive load increase on the operating unit will be less than 250 MVAR. The increased reactive load on the operating unit will result in a sensed drop in terminal voltage and the voltage regulator will increase the generator field current to return the terminal voltage to set-point.

B. Incorrect; The increased reactive load on the operating unit will result in a sensed drop in terminal voltage and the voltage regulator will increase the generator field current return the terminal voltage to set-point.

C. Incorrect; The reactive load that was being supplied by the tripped unit will be shared amongst the generators connected to the grid; therefore the reactive load increase on the operating unit will be less than 250 MVAR.

D. Correct; The reactive load that was being supplied by the tripped unit will be shared amongst the generators connected to the grid; therefore the reactive load increase on the operating unit will be less than 250 MVAR. The increased reactive load on the operating unit will result in a sensed drop in terminal voltage and the voltage regulator will increase the generator field current to return the terminal voltage to set-point.

Meets the K/A because it requires the examinee to know how the impact to reactive load of the generators when impacted by grid disturbances. Specifically a U2 RX trip wihich will provide over-excitation to the U1 generator.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(4) General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

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

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

I1-MP-XL-01 Ch 5; Main Generator; Rev 10 I1-MP-XL-02 Ch 6; Generator Excitation; Rev 8

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 22 of 166 25 July 2019 16 ID: 2019 NRC-2 Q 16 Points: 1.00 Unit 1 is at 100% power:

A reactor trip and safety injection occurred Abnormally high Auxiliary Building radiation trends are identified The crew has transitioned to 1BCA-1.2, LOCA Outside Containment RCS pressure is currently 1200 psig Which of the following describes why 1BCA-1.2 directs closing 1SI8809A, RH TO COLD LEGS 1A AND 1D ISOL VALVE, under these conditions?

A.

To establish a lineup for subsequent hot leg Injection.

B.

To determine if the leak is located in RHR discharge piping.

C.

To prevent RWST inventory from being lost to the environment.

D.

To maintain RHR discharge piping isolated when RH pumps are stopped.

Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 23 of 166 25 July 2019 Explanation: Per BD-CA-1.2, this step [BCA-1.2 step 2] instructs the operator to sequentially close and open all normally opened valves in paths that penetrate containment to identify and isolate the break outside containment the valving connecting the low pressure (600 psig) RH System to the high pressure (2500 psig) RCS is of primary interest, since the probability of a break occurring outside containment is most probable to occur in the low pressure RH System piping.

A.

Incorrect. Plausible because closing 1SI8809A isolates cold leg injection from RHR and would occur as part of aligning a RH flow path for hot leg injection. However, BCA-1.2 does not align hot leg injection (this alignment occurs in BEP ES-1.4).

B.

Correct. See explanation above.

C.

Incorrect. Plausible because BD-CA-1.2 does identify that the potential exists for RWST inventory loss. However, BD-CA-1.2 notes that the RWST could be drained to the auxiliary building if the RCS pressure is reduced to below the static head pressure in the RWST. In the question stem, RCS pressure is stated to be significantly greater than this value. Additionally, in such a scenario, the inventory loss would occur to the Auxiliary Building versus the environment.

D.

Incorrect. Plausible because RCS pressure is above RHR pump shutoff head and RHR pumps would be expected to be subsequently shut down in BEP-1. However, 1SI8809A is not closed due to securing RHR pumps. 1SI8809A would only remain closed in BCA-1.2 if closing the valve stopped the leak.

Meets the K/A because it requires the examinee to know the interaction between the RH system and the RCS. In addition, the examinee is required to understand the relationship of the system design and interaction with the impact of a leak in the system.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

1BCA-1.2, LOCA Outside Containment Unit 1, Rev. 300 BD-CA-1.2, LOCA Outside Containment, Rev. 300 N-BY-TQ-ILT-RH-18, Residual Heat Removal System, Rev. 10

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 24 of 166 25 July 2019 17 ID: 2019 NRC-2 Q 17 Points: 1.00 The crew is performing 2BCA-1.1, Loss of Emergency Coolant Recirculation.

RWST level is 35% and slowly lowering.

2B CV pump experiences oscillating discharge pressure and flow indications.

Which of the following actions should be taken FIRST, in accordance with 2BCA-1.1?

A.

Start the 2A CV pump.

B.

Secure the 2B CV pump to prevent potential pump damage.

C.

Transfer to 1BCA-1.3, Sump Blockage Control Room Guideline.

D.

Swap pump suction to the VCT and start a manual makeup to the VCT.

Answer:

B Answer Explanation Explanation:

A. Incorrect, step 13 of 1BCA-1.1 establishes one train of ECCS flow, in part by ensuring only one CV pump is running. At this point, the crew may elect to start the 2A CV pump, but the 2B CV pump would need to be secured due to a loss of suction.

B. Correct, the CAUTION prior to Step 1 states If suction is lost to any ECCS or CS pump(s),

the pump(s) should be stopped.

C. Incorrect, 1BCA-1.3 is entered if both RH trains are affected (unstable amps, flow or discharge pressure) such that at least one train of SI recirculation flow cannot be maintained.

D. Incorrect, this is done during Steps 28-30 of 1BCA-1.1 if RWST level drops below 9%.

Following the procedure requirements, the crew would have already Meets the K/A because it requires the examinee to know the actions of the emergency procedure to determine the proper course of action.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

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

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

2BCA-1.1 BD-CA-1.1

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 25 of 166 25 July 2019 18 ID: 2019 NRC-2 Q 18 Points: 1.00 Unit 1 was operating at 100% when a reactor trip occurred:

There is a loss of secondary heat sink.

The operating crew is establishing RCS bleed and feed in accordance with 1BFR-H.1, Response to Loss of Secondary Heat Sink.

PZR PORV 1RY455A has been opened.

PZR PORV 1RY456 failed to open and remains CLOSED.

Which of the following describes the consequences, if any, of the failure of PZR PORV 1RY456 to open under these conditions?

A.

An inadequate core cooling condition will occur regardless of further operator action.

B.

Adequate RCS heat removal will be provided by the single, open pressurizer PORV.

C.

Reactor head vent valves must be opened to provide additional bleed path capability.

D.

ECCS feed flow will provide adequate RCS heat removal regardless of bleed pathways.

Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 26 of 166 25 July 2019 Explanation: Per BFR-H.1 step 18, if both PZR PORVs cannot be opened then reactor head vent valves will be opened. Per BD-FR-H.1, if both PZR PORVs are not maintained open, the RCS may not depressurize sufficiently to permit adequate feed of subcooled ECCS flow to remove core decay heat. If core decay heat exceeds RCS bleed and feed heat removal capability, the RCS will repressurize rapidly, further reducing the feed of subcooled ECCS flow and resulting in a rapid reduction of RCS inventory. Although only one open PZR PORV may not be sufficient to maintain adequate RCS bleed flow, the operator should maintain one PZR PORV open, if possible, and open all RCS high point vents to provide additional bleed path capability.

A) Incorrect. Plausible because, ultimately, an inadequate core cooling condition (e.g. a red path on the core cooling critical safety function status tree) will be reached if actions within BFR-H.1 are not successful. However, additional BFR-H.1 actions are still available to the operators during the scenario presented in this question (i.e. opening reactor head vents, depressurizing steam generators, and feeding steam generators using condensate booster flow).

B) Incorrect. Plausible because an applicant may have the misconception that a single, open PZR PORV, absent any other RCS vent paths, will provide a sufficient vent pathway to sufficiently depressurize the RCS and permit adequate ECCS feed flow to remove core decay heat.

C) Correct. See explanation above.

D) Incorrect. Plausible because an applicant may have the misconception that ECCS feed flow, absent any RCS vent paths, will provide for adequate removal of core decay heat.

Meets the K/A because it requires the examinee to know the impact of operating the pressurizer PORVs for feed and bleed upon the RCS.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

1BFR-H.1, Response to Loss of Secondary Heat Sink Unit 1, Rev. 302 BD-FR-H.1, Response to Loss of Secondary Heat Sink, Rev. 301

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 27 of 166 25 July 2019 19 ID: 2019 NRC-2 Q 19 Points: 1.00 Unit 1 is at 45% power:

A ramp to 100% is in progress Control Bank "D" rods are at 185 steps on the step counters Rod Control is in AUTOMATIC The following sequence of events occurs:

TAVE CONT DEV LOW (1-14-E1)

Outward control rod movement is observed ROD CONT URGENT FAILURE (1-10-C6) alarms ROD DEV POWER RNG TILTS (1-10-A7) alarms Rod Control placed in MANUAL; Control Bank D step counters indicate 200 steps A blown Lift Coil fuse is discovered in Rod Control Power Cabinet 1BD

___________ Control Bank D rods are at 200 steps.

A.

No B.

All C.

All but one D.

Only Group 2 of Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 28 of 166 25 July 2019 Explanation:

A. Incorrect - Plausible since a Rod Control Urgent Failure due to a Logic Cabinet failure would terminate control rod motion. However the subsequent Rod Deviation/Power Range Tilts alarm indicates that control rod movement continued [Control Bank D Group 2 rods]

after the Urgent Failure.

B. Incorrect - The Rod Deviation/Power Range Tilts alarm indicates that one or more control rods in Control Bank D is lagging by at least 12 steps.

C. Incorrect - A blown Lift Coil fuse will affect more than one control rod.

D. Correct - The blown Lift Coil fuse will prevent movement of Control Bank D Group 1 rods.

Meets the K/A because it requires the examinee to know the impact of a failure to the rod control system and the resulting impact of that failure on rod portion.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

I1-RD-XL-01/ Chapter 28; Rod Control System; Rev 5a I1-PI-XL-01/ Chapter 29; Rod Position Indication System; Rev 7 BAR 1-10-C6; ROD CONT URGENT FAILURE; Rev 8 BAR 1-10-A7; ROD DEV POWER RNG TILTS; Rev 23

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 29 of 166 25 July 2019 20 ID: 2019 NRC-2 Q 20 Points: 1.00 During performance of the 1BOSR 1.4.2-1, MOVEABLE CONTROL ASSEMBLIES QUARTERLY SURVEILLANCE, Rod P4 became misaligned from the other rods in Shutdown Bank A.

Maintenance repairs and rod realignment are expected to be completed within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Rod P-4 indicates 208 steps on DRPI.

All other rods in Shutdown Bank A indicate 228 steps on both the Step Counters and DRPI.

1. Which of the following is the maximum power level allowed during recovery of Rod P4 per 1BOA ROD-3, DROPPED OR MISALIGNED ROD?

2. This power limit is established to ensure __ (2) __ when realigning the rod.

A.

1. 49%

2. reactor power remains < 100%

B.

1. 74%

2. reactor power remains < 100%

C.

1. 49%

2. localized power distribution limits are not exceeded D.

1. 74%

2. localized power distribution limits are not exceeded Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 30 of 166 25 July 2019 Explanation:

A) Incorrect. Plausible because reactor power would be reduced to < 50% for a rod misalignment of greater than 12 steps when recovery will take greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The reason for maintaining power below the correct value of 75% is not to ensure power remains below 100%,

but rather the impact on core power distribution, i.e., the potential for exceeding Power Distribution Monitoring System (PDMS) Limits when recovering a rod misaligned by an amount as significant as 20 steps.

B) Incorrect. Plausible as < 75% power is correct for the given conditions. The reason for maintaining power below 75% is not to ensure power remains below 100%, but rather the impact on core power distribution, i.e., the potential for exceeding PDMS Limits when recovering a rod misaligned by an amount as significant as 20 steps.

C) Incorrect. Plausible because reactor power would be reduced to < 50% for a rod misalignment of greater than 12 steps when recovery will take greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The second portion of the answer is correct.

D) Correct. IAW BOA ROD-3, maintain power <75% for a rod misaligned by greater than 12 steps when rod recovery will take less than four hours. Maintaining power below 75% during realignment of the rod ensures that PDMS Limits will not be exceeded.

Meets the K/A because it requires the examinee to know the necessary power limits to restore proper rod position.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

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

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

1BOA ROD-3, DROPPED OR MISALIGNED ROD UNIT 1

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 31 of 166 25 July 2019 21 ID: 2019 NRC-2 Q 21 Points: 1.00 Unit 2 operating at 100%

A reactor trip occurs.

Two control rods do not fully insert into the core.

Emergency boration is initiated per 2BOA PRI-2, EMERGENCY BORATION.

2CV8104, Emergency Boration Vlv indicates OPEN 2FI-183A, 2CV8104 downstream flow indicator reads 20 gpm 2FI-121A, charging header flow indicator reads 132 gpm What is the next required operator action?

A.

Continue with emergency boration as aligned until P-6 is LIT.

B.

Stop the boric acid transfer pump and realign charging pump suction to the RWST.

C.

Establish emergency boration via cold leg injection by aligning to the RWST and opening 2SI8801A.

D.

Direct equipment operators to manually bypass the boric acid filter by opening 2AB8458, Boric Acid Filter Supply to Boric Acid Tank Valve.

Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 32 of 166 25 July 2019 Explanation:

A) Incorrect. Plausible because 2BFR S.2 would have you continue to emergency borate with a slightly positive IR SUR, until the P-6 interlock is LIT. In this case, the present alignment is not delivering adequate boration flow (< 30 gpm) IAW 2BOA PRI-2.

B) Correct. With < 30 gpm emergency boration flow and indication of flow through 2CV121, 2BOA PRI-2 has the operator realign emergency boration from the RWST through the normal charging injection path.

C) Incorrect. Plausible as this is an alternate emergency boration pathway, but since flow exists through 2CV121 as indicated on 2FI-121A, emergency boration through the cold leg injection pathway is not required.

D) Incorrect. Plausible as this would be the correct answer if emergency boration from the RWST was not available to the operators.

Meets the K/A because it requires the examinee to identify the fault with the indicated flow and know that a manual alignment change is necessary.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

2BOA PR-2, EMERGENCY BORATION 2BFR-S.2, RESPONSE TO LOSS OF CORE SHUTDOWN

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 33 of 166 25 July 2019 22 ID: 2019 NRC-2 Q 22 Points: 1.00 Unit 1 tripped from 100% power 30 minutes ago.

While checking the Source Range High Flux trips per step 12 of 1BGP 100-5, Plant Shutdown and Cooldown:

Intermediate Range Channel N-35 indicates 6 x 10-5 % RTP and stable Intermediate Range Channel N-36 indicates 1 x 10-6 % RTP and stable Assuming that NO additional operator actions have been taken, which of the following describes BOTH:

1) The malfunction that has occurred.

AND

2) The current status of Source Range High Flux Trips.

A.

1. N-35 is not indicating correctly

2. Source Range High Flux Trips are unblocked.

B.

1. N-35 is not indicating correctly

2. Source Range High Flux Trips are blocked.

C.

1. N-36 is not indicating correctly

2. Source Range High Flux Trips are unblocked.

D.

1. N-36 is not indicating correctly

2. Source Range High Flux Trips are blocked.

Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 34 of 166 25 July 2019 Explanation: First part: the timeframe of 30 minutes post-trip provided in the question is sufficiently long that intermediate range readings would already be expected to be below 5 x 10-6 % RTP (based upon a -1/3 DPM post-trip SUR), thus indicating that a problem exists with N-35 (which is indicating abnormally high for the expected conditions). Second part: per 1BGP 100-5, step 12, when both Intermediate Range Channels indicate less than 5 x 10-6% RTP, ENSURE the Source Range High Flux Trips are automatically unblocked. N-BY-TQ-ILT-NI-31 discusses that the Source Range Hi Flux Trip is automatically reinstated during shutdown when both IR channels below are below the P-6 setpoint with P-6, in turn automatically resetting when 2 of 2 IR channels are < 5 x 10-6 % power.

A) First part is correct / second part is incorrect. Plausible because an applicant may incorrectly conclude that having only a single IR channel (N-36) at < 5 x 10-6 % power is sufficient to reset P-6 and unblock the Source Range High Flux Trips.

B) Correct. See explanation above.

C) First part is incorrect / second part is incorrect. Plausible because an applicant may conclude that the IR channel that is indicating a minimum value (N-36) has failed low. Additionally, an applicant may incorrectly conclude that having only a single IR channel (N-36) at < 5 x 10-6 %

power is sufficient to reset P-6 and unblock the Source Range High Flux Trips.

D) First part is incorrect / second part is correct. Plausible because an applicant may conclude that the IR channel that is indicating a minimum value (N-36) has failed low.

Meets the K/A because it requires the examinee know the required overlap of the SR and IR indication range in order to discern the answer.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(2) General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

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

References:

BGP 100-5, Plant Shutdown and Cooldown, Rev. 73 N-BY-TQ-ILT-NI-31, Gamma-Metric Source and Intermediate Range Nuclear Instrumentation, Rev. 6

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 35 of 166 25 July 2019 23 ID: 2019 NRC-2 Q 23 Points: 1.00 In 1BOA, SEC-3, Loss of Condenser Vacuum, Step 3.a RNO directs the operator to locally open Condensate Polisher Bypass valve 1CD210A (CD210A (401 +5 H5 TB1)).

If the grid below represents the Turbine Building Unit 1, the valve will be found somewhere in the square marked ___(1)___.

In order to locally open the valve, the operator must ___(2)___.

A.

1. X

2. Fail air to the valve B.

1. Y

2. Fail air to the valve C.

1. X

2. Declutch the MOV operator and use the handwheel D.

1. Y

2. Declutch the MOV operator and use the handwheel Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 36 of 166 25 July 2019 Explanation: This convention is defined in BAP 199-3, Procedure for Identifying Locations and Areas Local operation is described in BOP CD/CB-3, step 7.b A. (1) Correct (2) Correct B. (1) Incorrect. Credible because a corner of this square is also defined by the stated row and column.

(2) correct C. (1) Correct (2) Incorrect. Credible if the applicant believes the valve is an MOV.

D. (1) Incorrect. See B(1)

(2) Incorrect. See C(2)

Question meets K/A - The examinee must know what type of valve the 1CD210A is and how to locally manipulate the valve.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

BAP 199-3, Procedure for Identifying Locations and Areas Rev. 001 BOP CD-CB-3, Condensate/Condensate Booster System Fill and Vent, Rev. 36 step 7.b.

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 37 of 166 25 July 2019 24 ID: 2019 NRC-2 Q 24 Points: 1.00 Chemistry reports that a recent sample from the 0A Waste Gas Decay Tank indicates the tank contains 1 x 105 Curies of activity.

1. Based upon these conditions, entry into 0BOA RAD-3, DECAY TANK HIGH ACTIVITY, _____

required.

2. The gas decay tank activity limit ensures that the whole body exposure to a member of the public at the nearest site boundary will not exceed _____ in the event of a gas decay tank rupture.

A.

1. is

2. 0.5 rem B.

1. is

2. 25 rem C.

1. is NOT

2. 0.5 rem D.

1. is NOT

2. 25 rem Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 38 of 166 25 July 2019 Explanation:

A) Correct. 0BOA RAD-3 entry criteria is met when samples indicate that a gas decay tank contains greater than 5 x 104 Curies. Restricting the quantity of radioactivity contained in each gas decay tank provides assurance that, in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a member of the public at the nearest site boundary will not exceed 0.5 rem.

B) First part is correct / second part is incorrect. Restricting the quantity of radioactivity contained in each gas decay tank provides assurance that, in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a member of the public at the nearest site boundary will not exceed 0.5 rem. Plausible because the 10 CFR 100.11 whole body radiation dose limit for an individual located at the site boundary during a postulated fission product release is 25 rem.

C) First part is incorrect / second part is correct. 0BOA RAD-3 entry criteria is met when samples indicate that a gas decay tank contains greater than 5 x 104 Curies. Plausible because the entry criteria threshold may not be correctly recalled.

D) Both parts are incorrect. 0BOA RAD-3 entry criteria is met when samples indicate that a gas decay tank contains greater than 5 x 104 Curies. Restricting the quantity of radioactivity contained in each gas decay tank provides assurance that, in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a member of the public at the nearest site boundary will not exceed 0.5 rem. Plausible because the entry criteria threshold may not be correctly recalled and also because the 10 CFR 100.11 whole body radiation dose limit for an individual located at the site boundary during a postulated fission product release is 25 rem.

Meets the K/A because it requires the examinee to the know the limits for radiation allowed in the tank to limit/prevent a release to the public.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(8) Components, capacity, and functions of emergency systems.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

ILT - 060LSYS, Radioactive Waste Gas System, Rev. 5 0BOA RAD-3, Decay Tank High Activity, Rev. 100 TRM 3.11.f, Dose Rate for Gaseous Effluent TRM Appendix L, Explosive Gas & Storage Tank Radioactivity Monitoring Program 10 CFR 20.1201, Occupational Dose Limits for Adults 10 CFR.11, Determination of Exclusion Area, Low Population Zone, and Population Center Distance

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 39 of 166 25 July 2019 25 ID: 2019 NRC-2 Q 25 Points: 1.00 Which ONE of the following conditions would result in the Containment Air Lock Limiting Condition for Operation (LCO) NOT being satisfied?

A.

The Equipment Hatch Personnel Airlock outer door interlock is discovered to be non-functional.

B.

Maintenance personnel open the Equipment Hatch Personnel Airlock outer door without prior approval.

C.

CNMT HATCH DOOR SEAL TROUBLE (1-1-B2) alarms (local pressure indicator reads 3.0 psig and stable); the Equipment Hatch Personnel Airlock has NOT been entered.

D.

CNMT HATCH DOOR SEAL TROUBLE (1-1-B2) alarm takes 10 minutes to reset following closure of the Equipment Hatch Personnel Airlock outer door.

Answer:

A Answer Explanation Explanation:

A. Correct - The airlock door interlocks are required to be operable in Modes 1 - 4 per T.S.

3.6.2.

B. Incorrect - Plausible since there are administrative requirements (BAP 1450-1) for Containment entry. One airlock door at a time may be opened for normal entry / exit while in Modes 1 - 4 per T.S. 3.6.2.

C. Incorrect - Plausible since the alarm may indicate a door seal failure. The stable pressure reading is indicative of leak detection system malfunction. An inoperable leak detection system alone does not make the airlock inoperable.

D. Incorrect - Plausible since the alarm may indicate a door seal failure. No action is required unless the alarm does not reset within 15 minutes.

Meets the K/A because it requires the examinee to the know the operability conditions for the containment door hatches.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

Tech Spec 3.6.2 & Basis 3.6.2 BAP 1450-1, Access to Containment; Rev 44 BAR 1-1-B2; CNMT HATCH DOOR SEAL TROUBLE; Rev 5 1BOL PC-1; LCOAR CONTAINMENT AIRLOCK DOOR SEALS; Rev 6 I1-PC-XL-01/ Chapter 40

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 40 of 166 25 July 2019 26 ID: 2019 NRC-2 Q 26 Points: 1.00 Unit 1 is implementing 1BFR-C.2, Response to Degraded Core Cooling.

The crew has reached step 9, Depressurize all Intact SGs to 160 psig.

The cooldown should proceed at (1).

The reason for stopping the depressurization at 160 psig is to prevent (2)

A.

1. less than 100°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period

2. a RED priority on the Integrity Status Tree B.

1. less than 100°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period

2. accumulator nitrogen injection C.

1. the maximum rate available

2. a RED priority on the Integrity Status Tree D.

1. the maximum rate available

2. accumulator nitrogen injection Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 41 of 166 25 July 2019 Explanation:

The cooldown rate is limited to less than 100°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period per Step 9.a. The reason for stopping at 160 psig is in the Basis for Step 9. BD-FR-C.2, Bases page 18 of 37, states To prevent accumulator nitrogen injection, the operator should stop the secondary depressurization when SG pressure reaches the minimum SG pressure which prevents accumulator nitrogen injection and when at least two RCS hot leg temperatures fall below the associated temperature to prevent accumulator nitrogen injection. A steam generator pressure limit is set to preclude significant nitrogen injection into the RCS. To determine the steam generator pressure limit, an ideal gas expansion calculation should be performed The value of 160 psig is the calculated steam generator pressure limit.

A. (1) Correct.

(2) Incorrect. The caution before step 9 says that accumulator injection may cause a red path condition in INTEGRITY, but the cooldown should still be completed. The answer is plausible because the cooldown may lead to a red path in INTEGRITY.

B. (1) Correct. See explanation above.

(2) Correct. See explanation above.

C. (1) Incorrect. Plausible because this the required cooldown rate in 1BFR-C.1, Response to Inadequate Core Cooling.

(2) Incorrect. See A(2)

D. (1) Incorrect. See C(1)

(2) Correct.

Meets the K/A because it requires the examinee to know the reason the actions are taken per the procedure and their impact on the primary.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

1BFR-C.2, Response to Degraded Core Cooling. Rev. 300 BD-FR-C.2, Background for 1BFR-C.2, Rev. 300

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 42 of 166 25 July 2019 27 ID: 2019 NRC-2 Q 27 Points: 1.00 Unit 1 was operating at 100% power when a reactor trip and loss of offsite power occurred.

During the subsequent performance of a natural circulation cooldown, the formation of a steam void in the Reactor Vessel head was identified by the crew.

RVLIS is currently INOPERABLE.

Which of the following indications will be used to monitor the growth of the steam void per 1BEP ES-0.4, NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN VESSEL (WITHOUT RVLIS) UNIT 1?

A.

Pressurizer pressure B.

Core Exit Thermocouples C.

Pressurizer level D.

RCS Hot Leg temperatures Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 43 of 166 25 July 2019 Explanation:

A) Incorrect. The note preceding step 8 of 1BEP ES-0.4 states that the upper head region may void during RCS depressurization. This will result in a rapidly rising PZR level. Additionally, Pressurizer level is used to assess whether additional action is required due to the status of a steam void in the Reactor Vessel head in step 19 as well. Plausible because this same transient may also result in changes in pressurizer pressure which, although a monitored parameter during this procedure, is not the indication specified for this purpose.

B) Incorrect. The note preceding step 8 of 1BEP ES-0.4 states that the upper head region may void during RCS depressurization. This will result in a rapidly rising PZR level. Additionally, Pressurizer level is used to assess whether additional action is required due to the status of a steam void in the Reactor Vessel head in step 19 as well. Plausible because this same transient may also lead to less cooling of the core (thus causing CET temperatures to rise), which, although a monitored parameter during this procedure, is not the indication specified for this purpose.

C) Correct.

D) Incorrect. The note preceding step 8 of 1BEP ES-0.4 states that the upper head region may void during RCS depressurization. This will result in a rapidly rising PZR level. Additionally, Pressurizer level is used to assess whether additional action is required due to the status of a steam void in the Reactor Vessel head in step 19 as well. Plausible because this same transient may also lead to less cooling of the core (thus causing RCS Hot Leg temperatures to rise),

which, although a monitored parameter during this procedure, is not the indication specified for this purpose.

Meets the K/A because it requires the examinee to the know the instrumentation necessary for monitoring the initialization of Natural Circulation and how it is monitored during emergency conditions.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

1BEP ES-0.4, NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN VESSEL (WITHOUT RVLIS) UNIT 1, REV. 300

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 44 of 166 25 July 2019 28 ID: 2019 NRC-2 Q 28 Points: 1.00 Unit 1 is implementing 1BGP 100-1, Plant Heatup.

Reactor Coolant loops are filled and vented.

The crew is preparing to start the first Reactor Coolant Pump.

In accordance with BOP RC-1, Startup of a Reactor Coolant Pump, to start the RCP, the lowest allowable RCS pressure is ___(1)___ psig and the lowest allowable VCT pressure is ___(2)___ psig.

A.

1. 325

2. 5 B.

1. 325

2. 15 C.

1. 300

2. 5 D.

1. 300

2. 15 Answer:

D Answer Explanation Explanation:

These limits are stated in the RCP start procedure.

A. (1) Incorrect. Plausible because this is the minimum pressure if loops are not filled and vented.

(2) Incorrect. Plausible because the lesson plan says that 5 psig is adequate, but the procedure requires 15 psig.

B. (1) Incorrect. See A(1)

(2) Correct.

C. (1) Correct.

(2) Incorrect. See A(2)

D. (1) Correct (2) Correct.

Meets the K/A because it requires the examinee to the know the RCP start criteria.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(3) Mechanical components and design features of reactor primary system.

References:

BOP RC-1, Startup of a Reactor Coolant Pump, Revision 46

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 45 of 166 25 July 2019 29 ID: 2019 NRC-2 Q 29 Points: 1.00 The reactor is at full power.

The 1A Regenerative HX is in service.

The DC control power fuse for 1CV8324A, Chg to Regen HX 1A Isol Valve, opens.

Thirty (30) seconds later, total RCP seal injection flow has ___(1)____ and normal letdown flow is

___(2)____.

A.

1. risen

2. unchanged B.

1. remained constant

2. 0 gpm C.

1. risen

2. 0 gpm D.

1. remained constant

2. unchanged Answer:

C Answer Explanation Explanation:

The 1CV8324A fails closed which will cause all charging flow to go through the RCP seal injection line causing total seal injection flow to rise. The 1CV8324A is interlocked with the 1CV8389A (letdown line isolation to regen Hx valve), so when the 1CV8324A goes closed the 1CV8389A will go closed isolating letdown line flow.

A. Incorrect, plausible as the applicant may not understand the interlock between 1CV8324A and 1CV8389A.

B. Incorrect, plausible as the applicant may believe that 1CV8384A is an MOV that fails as-is, rather than an AOV that fails closed.

C. Correct, see explanation above.

D. Incorrect, plausible if the applicant does not understand the failure mode of 1CV8384A or the interlock between 1CV8384A and 1CV8389A.

Meets the K/A because it requires the examinee to the know the impact of the faailure of 1CV8324A has on RCP seal injection..

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

N-BY-TQ-ILT-CV-15A

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 46 of 166 25 July 2019 30 ID: 2019 NRC-2 Q 30 Points: 1.00 The crew is preparing to initiate letdown flow through a mixed bed demineralizer that has had new resin installed.

Per BOP CV-8, CV SYSTEM MIXED-BED DEMINERALIZER OPERATION, demineralizer effluent will be diverted to the Holdup Tank (HUT) when initially placing the demineralizer in service to...

A.

verify that the demineralizer decontamination factor is adequate.

B.

minimize subsequent effects on the pH of the RCS.

C.

borate the demineralizer to prevent an RCS dilution.

D.

prevent the introduction of resin into the RCS.

Answer:

C Answer Explanation Explanation:

A) Incorrect. While demineralizer decontamination factor is a relevant concept within this context, its verification is not required prior to placing a mixed bed demineralizer in service per BOP CV-8.

B) Incorrect. According to N-BY-TQ-ILT-CV-15A, while this a concern for Li-OH resin (note that the question is written assuming HOH resin), cation resin is Li saturated in order to minimize effects on pH as unwanted ions are removed from the coolant.

C) Correct. Borate the demineralizer to prevent an RCS dilution is correct per BOP CV-8 since leaving it aligned to the VCT would lower RCS boron concentration rapidly.

D) Incorrect. While the introduction of resin into the RCS is a concern, per BOP CV-8, the Reactor Coolant Filter shall be in service when demineralizers are in service to prevent introducing resin into the RCS.

Meets the K/A because it requires the examinee to predict the malfunction will remove boron and create an uncontrolled dilution, which is addressed in the BOP.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

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

References:

BOP CV-8, CV System Mixed-Bed Demineralizer Operation, Rev. 44 N-BY-TQ-ILT-CV-15A, Chemical and Volume Control System, Rev. 13

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 47 of 166 25 July 2019 31 ID: 2019 NRC-2 Q 31 Points: 1.00 Unit 1 is at 380°F with an RCS cooldown in progress.

Under these conditions, recirculating the 1B RH train back to the RWST using 1RH8735, RH Supply to SI Pump Suct Hdr, (RH Recirc to RWST), to equalize the RH and RCS boron concentrations is A.

acceptable, as long as the 1A RH train is aligned for injection.

B.

acceptable, as long as a dedicated operator is stationed at 1RH8735.

C.

NOT acceptable due to rendering BOTH ECCS trains inoperable.

D.

NOT acceptable due to the possible rise in RWST water temperature above the Tech Spec limit.

Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 48 of 166 25 July 2019 Explanation:

See BOP RH-5, Step 15 NOTE.

With Unit 1 at 380°F, the unit is in Mode 3 (>350°F). Both trains of RH are required to be operable in Mode 3. In this Mode, 1RH8735 is NOT allowed to be opened under any circumstances because it makes BOTH trains of RH inoperable. Both trains are inoperable because in the case of an SI actuation, full RH injection flow from either train is not possible, as with 1RH8716A&B open, some flow from both trains would flow back to the RWST instead of injecting into the RCS.

Below 350°F (Mode 4), 1RH8735 is allowed to be opened as long as an operator is stationed at the valve in continuous communication with the control room, and the RCS Thot suction isolation valves are closed.

A. Incorrect, this would be true in Mode 4, however in Mode 3, 1RH8735 cannot be opened.

B. Incorrect, this would be true in Mode 4, however in Mode 3, 1RH8735 cannot be opened.

C. Correct, opening 1RH8735 in Mode 3 would render both ECCS trains (both RH trains) inoperable as described above.

D. Incorrect, the RWST water would be recirculated through the RH system back to the RWST, while being cooled through the RH HX. Operators are required to throttle _RH606/607 as required to maintain RWST temperature between 35°F and 100°F, per BOP RH-5.

Meets the K/A because it requires the examinee to know the effects of the 1RH8735 positions and the impact to Tech Specs during modes of applicability.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

BOP RH-5

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 49 of 166 25 July 2019 32 ID: 2019 NRC-2 Q 32 Points: 1.00 Unit 1 is operating at 100%.

A LOCA occurs resulting in a Safety Injection.

Immediately thereafter, Bus 142 experienced an electrical fault and REMAINS de-energized.

RWST levels currently indicate:

1LT-930: 45%

1LT-931: 50%

1LT-932: 50%

1LT-933: 45%

The CURRENT expected state of the CNMT SUMP ISOL valves are:

1SI8811A ___(1)___ and 1SI8811B___(2)___?

A.

1. Closed

2. Closed B.

1. Opening

2. Closed C.

1. Closed

2. Opening D.

1. Opening

2. Opening Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 50 of 166 25 July 2019 Explanation (Optional):

A) First part is incorrect / second part is correct. 1SI8811A/B are automatically opened by 1LT-930/931/932/933 lowering to 46.7% (Lo-2) on a 2-out-of-4 coincidence. Since 1SI8811A has power available, it would be expected to be cycling open in response to the conditions provided.

B) Correct.

C) Both parts are incorrect. 1SI8811A/B are automatically opened by 1LT-930/931/932/933 lowering to 46.7% (Lo-2) on a 2-out-of-4 coincidence. Since 1SI8811A has power available, it would be expected to be cycling open in response to the conditions provided.

1SI8811B is powered from 132X4 which is electrically downstream of the de-energized Bus 142; while this valve would have a demand signal present for opening under the conditions provided, it has no power to do so and, thus, would remain in its normal, closed position.

D) First part is correct / second part is incorrect. 1SI8811A/B are automatically opened by 1LT-930/931/932/933 lowering to 46.7% (Lo-2) on a 2-out-of-4 coincidence. 1SI8811B is powered from 132X4 which is electrically downstream of the de-energized Bus 142; while this valve would have a demand signal present for opening under the conditions provided, it has no power to do so and, thus, would remain in its normal, closed position.

Meets the K/A because it requires the examinee to know the automatic actuation of the 1SI8811A/B valves due to an SI in conjunction with a low RWST level..

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

BAR 1-6-B7, RWST LEVEL LO-2, Rev. 7 1BEP ES-1.3, Transfer to Cold Leg Recirculation, Rev. 302 BOP SI-E1, Safety Injection System Electrical Lineup (Unit 1), Rev. 9 N-BY-TQ-ILT-ECCS-58, Emergency Core Cooling System, Rev. 10

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 51 of 166 25 July 2019 33 ID: 2019 NRC-2 Q 33 Points: 1.00 Given the following conditions on Unit 1:

RCS pressure is 330 psig and stable.

RCS temperature is 240°F and stable (as indicated on the CETC's).

1A RH train is operating in shutdown cooling.

1B RH train is in standby aligned for injection.

1A CV pump is operating with 125 gpm charging flow that had been manually raised by the operator attempting to maintain PZR level.

1B CV pump is administratively OOS.

RH letdown has been secured from the 1A RH train.

RCP Seal Return flow is approximately 5 gpm total.

Subcooling is acceptable per the iconic PZR Level has stabilized at 20%.

The source of the leak has NOT yet been determined.

Which of the following actions will the operator be procedurally directed to perform in response to these plant conditions?

A.

Start the 1B RH pump to restore core cooling.

B.

Open 1SI8801A and 1SI8801B to restore Pressurizer level to 25%.

C.

Secure the 1A RH pump and close the 1RH8701A or 1RH8701B to isolate the leak ONLY.

D.

Place the 1B RH train in shutdown cooling AND secure the 1A RH pump and close the 1RH8701A or 1RH8701B to isolate the leak.

Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 52 of 166 25 July 2019 Explanation:

With RH letdown isolated and charging flow > 120 gpm, charging flow is inadequate. The crew would enter 1BOA S/D-2 (or transition to from 1BOA PRI-1, Step 4) due to charging greater than 120 gpm (with letdown isolated) and being in Mode 4. Per 1BOA S/D-2, the mitigative strategy is to verify adequate PZR level and subcooling, secure the RH train operating in shutdown cooling, and check that train for leaks, while the opposite train remains aligned for injection.

A. Incorrect, this action is required if CETCs were rising. In this case, RCS temp is stable and subcooling is adequate. (1BOA S/D-2 Step 12&13)

B. Incorrect, aligning for high head injection through 1SI8801A/B is done when charging flow is inadequate (> 120 gpm) and IF PZR level or RCS subcooling can NOT be maintained. In this case, PZR level is stable and subscooling is adequate. (1BOA S/D-2 Step 4)

C. Correct, with adequate subcooling and PZR level, the correct action is to isolate the RH train operating in shutdown cooling and monitor subcooling (1BOA S/D-2 Step 11)

D. Incorrect, this action is required if a leak is determined to exist in the operating RH train (1BOA S/D-2 Step 15&16). In this case, 1B RH train is aligned for cold leg injection and is not aligned for shutdown cooling unless a leak is determined to exist.

Meets the K/A because it requires the examinee to know the mitigation actions required to address the loss of coolant accident with shutdown cooling in service.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

1BOA PRI-1 1BOA S/D-2

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 53 of 166 25 July 2019 34 ID: 2019 NRC-2 Q 34 Points: 1.00 Unit 1 is operating at 56% power.

1BOA PRI-1, EXCESSIVE PRIMARY PLANT LEAKAGE UNIT 1, is being implemented due to suspected leakage via Pressurizer Safety Valve 1RY8010A.

The crew is currently performing 1BOA PRI-1, step 28, CHECK PRT CONDITIONS.

Assuming NO further operator action, which of the following is the LATEST time at which the Pressurizer Relief Tank rupture disks would be expected to rupture?

A.

1156 B.

1148 C.

1143 D.

1131 Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 54 of 166 25 July 2019 Explanation: the PRT rupture disks are expected to rupture when a pressure of 100 psig is reached inside the PRT. Per the provided Plant Computer Data, PRT pressure is approximately 27 psig at time 1119. Using the data for PRT pressure, the rate of PRT pressure increase can be estimated to be approximately [(27.1 psig - 7.5 psig) / 10 minutes] = 1.96 psig/min. Based upon this, it will take approximately [(100 psig - 27.1 psig) / 1.96 psig/min] = 37.2 minutes for the PRT rupture disks to rupture, which corresponds to a time of 1156 (1119 + 37 minutes).

A. Correct. See explanation above.

B. Incorrect. Plausible because the applicant may incorrectly use the provided Plant Computer Data. The graphical data to the immediate left of PRT pressure on the Plant Computer Data handout is PRT level. Incorrectly referencing the PRT level graph for pressure values would result in a determination of approximately 82.9 psig at time 1119, with a rate of increase of approximately [82.9 psig - 77 psig) / 10 minutes] = 5.9 psig/min. Based upon this, it would take approximately [(100 psig - 82.9 psig) / 0.59 psig/min] = 28.98 minutes for the PRT rupture disks to rupture, which corresponds to a time of 1148 (1119 + 29 minutes).

C. Incorrect. Plausible because the applicant may incorrectly use the provided Plant Computer Data. The graphical data for PRT pressure contains highlighted values along its pressure axis, however, these represent the minimum and maximum values of the displayed scale. Incorrectly interpreting these highlighted values as the pressure values at times 1109 and 1119 respectively (which corresponds to the duration of the displayed data) would result in a determination of 0 psig at time 1109, 30 psig at a time of 1119, and a rate of pressure increase of approximately

[30 psig - 0 psig) / 10 minutes] = 3 psig/min. Based upon this, it would take approximately

[(100 psig - 27.1 psig) / 3 psig/min] = 24.3 minutes for the PRT rupture disks to rupture, which corresponds to a time of 1143 (1119 + 24 minutes).

D. Incorrect. Plausible because the applicant may not correctly recall the pressure value at which PRT rupture disks are expected to rupture. The PRT is designed to condense and cool a design basis discharge without pressure exceeding 50 psig. Per the provided Plant Computer Data, PRT pressure is approximately 27.1 psig at time 1119. Using the graphic data for PRT pressure, the rate of PRT pressure increase can be estimated to be approximately [(27.1 psig - 7.5 psig) /

10 minutes] = 1.96 psig/min. If 50 psig is incorrectly assumed to also be the pressure at which the PRT rupture disks will rupture, then it would take approximately [(50 psig - 27.1 psig) / 1.96 psig/min] = 11.68 minutes for the PRT rupture disks to rupture, which corresponds to a time of 1131 (1119 + 12 minutes).

Question meets K/A - The examinee must be able to interpret the information from the handout, which is a screen shot of the computer display, to discern the system issue and calculate the amount of time until the PRT rupture setpoint is reached.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

1BOA PRI-1, Excessive Primary Plant Leakage Unit 1, Rev. 110 N-BY-TQ-ILT-RY-14, Pressurizer, Rev. 11 References to be provided to applicants during examination:

Plant Computer Data Handout for PRT

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 55 of 166 25 July 2019 35 ID: 2019 NRC-2 Q 35 Points: 1.00 Unit 1 was at rated thermal power when:

CC SURGE TANK LEVEL HIGH LOW (1-2-A5) alarms CC surge tank level is 67% and slowly rising Unit 1 CC HX process monitor 1RT-PR009 shows High radiation VCT level is dropping at approximately 1%/min.

This leakage can be stopped by isolating the ___(1)___ heat exchanger.

Prior to isolating the leak, LCO 3.4.13, RCS Operational Leakage, ___(2)___ met.

A.

1. seal water

2. IS B.

1. seal water

2. IS NOT C.

1. letdown

2. IS D.

1. letdown

2. IS NOT Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 56 of 166 25 July 2019 Explanation:

According to 1BOA PRI-6, Attachment B, seal water HX leakage results in rising VCT level.

According to 1BOSR 4.13.1-2, each 1% of VCT level corresponds to 19.3 gallons. A drop in VCT level at 1%/min corresponds to a leak rate of > 19 gal/min. This exceeds all LCO 3.4.13 values. With the leakage in progress, SR 3.4.13.1 (water inventory balance) is not met so the LCO is not met. This is RO above the line knowledge.

A. (1) Incorrect. See explanation. Credible if applicant does not know pressure relationship between component cooling water and RCS at the seal water HX.

(2) Incorrect. See explanation. Credible if the applicant reasons that the letdown HX is not part of the RCS.

B. (1) Incorrect. See A(1)

(2) Correct C. (1) Correct (2) Incorrect. A(2)

D. (1) Correct. See explanation above.

(2) Correct. See explanation above.

Meets the K/A because it requires the examinee to determine the location of an inleakage from the RCS into the CC system and the impact of the leakage on Tech Specs for leakage.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

1BOA PRI-6, Component Cooling Malfunction, Rev. 113.

1BOSR 4.13.1-2, RCS Leakrate Manual Calculation, Rev. 6 LCO 3.4.13

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 57 of 166 25 July 2019 36 ID: 2019 NRC-2 Q 36 Points: 1.00 Byron Unit 2 startup is in progress with the plant in MODE 2.

The Bus 244 feed breaker supplying pressurizer heaters trips.

1) Which Pressurizer Heater groups are lost?

2) What is the status of the Technical Specification LCO for the Pressurizer?

A.

1. A and D

2. Met B.

1. A and D

2. NOT met C.

1. B and C

2. Met D.

1. B and C

2. NOT met Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 58 of 166 25 July 2019 Explanation:

A. Incorrect - Pressurizer Heater groups A and D are powered from Bus 243. The Tech Spec LCO is NOT met since Pressurizer Heater groups A and D are powered from the same bus and thus incapable of being power from redundant power sources.

B. Incorrect - Pressurizer Heater groups A and D are powered from Bus 243. The Tech Spec LCO is NOT met since Pressurizer Heater groups A and D are powered from the same bus and thus incapable of being power from redundant power sources.

C. Incorrect - Pressurizer Heat groups B and C are powered from Bus 244, but the Tech Spec LCO is NOT met since Pressurizer Heater groups A and D are powered from the same bus and thus incapable of being power from redundant power sources.

D. Correct - Pressurizer Heat groups B and C are powered from Bus 244. The Tech Spec LCO is NOT met since Pressurizer Heater groups A and D are powered from the same bus and thus incapable of being power from redundant power sources.

Meets the K/A because it requires the examinee to know the power supplies to the PZR heaters and how many are required in order to satisfy the LCO which is defined above the line.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

Technical Specification LCO 3.4.9 and associated basis.

N-BY-TQ-ILT-RY-14; Pressurizer (RY); Rev 11

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 59 of 166 25 July 2019 37 ID: 2019 NRC-2 Q 37 Points: 1.00 Unit 2 was operating at 22% power when a step load rejection occurred resulting in RCS pressure rising to 2370 psig.

RCS pressure is CURRENTLY 2320 psig and lowering.

Which of the following is a plant response CURRENTLY expected for these conditions?

A.

Pressurizer variable heaters modulating on.

B.

Pressurizer high pressure reactor trip.

C.

Pressurizer PORVs in the open position.

D.

Pressurizer spray valves modulating closed.

Answer:

C Answer Explanation Explanation:

A) Incorrect. Variable heaters modulate within the band of 2220 - 2250 psig.

B) Incorrect. A reactor trip on high pressurizer pressure occurs at 2385 psig.

C) Correct. The pressure setpoints associated with the operation of PORV 2RY456 are 2315 -

2335 psig. For PORV 2RY455A, the associated setpoints are 2315 - 2345 psig.

D) Incorrect. Spray valves modulate within the band of 2260 - 2310 psig.

Meets the K/A because it requires the examinee to know the automatic pressure setpoints for Pressurizer Heaters, Sprays and PORVs and how they respond when reaching the setpoint.

Also, what the setpoint for a Reactor trip is for Pressurizer pressure.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

Technical Specification LCO 3.4.9 and associated basis.

N-BY-TQ-ILT-RY-14; Pressurizer (RY); Rev 11

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 60 of 166 25 July 2019 38 ID: 2019 NRC-2 Q 38 Points: 1.00 At 10:00, Unit 2 is at 75% power with rod control in automatic.

At 10:02 Control rods automatically insert A Rod Stop annunciator alarms What failure would cause both conditions?

A.

Loop 2C Tave fails HIGH.

B.

PR channel N-42 fails HIGH.

C.

Tref programmer output fails HIGH.

D.

Pimp channel 2PT-505 fails HIGH.

Answer:

B Answer Explanation Explanation:

A. Incorrect, Tave channel failing high will cause rods to insert, but does not cause an immediate rod stop actuation (i.e. a not a direct result).

B. Correct, N-42 becomes auctioneered high, this causes a rate mismatch and rods will insert.

Rod stop C-2 will actuate due to 1 of 4 PR channels > 103%.

C. Incorrect, Tref programmer failing high would cause rods to withdraw as Tave > Tref.

D. Incorrect, Pimp failing high would result in a false high Tref, a power rate of change and Tave <

Tref, causing rods to withdraw.

Meets the K/A because it requires the examinee to understand the impact of the instrument /

channel failure has on the rod drive system.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

I1-NI-XL-03 Ch 33 Power Range NI Lesson Plan, N-BY-TQ-ILT-RD-28 Rod Control Lesson Plan

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 61 of 166 25 July 2019 39 ID: 2019 NRC-2 Q 39 Points: 1.00 Unit 1 is conducting a normal unit shutdown from rated thermal power.

At 30% power during the load decrease, Intermediate Range Nuclear Instrument channel N-36 fails FULL-SCALE HIGH.

NO corrective action is taken for the N-36 failure. Which of the following describes:

1) The immediate response of the Reactor Protection System to the failure; AND

2) The expected subsequent impact of the failure on the Reactor Protection and/or Nuclear Instrumentation Systems.

A.

1. The reactor trips immediately.

2. When power lowers to below 5E-6%, the Source Range High Flux trip is automatically unblocked.

B.

1. The reactor does not trip immediately.

2. The reactor will trip when power lowers to less than 5E-6%.

C.

1. The reactor trips immediately.

2. When power lowers to below 10%, the Source Range High Flux trip is automatically unblocked.

D.

1. The reactor does not trip immediately.

2. The reactor will trip when power lowers to less than 10%.

Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 62 of 166 25 July 2019 Explanation: The Intermediate Range High Flux Trip setpoint and coincidence logic is 1-out-of-2 Intermediate Range NIs at 25% power. The Intermediate Range High Flux Trip is manually blocked during a startup after the P-10 interlock has been met. This trip is automatically unblocked below the P-10 interlock setpoint. The P-10 interlock is actuated by 2-out-of 4 Power Range NIs detecting >10% power under conditions of rising power and is cleared by 3-out-of-4 Power Range NIs detecting <10% power under conditions of lowering power (thus reinstating the Intermediate Range trip, as well as the SR Hi Flux trip if also <P-6 interlock). The Source Range High Flux Trip is auto unblocked when power drops < P-6. Clearing the P-6 interlock requires 2-out-of-2 Intermediate Range NIs to detect <5 x 10-6% power.

A) Both parts incorrect. First part discussion: the Intermediate Range High Flux Trip would have previously been manually blocked (once the P-10 interlock had been met) in order for the reactor to have been at the power level provided in the stem. A reactor trip would not occur due to Intermediate Range High Flux until the P-10 interlock was subsequently cleared. Second part discussion: if both the P-10 interlock has been cleared (3-out-of-4 Power Range NIs detecting <10% power) and the P-6 interlock has also been cleared (2-out-of-2 Intermediate Range NIs detect <5 x 10-6%), then the Source Range High Flux Trip would be automatically unblocked. Since N-36 has failed high, the Source Range High Flux Trip would not be automatically unblocked.

B) First part correct / second part incorrect. See explanation above for first half discussion.

Second half discussion: if both the P-10 interlock has been cleared (3-out-of-4 Power Range NIs detecting <10% power) and the P-6 interlock has also been cleared (2-out-of-2 Intermediate Range NIs detect <5 x 10-6%), then both trains of the Source Range High Flux Trip would be automatically unblocked. However, under the given scenario, the reactor would instead have already tripped when power lowered <10% due to P-10 clearing and reinstating the Intermediate Range High Flux trip with failed high N-36 meeting the required 1-out-of-2 trip coincidence (setpoint is >25% RTP, which a failed high IR channel would satisfy).

C) Both parts incorrect. First part discussion: the Intermediate Range High Flux Trip would have previously been manually blocked (once the P-10 interlock had been met) in order for the reactor to have been at the power level provided in the stem. A reactor trip would not occur due to Intermediate Range High Flux until the P-10 interlock was subsequently cleared. Second part discussion: if both the P-10 interlock has been cleared (3-out-of-4 Power Range NIs detecting

<10% power) and the P-6 interlock has also been cleared (2-out-of-2 Intermediate Range NIs detect <5 x 10-6%), then both trains of the Source Range High Flux Trip would be automatically unblocked. Since the P-6 interlock would not have been cleared under this condition, the Source Range High Flux Trip would not have been automatically unblocked.

D) Both parts correct. See explanation above.

Meets the K/A because it requires the examinee to know the setpoints for the RPS and when the setpoint is active.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

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

References:

N-06OLSYS, Reactor Protection System, Rev. 5

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 63 of 166 25 July 2019 40 ID: 2019 NRC-2 Q 40 Points: 1.00 Unit 1 is performing a plant cooldown in accordance with 1BGP 100-5, Plant Shutdown and Cooldown.

RCS pressure 1800 psig.

RCS temperature 530°F.

All steam generator pressures 865 psig.

RCS pressure rose to 2000 psig, then returned to 1800 psig.

What actuations, if any, occurred?

A.

A Safety Injection ONLY.

B.

A Safety Injection and a Main Steam Isolation.

C.

A Main Steam Isolation ONLY.

D.

No actuations occurred.

Answer:

A Answer Explanation In order to be at the conditions in the stem, SI has been blocked IAW step F.32 of 1BGP 100-5.

The pressure excursion caused P-11 to clear, which unblocked SI and steam line isolation.

The return to 1800 psig caused SI to actuate, but steam pressure remained above the steam line isolation setpoint.

A. Correct.

B. Incorrect. Plausible because the steam line isolation/SI block is also removed.

C. Incorrect. Plausible because the pressurizer pressure SI was previously blocked.

D. Incorrect. Plausible due to failure to recognize the auto unblock of the SI actuation due to momentary loss of P-11.

Meets the K/A because it requires the examinee to know the setpoints for the actuation, the block value, and the unblock value of the Safety Injection function.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

1BGP 100-5, Plant Shutdown and Cooldown Rev 73 Lesson plan I1-EF-XL-01/Ch 61

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 64 of 166 25 July 2019 41 ID: 2019 NRC-2 Q 41 Points: 1.00 Unit 2 is at 100% rated thermal power when the following indications are received; Annunciator SOLID STATE PROT CAB GENERAL WARNING (2-4-B3) lights.

At Solid State Protection Cabinet A, the following lights are lit on the Tester Card:

GND EXT A red dot appears in the lens cap of the ground return fuse.

The crew would first ___(1)___. If conditions for a safety injection signal are subsequently generated, required train A components ___(2)___ operate properly.

A.

1. ENTER 2BEP 0, Reactor Trip or Safety Injection

2. WILL NOT B.

1. ENTER 2BEP 0, Reactor Trip or Safety Injection

2. WILL C.

1. REFER to BOP RP-6, Solid State Protection Cabinet Ground Return Fuse Replacement

2. WILL NOT D.

1. REFER to BOP RP-6, Solid State Protection Cabinet Ground Return Fuse Replacement

2. WILL Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 65 of 166 25 July 2019 Proposed references to be provided to applicants during examination:

BAR 2-4-B3 SOLID STATE PROT CAB GENERAL WARNING Explanation:

The alarm response shows that the GND and EXT lights means that the ground return fuse is open. This is effectively a loss of power to all 26 VDC input relays (K/A match). Byron Station Horsenotes SSPS-2 shows the effect of this blown fuse. In this condition, it is impossible for train A to automatically generate any output. It is system knowledge that no single failure that results in a general warning alarm will cause a reactor trip. This question tests the ability to predict the impact (no Rx trip generated, but all SSPS functions disabled). Selection of procedure BOP RP-6 will correct the consequences.

E. (1) Incorrect. Plausible because a general warning on both trains of SSPS will cause a reactor trip.

(2) Correct. See explanation F. (1) Incorrect. See A(1)

(2) Incorrect. Plausible because most general warning alarms do not mean loss of function.

G. (1) Correct. See explanation (2) Correct. See explanation H. (1) Correct. See explanation (2) Incorrect. See B(2)

Meets the K/A because it requires the examinee to know effects of a loss of DC inside the SSPS cabinet and the impact it will have on status of the unit.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

1BAR 2-4-B3 SOLID STATE PROT CAB GENERAL WARNING, Rev 5 I1-RP-XL-01 Chapter 60a, SSPS, p. 18 of 67 Horsenotes SSPS-2

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 66 of 166 25 July 2019 42 ID: 2019 NRC-2 Q 42 Points: 1.00 Unit 1 at 100% full power.

1A, 1B, 1C RCFCs are running in high speed.

1D RCFC is in standby.

Reactor trip and SI occur due to a Large Break LOCA.

After the SI actuation signal, what is the response of all RCFCs?

A.

ALL RCFCs start immediately in low speed.

B.

ALL RCFCs start 20 seconds later in low speed.

C.

1A, 1B, and 1C RCFCs start immediately in low speed. 1D RCFC starts after 20 seconds in low speed.

D.

1A, 1B, and 1C RCFCs start 20 seconds later in low speed. 1D RCFC starts immediately in low speed.

Answer:

B Answer Explanation Explanation:

BOP VP-5, Step E.5. There are two 20 second timers in the RCFC control circuit. An SI signal trips the high speed breaker, and at the same time, starts a 20 second timer to close the low speed breaker. The other timer is started by the high speed breaker being open (52/b contact). Both of these timers together guarantee that a fan running in high speed has 20 seconds to slow down before the low speed breaker closes. The result is that, either way (fan running in high speed or shutdown), all low speed breakers close 20 seconds after the SI signal.

A. Incorrect, 20 second time delay as described above.

B. Correct, see answer description above.

C. Incorrect, 20 second time delay as described above.

D. Incorrect, 20 second time delay as described above.

Meets the K/A because it requires the examinee to know the starting interlocks of the RCFCs in relation to an ESF (SI) actuation.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

6E-1-4030VP01-08, BOP VP-5

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 67 of 166 25 July 2019 43 ID: 2019 NRC-2 Q 43 Points: 1.00 Unit 1 was operating at 100% power when a LOCA occurred.

Annunciator 1-3-B4, CNMT PRESS HI-3, is lit.

1A CS pump is stopped with its control switch in the AFTER TRIP position.

1B CS pump is OOS for bearing replacement.

1CS019A, EDUCTOR 1A SPRAY ADD VALVE, is closed.

1CS007A, PUMP 1A HDR ISOL VALVE, is open.

Which of the following describes the MINIMUM control switch manipulation(s) required to start the 1A CS pump AND deliver CS flow to containment?

Place the...

A.

1A CS Test Switch to TEST, then place 1CS019A control switch to OPEN.

B.

1CS019A control switch to OPEN, then place 1A CS Pump control switch to START.

C.

1A CS Test Switch to TEST, then place 1A CS Pump control switch to START.

D.

1CS019A control switch to OPEN.

Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 68 of 166 25 July 2019 Explanation: In 1BEP-0 Attachment B (SI Verification), step 10, if containment pressure has risen to >20 psig (provided in the stem by the CNMT PRESS HI-3 condition) and Group 6 CS monitor lights are not lit, then CS flow will be manually actuated. Per 1BEP-0 Attachment C (Manual CS Actuation), step 1.c, if CS educator spray additive valves are not open, then the CS pump test switches are placed in TEST, followed by manually opening 1CS019A/B. CS Pump 1A/1B auto starts if the following interlock conditions are met: control switch in the AFTER TRIP or AFTER CLOSE positions, 1CS019A/B open, and a Containment Spray Actuation signal present. 1CS019A/B automatically opens upon CS Actuation, but can also be manually opened by placing the Test Switch (43TS A/B) in TEST and placing the valves control switch in OPEN. In the scenario given, as soon as the 1CS019A full open limit switch is made up, the last interlock required for a CS pump start is made up, and the 1A CS pump will start.

A. Correct. See explanation above.

B. Incorrect. For CS Pump 1A to start, 1CS019A must first be open. Manually opening 1CS019A is accomplished by first placing the Test Switch (43TS A) in TEST and then by placing the valves control switch in OPEN. This distractor is based upon a potential misconception regarding the required sequence of operations.

C. Incorrect. For CS Pump 1A to start, 1CS019A must first be open. Manually opening 1CS019A is accomplished by first placing the Test Switch (43TS A) in TEST and then by placing the valves control switch in OPEN. This distractor is based upon a potential misconception regarding the required sequence of operations.

D. Incorrect. For CS Pump 1A to start, 1CS019A must first be open. Manually opening 1CS019A is accomplished by first placing the Test Switch (43TS A) in TEST and then by placing the valves control switch in OPEN. This distractor is based upon a potential misconception regarding the required sequence of operations Meets the K/A because it requires the examinee to know how to manually actuate Containment Spray to containment.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

N-BY-TQ-ILT-CS-59, Containment Spray System, Rev. 8 1BEP-0, Reactor Trip or Safety Injection Unit 1, Rev. 302

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 69 of 166 25 July 2019 44 ID: 2019 NRC-2 Q 44 Points: 1.00 Unit 1 is at 9% power.

1BGP 100-3, Power Ascension is in progress RCS temperature is being controlled by the steam dumps.

What is the change in actual plant parameters if the controlling Main Steam Header Pressure transmitter (1PT-507) fails high?

Steam header pressure Reactor Power A.

Rises Lowers B.

Lowers Rises C.

Rises Rises D.

Lowers Lowers Answer:

B Answer Explanation Explanation:

A. Incorrect. Steam pressure does not rise and reactor power does not decrease.

B. Correct, at 9% power, the steam dump controls remain in steam pressure mode (1BGP 100-3T1). Actual steam header pressure will decrease as steam dump demand increases in an attempt to lower steam header pressure as the controlling channel has failed high.

Increasing steam demand will result in raising reactor power.

C. Incorrect, steam pressure change is not consistent with increased demand.

D. Incorrect, reactor power does not decrease.

Meets the K/A because it requires the examinee to know how to manually actuate Containment Spray to containment.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

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

References:

N-BY-TQ-ILT-DU-24 Steam Dump Lesson Plan 1BGP 100-3T1 1BGP 100-3

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 70 of 166 25 July 2019 45 ID: 2019 NRC-2 Q 45 Points: 1.00 Unit 1 was operating at 100% power when a reactor trip occurred.

1BEP ES-0.1, REACTOR TRIP RESPONSE UNIT 1, is being implemented.

The crew is currently performing 1BEP ES-0.1, step 5, CHECK FW STATUS.

RCS Tavg is currently 575°F and lowering.

Assuming NO further operator action, which of the following is the HIGHEST temperature at which Feedwater Regulating Valves 1FW510, 1FW520 1FW530, and 1FW540 would be expected to close automatically?

A.

564°F B.

557°F C.

550°F D.

545°F Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 71 of 166 25 July 2019 Explanation: an automatic feedwater isolation is designed to occur on a reactor trip (P-4 interlock signal) in coincidence with a Low Tavg of < 564°F (on 2/4 RCS loops). This feedwater isolation limits cold water addition to the SGs and the cooldown rate of RCS following a reactor trip, thus limiting thermal stress and reactivity addition. This feedwater isolation signal will result, in part, in the automatic closure of Feedwater Regulating (Control) Valves 1FW510, 1FW520, 1FW530, and 1FW540. 1BEP ES-0.1, step 5, checks if a feedwater isolation has occurred with TAVE < 564°F and directs subsequent manual actions to isolate feedwater if needed.

A) Correct. See explanation above.

B) Incorrect. 557°F is the RCS Tavg value used in 1BEP ES-0.1, step 1 RNO, as a decision point to determine if the dumping of steam should be secured and total feed flow should be controlled (i.e. reduced). This value is plausible because it is both relatively close to the correct answer and is an important temperature value associated with plant operations.

C) Incorrect. 550°F is the minimum Tavg value for criticality per TS 3.4.2. This value is plausible because it is both relatively close to the correct answer and is an important temperature value associated with plant operations.

D) Incorrect. 545°F is the RCS Tavg value used in 1BEP ES-0.1, step 1 RNO, as a decision point to determine if an emergency boration should performed. This value is plausible because it is both relatively close to the correct answer and is an important temperature value associated with plant operations.

Meets the K/A because it requires the examinee to know the RCS temperature impact on feedwater.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(4) Secondary coolant and auxiliary systems that affect the facility.

(5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

N-BY-TQ-ILT-CD-FW-25, Cond/FW System, Rev. 12 1BEP ES-0.1, REACTOR TRIP RESPONSE UNIT 1, Rev. 300 Technical Specification 3.4.2, RCS Minimum Temperature for Criticality, Amendment 171

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 72 of 166 25 July 2019 46 ID: 2019 NRC-2 Q 46 Points: 1.00 Unit 1 is at full power.

The main feedwater line experienced a large rupture immediately upstream of 1FW510 1A Steam Generator Feedwater Reg Valve.

Unit 1 tripped due to Lo2 Steam Generator level

1) What state will the contents of the Main FW Line enter the turbine building?

2) After completing 1BEP-0, Reactor Trip or Safety Injection, to what procedure will the crew transition?

A.

1. All steam

2. 1BEP-2, Faulted Steam Generator Isolation B.

1. All steam

2. 1BEP ES-0.1, Reactor Trip Response C.

1. Part steam, part liquid

2. 1BEP-2, Faulted Steam Generator Isolation D.

1. Part steam, part liquid

2. 1BEP ES-0.1, Reactor Trip Response Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 73 of 166 25 July 2019 Explanation:

For the feedwater conditions, hfg is larger than hf. This means that there is not enough energy in the liquid for it all to be converted to steam. This question requires system knowledge to determine that when the reactor trips, the steam generators will be isolated from the break, so SG pressure should remain stable. The question requires procedure knowledge to determine that a reactor trip in this condition will lead the crew to the reactor trip response procedure.

A. (1) Incorrect. Plausible if the applicant believes that the hot feedwater will all flash to steam.

(2) Incorrect. Plausible if the applicant believes that the steam generator(s) will continue to feed the break.

B. (1) Incorrect. See A(1)

(2) Correct C. (1) Correct.

(2) Incorrect. See A(2)

D. (1) Correct.

(2) Correct.

Meets the K/A because it requires the examinee to know how the rupture will manifest itself in the Turbine Building and which procedure will be required to mitigate this event.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(4) Secondary coolant and auxiliary systems that affect the facility.

References:

Steam tables.

BEP-0, Reactor Trip or Safety Injection

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 74 of 166 25 July 2019 47 ID: 2019 NRC-2 Q 47 Points: 1.00 At 10:00, Unit 2 was at 5% rated thermal power, starting up from a refueling outage.

At 10:02, the crew manually tripped the reactor because of multiple dropped control rods.

Compared to a trip from 100%, the feedwater flow required for temperature control is ___(1)___.

For the expected conditions, 2BEP ES-0.1, Reactor Trip Response, allows reducing total feed flow to a minimum of ___(2)___ gpm.

A.

1. lower

2. zero B.

1. the same

2. zero C.

1. lower

2. >500 D.

1. the same

2. >500 Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 75 of 166 25 July 2019 Explanation:

The heat removal requirement for feedwater flow is a function of decay heat plus RCP heat.

For the listed conditions, there is very little decay heat compared to a trip from RTP. Since the initial power level was so low, there is little to no shrink on the trip, so SG levels are expected to be > 14 NR. In this case, total flow may be reduced to zero if necessary to control SG level.

A. (1) Correct (2) Correct B. (1) Incorrect. Plausible if the applicant does not know that decay heat is much lower at these conditions.

(2) Correct C. (1) Correct.

(2) Incorrect. Plausible if the applicant thinks that the post-trip SG levels will be < 14% NR.

D. (1) Incorrect. See B(1)

(2) Incorrect. See C(2)

Meets the K/A because it requires the examinee to know that the low decay heat from a RX trip is significantly affected by the amount of AF flow and the plant design and procedure actions to limit that impact.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

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

References:

2BEP ES-0.1, Reactor Trip Response

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 76 of 166 25 July 2019 48 ID: 2019 NRC-2 Q 48 Points: 1.00 Unit 1 tripped from rated thermal power due to a loss of normal feedwater.

Following the trip, alarm AF PUMP DSCH HIGH (1-3-B7) is received.

The below indications are present.

1AF005A CLOSED 1AF005B CLOSED 1AF005C CLOSED 1AF005D CLOSED 1AF005A OPEN 1AF005B OPEN 1AF005C OPEN 1AF005D OPEN A possible reason for this indication is ___(1)___.

In this condition, the operator can open 1AF005A ___(2)___.

A.

1. 1AF005A flow control valve potentiometer has lost instrument power

2. from the control room B.

1. 1AF005A flow control valve potentiometer has lost instrument power

2. only from outside the control room C.

1. flow transmitter 1FT-AF011A has failed high

2. only from outside the control room D.

1. flow transmitter 1FT-AF011A has failed high

2. from the control room Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 77 of 166 25 July 2019 Explanation:

Of these, the cause must be a failure of the flow transmitter. 1AF005A is closed because of the system response to the failed flow transmitter. Loss of instrument power to the potentiometer would cause 1AF005A to close, but this would not result in high flow. BAR 1-3-B7 says AF Flow indication is electronically limited to 240 gpm, so if a 1AF005 valve is closed due to an erroneous high flow signal, OPEN the affected 1AF005 valve by raising its respective controller to maximum. In this situation, the affected 1AF005 valve can only be cycled full open or full closed and will operate very slowly (>> 10 minute stroke time).

A. (1) Incorrect. Plausible if the applicant does not understand the effect of loss of instrument power to the potentiometer, and generally, a transmitter failure causes full flow meter deflection.

(2) Correct B. (1) Incorrect. See A(1)

(2) Incorrect. Plausible because local control is also possible.

C. (1) Correct.

(2) Incorrect. See B(2)

D. (1) Correct (2) Correct Meets the K/A because a flow controller is defined as a device used to measure and control the flow of liquids or gases. The transmitter is part of a control loop.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

BAR 1-3-B7, AF PUMP DSCH HIGH Rev. 3

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 78 of 166 25 July 2019 49 ID: 2019 NRC-2 Q 49 Points: 1.00 BOP DG-11, Diesel Generator Startup, is in progress with the crew taking actions to parallel the 1A DG to Bus 141.

1A DG indications on 1PM01J are as follows:

Incoming Voltage: 122 volts Running Voltage: 124 volts Synchroscope is rotating slowly in the SLOW direction If the DG Output Breaker were closed, which of the following would occur and/or be observed on the 1PM01J indicators below?

DG 1A Output VARs DG 1A Output Watts A.

IN At or near 0 B.

IN Rises C.

OUT At or near 0 D.

OUT Rises Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 79 of 166 25 July 2019 Explanation:

Operating the EDG in the manner described in the question is different than the procedure states. The procedure (BOP DG-11, DG S/U) has the sync scope rotating slowly in the fast direction. If the machine were operated as described in the stem the potential for reverse powering the EDG is more likely because KW stays at or near 0, as opposed to rising to 1000 kw and further from the reverse power trip. The question stem requires the examinee to assess that the DG is running "slower" than the grid, hence it will not pick up load. This is determined by the sync scope running in the slow direction. The machine is also "under excited" because incoming voltage is less than running voltage. Under these conditions the machine will become a VAR load, or VARs will be negative on the control board indicator, and it will not pick up load as previously stated.

A. Correct, see explanation.

B. Incorrect, see explanation.

C. Incorrect, see explanation.

D. Incorrect, see explanation.

Meets the K/A because it requires the examinee to know how the synchroscope rotation and associated voltages impact the DG after the breakers closes.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

BOP DG-11

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 80 of 166 25 July 2019 50 ID: 2019 NRC-2 Q 50 Points: 1.00 Unit 1 is operating at 100% power when the following indications are observed:

BAR 1-21-E10, 125V DC PNL 111/113 VOLT LOW, is ILLUMINATED.

BAR 1-22-E10, 125V DC PNL 112/114 VOLT LOW, is CLEAR.

Main Control Board DC Bus 111 voltage indicates 130 VDC.

Main Control Board DC Bus 112 voltage indicates 130 VDC.

Based upon the above conditions, which of the following electrical buses would be expected to have lost CONTROL POWER?

A.

Buses 141 and 143 B.

Bus 143 ONLY C.

Buses 142 and 144 D.

Bus 144 ONLY Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 81 of 166 25 July 2019 Explanation: annunciator BAR 1-21-E10 is caused by 125V DC Bus 111 and/or 125V DC Bus 113 being de-energized. The indications of a loss of a 125 VDC ESF bus (i.e. bus 111) would be the 125 VDC Bus volt low annunciator lit in conjunction with MCB DC Bus voltage at zero volts.

However, MCB DC Bus 111 voltage indication remains normal at 130 VDC. Based upon this, the conditions in the stem indicate that a loss of 125V DC Bus 113 has occurred. Since DC Bus 113 supplies control power to Bus 143, this bus would be expected to have lost its control power supply in the given scenario. The indications associated with DC Buses 112 and 114 can be explained in a comparable manner, except in that case, both buses remain energized and would be expected to be suppling control power to Buses 142 and 144 respectively.

A) Incorrect. DC Bus 111 supplies control power to Bus 141. The conditions provided in the stem indicate that DC Bus 111 remains energized and, therefore, that Bus 141 would be expected to have control power available.

B) Correct. See explanation above.

C) Incorrect. DC Bus 112 supplies control power to Bus 142. The conditions provided in the stem indicate that DC Bus 112 remains energized and, therefore, that Bus 142 would be expected to have control power available.

D) Incorrect. DC Bus 114 supplies control power to Bus 144. The conditions provided in the stem indicate that DC Bus 114 remains energized and, therefore, that Bus 144 would be expected to have control power available.

Meets the K/A because it requires the examinee to know which major bus control power is feed from which DC train.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

BAR 1-21-E10, 125V DC PNL 111/113 VOLT LOW, Rev. 3 BAR 1-22-E10, 125V DC PNL 112/114 VOLT LOW, Rev. 3 BOP DC-1, 125V DC ESF Battery Chargers Start-up, Rev. 18 N-BY-TQ-ILT-DC-8A, 125VDC Power Systems, Rev. 7

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 82 of 166 25 July 2019 51 ID: 2019 NRC-2 Q 51 Points: 1.00 At 10:00, the 1A DG was started using a slow mode start per 1BOSR 8.1.2-1, Unit One 1A Diesel Generator Operability Surveillance, for a 4-hour monthly loaded run.

At 12:30, the EO reported a loud noise and that BOTH starting air bank pressures were rapidly approaching 0 psig.

What is the immediate impact on the operating 1A DG?

A.

The DG will ONLY trip on overspeed.

B.

ONLY the DG starting capability is lost.

C.

The DG will trip due to a loss of starting air.

D.

ONLY the DG non-emergency trip capabilities are lost.

Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 83 of 166 25 July 2019 Explanation:

A. Correct, on a loss of starting air pressure, the engine can operate without control air but will not trip (except overspeed, which is a mechanical trip) on any automatic or manual signal. An overspeed signal will shut the air intake butterfly, which will stop the engine.

B. Incorrect, while starting capability is lost, the DG is running at full load and will not trip as described above. This choice is incorrect as this is not the ONLY impact.

C. Incorrect, the DG will not trip on a loss of starting air, but will lose trip capabilities. The applicant may confuse the loss of the associated DC control power (will cause an immediate DG trip) with a loss of starting air.

D. Incorrect, the non-emergency trips (see below) are only lost on an SI or UV signal.

Non-emergency trips:

Low Lube oil pressure Low turbocharger oil pressure Turbocharger thrust bearing failure High crankcase pressure Main bearing high temperature Jacket water high temperature Connecting rod bearing high temperature Generator bearing high temperature Meets the K/A because it requires the examinee to know that the loss of the Starting Air Receivers disables most of the trip functions of the EDGs.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

1BOSR 8.1.2-1 N-BY-TQ-ILT-DG-9 Diesel Generator & Auxiliaries System Lesson Plan

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 84 of 166 25 July 2019 52 ID: 2019 NRC-2 Q 52 Points: 1.00 Unit 1 is operating at 100% power.

A liquid waste release is in progress.

0WX353, RELEASE TK DSCH TO CIRC WATER BLOWDOWN RAD MONITOR FLOW valve, has been opened.

Subsequently, an alarm is received for RMS Monitor 0PR01J, LIQ RADWASTE EFF.

The 0PR01J alarm color is MAGENTA.

1) What type of failure has occurred for 0PR01J?

2) What is the expected current position of 0WX353?

A.

1. OPERATE failure

2. OPEN B.

1. OPERATE failure

2. CLOSED C.

1. COMMUNICATIONS failure

2. OPEN D.

1. COMMUNICATIONS failure

2. CLOSED Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 85 of 166 25 July 2019 Explanation: per BOP AR/PR-11A2, the magenta alarm color indicates that a communications failure has occurred. No automatic actions are expected to occur provided that failure is strictly a communications failure.

A) First part incorrect / second part correct. First part explanation: per BOP AR/PR-11A2, the magenta alarm color indicates that a communications failure has occurred. Per BOP AR/PR-11A13, an alarm color of blue would indicate that an operate failure had occurred.

B) Both parts incorrect. First part explanation: per BOP AR/PR-11A2, the magenta alarm color indicates that a communications failure has occurred. Per BOP AR/PR-11A13, an alarm color of blue would indicate that an operate failure had occurred. Second part explanation: an operate failure will result in an automatic action where the affected monitor will be placed into an interlock condition. Per BOP AR/PR-11T1, an interlock function of 0PR01J is to close 0WX353.

C) Correct. See explanation above.

D) First part correct / second part incorrect. Second part explanation: an operate failure will result in an automatic action where the affected monitor will be placed into an interlock condition. Per BOP AR/PR-11T1, an interlock function of 0PR01J is to close 0WX353.

Meets the K/A because it requires the examinee to know the actuation of the 0PR01J and how that impacts the ability to perform a liquid release.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

BAR RMS-1-0PR01J, LIQ RADWASTE EFF, Rev.1 BOP AR/PR-11A2, RMS COMMUNICATIONS MONITOR COMMUNICATIONS FAILURE, Rev. 52 BOP AR/PR-11A13, OPERATE FAILURE CHANNEL OPERATE FAILURE, Rev. 52 BOP AR/PR-11T1, RAD MONITOR INTERLOCK FUNCTION TABLE, Rev. 10 N-06OLSYS, Radiation Monitors, Rev. 8 06OLSYS/NPOSYS, Liquid Radwaste System, Rev. 5

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 86 of 166 25 July 2019 53 ID: 2019 NRC-2 Q 53 Points: 1.00 A fault on Bus 142 makes the ______ pump unavailable.

A.

1A Essential Service Water (SX)

B.

1A Auxiliary Feedwater (AF)

C.

1B Essential Service Water (SX)

D.

1B Auxiliary Feedwater (AF)

Answer:

C Answer Explanation Explanation:

E. Incorrect. Plausible if the applicant does not know the train relationship between the buses and the pump.

F. Incorrect. Plausible if the applicant does not know the train relationship between the buses and the pump.

G. Correct.

H. Incorrect. Plausible if the applicant does not know that the B AF pump is diesel driven and can supply its own cooling if the associated SX pump is not running.

Meets the K/A because it requires the examinee to know which pump is energized by which bus.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

Chapter 20/I1-SX-SL-01

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 87 of 166 25 July 2019 54 ID: 2019 NRC-2 Q 54 Points: 1.00 The crew has entered 1BEP ES-0.1, Reactor Trip Response, with the follow indications:

Annunciator 0-39-C3, IA RCVR 2 PRESS LOW is LIT Annunciator 2-1-C7, REMOTE S/D PANEL TROUBLE is LIT Annunciator 2-9-D3, CNG LINE FLOW HIGH LOW is LIT The Crew has concurrently entered 2BOA SEC-4, Loss of Instrument Air Unit 2 To control RCP Seal Injection flow, the A.

NSO will manually throttle 2CV121.

B.

NSO will manually isolate 2CV121 and manually throttle open the Bypass valve.

C.

EO will locally throttle 2CV121.

D.

EO will locally isolate 2CV121 and locally throttle open the Bypass valve.

Answer:

D Answer Explanation Explanation: The indications presented in the stem represent a Loss of Instrument Air for Unit 2 due to a large leak in the Aux Building. On a loss of Instrument Air, 2CV121 will fail to the OPEN position. Step 3 of 2BOA SEC-4, Loss of Instrument Air Unit 2, gives direction for controlling RCP Seal Injection by locally bypassing and isolating the 2CV121.

A) Incorrect. 2CV121 fails open on a loss of IA. Plausible because 2BOA SEC-4 does describe the operation of the PZR PORVs, that can still operate for a limited number of cycles with a loss of IA. Also, the NSO would normally throttle 2CV121 to control charging and seal injection flows.

B) Incorrect. To isolate and bypass 2CV121 requires local actions. Plausible because 2BOA SEC-4, Loss of Instrument Air Unit 2, does give direction to control some of the valves from the Control Room such as the 2GS004, 2GS001 and 2GS025. Also 2BOA SEC-4 gives direction for controlling RCP Seal Injection by bypassing and isolating 2CV121 C) Incorrect. 2CV121 fails open on a loss of IA. Plausible because there are valves that are controlled locally by failing IA and then locally throttling the valve, such as the 2AF005s.

D) CORRECT. See explanation above.

Meets the K/A because it requires the examinee to know the necessary actions to be taken in the field to address the loss of IA to various CV control valves, specifically 2CV121.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(6) Design, components, and function of reactivity control mechanisms and instrumentation.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

2BOA SEC-4, Loss of Instrument Air Unit 2

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 88 of 166 25 July 2019 55 ID: 2019 NRC-2 Q 55 Points: 1.00 Which of these actions can be completed without resetting an active safety injection signal?

A.

Reset feedwater isolation via the Push Buttons.

B.

Perform a normal shutdown of an emergency diesel generator.

C.

Close (and maintain closed) the reactor trip breakers.

D.

Establish Post-LOCA purge flow damper alignment.

Answer:

D Answer Explanation Explanation: An SI causes a containment ventilation isolation, which isolates post-LOCA purge by closing VQ003. But a containment ventilation isolation can be reset with SI present. None of the other actions can be done without resetting SI.

A. Incorrect. Plausible because a SI actuation causes a FW Isolation, but the SI actuation signal must be reset/blocked in order to reset the FW isolation.

B. Incorrect. Plausible because a SI actuation causes an automatic start of the DG, but the SI actuation signal must be reset/blocked in order to shut down the EDGs.

C. Incorrect. Plausible because a SI actuation causes a reactor trip, but the SI signal must be reset/blocked in order to close (and maintain closed) the reactor trip breakers.

D. Correct. An SI causes a containment ventilation isolation, including the post-LOCA purge path. But a containment ventilation isolation can be reset with SI present.

Meets the K/A because it requires the examinee to know the necessary actions to reset Containment Ventilation Isolation after and SI Actuation.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

Module Ch 61, I1-EF-SL091 Horse Notes EF-2

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 89 of 166 25 July 2019 56 ID: 2019 NRC-2 Q 56 Points: 1.00 Unit 1 is making preparations for startup following a refueling outage.

The 1A Control Rod Drive MG Set was started earlier and its associated Generator Breaker is currently closed.

The 1B Control Rod Drive MG Set has just been started.

The 1A and 1B Control Rod Drive MG Sets have NOT yet been paralleled.

The Synchronizer Switch for the INCOMING Control Rod Drive MG Set has been just been turned ON at the Control Rod Drive MG Set Control Cabinet.

1. How is the 1B Control Rod Drive MG Set Generator Output Breaker expected to subsequently be closed?

2. What is the expected plant impact if the 1B Control Rod Drive MG Set subsequently de-energizes after the reactor startup is complete?

A.

1. Manually

2. A reactor trip will occur B.

1. Manually

2. A reactor trip will NOT occur C.

1. Automatically

2. A reactor trip will occur D.

1. Automatically

2. A reactor trip will NOT occur Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 90 of 166 25 July 2019 Explanation: per BOP RD-5, after the Synchronizer Switch for the incoming Control Rod Drive MG Set has been turned on at the Control Rod Drive MG Set Control Cabinet, the incoming and running Control Rod Drive MG Sets should parallel automatically, as indicated by closure of the incoming (e.g. 1B) Control Rod Drive MG Set Generator Output Breaker. Per BOP RD-2, shutdown of both Control Rod Drive MG Sets will result in a Reactor Trip.

A) Both parts are incorrect. First part explanation: with the Synchronizer Switch turned on for the 1B Control Rod Drive MG Set, the 1B Control Rod Drive MG Set Generator Output Breaker is expected to close and parallel the MG sets automatically. Second part explanation: the de-energization of a single MG set is not expected to result in a reactor trip provided that the other MG set remains in service.

B) First part incorrect / second part correct. First part explanation: with the Synchronizer Switch turned on for the 1B Control Rod Drive MG Set, the 1B Control Rod Drive MG Set Generator Output Breaker is expected to close and parallel the MG sets automatically.

C) First part correct / second part incorrect. Second part explanation: the de-energization of one MG set is not expected to result in a reactor trip provided that the other MG set remains energized.

D) Correct. See explanation above.

Meets the K/A because it requires the examinee to know the automatic synch capabiliaties of the MG Sets.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

BOP RD-1, CONTROL ROD DRIVE MG SET STARTUP, Rev. 10 BOP RD-2, Control Rod Drive MG Set Shutdown, Rev. 6 BOP RD-5, Control Rod Drive MG Set Start Up and Paralleling to Operating Control Rod Drive MG Set, Rev. 11 N-06OLSYS, Rod Control System, Rev. 05a

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 91 of 166 25 July 2019 57 ID: 2019 NRC-2 Q 57 Points: 1.00 Initial conditions in Unit 1:

A reactor startup was in progress.

Reactor power was at 1x10-7% in the intermediate range.

Control rods were being withdrawn.

When:

ROD CONT URGENT FAILURE (1-10-C6) alarms.

DRPI Urgent Failure lights all flashing.

Rod Bottom LEDs light on rod B8.

General Warning LEDs light on rod B8.

All Data A and Data B LEDs are flashing.

These indications could be caused by A.

a blown stationary gripper fuse on control rod B8.

B.

a blown fuse in the control rod power cabinet.

C.

bad DRPI DATA A for control rod B8.

D.

a failed Detector/Encoder Card in DRPI Data Cabinet A.

Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 92 of 166 25 July 2019 Explanation:

D is the only correct answer of the choices given for the following reasons:

These alarms mean that DRPI has no reliable data on the position of rod B8. A DRIP General Warning is caused by bad DRIP data:

An Urgent Failure alarm occurs when both Data A and Data B are bad A Non-Urgent Failure when either Data A or Data B is bad.

There are no inputs from the Rod Control system that cause a DRIP general warning, so distractors A and B are incorrect. Bad Data A would bring in the Data A failure LEDs, but would NOT cause an Urgent Failure Alarm, so distractor C is incorrect.

This event happened in 1987 at Braidwood. At that time, the Detector/Encoder card error caused a > 1 bit deviation between data A and Data B.

A. Incorrect. Credible because this is one possible cause of alarm 1-10-C.

B. Incorrect. Credible because this is one possible cause of alarm 1-10-C.

C. Incorrect. Credible if the applicant does not know what causes a DRIP Urgent Failure.

D. Correct. A failed Detector/Encoder Card in DRIP Data Cabinet A is the only possibility of those listed.

Meets the K/A because it requires the examinee to know the failure indications of DRPI.

RO level question because:

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(6) Design, components, and function of reactivity control mechanisms and instrumentation.

References:

I1-PI-XL-01, Chapter 29, Rod Position Indication System BAR 1-10-C6, ROD CONT URGENT FAILURE, Rev. 8

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 93 of 166 25 July 2019 58 ID: 2019 NRC-2 Q 58 Points: 1.00 Unit 1 is at 8% power during a plant startup.

Vital instrument bus 111 BYPASS SOURCE TO LOAD pushbutton is pressed.

Which of the following describes the resulting plant response and reason?

A.

The reactor does NOT trip because power is rapidly transferred to the CVT.

B.

The reactor does NOT trip because power is still below P-10.

C.

A reactor trip occurs because N35 momentarily de-energizes.

D.

A reactor trip occurs because N31 momentarily de-energizes.

Answer:

A Answer Explanation Explanation:

A) Correct. When the BYPASS SOURCE TO LOAD pushbutton is depressed the static switch will rapidly transfer the power supply for Instrument Bus 111 from the inverter to the CVT without causing the IRNI Channel N35 to cause a trip signal.

B) Incorrect. Plausible as IRNI N35 and PRNI N41 are powered from Instrument Bus 111, but with power below P-10, if IRNI N35 was affected by the loss of power a reactor trip would occur.

C) Incorrect. Plausible as the static switch is break before make, but is designed to rapidly transfer power such that a reactor trip signal will not be generated by IRNI N35.

D) Incorrect. Plausible as SRNI N31 is powered by Instrument Bus 111, but with power above P-6 any trip signal developed by N31 would be blocked.

Meets the K/A because it requires the examinee to know the operation of the Instrument Inverter.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(6) Design, components, and function of reactivity control mechanisms and instrumentation.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

1BOA ELEC-2, LOSS OF INSTRUMENT BUS UNIT 1

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 94 of 166 25 July 2019 59 ID: 2019 NRC-2 Q 59 Points: 1.00 Unit 1 has experienced an automatic Reactor Trip and Safety Injection actuation.

During the performance of 1BEP-0, Reactor Trip or Safety Injection, the RO checks for Adverse Containment using the 1PM06J gauges, and RMS trends shown below:

The RO will report that Containment is...?

A.

NOT ADVERSE.

B.

ADVERSE based on containment pressure ONLY.

C.

ADVERSE based on radiation levels ONLY.

D.

ADVERSE based on BOTH containment pressure AND radiation levels.

Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 95 of 166 25 July 2019 Explanation: ADVERSE Containment is defined as: CNMT pressure GREATER THAN 5 PSIG or CNMT radiation level GREATER THAN 105 R/HR. Only a single indication, either pressure or radiation monitor, is required to declare Containment Adverse. 4AS120 does indicate over 100000 (105) R/HR per the trend and this is sufficient to declare Containment Adverse due to radiation levels.

A. Incorrect: The indicating pressure gauges do show that CNMT pressure is less than 5 psig.

However, the radiation monitors indicate 2.26 x 105 and 2.7 x 105. This exceeds the required radiation level to declare CNMT Adverse.

B. Incorrect: The indicating pressure gauges do show that CNMT pressure is less than 5 psig.

This answer is plausible because the indications do show the value greater than the 3.4 psig required for the SI.

C. CORRECT. See explanation above.

D. Correct: The gauges clearly show CNMT pressure less than 5 psig.

Question meets K/A - The examinee must know the criteria for Adverse CNMT which is met by the occurrence of a RCS LOCA.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

1BEP 0 N-BY-TQ-ILT-0 1BEP-0 Reactor Trip or Safety Injection BEP-0

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 96 of 166 25 July 2019 60 ID: 2019 NRC-2 Q 60 Points: 1.00 What feature of the spent fuel pool cooling system is designed to prevent draining the pool?

A.

Check valves installed at the discharge of each spent fuel pit pump.

B.

A check valve installed in the common return line.

C.

A hole drilled in each of the spent fuel pit pump suction lines.

D.

A hole drilled in the common return line.

Answer:

D Answer Explanation Explanation: Return line has anti-siphon hole (1/2 diameter, 4 below water level) to prevent draining SFP.

A. Incorrect. Plausible because there is a check valve as described, but the dewatering protection on the suction line is that the suction line is well above the fuel.

B. Incorrect. Plausible because check valves are used in some applications as anti-siphoning devices.

C. Incorrect. Plausible because this would work if the hole were below the water line, but there is no such hole.

D. Correct. See explanation above.

Question meets K/A - The examinee must know the design features for the Spent Fuel Pool and associated equipment.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

Module Ch 51, I1-FC-XL-01

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 97 of 166 25 July 2019 61 ID: 2019 NRC-2 Q 61 Points: 1.00 Unit 1 Reactor power is 90% and rising slowly.

RCS TAVE is 576 °F and lowering slowly.

Pressurizer pressure is 2210 psig and lowering slowly.

Main generator load is 1058 MW and stable.

ALL Steam Generator narrow range levels are 62% and rising slowly.

ALL Steam Generator pressures are 900 psig and lowering slowly.

Based upon the above conditions, a leak currently exists in...

A.

a main steamline.

B.

a main feedline.

C.

an RCS loop.

D.

the PZR vapor space.

Answer:

A Answer Explanation Explanation:

Main steamline is correct; TAVE lowers due to TC lowering, which also causes Rx Power to rise and PZR Level to lower. SG pressure also lowers due to the leak; while this will cause Main Generator MW output to lower initially, the governor valves will respond by opening to compensate. Additionally, SG levels will rise due to the effects of swell.

A) Correct. See explanation above.

B) Incorrect. A main feedline break is incorrect because SG level is rising and TAVE is lowering.

Plausible because TAVE could lower in a feed line break scenario if the pressure source were the SG instead of the FW pumps, however, in this case the plant is still at power with the FW pumps running.

C) Incorrect. RCS loop is incorrect due to TAVE lowering, as well as the secondary plant indications. Plausible because pressure is dropping slowly, however, this in itself does not indicate a primary leak.

D) Incorrect. PZR vapor space is incorrect due to the secondary plant indications. Plausible because pressure is dropping slowly, however, this in itself does not indicate a primary leak.

Question meets K/A - The examinee must know the concepts of shrink and swell to properly identify the impact of the leak on the main steam line.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

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

References:

Module Ch 51, I1-FC-XL-01

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 98 of 166 25 July 2019 62 ID: 2019 NRC-2 Q 62 Points: 1.00 Initiating a CD/FW runback with DEHC in Operator Automatic Mode will cause the turbine to run back at a rate of ___(1)___. The runback will automatically stop at ___(2)___.

A.

1. 250 MW/min

2. 780 MW B.

1. 250 MW/min

2. 700 MW C.

1. 200% min

2. 780 MW D.

1. 200% min

2. 700 MW Answer:

B Answer Explanation Explanation:

A. (1) Correct (2) Incorrect. Plausible because this is where a Heater Drain (HD) runback will automatically stop.

B. (1) Correct (2) Correct C. (1) Incorrect. Plausible because this is the rate when DEHC is in Manual Mode (2) Incorrect. See A(2)

D. (1) Incorrect. See C(1)

(2) Correct Question meets K/A - The examinee must know the automatic and manual operation of the DEHC controls to identify the correct operation.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

Module Ch 37a, I1-EH-XL-01

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 99 of 166 25 July 2019 63 ID: 2019 NRC-2 Q 63 Points: 1.00 2PR08J, S/G Blowdown line radiation monitor, received a spurious high radiation alarm (spiked above the high alarm setpoint).

Which of the following is a correct response to the spurious high radiation signal?

Note:

2SD002A/B/C/D; AOV S/G BlowDown Isolation Valves, 2SD005A/B/C/D; AOV S/G BlowDown Sample Valves, 2PS179A/B/C/D; S/G BlowDown Sample Valves A.

Only the 2PS179A/B/C/D valves close.

B.

Only the 2SD002A/B/C/D valves close.

C.

All 2PS179A/B/C/D and 2SD002A/B/C/D valves close.

D.

All 2SD002A/B/C/D and 2SD005A/B/C/D valves close.

Answer:

A Answer Explanation Explanation:

A. Correct, when 2PR08J senses a high radiation condition, the 2PS179_ valves close.

B. Incorrect, a HELB actuation would cause only the 2SD002_ valves to close (as well as the unmentioned 2SD054_ valves). A containment phase A isolation would close the 2SD005_

valves in addition to the 2SD002_ valves.

C. Incorrect, this would be the result of a 2PR08J high rad condition and a HELB actuation.

D. Incorrect, a containment phase A isolation would cause the 2SD002_ and 2SD005_ valves to close.

Question meets K/A - The examinee must know the correct components that actuate with a high radiation signal for the 2PR08J monitor.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

References:

BAR RMS-1-2PR08J BOP SD-100

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 100 of 166 25 July 2019 64 ID: 2019 NRC-2 Q 64 Points: 1.00 Unit 1 is at 100% power.

The 0B Gas Decay Tank reached its maximum Curie limitation and was isolated.

Subsequently, the 0B Gas Decay Tank relief valve inadvertently lifted and is stuck open.

Based on this indication:

A.

ONLY 1/2PR28J, Vent Stack Effluent, will provide monitoring indication for the release.

If the HIGH alarm is reached, an interlock will isolate the release.

B.

ONLY 1/2PR28J, Vent Stack Effluent, will provide monitoring indication for the release.

If the HIGH alarm is reached, an interlock will NOT isolate the release.

C.

0PR02J, Gas Decay Tank Effluent, and 1/2PR28J Vent Stack Effluent, will provide monitoring indication for the release.

If a HIGH alarm occurs on ANY of these radiation monitors, an interlock will isolate the release.

D.

0PR02J, Gas Decay Tank Effluent, and 1/2PR28J Vent Stack Effluent, will provide monitoring indication for the release.

If a HIGH alarm occurs on ANY of these radiation monitors, an interlock will NOT isolate the release.

Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 101 of 166 25 July 2019 Explanation:

A) Incorrect. Plausible as the 1/2PR28J will provide monitoring of the release, but the 0PR02J will also provide monitoring. The relief valve tailpipe enters the ventilation pathway downstream of the 0GWRCV014 Gas Decay Tank Discharge Release Control Valve which will close on the 0PR02J High alarm, but as such will be ineffective at stopping the release.

B) Incorrect. Plausible as the 1/2PR28J will provide monitoring of the release, but the 0PR02J will also provide monitoring. The release will also not isolate on any interlock.

C) Incorrect. Plausible as both 0PR02J and 1/2PR28J will provide monitoring for the release, but the relief valve tailpipe enters the ventilation pathway downstream of the 0GWRCV014 Gas Decay Tank Discharge Release Control Valve which will close on the 0PR02J High alarm, but as such will be ineffective at stopping the release.

D) Correct. 0PR02J, Gas Decay Tank Effluent and 1/2PR28J Vent Stack Effluent will provide monitoring indication for the release. A HIGH alarm on any of these monitors will NOT actuate an interlock to isolate the release. The relief lifting bypasses the 0GWRCV014 Gas Decay Tank Discharge Release Control valve (which DOES automatically close on high rad alarm of 0PR02J).

Question meets K/A - The examinee must know the system design of the Waste Gas system and the points in which it has radiation monitors.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

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

References:

BOP GW-1, GASEOUS WASTE SYSTEM STARTUP AND OPERATION 0BOA RAD-3, DECAY TANK HIGH ACTIVITY

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 102 of 166 25 July 2019 65 ID: 2019 NRC-2 Q 65 Points: 1.00 Byron Unit 1 is operating in MODE 1 1PM09J-G1 alarm - LCSR B-1 (D-52) FIRE WIRE OPEN - was received last shift and an hourly fire watch patrol was implemented.

1PM09J G2 alarm - LCSR B-1 (D-51) FIRE - just alarmed.

The Control Room Operators will A.

Actuate the station fire alarm and Dispatch an EO to manually initiate CO2 to the Unit 1 LCSR Zone B-1.

B.

Verify automatic initiation of CO2 to the Unit 1LCSR Zone B-1 and Actuate the station fire alarm.

C.

Verify that the LCSR B-1 (D-52) FIRE is NOT illuminated and dispatch an EO to the Unit 1 LCSR to place the Zone B-1 abort switch in ABORT.

D.

Verify that the LCSR B-1 (D-52) FIRE is NOT illuminated and dispatch an EO to the Unit 1 LCSR to shut the Zone B-1 CO2 Block Valve.

Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 103 of 166 25 July 2019 Explanation:

A. Incorrect; the LCSR Zone B-1 suppression system will automatically actuate with the specified alarms.

B. Correct; the LCSR Zone B-1 suppression system will automatically actuate with the specified alarms.

C. Incorrect; automatic actuation is expected even if the LCSR B-1 (D-52) FIRE is NOT illuminated and momentary operation of the abort switch will not prevent the automatic actuation of the system; switch must be HELD in the ABORT position to prevent actuation, and is effective only if operated prior to the delay timer timing out.

D. Incorrect; automatic actuation is expected even if the LCSR B-1 (D-52) FIRE is NOT illuminated and there is no procedural direction shut the block valve once the system has actuated.

Question meets K/A - The examinee must know the consequences of a malfunction coincident with a Fire Wire Open system degradation.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

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

References:

BAP 1100-10; Response Procedure for Fire/Fire Alarm; Rev 18 BAR 1PM09J-G1; LCSR B-1 (D-51); Rev 6 BAR 1PM09J-G2; LCSR B-1 (D-52); Rev 6 BAR 0-37-A4; Unit 1 Area Fire; Rev 11

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 104 of 166 25 July 2019 66 ID: 2019 NRC-2 Q 66 Points: 1.00 Unit 1 is in Mode 3.

1BGP 100-5, Plant Shutdown and Cooldown is in progress in preparation for refueling.

The crew was preparing to Borate the RCS to the target refueling boron concentration.

The boration will be through the emergency boration flowpath.

Initial Boric Acid Storage Tank level is 80%.

A 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> hold will be required to complete check valve testing at NOP and NOT.

If no makeup is added to the Boric Acid Storage Tank, the amount of boration time for the boric acid storage tank level to drop below the applicable TLCO for Borated Water Sources is ___(1)___ gpm boration for between ___(2)____ hours.

A.

1. 40

2. 7 and 8 B.

1. 40

2. 12 and 13 C.

1. 55

2. 5 and 6 D.

1. 55

2. 9 and 10 Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 105 of 166 25 July 2019 REFERENCE PROVIDED - BTC 1.3, Boric Acid Storage Tanks 1/2AB03T.

Explanation: To get this question correct, the applicant must:

Properly read the tank curve (meeting the generic K/A)

The slope (rise/run) of the curve is 1%/440 gallons The current volume is 80% x 440 gallons/% = 35,200 gallons The TLCO is 40% x 440 gallons/% = 17600 gallons The difference is also 17,600 gallons Know that the plant will still be greater than MODE 5 when the boration is completed. (plant specific knowledge)

The TLCO will be reached when 17,600 gallons are pumped.

17,600 gallons/40 gallons/minute = 440 min or ~7.33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> 17,600 gallons/55 gallons/minute = 320 min or ~ 5.33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> 29,920 gallons is the volume from 80% down to 12% (provided 440 gallons per percent).

29,920 gallons/40 gallons/minute = 748 min or ~12.46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> 29,920 gallons/55 gallons/minute = 544 min or ~ 9.06 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> A. Incorrect. Credible if the applicant uses the normal boration flowrate (40 gpm) to meet 40%

TLCO limit (Mode 1 - 4).

B. Incorrect. Credible if the applicant uses the normal boration flowrate (40 gpm) to meet the 12 % TLCO limit (MODE 5-6).

C. Correct.

D. Incorrect. Credible if the applicant uses the emergency boration flowrate (55 gpm) to meet the 12% YLCO limit (MODEs 5-6).

Question meets K/A - The examinee must know the consequences of a malfunction coincident with a Fire Wire Open system degradation.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

BTC 1 3 Boric Acid Storage Tanks 1/2AB03T rev 2

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 106 of 166 25 July 2019 67 ID: 2019 NRC-2 Q 67 Points: 1.00 In accordance with OP-AA-111-101 Operating Narrative Logs and Records, if an erroneous log entry is on a hand written or printed record, the correction should be made by...

A.

discarding the erroneous record and recreating a correct record.

B.

completely blacking out the erroneous entry and inserting corrected information.

C.

drawing a single line through the erroneous entry and inserting corrected information.

D.

using correction fluid (White-out) to remove the erroneous entry and inserting the corrected information.

Answer:

C Answer Explanation Explanation:

OP-AA-111-101, Section 4.1.10 and RM-AA-101-1008, specifically direct drawing a single line through erroneous log entries and that the erroneous entry NOT be obliterated.

Choice A is incorrect, see explanation above.

Choice B is incorrect, see explanation above.

Choice C is correct, see explanation above.

Choice D is incorrect, see explanation above.

Question meets K/A - The examinee must know the administrative requirements for narrative logs and record keeping.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

OP-AA-111-101; Operating Narrative Logs and Records; Rev 6 RM-AA-101-1008; Processing and Storage of Records; Revision 9

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 107 of 166 25 July 2019 68 ID: 2019 NRC-2 Q 68 Points: 1.00 Unit 2 is operating at 100% power.

A plant transient occurs.

While responding to the transient, the At-The-Controls (ATC) RO observes reactor control systems behaving unexpectedly.

Which of the following practices by the ATC RO would be consistent with OP-AA-101-111-1001 OPERATIONS STANDARDS AND EXPECTATIONS for conservative decision-making and action?

A.

NOT becoming totally involved in any single operation.

B.

Implementing immediate actions only AFTER verbalizing them.

C.

Reducing power OR initiating a reactor scram without delay.

D.

Referencing hard cards BEFORE taking immediate operator actions.

Answer:

C Answer Explanation Explanation: OP-AA-101-111-1001, section 4.11.3, states that when faced with unexpected or anomalous behavior of the reactor, or its control systems, take conservative action, including reducing power or a initiating a reactor scram without hesitation.

A) Incorrect. Per OP-AA-101-111, section 4.1.7, the Shift Manager shall not become totally involved in any single operation to the extent a loss of overall plant perspective occurs.

B) Incorrect. Per OP-AA-101-111, section 4.7.2, the ROs immediate actions to STABILIZE the plant during transient conditions take priority over verbalization to the Unit Supervisor. If possible, verbalization should be accomplished to inform the Unit Supervisor of actions being taken.

C) Correct. See explanation above.

D) Incorrect. Per OP-AA-101-111, section 4.7.2, hard cards or procedures shall be used to take mitigating actions that are not deemed immediate operator actions.

Question meets K/A - The examinee must know the administrative requirements for Operations Standards and Roles and Responsbilities.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

OP-AA-101-111, Roles and Responsibilities of On-Shift Personnel, Rev. 12 OP-AA-101-111-1001, Operations Standards and Expectations, Rev. 21

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 108 of 166 25 July 2019 69 ID: 2019 NRC-2 Q 69 Points: 1.00 The Unit 1 limitation in establishing feed flow to a NON-DRY steam generator in response to a Loss of Secondary Heat Sink is a function of ___(1)___.

The Unit 1 reactor trip setpoint for low-low steam generator level is ___(2)___.

A.

1. the duration of the loss of feed flow

2. 18%

B.

1. the duration of the loss of feed flow

2. 36.3%

C.

1. subcooling in the tempering line

2. 18%

D.

1. subcooling in the tempering line

2. 36.3%

Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 109 of 166 25 July 2019 Explanation:

Both parts of this question have different answers for Unit 1 (U1) and Unit 2 (U2), which ensures the K/A is being appropriately tested. U1 only has a FW input at the top. U2 SGs are different in that they have what is referred to as a tempering line. Specifically, U2 SGs have 2 FW lines coming in, one at the bottom and one at the top. This design is to prevent water hammer and tube ruptures that may result from that water hammer. Under low flow conditions (up to

~20% power), flow is directed through the feed reg bypass valve, through the 2FW039_ valve and into the aux nozzle. After the 2FW009_ valves are opened, most of the flow is directed into the bottom of the S/G, but some flow (100 gpm) still remains through the tempering line through the 2FW034/35 valves to keep the line from stagnating. This explains, in part, why U2 still has a water hammer prevention system (WHPS) and U1 does not. Limitations for establishing feed flow to a non-dry steam generator in response to a Loss of Secondary Heat Sink are found in 1(2)BFR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, ATTACHMENT B, for the respective unit, and are reflected in Part (1) of each answer. As noted, the incorrect distractors are correct for U2, which makes them plausible.

A. (1) Correct. See above explanation.

(2) Correct. See above explanation.

B. (1) Correct (2) Incorrect. This is the Unit 2 setpoint.

C. (1) Incorrect. This is the Unit 2 limitation (2) Incorrect. This is the Unit 2 setpoint D. (1) Incorrect. This is the Unit 2 limitation (2) Correct Question meets K/A - The examinee must know the differences between Unit 1 and Unit 2 SG Level setpoint actuations to correctly answer the question.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(3) Mechanical components and design features of reactor primary system.

References:

1BFR H.1, Response to Loss of Secondary Heat Sink, Rev. 302 2BFR H.1, Response to Loss of Secondary Heat Sink, Rev. 301 1BEP 0, Reactor Trip or Safety Injection, Rev 302 2BEP 0, Reactor Trip or Safety Injection, Rev 302

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 110 of 166 25 July 2019 70 ID: 2019 NRC-2 Q 70 Points: 1.00 Proposed Question: Question #70 Unit 1 Conditions:

The reactor trip breakers are open.

Average reactor coolant temperature is 130°F.

One reactor coolant pump is running.

One source range neutron flux monitor is INOPERABLE.

Which of these changes is NOT ALLOWED by technical specifications?

A.

Raising average reactor coolant temperature to 210°F.

B.

Stopping the running reactor coolant pump.

C.

De-tensioning any reactor vessel head closure bolt.

D.

Closing ONE reactor trip breaker for testing.

Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 111 of 166 25 July 2019 Explanation: With one INOPERABLE source range neutron flux monitor, LCO 3.9.3 is not MET. MODE 6 entry is not allowed.

To get this question correct, the applicant has to know that:

LCO 3.9.3 applies in MODE 6 (above the line RO knowledge).

LCO 3.9.3 requires two source range neutron flux monitors to be OPERABLE Know that de-tensioning any reactor vessel head closure bolt constitutes MODE 6 entry.

This meets the K/A requirement to determine MODE of operation.

A. Incorrect. This would be MODE 4 entry, and no SR instruments are required in MODE 4 with trip breakers open.

B. Incorrect. This would cause LCO 3.3.9 to not be MET, but intentional action statement entry is allowed (this action statement is usually entered when going to MODE 6).

C. Correct.

D. Incorrect. The requirement of LCO 3.3.1 in MODE 5 is for the rod control system to be not capable of rod withdrawal. With only one trip breaker closed, rods withdrawal is still not possible.

Question meets K/A - The examinee must know the Technical Specifications for each mode.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

Unit 1 Core Operating Limits Report, Rev, 13 LCO 1.1, Definitions LCO 3.9.3, Nuclear Instrumentation 1BGP 100-6, Unit 1 Refueling Outage LCO 3.3.1 LCO 3.3.9

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 112 of 166 25 July 2019 71 ID: 2019 NRC-2 Q 71 Points: 1.00 With respect to Technical Specification required actions, which of the following require action(s) to be completed within one hour?

A.

A LCO is NOT met and an associated action is not provided.

B.

Trip of a Reactor Coolant Pump with Thermal Power at 25%.

C.

One Intermediate Range Neutron Flux channel inoperable at 5% RX power.

D.

Axial Flux Difference NOT within limits with Reactor Thermal Power at 30%.

Answer:

A Answer Explanation Explanation:

A. Correct; TS LCO 3.0.3 requires that action shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit in a MODE or other specified condition in which the LCO is not applicable.

B. Incorrect; TS LCO 3.4.4 requires unit to be placed in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

C. Incorrect. TS LCO 3.3.1 requires that Thermal Power be raised greater than P-10 or less than P-6 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

D. Incorrect; TS LCO 3.2.3 is not applicable less than 50% RX Power.

Question meets K/A - The examinee must know the required one hour or less action statements for all Tech Specs.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

Technical Specifications 2.2 LCO 3.0.3 LCO 3.2.3 LCO 3.4.4

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 113 of 166 25 July 2019 72 ID: 2019 NRC-2 Q 72 Points: 1.00 Units 1 and 2 are at 100% power.

A leak must be repaired on a pipe in the Aux. Bldg. pipe tunnel.

The general area dose rate in the location of the repair is 600 mrem/hr.

In order to reach the location of the repair the worker must transit through a 6 Rem/hr high radiation area for 2 minutes and return via the same path.

The worker currently has an accumulated annual dose of 400 mrem.

Calculate the MAXIMUM allowable time that the worker can participate in the repairs AT THE JOB SITE and NOT exceed the Current Annual TEDE Administrative Dose Control Level IAW RP-AA-203, EXPOSURE CONTROL AND AUTHORIZATION.

A.

70 minutes B.

120 minutes C.

140 minutes D.

160 minutes Answer:

B Answer Explanation Explanation:

A) Incorrect. Answer is based on an Administrative Dose Control Level of 1500 mrem instead of a 2000 mrem limit.

B) Correct. The TEDE Administrative Dose Control Level is 2000 mrem annually. The operator possesses 400 mrem already + will receive 400 mrem during 4 minutes of transiting a 6 Rem/hr field and therefore will be limited to 120 minutes in a 600 mrem/hr field to conduct the repairs.

C) Incorrect. Answer is plausible for a 2000 mrem/yr limit, but only considers transiting in one direction.

D) Incorrect. Answer is plausible for a 2000 mrem/yr limit, but forgets to include transit time or the operators previously accrued dose.

Question meets K/A - The examinee must know the required one hour or less action statements for all Tech Specs..

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(12) Radiological safety principles and procedures.

References:

RP-AA-203, EXPOSURE CONTROL AND AUTHORIZATION

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 114 of 166 25 July 2019 73 ID: 2019 NRC-2 Q 73 Points: 1.00 You have been assigned to hang a Clearance in a congested radiologically controlled room. The room was recently surveyed and the following radiological conditions are listed on the RWP survey map:

General area radiation levels: 110 mR/Hr.

Airborne radiation is 0.25 DAC.

Contamination levels of 250dpm/100cm2 Beta/Gamma and 25 dpm/100cm2 Alpha.

Which of the following requirements must be met in order to perform the task?

1. High Rad Area (HRA) Briefing

2. Protective Clothing (PC)

3. Respirator A.

1 ONLY B.

2 ONLY C.

1 and 2 ONLY D.

1, 2, and 3 Answer:

C Answer Explanation Explanation:

A. Incorrect; PCs required due to contamination levels.

B. Incorrect; HRA Briefing required since general area radiation levels exceed 100 mr/hr.

C. Correct; HRA Briefing required since general area radiation levels exceed 100 mr/hr and PCs required due to contamination levels. Respirator not required since area does not meet the requirements to be considered an airborne radiation hazard; airborne rad level is less than 0.3 DAC.

D. Incorrect; Respirator not required since airborne rad level is less than 0.3 DAC.

Question meets K/A - The examinee must know the requirements for access to a radiological area during normal working conditions.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(12) Radiological safety principles and procedures.

References:

RP-AA-376; Radiological Posting, Labeling, and Markings; Rev 14 RP-AA-403; Administration of the Radiation Work Permit Program; Rev 9 RP-AA-410; Selection, Use, and Control of Protective Clothing; Rev 8 RP-AA-1008, Unescorted Access to and Conduct in Radiologically Posted Areas

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 115 of 166 25 July 2019 74 ID: 2019 NRC-2 Q 74 Points: 1.00 Which of the following parameters are required, by Technical Specifications, to have OPERABLE indications available in the Main Control Room for Post-Accident Monitoring?

(1) Reactor Coolant System (RCS) Pressure (2) Reactor Vessel Water Level (3) RCS Average Loop Temperature (Tave)

(4)

Pressurizer Water Level A.

1, 2, and 3 B.

2, 3, and 4 C.

1, 3, and 4 D.

1, 2, and 4 Answer:

D Answer Explanation Explanation:

A. Incorrect; Technical LCO 3.3.3 requires RCS Hot and Cold Leg temperatures but not loop Tave.

B. Incorrect; Technical LCO 3.3.3 requires RCS Hot and Cold Leg temperatures but not loop Tave.

C. Incorrect; Technical LCO 3.3.3 requires RCS Hot and Cold Leg temperatures but not loop Tave.

D. Correct; Technical LCO 3.3.3 requires OPERABLE instrumentation for RCS Pressure, Reactor Vessel Water Level and Pressurizer Water Level.

Question meets K/A - The examinee must know which instrumentation is required for PAMS as determined by LCO 3.3.3.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(6) Design, components, and function of reactivity control mechanisms and instrumentation.

References:

Technical Specification LCO 3.3.3 (Table 3.3.3-1)

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 116 of 166 25 July 2019 75 ID: 2019 NRC-2 Q 75 Points: 1.00 The Nuclear Accident Reporting System (NARS) is used to transmit information to ___(1)___. ___(2)___

must be notified FIRST following an initial emergency classification.

A.

1. the Nuclear Regulatory Commission (NRC)

2. the Nuclear Regulatory Commission (NRC)

B.

1. the Nuclear Regulatory Commission (NRC)

2. state and local agencies C.

1. state and local agencies

2. the Nuclear Regulatory Commission (NRC)

D.

1. state and local agencies

2. state and local agencies Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 117 of 166 25 July 2019 Explanation:

A. (1) Incorrect. Credible because there is nothing in the name that would make either state and local agencies or the Nuclear Regulatory Commission an obvious choice.

(2) Incorrect. Credible because the NRC has an important oversight function in emergencies.

B. (1) Incorrect. See A(1)

(2) Correct C. (1) Correct (2) Incorrect. See B(2)

D. (1) Correct. See above explanation.

(2) Correct. See above explanation.

Question meets K/A - The examinee must know the required notification times as outlined by the EP procedure.

10CFR55.41:

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

EP-AA-114, Notifications, Rev. 15

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 118 of 166 25 July 2019 76 ID: 2019 NRC-2 Q 76 Points: 1.00 A reactor startup was in progress.

Reactor power was holding at 7% in accordance with 1BGP 100-2, Plant Startup.

From these conditions, the following alarms were then received; S/G 1B FLOW MISMATCH FW FLOW LOW (1-15-B4)

S/G 1B LEVEL DEVIATION HIGH LOW (1-15-B9)

S/G 1B LEVEL LOW (1-15-B6)

S/G 1B LVL LO-2 RX TRIP ALERT (1-15-B5)

A manual reactor trip was unsuccessful.

The next required transition from 1BEP 0, Reactor Trip or Safety Injection, should be to ___(1)___ AND the emergency plan classification should be ___(2)___.

A.

1. 1BFR-S.1, Response to Nuclear Power Generation/ATWS

2. Alert B.

1. 1BFR-S.1, Response to Nuclear Power Generation/ATWS

2. no classification C.

1. 1BEP ES-0.1, Reactor Trip Response

2. Alert D.

1. 1BEP ES-0.1, Reactor Trip Response

2. no classification Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 119 of 166 25 July 2019 Reference provided: EAL Matrix Explanation: Since reactor power is steady at 7%, the SUBCRITICALITY CSFST is RED. The condition of Step 1 RNO of 1BEP 0 (Rx power >5%) is met, so the crew should proceed to 1BFR-S.1. The Alert condition in the EAL matrix associated with RPS failure applies if Reactor Power is > 5%.

A. (1) Correct. See above explanation.

(2) Correct. See above explanation.

B. (1) Incorrect. See D (1)

(2) Correct C. (1) Correct (2) Incorrect. See D (2)

D. (1) Incorrect. Credible because this is the procedure entry to be made without going to 1BFR-S.1.

(2) Incorrect. Credible because, without entry to 1BFR-S.1, no EAL classification would be required.(2) Correct. See above explanation.

Question meets K/A - The examinee must know the entry conditions for 1BSR S.1, Response to Nuclear Power Generation/ATWS.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

References:

1BEP 0 R t

T i S f t I j ti R

302

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 120 of 166 25 July 2019 77 ID: 2019 NRC-2 Q 77 Points: 1.00 The Bases for LCO 3.4.13.d, RCS Operational Leakage, Primary to Secondary LEAKAGE through any one SG, is ___(1)___.

LCO 3.4.13.d compliance is periodically confirmed by ___(2)___.

A.

1. meeting the assumptions of the steam generator tube rupture analysis.

2. performing an RCS water inventory balance.

B.

1. meeting the assumptions of the steam generator tube rupture analysis.

2. evaluating radiation monitor readings and SG sample results.

C.

1. minimizing the frequency of tube ruptures.

2. performing an RCS water inventory balance.

D.

1. minimizing the frequency of tube ruptures.

2. evaluating radiation monitor readings and SG sample results.

Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 121 of 166 25 July 2019 Explanation:

TS Bases Section B 3.13.d states:

"The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

A. (1) Incorrect. Plausible because initial leakage is a consideration in SGTR dose calculation, but this this value is not limiting. The LCO requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis.

(2) Incorrect. Plausible because this is how other leak rates are determined. However, the basis states that LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

B. (1) Incorrect. See A(1)

(2) Correct. Only rad monitor readings and sample results are useful to measure leak rates this small.

C. (1) Correct. See explanation above.

(2) Incorrect. See A(2)

D. (1) Correct. See explanation above.

(2) Correct. Only rad monitor readings and sample results are useful to measure leak rates this small.

Question meets K/A - The examinee must know the Tech Spec requirements for SG Primary to Secondary leakage. This is a SRO ONLY question due to the bases knowledge required to determine the compliance needs.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(2) Facility operating limitations in the technical specifications and their bases.

References:

TS basis for LCO 3.4.13

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 122 of 166 25 July 2019 78 ID: 2019 NRC-2 Q 78 Points: 1.00 Unit 1 is in MODE 4 with RCS cooldown and depressurization in progress.

1-7-C2, RCP LOWER BRNG TEMP HIGH is in alarm The following is observed on 1C Reactor Coolant Pump (RCP).

Seal Leak-off Flow is 3.0 gpm and stable.

Seal Differential Pressure 300 psid and stable.

Seal Outlet Temperature is 165°F and rising slowly.

Lower Radial Bearing Temperature185°F and rising slowly.

Which of the following describes the operational status of 1C RCP?

A.

Operation may continue since Seal Leak-off Flow is < 5 gpm.

B.

Operation may continue since the Seal Outlet Temperature is < 235°F.

C.

Operation may continue since the Lower Radial Bearing Temperature is < 225°F D.

An immediate shutdown of 1C RCP is required, since one or more operational limits has been exceeded.

Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 123 of 166 25 July 2019 Explanation:

A. Incorrect; plausible since 5 gpm is the upper limit for #1 Seal Leak-off flow during at power operation.

B. Incorrect; plausible since 235°F Seal Outlet temperature is upper limit for pump operation if the temperature is stable.

C. Incorrect; plausible since 225°F Lower Radial Bearing temperature is the upper limit for pump operation if the temperature is stable.

D. Correct; seal leak-off flow is abnormally high for the specified differential pressure and Lower Radial Bearing and Seal Outlet temperatures are not stable.

Question meets K/A - Question is a K/A match since the stated parameters are used to monitor RCP performance and abnormal values could lead to trip or shutdown of a RCP by assessing the abnormal conditions..

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References:

1(2)BOA RCP-1; REACTOR COOLANT PUMP SEAL FAILURE UNIT 1(2); Rev 105(107)

N-BY-TQ-ILT-RC-13; Reactor Coolant Pump; Rev 10 I1-OA-XL-27; 1BOA RCP-1, RCP Seal Failure; Rev 9 Proposed references to be provided to applicants during examination:

FIGURE 1BOA RCP-1-1, NO. 1 SEAL NORMAL OPERATING RANGE

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 124 of 166 25 July 2019 79 ID: 2019 NRC-2 Q 79 Points: 1.00 Unit 1 was at rated thermal power when:

SAT 142-1 faults.

The 1B Essential Service water pump did not automatically start, and was manually started by the crew.

If SAT fault is not repaired, Unit 1 must be in MODE 5 in ______ hours from the time of the power loss.

A.

37 B.

48 C.

60 D.

108 Answer:

D Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 125 of 166 25 July 2019 Explanation: To select the correct answer, the applicant must understand:

Unit 1 has ONE qualified circuit. The SAT fault will trip both SATs, however, a qualified circuit is still available from U2 via the STA, ESF Bus, and the crosstie breakers.

Failure of the 1B Essential Service water pump to automatically start makes this train inoperable, since SR 3.7.8.5 (auto start) is not met.

Even though 1B SX Pump is INOPERABLE due to the failure to auto start, the cascade to the 1B DG is not applicable due to the SX crosstie valves being normally open. This allows an SX supply source from an OPERABLE train to support the DG.

A. Incorrect. This is the time if Action H is limiting (Action A, H). This would be applicable if the 1B DG was inoperable and the candidate failed to identify the second qualified circuit comes from U2.

B. Incorrect. This is the time if Action E is limiting (Action E, then G). This would be applicable if the 1B DG was inoperable and the candidate correctly identified the loss of one qualified circuit.

C. Incorrect. This is the time if Action D is limiting (Action D, then G). This would be applicable by identifying the 1B DG was OPERABLE and failure to identify the second qualified circuit from U2.

D. Correct. This is the time if Action A is limiting (action A, then G). The loss of the U1 SATs only creates a loss of ONE qualified circuit for each bus.

Question meets K/A - This question is SRO only because the status of the 1B DG has to be assessed against the impact of the SX pump inoperability and then evaluated against the lost qualified circuit. These assessment and evaluation requires bases knowledge and evaluation between LCO 3.0.6, 3.7.8, and 3.8.1.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(2) Facility operating limitations in the technical specifications and their bases.

References:

LCO 3.8.1, AC Sources - Operating LCO 3.7.8, Essential Service Water (SX) System LCO 3.0.6, Support Systems Proposed references to be provided to applicants during examination:

LCO 3.8.1 AC Sources - Operating

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 126 of 166 25 July 2019 80 ID: 2019 NRC-2 Q 80 Points: 1.00 Unit 1 was at 100%.

1-22-E10, 125V DC PNL 112/114 VOLT LOW is LIT DC Bus 112 Voltage indicates 0 volts With no operator action, the plant will trip on ___(1)___. After the trip, normal PZR level control can be established by first performing ___(2)___.

A.

1. steam generator water level low low

2. 1BOA ESP-2, Re-establishing CV Letdown, then 1BOA ELEC-1, Loss of DC Bus B.

steam generator water level low low 1BOA ELEC-1, Loss of DC Bus, then 1BOA ESP-2, Re-establishing CV Letdown C.

1. reactor coolant flow low

2. 1BOA ESP-2, Re-establishing CV Letdown, then 1BOA ELEC-1, Loss of DC Bus D.

1. reactor coolant flow low

2. 1BOA ELEC-1, Loss of DC Bus, 1BOA ESP-2, Re-establishing CV Letdown Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 127 of 166 25 July 2019 Explanation: When either bus 111 or 112 is lost, feed regulating and bypass valves will fail closed, reducing S/G inventory to the Low-Low S/G level reactor trip setpoint.

The loss of letdown is caused by loss of instrument air to containment. 1IA066, IA CNMT inside isolation valve, has lost power to its solenoid and must be opened locally. This instruction is only found in 1BOA ELEC-1 and must be performed first to enable letdown restoration. The applicant has to know that IA is isolated to containment to get the correct answer and that of these two, only 1BOA ELEC-1 contains instructions to open it. It is SRO only because the applicant must prioritize procedure use.

A. (1) Correct (2) Incorrect. Plausible because this is how letdown is usually restored.

B. (1) Correct (2) Correct C. (1) Incorrect. Plausible because one reactor coolant pump will lose power after the trip because of failure to transfer power to offsite.

(2) Incorrect. See A(2)

D. (1) Incorrect. See C(1)

(2) Correct Question meets K/A - This question meets the K/A as the examinee is required to know the impact of a loss of DC to the RCS. This is an SRO ONLY question due to the required assessment of the conditions and determination of the necessary procedure and actions to mitigate the issue.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References:

Chapter 8A/I1-DC-XL-01 1BOA ESP-2, Re-establishing CV Letdown Rev. 1 1BOA ELEC-1, Loss of DC Bus Unit 1, Rev. 104

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 128 of 166 25 July 2019 81 ID: 2019 NRC-2 Q 81 Points: 1.00 Unit 1 was at 100% rated thermal power when; The following alarms came in and remained:

PMG OUTPUT BRKR TRIP (1-19-A8)

GENERATOR FIELD FORCING (1-19-B6)

Exciter current stabilized at 106 amps.

The ___(1)___ Alarm is NOT VALID for plant conditions.

If the Unit Supervisor subsequently decides to disable the alarm that is INCONSISTENT with plant conditions, this must be done in accordance with the instructions of ___(2)___.

A.

1. PMG OUTPUT BRKR TRIP

2. OP-AA-103-102, Watchstanding Practices B.

1. GENERATOR FIELD FORCING

2. OP-AA-109-101, Clearance and Tagging C.

1. PMG OUTPUT BRKR TRIP

2. OP-AA-103-102, Watchstanding Practices D.

1. GENERATOR FIELD FORCING

2. OP-AA-109-101, Clearance and Tagging Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 129 of 166 25 July 2019 Explanation: The Generator Field Forcing alarm comes in when exciter current exceeds 100 amps. This alarm is consistent with plant conditions.

When the PMG output breaker trips, excitation goes to zero (and the generator trips). This alarm is NOT consistent with plant conditions.

OP-AA-103-102, p. 9 of 16 contains instructions regarding disabling nuisance alarms.

A. (1) Incorrect. It is credible that a breaker trip could cause high excitation because loss of the current sensing signal would result in high excitation current. The applicant has to know the function of the PMG to eliminate this answer.

(2) Correct B. (1) Incorrect. See A(1)

(2) Incorrect. Credible because disabling an alarm involves applying an Equipment Tracking Tag. OP-AA-109-1 contains the instructions for WTO Tracking list tags.

There is nothing about these procedure titles that make it obvious which one would contain these instructions.

C. (1) Correct. See explanation above.

(2) Correct. See explanation above.

D. (1) Correct (2) Incorrect. See B(2).

Question meets K/A - This question meets the K/A because the examinee is required to understand the implications of the associated alarms as to the status of the generator.

Question is SRO only because it tests administrative processes for disabling annunciators and also because it requires the applicant to assess and select a procedure.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(3) Facility licensee procedures required to obtain authority for design and operating changes in the facility.

References:

BAR 1-19-B6, GENERATOR FIELD FORCING, Rev 010 BAR 1-19-A8, PMG OUTPUT BRKR TRIP, Rev. 4 OP-AA-109-101, Clearance and Tagging, Rev 012 OP-AA-103-102, Watchstanding Practices, Rev. 18 Chapter 6, Generator Excitation, i1-MP-XL-02.

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 130 of 166 25 July 2019 82 ID: 2019 NRC-2 Q 82 Points: 1.00 Unit 1 was in a refueling outage when a fuel assembly was dropped in the spent fuel pool.

Fuel Building Fuel Handling Incident Radiation monitor readings peaked at; 0RE-AR55 - 1115 mRem/hr 0RE-AR56 - 1075 mr/hr The radiation monitors in 1BOA REFUEL-1, Fuel Handling Emergency, are used to determine ___(1)___.

This event should be classified as an ___(2)___.

A.

1. the location of the accident

2. Unusual Event B.

1. the location of the accident

2. Alert C.

1. if the affected area must be evacuated

2. Unusual Event D.

1. if the affected area must be evacuated

2. Alert Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 131 of 166 25 July 2019 Explanation: 1BOA REFUEL-1 uses the incident monitors to determine the location of the accident (The K/A is to determine and interpret ARM indications). The affected area is automatically evacuated regardless of area rad monitor readings.

RA2, criteria is MET for an Alert classification.

A. (1) Correct (2) Incorrect. Credible because an Unusual event also uses the rad monitors in Table R1 as a criterion.

B. (1) Correct (2) Correct C. (1) Incorrect. Credible because area rad monitors could indicate habitability.

(2) Incorrect. See A(2)

D. (1) Incorrect. See C(1)

(2) Correct Question meets K/A - This meets the K/A as it requires the examinee to determine the applicability of the radiation monitor for the associated accident. This question is SRO only because requires E-plan classification.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(7) Fuel handling facilities and procedures.

References:

1BOA REFUEL-1, Fuel Handling Emergency Chapter 49, I1-AR-XL-01 EAL Matrix, Byron Annex Proposed references to be provided to applicants during examination:

EAL Matrix, Byron Annex

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 132 of 166 25 July 2019 83 ID: 2019 NRC-2 Q 83 Points: 1.00 Unit 2 was at 100% rated thermal power when; STEAM GENERATOR LEAKAGE EXCEEDED (2-16-A5) alarmed SJAE/GLAND STEAM EXHAUSTER HIGH (2PR27J) alarmed SG BLOWDOWN RADIATION HIGH (2PR08J) alarmed The crew entered 2BOA SEC-8, Steam Generator Tube Leak The crew determined that the initial leak rate was 90 gpd After one hour, the leak rate was confirmed by two independent indications to be steady at 90 gpd Over this hour, the ___(1)___ radiation monitor provided a more accurate indication of the leak rate trend.

After this hour has elapsed, the crew is required to reduce power per ___(2)___.

A.

1. SG BLOWDOWN

2. 2BGP 100-4T1, Power Descension Flowchart B.

1. SG BLOWDOWN

2. 2BGP 100-4T1.1, Rapid Power Reduction Flowchart C.

1. SJAE/GLAND STEAM EXHAUSTER

2. 2BGP 100-4T1, Power Descension Flowchart D.

1. SJAE/GLAND STEAM EXHAUSTER

2. 2BGP 100-4T1.1, Rapid Power Reduction Flowchart Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 133 of 166 25 July 2019 Explanation:

The NOTE prior to step 6 of 2BOA SEC-8 says that the blowdown monitors may rise for 6 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with a steady leak rate as the SG activity stabilizes. The trend on the monitor is not useful to determine if leak rate is changing.

If leak exceeds 75 gpd for > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, power must be reduced per 2BGP 100-4T1. The rapid power reduction flowchart would only be used if leak rate reached 100 gpd.

A. (1) Incorrect. See explanation. Credible because secondary radiation is a key indicator of SG tube leakage.

(2) Correct. See C(2)

B. (1) Incorrect. See A(1).

(2) Incorrect. See explanation. Credible because a rapid power reduction is required for a greater leak rate.

C. (1) Correct. Steps 7.d. and e. of 2BOA SEC-8 uses the SJAE trend to decide if leak rate is stable or rising.

(2) Correct. Step 10.a RNO directs use of 2BGP 100-4T1 if leak rate is < 100 gpd.

D. (1) Correct. See C(1)

(2) Incorrect. See B(2).

Question meets K/A - The examinee is required to know the best radiation monitor to determine flow rate. This is SRO ONLY as it requires detailed assessment of the procedure in order to determine the specific power reduction and the necessary procedure for execution.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References:

2BOA SEC-8, Steam Generator Tube Leak, Rev. 111 2BGP 100-4, Power Descension, Rev, 52 2BOSR SG-1, Steam Generator Primary to Secondary Leakage Estimation.

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 134 of 166 25 July 2019 84 ID: 2019 NRC-2 Q 84 Points: 1.00 The control room has been evacuated in accordance with 1BOA PRI-5, Control Room inaccessibility.

1B hot leg instrument 1B TI-RC06A on remote shutdown panel 1PL05J is reading off-scale low (0°F).

1B hot leg instrument 1TI-RC023A on Fire Hazards Panel 1PL10J is indicating the same as the other hot legs (~572°F).

LCO 3.3.4, Remote Shutdown System, ___(1)___ MET.

If an SI signal is received during performance of 1BOA PRI-5, the crew is expected to terminate SI in accordance with ___(2)___.

A.

1. IS NOT

2. Attachment F of 1BOA PRI-5, Control Room inaccessibility B.

1. IS NOT

2. 1BEP ES-1.1, SI Termination C.

1. IS

2. 1BEP ES-1.1, SI Termination D.

1. IS

2. Attachment F of 1BOA PRI-5, Control Room inaccessibility Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 135 of 166 25 July 2019 Explanation: The requirement for hot leg temperature is one per loop. No credit can be taken for panel 1PL10J according to 1BOSR 3.4.1.1, Remote Shutdown Instrumentation Monthly Surveillance. LCO 3.3.4 is not met.

The note prior to step 1 of 1BOA PRI-5 says Entry into 1BEP-0 is not required. E-0 is designed to be performed from the control room and cannot be performed at the remote shutdown panel. For example, there are no rod bottom lights at the RSP. There is no way to get to 1BEP ES-1.1 if 1BEP 0 is not entered.

A. (1) Correct. See explanation above.

(2) Correct. See explanation above. Step 11 of 1BOA-PRI-5 directs use of attachment F if SI actuates.

B. (1) Correct (2) Incorrect. See explanation. Credible because emergency procedures are normally used for SI response C. (1) Incorrect. See explanation. Credible because information on panel 1PL10J is available and will be used for info not available on the remote shutdown panel.

(2) Incorrect. See B(2)

D. (1) Incorrect. See C(1)

(2) Correct. See A(1)

Question meets K/A - The examinee is required to know the specific instrumentation that is available from the Remote Shutdwon Panel vs the Main Control Room. This is SRO ONLY as an analysis and determination is required for securing the Safety Injection.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References:

LCO 3.3.4 and Bases 1BOA PRI-5, Control Room Inaccessibility, Rev. 109.

1BOSR 3.4.1-1, Remote Shutdown Instrumentation Monthly Surveillance, Rev 9.

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 136 of 166 25 July 2019 85 ID: 2019 NRC-2 Q 85 Points: 1.00 Unit 1 is in MODE 4.

• Cooldown of the RCS and Pressurizer is in progress, using RHR with 1D RCP running.

• The NSO has recorded the following temperatures:

RCS Tcold PZR Liquid 08:00 320°F 460°F 08:30 260°F 400°F 09:00 210°F 340°F In accordance with TS and TRM requirements, the SRO will....

A.

allow the cooldown of both the RCS and PZR to continue B.

stop the cooldown of both the RCS and PZR C.

stop the cooldown of the RCS and allow the cooldown of the PZR to continue D.

stop the cooldown of the PZR and allow the cooldown of the RCS to continue Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 137 of 166 25 July 2019 Explanation:

A) Incorrect. Plausible because the PZR cooldown rate is 120°F/hour which < then the TRM 3.4.c limit of 200°F/hour and is therefore acceptable. The RCS cooldown is 110°F/hour, which is higher than the limit of 100°F/hour found in the PTLR and TS 3.4.3, therefore RCS cooldown must be stopped.

B) Incorrect. Plausible as the RCS cooldown must be stopped, but the PZR may continue.

C) Correct. The PZR cooldown rate is 120°F/hour which < then the TRM 3.4.c limit of 200°F/hour and is therefore acceptable. The RCS cooldown is 110°F/hour, which is > than the limit of 100°F/hour found in the PTLR and TS 3.4.3, therefore RCS cooldown must be stopped.

D) Incorrect. Plausible as one of the cooldowns can continue, while the other must be stopped.

In addition, the applicant may note that the cooldown rate for the PZR is higher than that of the RCS and assume it is the limiting condition for operation.

Question meets K/A - Question meets K/A/SRO Only - SRO candidate must evaluate the cooldown rate, and apply the limits from the TS and TRM in order to provide the proper direction to the crew.

10CFR55.43(b)(2) 10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(2) Facility operating limitations in the technical specifications and their bases.

References:

1BGP 100-5. PLANT SHUTDOWN AND COOLDOWN Technical Specification 3.4.3, RCS Pressure and Temperature (P/T) Limit Technical Requirement Manual 3.4.c, Pressurizer Temperature Limits Unit 1 Pressure Temperature Limits Report

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 138 of 166 25 July 2019 86 ID: 2019 NRC-2 Q 86 Points: 1.00 Unit 1 is performing 1BGP 100-1, Plant Heatup.

RCS temperature is 150°F.

A Precaution in 1BGP 100-1 to apply a correction factor to RCS pressure when starting a RCP is required because indicated Wide Range pressure will be ___(1)____ than actual RCS pressure. If the limit of Pressure and Temperature Limits Report Figure 2.1 is exceeded, technical specifications requires restoring parameters to within limits AND ___(2)___.

A.

1. less

2. establishing an RCS vent of 2.0 square inches B.

1. less

2. performing an engineering evaluation prior to MODE 4 entry C.

1. greater

2. establishing an RCS vent of 2.0 square inches D.

1. greater

2. performing an engineering evaluation prior to MODE 4 entry Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 139 of 166 25 July 2019 Explanation: This is a K/A match because it tests the operation of Reactor Coolant Pumps in an integrated plant procedure. 1BGP 100-1 has a repeating caution and separate step that margin must be added to the indicated pressure when RCPs are running (35 psi for two RCPs).

Starting a reactor coolant makes RCS wide range pressure not conservative with respect to cold over-pressure conditions.

Action C.2, as explained in the TS 3.4.3 bases, requires an engineering evaluation prior to resuming RCS operation:

Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify that the RCPB integrity remains acceptable and must be completed prior to entry into MODE 4.

It is part 2 of the question that makes it SRO only. It requires knowledge of the actions statements, which is below the line.

A. (1) Correct.

(2) Incorrect. Credible because this is a required action found in LCO 3.4.12, Low Temperature Over Pressure. It is a reasonable because it ensures that the RCS cannot be pressurized, but it is not required in this case.

B. (1) Correct. See above explanation.

(2) Correct. See above explanation.

C. (1) Incorrect. Credible if the pressure instrument is nearer the discharge of the RCP than to the vessel.

(2) Incorrect. See A(2)

D. (1) Incorrect. See C(1)

(2) Correct Question meets K/A - The examinee is required to know the applicable Tech Spec for the PTLR and the applicable mode applicable. This is SRO ONLY because it requires an in depth knowledge of the bases to determine the required action for exceeding the PTLR limits.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(2) Facility operating limitations in the technical specifications and their bases.

References:

1BGP 100-1, Plant Heatup Rev 064.

Byron Unit 1 Pressure and Temperature Limits Report, November 2015.

LCO 3.4.3

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 140 of 166 25 July 2019 87 ID: 2019 NRC-2 Q 87 Points: 1.00 Unit 2 is at 100%.

• 2A SX pump RUNNING.

• 2A, C, and D RCFC are running in FAST speed.

Subsequently, the following annunciators are received:

• 2-1-A2, CNMT DRAIN LEAK DETECT FLOW HIGH

• 2-2-A2, SX PUMP DSCH HDR PRESS LOW As Unit Supervisor, you direct entry into the appropriate annunciator response procedures and enter 2BOA PRI-7, ESSENTIAL SERVICE WATER MALFUNCTION UNIT 2. What operator actions will you direct to address the plant conditions?

A.

Start the 2B RCFC as necessary to maintain containment temperature and pressure conditions.

B.

Shutdown the 2D RCFC and isolate SX flow to the 2B and 2D RCFCs. If conditions do not improve, restore SX flow to the 2B and 2D RCFCs and swap SX pumps before isolating the 2A and 2C RCFCs.

C.

Stop all RCFCs running in FAST speed, isolate and bypass the 2A and 2B containment chillers, ensure all RCFC inlet and outlet valves are open and start all RCFCs in SLOW speed.

D.

Manually actuate PHASE A containment isolation to stop the source of the leak and monitor containment parameters every 30 minutes.

Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 141 of 166 25 July 2019 Explanation:

A) Incorrect. Plausible because this is what operators would be expected to do following an RCS leak into the primary containment IAW 2BOA PRI-1.

B) Correct. 2BOA PRI-7 has the crew stop the RCFCs associated with the non-running 2B SX pump (2B and 2D) and then isolate SX to them. If this does not work, then the RCFC are restored, SX is swapped and the A and C RCFC would be stopped and have SX isolated to them.

C) Incorrect. Plausible as this is what operators would have to do if RCFCs did not shift to Accident Mode following a Safety Injection on an RCS leak per 2BEP-0.

D) Incorrect. Plausible as this is what operators would do if a PHASE A containment isolation would be expected and did not occur following an RCS leak into containment IAW 2BEP-0, and containment parameter monitoring would be performed IAW 2BOSR RF-1 as directed by annunciator response procedure 2-1-A2 Question meets K/A - The examinee is required to know and understand the impact of a SX leak inside Containment. SRO Only Justification: Tests applicant on the content of an Abnormal operating procedure and not just its overall mitigative strategy.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References:

2BOA PRI-7, ESSENTIAL SERVICE WATER MALFUNCTION UNIT 2 BAR 2-1-A2, CNMT DRAIN LEAK DETECT FLOW HIGH 2BOA PRI-1, EXCESSIVE PRIMARY PLANT LEAKAGE UNIT 2 BAR 2-2-A2, SX PUMP DSCH HDR PRESS LOW

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 142 of 166 25 July 2019 88 ID: 2019 NRC-2 Q 88 Points: 1.00 Both Unit 1 and Unit 2 were at rated thermal power.

1BOSR 8.6.1-1, 125V DC ESF Battery Bank and Charger 111 Operability Weekly Surveillance, was just completed.

At 0800, May 1, the following results are provided to the control room.

DESCRIPTION RESOLUTION READING TO RECORD READING T.S.

VALUE PILOT CELL VOLTAGE 10VDC(X.XXXXX)

LOWEST READING 2.1 VDC

>2.07 VDC BATTERY TERMINAL VOLTAGE 1000VDC (XXX.XXX)

LOWEST READING 126.9 VDC

> 127.6 VDC CURRENT SHUNT RESISTOR VOLTAGE 100Mvdc (X.XXX)

HIGHEST READING 0.1 m VDC

< 0.3 mVDC If no repairs are made, Unit 1 must be in MODE 3 no later than A.

1600 May 1 B.

1400 May 8 C.

1600 May 8 D.

0400 May 10 Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 143 of 166 25 July 2019 Explanation: This tests the K/A (knowledge of surveillance procedures) because the applicant must understand the purpose of the procedure in question to get the correct answer. The bases for SR 3.8.4.1 (Terminal voltage) says Verifying battery terminal voltage while on float charge helps to ensure the effectiveness of the battery chargers. If terminal voltage fails to meet its SR, it is the battery charger that is inoperable. This requires entry into LCO 3.8.4 Action A (7 days to restore), then Action E (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to MODE 3), or May 8 @ 1400. The question is SRO only because it tests TS bases.

A. Incorrect. Credible if applicant enters actions D and E.

B. Correct C. Incorrect. Credible if applicant believes that when cross tie is closed, only actions C and E must be finished (2 hr to crosstie, then enter action C, 7days to restore, then enter action E, 6hr to MODE 3).

D. Incorrect. Credible if applicant believes that when cross tie is closed, only actions B and E must be finished (2 hr to cross tie, 8.5 days to restore, then action E, 6 hr to MODE 3.

Question meets K/A - requires the examinee to know the requirements of the surveillance and the applicable Tech Spec when the surveillance requirement is not met.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(2) A Facility operating limitations in the technical specifications and their bases.

References:

LCO 3.8.4 B 3.8.4 1BOSR 8.6.1-1, 125V DC ESF Battery Bank and Charger 111 Operability Weekly Surveillance, Rev. 20 References to be provided to applicants during examination:

LCO 3.8.4, DC Sources - Operating

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 144 of 166 25 July 2019 89 ID: 2019 NRC-2 Q 89 Points: 1.00 Unit 1 was at 100% rated thermal power.

On May 1, the crew was performing 1BOSR 8.1.2-1, 1A Diesel Generator Operability Surveillance This 31 day surveillance interval was due to expire on May 2 @ 0500.

On May 1 @ 0500, the Diesel Gen 1A Feed to 4KV Bus 141 Sync ACB 1413 was turned to ON, but the synchscope did not respond to manipulations of the 1A Gov Adj control.

Subsequent investigation revealed that the synchronizing circuit had failed.

Repair parts are not on site and will have to be ordered.

If the condition is not repaired, the SRO must declare LCO 3.8.1 NOT MET on A.

May 1 @ 0500 B.

May 2 @ 0500.

C.

May 9 @ 2300.

D.

June 9 @ 2300.

Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 145 of 166 25 July 2019 Surveillance Frequency Control Program Explanation: With the synchronizing circuit failed, SR 3.8.1.3 cannot be met. SR 3.0.1 says Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO.

The LCO was not met from the time of discovery that synchronizing was not possible. This is SRO only because it involves an application of SR 3.0.1.

A. Correct B. Incorrect. Plausible because this is when the surveillance interval expires.

C. Incorrect. Plausible because this is when the surveillance interval + 25% extension expires.

D. Incorrect. Plausible because SR 3.0.3 (missed surveillance) would allow an additional 31 days + 25% extension if it applied (it does not apply).

Question meets K/A - requires the examinee to the failures impact on operability and then evaluate SR 3.0.1 for possible extension of surveillance requirements.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(2) A Facility operating limitations in the technical specifications and their bases.

References:

LCO 3.8.1 SR 3.0.1 Surveillance Frequency Control Program References to be provided to applicants during examination: LCO 3.8.1

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 146 of 166 25 July 2019 90 ID: 2019 NRC-2 Q 90 Points: 1.00 Refueling operations were in progress in Unit 1 when:

1-1-C1, SPENT FUEL PIT LEVEL HIGH LOW, is LIT.

The local level indication showed 23 6 1/2 on ABOVE SPENT FUEL ruler.

This condition could have been caused by trip of the running Containment Mini-purge ___(1)___ fan.

The limiting basis of LCO 3.7.14, Spent Fuel Pool Water Level is ___(2)___.

A.

1. Supply

2. meeting an assumption of the fuel handling accident analysis B.

1. Supply

2. ensuring adequate heat removal for the spent fuel C.

1. Exhaust

2. meeting an assumption of the fuel handling accident analysis D.

1. Exhaust

2. ensuring adequate heat removal for the spent fuel Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 147 of 166 25 July 2019 Explanation: 23 6 1/2 on ABOVE SPENT FUEL ruler is a low level condition. The first part of the question tests the precaution from BOP VQ-5, Containment Purge System Operation -

Startup and Shutdown of the Containment Purge System may impact Spent Fuel Pool Level if the Spent Fuel Pool is connected to the Refueling Cavity via the Transfer Canal. Trip of the supply fan will lower containment pressure, causing water to flow from the spent fuel pit to the refueling cavity through the transfer canal. The second part tests TS bases, which makes it SRO-only.

A. (1) Correct. See explanation above.

(2) Correct. See explanation above.

B. (1) Correct (2) Incorrect. Plausible because heat removal is a function of the water, but that function is not limiting for the LCO.

C. (1) Incorrect. Plausible because the alarm does not specify is this level is high or low.

(2) Correct D. (1) Incorrect. See C(1)

(2) Incorrect. See B(2)

Question meets K/A - requires the examinee to understand the impact of the failure and to be able to apply Tech Specs for the failure.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(2) AFacility operating limitations in the technical specifications and their bases.

References:

B 3.7.14, Spent Fuel Pool Water Level bases BOP VQ-5, Containment Purge System Operation, Rev. 6

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 148 of 166 25 July 2019 91 ID: 2019 NRC-2 Q 91 Points: 1.00 In compliance with TRM Surveillance Requirement 3.9.c.1, 1BOSR 9.c.1-1, Refueling Machine Operability Surveillance, must be performed within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of movement of fuel assemblies within the reactor vessel to test, in part, fuel handling system protective interlocks.

Which of the following Interlocks is required by the TRM?

A.

Excessive suspended weight - stop upward hoist motion with a load 2850lb B.

Hoist-Gripper Position - permit upward hoist motion only when gripper indicating switch shows open or closed C.

Gripper-Weight Indicator - prevents gripper from opening until it senses 1200lb on refueling machine D.

Bridge-Trolley-Hoist - Permits only ONE drive to be operable at any one time Answer:

A Answer Explanation Explanation:

A) Correct. Two excessive suspended weight switches open the hoist drive circuit in the up direction to protect the fuel assembly and the refueling machine. The first switch is set to open the hoist drive circuit in the up direction, if the suspended weight is in excess of 2850 pounds (approximately 110% of the combined weight of a fuel assembly, RCCA and the gripper mast).

This is a concern if a fuel assembly is binding with adjacent components. The second switch is set to open the hoist drive circuit in the up direction if the load is in excess of 3400 pounds (approximately 125% of the combined load of a fuel assemble, RCCA, and the gripper mast).

B) Incorrect. Plausible as this is an actual interlock of the fuel handling system.

C) Incorrect. Plausible as this is an actual interlock of the fuel handling system.

D) Incorrect. Plausible as this is an actual interlock of the fuel handling system.

Question meets K/A - SRO Only Justification: Requires knowledge of TRM surveillance requirements.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(7) Fuel handling facilities and procedures.

References:

TRM 3.9.c, REFUELING MACHINE/AUXILAIRY HOIST/CAVITY MAINTENANCE CRANE 1BOSR 9.c.1-1, REFUELING MACHINE OPERABILITY SURVEILLANCE

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 149 of 166 25 July 2019 92 ID: 2019 NRC-2 Q 92 Points: 1.00 The Unit 1 crew took the following actions in accordance with 1BGP 100-3, Power Ascension.

Raised Steam dump demand to 40% on 1UI-500.

Synchronized the main turbine.

Began raising turbine load to 125 MW.

At this time, steam dump valve MS004A stuck in the fully open position.

The Steam Dumps will initially control steam header pressure ___(1)___ the setpoint.

Manual isolation of the stuck open steam dump valve can be done in accordance with the instructions found in ___(2)___.

A.

1. BELOW

2. OP-AA-103-102-1001, Strategies for Successful Transient Mitigation B.

1. BELOW

2. BOP MS-100 Main Steam Dump Operation C.

1. AT

2. OP-AA-103-102-1001, Strategies for Successful Transient Mitigation D.

1. AT

2. BOP MS-100 Main Steam Dump Operation Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 150 of 166 25 July 2019 Explanation: The stream pressure controller is a ProportionaI-Integral (PI) controller. The integral feature will cause the controller to change its output as necessary until steam header pressure is being controlled at the setpoint.

No procedure directs isolation of a stuck open steam dump valve. Thus, section 4.1.2.3, B of OP-AA-103-102-1001 applies.

A. (1) Incorrect. Plausible because a stuck open steam dump valve would tend to make steam pressure too low. However, the PI controller will overcome this.

(2) Correct B. (1) Incorrect. See A(1)

(2) Incorrect. Plausible because this is a likely place for such instructions.

C. (1) Correct (2) Correct D. (1) Correct (2) Incorrect. See B(2)

Question meets K/A - requires the examinee to know how the steam dumps will respond to the failure and where the procedural guidance is for this situation.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References:

1BGP 100-3, Power Ascension Horse notes MS-4, Main Steam Dumps OP-AA-103-102-1001, Strategies for Successful Transient Mitigation, Rev 2 BOP MS-100 Main Steam Dump Operation, Rev. 0

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 151 of 166 25 July 2019 93 ID: 2019 NRC-2 Q 93 Points: 1.00 Both Byron Units 1 and 2 are initially at 100% power.

1A Service Air Compressor (SAC) is out of service for major maintenance.

1B SAC is aligned for STANDBY 1 operation 2A SAC is aligned for LAG operation 2B SAC is aligned for LEAD operation Unit 2 experiences a complete Loss of Off-Site Power followed by a trip of Unit 2.

Instrument Air (IA) Header Pressure is observed to be 100 psig and dropping at a rate of 0.5 psig/minute.

With respect to UNIT 1, (1) Identify the automatic actions expected to take place within the next 30 minutes.

(2) Identify the mitigating actions that should be directed.

AUTOMATIC ACTIONS MITIGATING ACTIONS

1. SAC 1B automatic start

2. In-service IA dryer(s) automatic bypass

3. Letdown system isolation X. Isolate IA to Unit 2 to maximize IA availability to Unit 1 Y. Restore power to Bus 244 Z. Control Charging flow to maintain Pressurizer Level less than 60%

A.

1. 1 ONLY

2. X ONLY B.

1. 2 ONLY

2. Y ONLY C.

1. 1 and 2 ONLY

2. X and Y ONLY D.

1. 1, 2, and 3

2. X, Y, and Z Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 152 of 166 25 July 2019 Explanation:

Each of the automatic actions listed are expected to occur as air pressure drops. The actions are listed in order of descending air pressure. Dropping air pressure is an entry system for each of the _BOA SEC-4 procedures and each of the procedures direct entry into the other SEC-4 procedures. The loss of offsite power to Unit 2 results in SAC 1B being the only available and operating air compressor. _BOA ELECT-4 provides the guidance for responding to LOOP and restoring NON-ESF Loads, as necessary. The mitigating actions (none of which are identified as immediate actions) are directed by the associated BOAs.

A. Incorrect; 1B SAC would have started approximately 8 minutes earlier at 104 psig, given the 0.5 psig/minute rate of depressurization. The rate of depressurization is indicative of normal air demands and not indicative of a leak which might require leak isolation actions (0BOA SEC-4, Step 4). Additionally, isolation of IA to Unit 2 would unnecessarily complicate the response to the loss of offsite power. Pressure would only drop to approximately 85 psig in the next 30 minutes which is well above the pressure where Letdown would isolate.

B. Correct; the in-service IA dryers automatically bypass at 90 psig, which would be within the next 20 minutes, given the 0.5 psig/minute rate of depressurization. Restoring power to Bus 244 in accordance with 2BOA ELECT-4, Step 9, would allow start of the 2B SAC using either 2BOA ELECT-4, Step 12.d OR 0BOA SEC-4, Step 2.b RNO, which would recover pressure. Refer also to explanation for choice A.

C. Incorrect; refer to explanations A and B D. Incorrect refer to explanations A and B. Control of charging flow to control PZR level (_BOA SEC-4, step 5) would only be needed if Letdown had isolated.

Question meets K/A - This is SRO ONLY because it requires examinee to have detail knowledge of 2BOA ELEC-4 to answer the question.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References:

N-BY-TQ-ILT-SA-53; Service Air and Instrument Air; Rev 11 0BOA SEC-4; Loss of Instrument Air Unit 0; Rev 110 1BOA SEC-4; Loss of Instrument Air Unit 1; Rev 108 2BOA SEC-4; Loss of Instrument Air Unit 2; Rev 106 2BOA ELECT-4; Loss of Offsite Power Unit 2; Rev 113

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 153 of 166 25 July 2019 94 ID: 2019 NRC-2 Q 94 Points: 1.00 Unit 1 was loading fuel when the audible source range indication inside the containment building became silent.

Both source range instruments were otherwise functioning normally.

Because of the failure of the containment building audible source range monitor, fuel loading may A.

continue only if an Inverse Count Rate Ratio (ICCR) is performed.

B.

continue only if the Source Range High Flux at Shutdown alarm is OPERABLE.

C.

continue only if the Boron Dilution Protection System (BDPS) is OPERABLE.

D.

NOT continue until the audible source range indication is restored to the containment building.

Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 154 of 166 25 July 2019 Explanation: LIMITATIONS AND ACTIONS OF 1BGP 100-6, 8.d.2).

This is not RO knowledge because it does not involve operability requirements (there are no operability requirements for the audible count rate) and it does not involve immediate actions.

It is the SROs responsibility to ensure the limitations and actions are complied with during the refueling process.

A. Correct. See explanation above.

B. Incorrect. The requirements for the High Flux at Shutdown alarm are not affected by the availability of the audible source range indication. Plausible because this would be another indication of rising count rate.

C. Incorrect. This system is not required in MODE 6. Plausible because this would provide an alarm for rising count rate.

D. Incorrect. Plausible because some source range instrument failures require fuel movement to stop.

Question meets K/A - This is SRO ONLY because it requires knowledge of the definition of operable components for the source range monitors as defined in the bases.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(7) Fuel handling facilities and procedures.

References:

1BGP 100-6, Refueling Outage

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 155 of 166 25 July 2019 95 ID: 2019 NRC-2 Q 95 Points: 1.00 Unit 1 is at EOL, in coastdown, at 90% power. In accordance with 1BGP 100-8, Generic Reactor Control Guidance, which of the following is correct concerning I control?

Control rods are used to maintain I ____(1)____ the target value and power should not be reduced by more than ____(2)_____ per shift, if possible.

A.

1. slightly below

2. 1%

B.

1. slightly below

2. 2%

C.

1. slightly above

2. 1%

D.

1. slightly above

2. 2%

Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 156 of 166 25 July 2019 Explanation:

Per 1BGP 100-8, Step F.7.c.1, During coastdown rods will be maintained at approximately 220 to 223 steps withdrawn in manual control. This partial insertion will give the operators control over I during Xenon oscillations. RODS SHOULD NOT BE USED TO MAINTAIN I ON THE TARGET VALUE. Instead, operators should allow I to slowly drift positive (up to approximately 2% of the positive limit) while rods are held constant. Control rods may then be inserted to maintain I at least 2% from its positive limit. If PDMS is inoperable, I may be allowed to trend up to the positive RAOC limit.

Step F.7.c.4 states Power should not be reduced by more than 1% per shift if possible. This will minimize the change to I. Also, remember that Xenon reactivity will initially increase following a load reduction.

A. Incorrect, I is maintained slightly above the target value.

B. Incorrect, I is maintained slightly above the target value. Power should not be reduced by more than 1% per shift. The applicant may confuse this with the allowed I drift up to 2%

of the positive limit.

C. Correct, see explanation above.

D. Incorrect, Power should not be reduced by more than 1% per shift. The applicant may confuse this with the allowed I drift up to 2% of the positive limit.

Question meets K/A - requires the examinee to know the methods of controlling I as specified in 1BGP 100-8. This is SRO ONLY because it requires assessment of the conditions and a reactivity strategy development as provided by procedural guidance.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(6) Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

References:

1BGP 100-8

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 157 of 166 25 July 2019 96 ID: 2019 NRC-2 Q 96 Points: 1.00 At 1200 on November 1, it was determined that a weekly Tech Spec surveillance, scheduled to be completed on October 22, had not been completed. A risk evaluation was completed and compensatory measures to manage the risk were implemented at 0000 on November 2.

Declaration of the LCO not being met, may be delayed until no later than A.

1200 on November 2.

B.

0000 on November 3.

C.

1200 on November 8.

D.

0000 on November 9.

Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 158 of 166 25 July 2019 Explanation:

A. Incorrect; plausible since this is the allowable delay time in SR 3.0.3 if a risk evaluation is NOT performed.

B. Incorrect; SR 3.0.3 allows a delay of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if a risk evaluation is not performed, but the delay is from the time of discovery, not from completion of the risk evaluation.

C. Correct; Per SR 3.0.3, the declaration may be delayed up to the limit of the specified Frequency (7 days) provided a risk evaluation is completed and the associated risk is managed.

D. Incorrect; SR 3.0.3 allows a delay up to the limit of the specified Frequency (7 days) provided a risk evaluation is completed and the associated risk is managed, but the delay is from the time of discovery, not from completion of the risk evaluation.

Question meets K/A - This is SRO ONLY because it requires the examinee to know that declaration of not meeting the LCO may be delayed as stated in SR 3.0.3.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References:

Technical Specification SR 3.0.3 N-BY-TQ-ILT-TS-3 Introduction to Technical Specifications

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 159 of 166 25 July 2019 97 ID: 2019 NRC-2 Q 97 Points: 1.00 Unit 1 is operating at 90% power.

An NSO reports a failure of Unit 1 annunciators to test.

1-4-D7, AN SYS GROUND, is ILLUMINATED 1-4-B7, SER TROUBLE, is CLEAR 1-4-A6, COMPUTER TROUBLE, is CLEAR Based upon the above conditions, the Unit Supervisor should direct operators to monitor A.

affected parameters continuously on the MCB ONLY.

B.

the SER closely for changes of alarm status.

C.

affected MCB indications at 30 minute intervals.

D.

local panels and equipment at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> intervals.

Answer:

B Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 160 of 166 25 July 2019 Explanation: both 1-4-D7 being illuminated and a failure of Unit 1 annunciators to test are entry criteria for 1BOA ELEC-7. The caution prior to 1BOA ELEC-7 step 3 states that [l]oss of BOTH the alarm windows and the SER points require close monitoring of affected parameter(s). Loss of the alarm window only requires close monitoring of the SER. Step 3 directs the Unit Supervisor to assign personnel to monitor affected plant parameters, to monitor the SER, and to report all changes of alarm status.

A) Incorrect. The caution prior to 1BOA ELEC-7 step 3 states that [l]oss of BOTH the alarm windows and the SER points require close monitoring of affected parameter(s). Loss of the alarm window only requires close monitoring of the SER. Step 3.b directs the monitoring MCB indications at 10-20 minute intervals.

B) Correct. See explanation above.

C) Incorrect. 1BOA ELEC-7, step 3.b, directs the monitoring MCB indications at 10-20 minute intervals. Per step 3.c, local panels and equipment are directed to be monitored at 30 minute intervals.

D) Incorrect. BAR 1-4-A6 contains a note stating that BOP CX-16, Interrogation of the Process Computer, has some short-term actions (within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) that need to be addressed. However, 1BOA ELEC-7, step 3.c, directs that local panels and equipment are to be monitored at 30 minute intervals.

Question meets K/A - SRO-only Justification: this question requires the assessment of plant conditions (e.g. abnormal) and then the selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References:

1BOA ELEC-7, Loss of Annunciators Unit 1, Rev. 4 BAR 1-4-B7, SER TROUBLE, Rev. 3 BAR 1-4-A6, COMPUTER TROUBLE, Rev. 12 BAR 1-4-D7, AN SYS GROUND, Rev. 52

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 161 of 166 25 July 2019 98 ID: 2019 NRC-2 Q 98 Points: 1.00 A steam generator tube rupture is occurring concurrently with a stuck open safety valve on the same steam generator.

(1) The primary contributor of the radioactivity being released to the environment will be __________.

(2) The expected implementation of the BEPs following entry into 1BEP-0 will be __________.

1BEP-2, Faulted Steam Generator Isolation Unit 1 1BEP-3, Steam Generator Tube Rupture Unit 1 1BCA-3.1, SGTR With Loss of Reactor Coolant - Subcooled Recovery Desired Unit 1 A.

1. Airborne Gas

2. 1BEP-2, 1BEP-3, then 1BCA-3.1 B.

1.

Airborne Particulate

2. 1BEP-2, 1BEP-3, then 1BCA-3.1 C.

1. Airborne Gas

2. 1BEP-3, 1BEP-2, then 1BCA-3.1 D.

1. Airborne Particulate

2. 1BEP-3, 1BEP-2, then 1BCA-3.1 Answer:

A Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 162 of 166 25 July 2019 Explanation:

The tube rupture will result in primary coolant to the SG and since the release is from the SG steam space little or no exposure is expected from particulate. 1BEP-0 checks for faulted generator status before checking for rupture status; once a faulted SG is detected in 1BEP-0, the transition to 1BEP-2 is made. The transition to 1BEP-3 is made from 1BEP-2 when it is determined that the SG tubes are not intact. Finally, the transition from 1BEP-3 to 1BCA-3.1 is made when it is determined that the ruptured SG is also faulted (SG pressure < 330 psig).

There is no transition from 1BEP-2 to 1BCA-3.1.

A. (1) Correct contributor; (2) correct procedure sequence B. (1) Incorrect contributor; (2) correct procedure sequence C. (1) Correct contributor; (2) incorrect procedure sequence D. (1) Incorrect contributor; (2) incorrect procedure sequence Question meets K/A - K/A is linked to 10 CFR 55.43 topic; specifically, 10 CFR 55.43(b)4.

Part (1) requires an assessment of how the leak location affects the release type, and part (2) requires an understanding the progression through the mitigating procedures.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

References:

1BEP 0; Reactor Trip or Safety Injection Unit 1; Rev 302 1BEP-2, Faulted Steam Generator Isolation Unit 1; Rev 300 1BEP-3, Steam Generator Tube Rupture Unit 1; Rev 301 1BCA-3.1, SGTR With Loss of Reactor Coolant - Subcooled Recovery Desired Unit 1; Rev 300

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 163 of 166 25 July 2019 99 ID: 2019 NRC-2 Q 99 Points: 1.00 Unit 1 has experienced an event resulting in the declaration of a GENERAL EMERGENCY. The Shift Manager has command and control of the event.

The Technical Support Center (TSC) is NOT functional.

The Emergency Offsite Facility (EOF) is staffed and ready to assume command and control.

In accordance with EP-AA-112, Emergency Response Organization/Emergency Response Facility Activation and Operation, which of the following responsibilities can be transferred from the Shift Manager to the EOF?

1. Event Classification

2. Protective Action Recommendations

3. Notification of Offsite Agencies A.

ONLY 1 AND 2 B.

ONLY 1 AND 3 C.

ONLY 2 AND 3 D.

1, 2, AND 3 Answer:

C Answer Explanation Explanation:

A. Incorrect: Classification would normally be transferred from the MCR to the TSC, but in the case where the TSC is not functional the responsibility remains with the MCR.

B. Incorrect: Classification remains with the MCR.

C. CORRECT: EP-AA-112, steps 3.1.3 and 3.1.4, PARs and Notifications are transferred to EOF, Classification remains with the MCR.

D. Incorrect: Classification remains with the MCR.

Question meets K/A - This meets the K/A - The examinee must know the emergency plan procedure's assignment of the emergency director's duties. This is SRO ONLY because of the required knowledge for Shift Manager actions.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References:

EP-AA-112; Emergency Response Organization/Emergency Response Facility Activation and Operation; Rev 20

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 164 of 166 25 July 2019 100 ID: 2019 NRC-2 Q 100 Points: 1.00 You are the Unit Supervisor, on Unit 1.

You are implementing 1BCA 1.3, Sump Blockage Control Room Guideline.

The STA reports that a BST has turned ORANGE (all other BSTs are GREEN or YELLOW).

In accordance with BAP 1310-10, HU-AA-104-101, Procedure Use and Adherence, Byron Addendum, you are expected to A.

exit 1BCA 1.3 and implement the associated BFR.

B.

continue in 1BCA 1.3 while concurrently implementing the BFR.

C.

complete 1BCA 1.3 before implementing the BFR for the ORANGE BST.

D.

continue in 1BCA 1.3; BFR implementation is ONLY required upon receipt of a RED BST.

Answer:

C Answer Explanation

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 165 of 166 25 July 2019 Explanation: Both Red and Orange BST conditions normally require departure from any other (non BFR) procedure. There are some exceptions, and these exceptions are delineated in the specific procedures, and SROs are expected to know which BCAs contain these exceptions.

_BCA 1.3 is one of these exceptions.

A. Incorrect; plausible since this normally the expected action.

B. Incorrect; plausible since this may be true for a YELLOW path BFR if time and resources permit concurrent implementation.

C. Correct; per the NOTE at the beginning of 1BCA 1.3 D. Incorrect; plausible since BFR implementation would be required if an ORANGE or RED BST existed when the actions of 1BCA 1.3 are completed.

Question meets K/A - This SRO ONLY because the examinee is required to know the exceptions to the normal BST usage rules.

10CFR55.43:

The written examination for an operator for a facility will include a representative sample from among the following 7 items, to the extent applicable to the facility.

(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

References:

BAP 1310-10, HU-AA-104-101, Procedure Use and Adherence, Byron Addendum; Rev 21 1BCA 1.3, Sump Blockage Control Room Guideline Unit 1; Rev 301

EXAMINATION ANSWER KEY 2019 NRC-2 BYR OPS ILT EXAM Page: 166 of 166 25 July 2019