Text

8-Columbia-2017-02 Final Written Exam
	ML17081A533
Person / Time
Site:	Columbia
Issue date:	03/09/2017
From:	Vincent Gaddy; Operations Branch IV
To:	; Energy Northwest
References
	Download: ML17081A533 (541)
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2017 NRC RO/SRO Written Exam Key 1 A B C D 26 A B C D 51 A B C D 76 A B C D 2 A B C D 27 A B C D 52 A B C D 77 A B C D 3 A B C D 28 A B C D 53 A B C D 78 A B C D 4 A B C D 29 A B C D 54 A B C D 79 A B C D 5 A B C D 30 A B C D 55 A B C D 80 A B C D 6 A B C D 31 A B C D 56 A B C D 81 A B C D 7 A B C D 32 A B C D 57 A B C D 82 A B C D 8 A B C D 33 A B C D 58 A B C D 83 A B C D 9 A B C D 34 A B C D 59 A B C D 84 A B C D 10 A B C D 35 A B C D 60 A B C D 85 A B C D 11 A B C D 36 A B C D 61 A B C D 86 A B C D 12 A B C D 37 A B C D 62 A B C D 87 A B C D 13 A B C D 38 A B C D 63 A B C D 88 A B C D 14 A B C D 39 A B C D 64 A B C D 89 A B C D 15 A B C D 40 A B C D 65 A B C D 90 A B C D 16 A B C D 41 A B C D 66 A B C D 91 A B C D 17 A B C D 42 A B C D 67 A B C D 92 A B C D 18 A B C D 43 A B C D 68 A B C D 93 A B C D 19 A B C D 44 A B C D 69 A B C D 94 A B C D 20 A B C D 45 A B C D 70 A B C D 95 A B C D 21 A B C D 46 A B C D 71 A B C D 96 A B C D 22 A B C D 47 A B C D 72 A B C D 97 A B C D 23 A B C D 48 A B C D 73 A B C D 98 A B C D 24 A B C D 49 A B C D 74 A B C D 99 A B C D 25 A B C D 50 A B C D 75 A B C D 100 A B C D

ES-401 Site-Specific RO Written Examination Form ES-401-7 Cover Sheet U. S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Name:

Date: March 9, 2017 Facility / Unit: Columbia Generating Station Region: I II III IV Reactor Type: W CE BW GE Start Time: 0900 Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the examination begins Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

_______________________________

Applicant's Signature Results Examination Value 74 75 Points Applicant's Score ________ Points Applicant's Grade ________ Percent

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: RO-1 Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 1 K/A 295001.AA1.01 Level of Difficulty: 3 Importance Rating 3.5 Partial or Complete Loss of Forced Core Flow Circulation: Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Recirculation system Question # 1 CGS is operating in Mode 1. Reactor power 100%.

Reactor Recirculation (RRC) Pump B, RRC-P-1B, tripped.

The crew has entered ABN-RRC-LOSS, Loss of Reactor Recirculation flow.

RRC A Drive Flow (RRC-FI-M/A/R676A) - 43,500 gpm

RRC Jet Pump Loop A Flow (MS-FI-611A) - 56 Mlb/hr

RRC Pump A frequency (RRC-HZM-R670A) - 57 hz

Control rods are inserted to the 100% rod line.

What actions should the crew take to stabilize the plant?

The crew should verify the Loop A Auto/Manual controller, RRC-M/A-676A, is in MANUAL, then A. raise RRC Jet Pump Loop A flow to GE 57.5 Mlb/hr.

B. Immediately attempt to restart RRC-P-1B.

C. Lower RRC Loop A Drive flow to LT 41,725 gpm.

D. raise RRC Loop A Drive flow by raising RRC Pump A frequency to 60 hz.

Answer: C Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The K/A discusses operating the recirculation system during a full or partial loss of flow. This question asks the student to understand how recirculation flow is adjusted after losing one recirculation pump.

SRO Only:

N/A.

Explanation:

ABN-RRC-LOSS step 4.3.4 directs reducing RRC Loop A drive flow to LT 41,725GPM.

A. Incorrect. This distractor is plausible since actions to restore boiling boundary to GE 4.0 is referenced in step 4.3.13 of ABN-RRC-LOSS. The distractor is also plausible because the maximum jet pump flow allowed in single look operations is 57.5Mlb/hr. The distractor is incorrect because RRC Loop Flow should be lowered, not raised.

B. Incorrect. Plausible because starting RRC-P-1B restore the plant to its previous normal lineup, but is incorrect because an RRC pump should not be restarted until it is understood why the pump tripped and appropriate procedures are followed to control reactor power.

C. Correct. ABN-RRC-LOSS step 4.3.4 directs reducing RRC Loop A drive flow to LT 41,725GPM.

D. Incorrect. This distractor is plausible because 60Hz is the maximum frequency that the ASD system and step 4.3.5 states that RRC flow should be maximized without exceeding 57.5Mlb/hr Jet Pump Flow or 41,725GPM Drive Flow. The distractor is incorrect, because RRC flow should be lowered, not raised, to keep drive flow less than 41,725GPM.

Technical Reference(s) Attached w/ Revision # See ABN-RRC-LOSS, Loss of Reactor Recirculation Flow Comments / Reference Proposed references to be provided during examination: None Learning Objective: RO-1484: Respond to a loss of RRC-P-1A or RRC-P-1B Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-RRC-LOSS, step 4.3.4 Revision: Major 013 Minor 001 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-RRC-LOSS, step 4.3.13 Revision: Major 013 Minor 001 Comments /

Reference:

ABN-RRC-LOSS, step 4.3.5 Revision: Major 013 Minor 001 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-RRC-LOSS, Notes prior to step 4.3.1 Revision: Major 013 Minor 001 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 1 K/A 295003.AA2.01 Level of Difficulty: 3 Importance Rating 3.4 Partial or Complete Loss of A.C. Power: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Cause of partial or complete loss of A.C. power Question # 2 CGS is operating in Mode 1, 100% core thermal power.

A plant transient occurs.

Current conditions:

Reactor power approximately 65%.

Reactor Recirculation (RRC) pumps running at 30 hz.

HPCS Service Water pump, HPCS-P-2, running.

These conditions are caused by a loss of power to A. SM-1.

B. SM-2.

C. SM-3.

D. SM-4.

Answer: B Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The candidate must analyze conditions in the stem and determine the bus loss (loss of power) that caused the conditions.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since a loss of SM-1 will cause a loss of Condensate Pump 1A and Condensate Booster Pump 2A, which will cause a trip of Reactor Feedwater Pump 1B on low suction pressure. (See ABN-ELEC-SM1/SM7, Attachment 7.1). The loss of a feed pump will cause RPV level to lower. When RPV level reaches less than 31.5 (Level 4) with one feed pump offline, Reactor Recirculation (RRC) pumps will run back to 30 hz to prevent a low RPV level scram. The distractor is incorrect because E-SM-4 is normally powered from SM-2, a loss of SM-1 will not cause a start of 4-DG3 and the HPCS Service Water pump will not be running.

B. Correct. A loss of power to SM-2 will cause a loss of Condensate Pump 1B and Condensate Booster Pump 2B, which will cause a trip of Reactor Feedwater Pump 1B on low suction pressure.

(See ABN-ELEC-SM1/SM7, Attachment 7.1). The loss of a feed pump will cause RPV level to lower. When RPV level reaches less than 31.5 (Level 4) with one feed pump offline, Reactor Recirculation (RRC) pumps will run back to 30 hz to prevent a low RPV level scram. Additionally, since SM-2 normally powers SM-4, a loss of SM-2 will cause the HPCS emergency diesel generator, 4-DG3, to start. The HPCS Service Water pump, HPCS-P-2, will start to support emergency diesel generator operations.

C. Incorrect. Plausible since a loss of SM-3 will cause a loss of Condensate Pump 1C and Condensate Booster Pump 2C, which will cause a trip of Reactor Feedwater Pump 1B on low suction pressure. (See ABN-ELEC-SM3/SM8, Attachment 7.1). The loss of a feed pump will cause RPV level to lower. When RPV level reaches less than 31.5 (Level 4) with one feed pump offline, Reactor Recirculation (RRC) pumps will run back to 30 hz to prevent a low RPV level scram.

However, since E-SM-4 is normally powered from SM-2, a loss of SM-3 will not cause a start of 4-DG3 and the HPCS Service Water pump will not be running.

D. Incorrect. Plausible since tripping CB-2/4 will cause the HPCS diesel generator, 4-3DG to start.

This will start the HPCS Service Water pump. However, a loss of power to SM-4 will not affect the condensate and feed system. Subsequently, reactor power will remain at 100% and RRC pumps will be operating at their normal frequency for full power.

Technical Reference(s)

ABN-ELEC-SM3/SM8, SM-3, SM-8, SM-85, SM-82, SL-81, SL-83 &

SL-31 Distribution System Failures Attached w/ Revision # See ABN-ELEC-SM1/SM7, SM-1, SM-7, SM-75, SM-72, SL-71, SL-73 & Comments / Reference SL-11 Distribution System Failures ABN-ELEC-SM2/SM4, SM-2, SM-4 and SL-21 Distribution System Failures Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: 6809- Given a loss of SM-2, identify those automatic actions that may have occurred.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Comments /

Reference:

ABN-ELEC-SM1/SM7, Attachment 7.1 Revision: Major 018, Minor 001 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-ELEC-SM2/SM4, sections 7.2 and 7.4 Revision: Major 006, Minor 003 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-ELEC-SM3/SM8, section 7.1.1 Revision: Major 018 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 1 K/A 295004.2.4.46 Level of Difficulty: 3 Importance Rating 4.2 Partial or Complete Loss of D.C. Power: Ability to verify that the alarms are consistent with the plant conditions.

Question # 3 CGS is operating in Mode 1.

A fault causes multiple annunciators to alarm, including:

DEH TROUBLE (H13-P820.B1 5)

CRD PUMPS ABNORMAL OPERATION (H13-P603.A7 6)

RCIC TURBINE TRIP (H13-P601.A4 5)

REACTOR FEEDWATER CONTROL SYSTEM TROUBLE (H13-P603.A8 7)

RCIC DIV 1 OOS (H13-P601.A4.6-8)

These alarms are consistent with a loss of voltage on A. DP-S1/1.

B. DP-S1/2.

C. DP-S2/1.

D. DP-S1/7.

Answer: A Page 1 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question requires an understanding of systems/components lost with a loss of DC power and the subsequent annunciators that will alarm.

SRO Only:

N/A Explanation:

A. Correct. A loss of voltage on DP-S1/1 (Div. 1 125 vdc) will cause a DEH TROUBLE annunciator due to a loss of one DEH power supply. CRD pump 1A indication and control are lost causing the CRD PUMPS ABNORMAL OPERATION annunciator alarm. The REACTOR FEEDWATER CONTROL SYSTEM TROUBLE alarm is received since RPV Narrow Range Level C fails downscale with a loss of Div. 1 125 vdc. Indication and control is lost for the majority of the RCIC controls at H13-P60, causing the RCIC TURBINE TRIP and RCIC DIV 1 OUT OF SERVICE annunciators to alarm.

B. Incorrect. Plausible since a loss of Division 2 125 vdc (DP-S1/2) will cause the DEH TROUBLE annunciator due to a loss of one DEH power supply. Additionally, CRD pump 1B indication and control are lost. This should cause the CRD PUMPS ABNORMAL OPERATION annunciator to alarm. Incorrect because a loss of Division 2 125 vdc causes a loss of annunciator power to H13-P601, H13-P602 and H13-P603 except ANNUNCIATOR 125 VDC LOSS H13-P603.A7-1.1, which alarms.

C. Incorrect. Plausible since a loss of 250VDC will cause a loss of indication and control for RCIC components and RFP Turbine Emergency Oil Pumps. A RCIC DIV 1 OUT OF SERVICE annunciator will occur. Incorrect because a DEH trouble annunciator would not be received.

D. Incorrect. Plausible because a loss of DP-S1/7 affects equipment that is affected by a loss of DP-S1/1. Both RFW pumps will trip on loss of control power. Annunciators received include 3-ELEMENT RFW CONTR. INOP (H13-P603.A8). CRD CHARGE WATER PRESS LOW (H13-P603.A7.3-8) will alarm. The RCIC DISCH PRESS LOW (H13.P601.A4.1-4) and RCIC PUMP DISCH FLOW FLOW (H13.601.A4.3-7) annunciators will alarm. However, the RCIC turbine will not trip and the DEH TROUBLE annunciator will not alarm.

Technical Reference(s)

ABN-ELEC-125VDC, Plant BOP, DIV 1,2 & 3 125 VDC Distribution System Failures Attached w/ Revision # See Comments / Reference ABN-ELEC-250VDC, Plant 250 VDC Distribution System Failures Proposed references to be provided during examination: None Learning Objective: 7652 - Predict the effects(s) a failure of 125VDC bus S1-1 will have on: (l) CR Annunciators Question Source: Bank #

Modified Bank # (Note changes or attach parent)

Page 2 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 3 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-ELEC-125VDC Revision: 14 Page 4 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 5 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-ELEC-250VDC Revision: Major 004 Minor 001 Page 6 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 3 Date: 2/2/2017 Tier 1 Group 1 K/A 295005.AK1.03 Level of Difficulty: 2 Importance Rating 3.5 Main Turbine Generator Trip: Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP : Pressure effects on reactor level Question # 4 CGS is operating in Mode 1. Reactor power is 100%.

A loss of lube oil causes a main turbine trip How does in-shroud RPV level initially respond to this event?

In-shroud RPV level will initially A. rise due to loss of recirculation pumps.

B. lower due to the collapse of voids in the core.

C. rise due to the rapid reduction in steam flow.

D. lower due to initiation of RPV level setpoint setdown.

Answer: B Page 1 of 8

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This question requires candidates to identify RPV level response due to the pressure transient caused by a main turbine trip.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since recirculation pumps will trip on a main turbine trip from greater than 30%

reactor power. Tripping RRC pumps alone will raise both indicated and in-shroud RPV level during the first few seconds after the turbine trip (see attached FSAR figure 15.3-2.3). However, in this case, the effect is overridden by the rapid collapse of voids in the core due to the immediate increase in RPV pressure when the main turbine is tripped. The overall effect is that in-core RPV level will initially lower (see attached FSAR figure 15-2.4).

B. Correct. A main turbine trip will cause a large RPV pressure spike. Although indicated RPV level will initially rise due to tripping RRC pumps, a rapid collapse of voids in the core will cause in-shroud RPV level to initially lower. See explanation for distractor A.

C. Incorrect. Plausible since steam flow will rapidly lower on a main turbine trip while feed flow will gradually lower as the level control system compensates for reduced steam flow. This effect would cause RPV level to rise initially while feed flow is greater than steam flow. However, the rapid collapse of voids in the core due to RPV pressure rise will override this effect and RPV level will initially lower. See explanation for distractor A.

D. Incorrect. Plausible since the RPV level setpoint setback feature is in service on a RPS initiated scram. However, this feature determines where RPV level will be controlled steady state. The initial RPV level reduction occurs as core voids collapse as RPV pressure rapidly rises. See explanation for distractor A.

Technical Reference(s) Attached w/ Revision # See CGS Final Safety Analysis Report (FSAR) Comments / Reference T.S. Bases, 3.3.4.1 Proposed references to be provided during examination: None Learning Objective: 11647 - Explain the reasons for the following responses as they apply to Main Turbine trip: d. Generator Trip Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Page 2 of 8

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 10 CFR Part 55 Content: 55.41 41.5 55.43 Page 3 of 8

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FSAR Revision: Major 63 Minor 008 Page 4 of 8

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 5 of 8

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

T.S. Bases, 3.3.4.1 Revision: 92 Page 6 of 8

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FSAR Revision: Major 63 Minor 008 Page 7 of 8

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 8 of 8

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 1 K/A 295006.AK2.04 Level of Difficulty: 2 Importance Rating 3.6 SCRAM: Knowledge of the interrelations between SCRAM and the following: Turbine trip logic Question # 5 The reactor is operating at 35%.

You see the indications pictured below for DEH-PI-21(Hydraulic Oil Header Pressure), and RFW-LI-606A (RPV Level).

Which of the following should have occurred?

A. Turbine Trip, Reactor SCRAM B. Turbine Trip, Reactor does not SCRAM C. No Turbine Trip, Reactor SCRAM D. No Turbine Trip, Reactor does not SCRAM Answer: A Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

A turbine trip above 30% will result in a reactor SCRAM. A turbine trip below 30% will not result in a reactor SCRAM. The question determines whether the student knows the power value at which a SCRAM will occur due to a turbine trip and whether they know a turbine trip can cause a SCRAM.

SRO Only:

N/A Explanation:

The candidate must analyze the indication for DEH pressure and recognize it is below the turbine trip value of 1600 psig. They must then determine whether or not a SCRAM occurs based on reactor power exceeding 30%. The correct answer is A.

A. Correct. DEH pressure is below 1600psig resulting in a turbine trip. Because reactor power is above 30%, a reactor scram also occurs.

B. Incorrect. Plausible since a turbine trip occurs, but the distractor is incorrect since a SCRAM will also occur.

C. Incorrect. Plausible because RPV level is lower than normal and approaching the SCRAM setpoint of 13 inches. However, a turbine trip will occur.

D. Incorrect. Plausible because RPV level is low, but has not reached the SCRAM setpoint of 13.

However, a turbine trip will occur and therefore, the reactor will scram.

Technical Reference(s) Attached w/ Revision # See SD000129 r12 Main Turbine System Description Comments / Reference 4.603.A8 5-4 Annunciator Response Procedure Proposed references to be provided during examination: None Learning Objective: 5566 - List all parameters and setpoints that will cause a turbine trip Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000129 Revision: 12 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

4.603.A8 Revision: 36 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 1 K/A 295016.AA1.06 Level of Difficulty: 2 Importance Rating 4.0 Control Room Abandonment: Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT: Reactor water level Question # 6 Which of the following describes the RPV level indications available at the RSD panel and ARSD panel?

At the RSD panel, (1) RPV level indication(s) is/are available. At the ARSD panel, (2)

RPV level indication(s) is/are available.

A. (1) Wide Range only (2) Wide Range only B. (1) Wide Range only (2) Wide Range and Fuel Zone C. (1) Wide Range and Fuel Zone (2) Wide Range only D. (1) Wide Range and Fuel Zone (2) Wide Range and Fuel Zone Answer: A Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question evaluates the candidates understanding of what indications are available to monitor reactor water level during control room abandonment which would impact their ability to correctly monitor RPV level during an accident when control room abandonment is required.

SRO Only:

N/A.

Explanation:

Only wide range indication is available at the RSD and ARSD panels.

A. Correct. Only wide range indication is available at the RSD and ARSD panels.

B. Incorrect. Plausible because fuel zone indication is an actual RPV level indication available in the control room and is designed for accident conditions and mitigating severe transients. Plausible because fuel zone indication would be desirable if a LOCA occurred while control room abandonment is necessary. Incorrect because fuel zone indication is not available outside the control room.

C. Incorrect. Plausible because fuel zone indication is an actual RPV level indication available in the control room and is designed for accident conditions and mitigating severe transients. Plausible because fuel zone indication would be desirable if a LOCA occurred while control room abandonment is necessary. Incorrect because fuel zone indication is not available outside the control room.

D. Incorrect. Plausible because fuel zone indication is an actual RPV level indication available in the control room and is designed for accident conditions and mitigating severe transients. Plausible because fuel zone indication would be desirable if a LOCA occurred while control room abandonment is necessary. Incorrect because fuel zone indication is not available outside the control room.

Technical Reference(s) Attached w/ Revision # See SD000210, RSD/ARSD System Text (pages 19,20) Comments / Reference Proposed references to be provided during examination: None Learning Objective: RO-1057 Perform actions for a control room evacuation 5582 List calibration conditions and nominal ranges for each of the five ranges of level instruments Question Source: Bank # LO01281 Modified Bank # (Note changes or attach parent)

New Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.7 55.43 Comments /

Reference:

SD000126 r13 mr1 Revision:

Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 1 K/A 295018.AK3.01 Level of Difficulty: 2 Importance Rating 2.9 Partial or Complete Loss of Component Cooling Water: Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : Isolation of non-essential heat loads Question # 7 CGS is operating in Mode 1.

How does the Reactor Component Cooling (RCC) system respond to a partial loss of flow?

Degraded RCC flow, as sensed by (1) , causes an automatic isolation of RCC supply to the (2) in order to maximize available cooling to the Drywell.

A. (1) RCC pump breaker position (2) Radwaste and Reactor buildings B. (1) RCC pump breaker position (2) Radwaste building only C. (1) RCC pump discharge pressure (2) Radwaste and Reactor buildings D. (1) RCC pump discharge pressure (2) Radwaste building only Answer: A Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question determines if the candidate understands the reason RCC-V-6 closes on low flow. It determines if they know which component is protected when the isolation occurs.

SRO Only:

N/A Explanation:

On low flow in the RCC system for 10 seconds, RCC-V-6 will close isolating the reactor building and radwaste buildings from the system. The only components with flow remaining will be RRC pumps and drywell air cooling units.

A. Correct. When less than two RCC pump breakers are closed for greater than 10 seconds, RCC-V-6, Radwaste/Reactor Building Supply valve, will close to maximize RCC flow to components inside primary containment.

B. Incorrect. Plausible since low RCC flow is sensed by RCC pump breaker position. However, RCC flow to the Radwaste building AND the Reactor building is isolated.

C. Incorrect. Plausible since RCC flow to the Radwaste building AND the Reactor building is isolated on a low flow condition. However, low RCC flow is sensed by RCC pump breaker position, not RCC pump discharge pressure.

D. Incorrect. Plausible since low RCC pump discharge pressure will cause a RCC pump discharge pressure low alarm. Additionally, RCC flow to the Radwaste building will be isolated on a partial loss of RCC flow. However, low RCC flow is sensed by RCC pump breaker position and RCC flow to the Radwaste building AND the Reactor building is isolated.

Technical Reference(s) Attached w/ Revision # See SD000196, RCC System Text Comments / Reference Proposed references to be provided during examination: None Learning Objective: 5705 - State the purpose of the following components: (b) RCC-V-6 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.7 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 55.43 Comments /

Reference:

SD000196 Revision: r14 mr1 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 1 K/A 295019.AA2.01 Level of Difficulty: 2 Importance Rating 3.5 Partial or Complete Loss of Instrument Air: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR : Instrument air system pressure Question # 8 Given the following:

A leak has developed on the control air header.

Control air header pressure is 92 psig and down slow.

As pressure continues to drop, select the next automatic action that will occur in the control and service air system.

A. Standby control air compressor STARTS.

B. Control/service air crosstie valve, SA-PCV-2, CLOSES.

C. Service air compressor, SA-C-1, STARTS.

D. Desiccant dryer bypass valve, CAS-PCV-1, OPENS.

Answer: B Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question determines if the candidate understands automatic actions in the CAS system related to system pressure.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since the standby Control Air Compressor will automatically start as control air header pressure lowers. However, the standby Control Air Compressor starts at 100 psig.

Therefore, the standby compressor will already be running when header pressure 92 psig and lowering.

B. Correct. SA-PCV-2 closes when control air system pressure drops to LT 80 psig to isolate the service air header from the control air header.

C. Incorrect. Plausible since the service air compressor is used to supply air to the control air system when required. However, the service air compressor is continuously running. It is loaded and unloaded as necessary to maintain header pressure. There is no auto start feature for this compressor.

D. Incorrect. Plausible since CAS-PCV-1 opens when control air header pressure drops to LT 75 psig to prevent a loss of control air due to desiccant dryer or filter blockage. However, this is not the next automatic action to occur since SA-PCV-2 closes at LT 80 psig.

Technical Reference(s) Attached w/ Revision # See ABN-CAS,Control Air System Failure Comments / Reference Proposed references to be provided during examination: None Learning Objective: 5878 - List the expected automatic Control Air system response due to a leak in the Control Air system.

Question Source: Bank # LO00435 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 1996 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.4 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-CAS Revision: 9 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 1 K/A 295021.2.4.11 Level of Difficulty: 3 Importance Rating 4.0 Loss of Shutdown Cooling: Knowledge of abnormal condition procedures.

Question # 9 CGS is in Mode 4.

RHR Loop A is in Shutdown Cooling (SDC) mode.

RHR Loop B is in Suppression Pool Cooling (SPC) mode.

OSP-ELEC-M701, DG1 Monthly Operability Test is in progress.

The following indications are observed:

Annunciator H13-P601.A4.2-3, RHR A PUMP TRIP, is in alarm.

RHR-V-8, Shutdown Cooling Suction Isolation Valve, is CLOSED.

Field operators report that RHR-V-8 cannot be manually opened.

RPV Pressure indicates 2 psig

Reactor coolant temperature is 195°F, up slow.

Both RRC pumps are secured.

What actions should be taken to restore SDC?

A. Restore RHR Loop A to SDC mode with suction from the spent fuel pool per ABN-RHR-SDC-LOSS, Loss of Shutdown Cooling.

B. Shift RHR-P-2B to SDC mode with suction from the spent fuel pool per ABN-RHR-SDC-ALT, Residual Heat Removal Alternate Shutdown Cooling.

C. Initiate shutdown cooling through main steam line drain valves per ABN-ADHR, Alternate Decay Heat Removal.

D. Place RHR Loop B in the Fuel Pool Cooling Assist mode per ABN-FPC-ASSIST-ALT, Alternate Fuel Pool Cooling Assist.

Answer: B Question #9 Deleted per Post-Exam Comments Page 1 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question determines whether a candidate can determine that shutdown cooling has been lost based on plant indications and select the correct abnormal operating procedure based on those conditions.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since RHR Loop A would be returned to SDC mode using ABN-RHR-SDC-LOSS. However, with RHR-V-8 stuck closed, an alternate method of SDC must be employed.

RHR Loop A cannot take a suction form the spent fuel pool.

B. Correct. With RHR-V-8 stuck closed, the normal SDC is lost. Additionally, reactor coolant temperature is in jeopardy of rising above 200°F. Therefore, an alternate method of SDC must be employed. ABN-RHR-SDC-LOSS, step 4.1.9, refers operators to ABN-RHR-SDC-ALT. Step 4.3 directs operators to lineup RHR Loop B for SDC with suction from Fuel Pool Cooling (Attachment 7.2).

C. Incorrect. Plausible since SDC through the main steam line drains is an alternate SDC method per step 4.1.9 of ABN-RHR-SDC-LOSS. However, the initial conditions of ABN-ADHR require the reactor to be in mode 3.

D. Incorrect. Plausible since SDC with RHR in the Fuel Pool Cooling (FPC) Assist mode is a valid method per step 4.1.9 of ABN-RHR-SDC-LOSS. However, SOP-FPC-ASSIST-ALT requires that RHR is first lined up for FPC Assist with Suction from the RPV (RHR-V 8 & 9) and Discharging to the SFP Diffusers (section 5.1), which requires RHR-V-8 to be opened.

Technical Reference(s)

ABN-RHR-SDC-LOSS, : Loss of Shutdown Cooling ABN-RHR-SDC-ALT, Residual Heat Removal Alternate Shutdown Attached w/ Revision # See Cooling Comments / Reference ABN-ADHR, Alternate Decay Heat Removal SOP-FPC-ASSIST-ALT, Alternate Fuel Pool Cooling Assist Proposed references to be provided during examination: None Learning Objective: RO-1302 - Respond to a loss of shutdown cooling Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Page 2 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Comments /

Reference:

ABN-RHR-SDC-LOSS Revision: 6 Page 3 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-RHR-SDC-ALT Revision: 13 Page 4 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 5 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-ADHR Revision: 0 Page 6 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-FPC-ASSIST-ALT Revision: 010 Page 7 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/1/2017 Tier 1 Group 1 K/A 295023.AK1.03 Level of Difficulty: 2 Importance Rating 3.7 Refueling Accidents: Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS : Inadvertent criticality Question # 10 CGS is in Mode 5. Fuel Shuffling is in progress.

Initial source range counts:

SRM-A: 75 cps, steady

SRM-B: 105 cps, steady

SRM-C: 95 cps, steady

SRM-D: 93 cps, steady As a fuel bundle is moved into the core, the control room operator notes the following source range response:

SRM-A: 153 cps, steady

SRM-B: 188 cps, up slow

SRM-C: 275 cps, up slow

SRM-D: 156 cps, steady The bundle is not adjacent to a SRM.

What actions should be taken per PPM 6.3.2, Fuel Shuffling and/or Offloading and Reloading?

Immediately stop fuel bundle insertion. Then A. withdraw the fuel bundle from the core and evaluate the cause of the source range level increase.

B. bundle insertion may be continued slowly with close observation of subcritical multiplication behavior.

C. verify all control rods are inserted and obtain permission from the Station Nuclear Engineer prior to moving the fuel bundle.

D. verify SRM signal to noise ratio and obtain permission from the Reactor Engineering Manager prior to moving the fuel bundle.

Answer: B Page 1 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question determines if candidates are able to identify that inadvertent criticality has occurred and what actions must be taken based on this.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since the fuel bundle must be removed from the core if a loss of communications with the control room occurs (see note prior to PPM 6.3.2, step 6.1.1. However, when the average source range count rate doubles, fuel moves may continue slowly if SRM response appears normal.

B. Correct. Initial average SRM count rate: 92 cps; final SRM count rate: 193 cps. Per step 6.1.1.c of PPM 6.3.2, If an unexpected doubling in average SRM count rate occurs, or two doublings of any single SRM count rate occurs, then stop the insertion of the bundle. Bundle insertion may then be continued slowly with close observation of subcritical multiplication behavior.

C. Incorrect. Plausible because TS LCO 3.9.3 requires all control rods to be fully inserted when loading fuel into the core and a withdrawn control rod could cause SRM counts to increase..

However, resuming fuel shuffling requires reactivity managers concurrence. Additionally, the move may continue as stated in answer B explanation above.

D. Incorrect. Plausible since satisfactory SRM signal to noise ratio is required by TS LCO 3.3.1.2.

This is verified every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during core alterations. However, resuming fuel shuffling requires reactivity managers concurrence. Additionally, the move may continue as stated in answer B explanation above.

Technical Reference(s)

PPM 6.3.2, Fuel Shuffling and/or Offloading and Reloading Attached w/ Revision # See TS LCO 3.3.1.2 Comments / Reference TS LCO 3.9.3 Proposed references to be provided during examination: None Learning Objective: 7700 - State the indications used to identify criticality during fuel loading Question Source: Bank #

Modified Bank # LX00330 (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Page 2 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 10 CFR Part 55 Content: 55.41 41.10 55.43 Comments /

Reference:

PPM 6.3.2 Section 6.1 Revision: 24 Page 3 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

TS LCO 3.9.3 Revision: Amendment 237 Page 4 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

TS LCO 3.3.1.2 Revision: Amendment 237 Page 5 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Original Question LX00330 Revision:

Identify the statement below that describes the indication used to avoid inadvertant criticality during fuel load.

A. An unexpected doubling in average SRM count rate, or two doublings of any SRM count rate occurs.

B. Reactor period is less negative than the value seen during conduct of the shutdown margin test.

C. SRM count rate on the detector closest to the new bundle is greater than on the other SRM's.

D. Shutdown margin test done after each bundle is inserted demonstrates Keff getting closer to 1.0.

Answer: A Page 6 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 3 Date: 2/8/2017 Tier 1 Group 1 K/A 295024.EK2.08 Level of Difficulty: 3 Importance Rating 4.0 High Drywell Pressure: Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following: ADS Question # 11 CGS is operating in Mode 1.

A large LOCA into primary containment occurred.

Wetwell pressure cannot be restored and maintained less than PSP.

Emergency depressurization is required.

Why is it preferable to open ADS SRVs with the current plant conditions?

When compared to other SRVs, one reason ADS SRVs are preferred is that A. pressure will be lowered more rapidly due to their larger capacity.

B. the instantaneous trip channel results in faster initiation.

C. they can be opened with a loss of containment instrument air.

D. they are qualified for long-term cooling post-LOCA.

Answer: D Page 1 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The K/A asks for understanding of the relationship between high drywell pressure and SRVs. The correct answer is based on the fact that ADS SRVs are designed to operate in more adverse conditions (higher drywell pressure and temperature) than other SRVs.

SRO Only:

N/A Explanation:

Per PPM 5.0.10, ADS SRVs are preferred because of their qualifications.

A. Incorrect. ADS SRVs due not have a higher capacity than other SRVs. Plausible because ADS SRVs are designed differently (logic, qualifications, etc.) than other SRVs.

B. Incorrect. The instantaneous trip channel does NOT result in faster initiation. Plausible because an instantaneous trip channel exists in the system.

C. Incorrect. Plausible since all SRVs will operate upon a loss of containment instrument air.

Therefore, this is not a difference between ADS SRVs and non-ADS SRVs.

D. Correct. All SRVs are environmentally qualified the same. The support systems for the ADS SRVs are designed to provide sufficient nitrogen to allow the ADS SRVs to remain operable through an entire post-LOCA cooldown, while non-ADS SRVs are only qualified to be operable up to 2 days.

There are two N 2 banks that are credited in accident analysis. One bank supplies 3 ADS SRVs while the other supplies four ADS SRVs. The banks will automatically shift to supply the ADS SRVs if the N 2 header pressure lowers to a specific value. The control power required to shift the banks is divisionally separated. Therefore, if one division of power is lost, at least 3 ADS SRVs will remain available.

Technical Reference(s) Attached w/ Revision # See PPM 5.0.10, Flowchart Training Manual Comments / Reference CGS System Description, Vol. 7, Chap. 5, Automatic Depressurization.

Proposed references to be provided during examination: None Learning Objective: 11874 - Describe the physical connection and/or cause-and-effect relationship between the Automatic Depressurization System and the following: Drywell pressure Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Page 2 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.7 55.43 Comments /

Reference:

PPM 5.0.10 Revision: 21 mr 1 Page 3 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

CGS System Description, Automatic Revision: 12 Depressurization, section II, Design Bases Page 4 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 5 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 6 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 7 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/2/2017 Tier Group K/A 295025.EK3.02 Level of Difficulty: 2 Importance Rating 3.9 High Reactor Pressure: Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE : Recirculation pump trip Question # 12 CGS is operating in Mode 1.

The reactor is in an ATWS condition following an event.

Reactor pressure has risen to 1153 psig.

Both Reactor Recirculation (RRC) Pumps have tripped.

Which of the following describes the reason for the trip of the RRC pumps?

Tripping the RRC pumps A. raises core inlet subcooling which reduces reactor power.

B. increases voiding in the core which adds negative reactivity.

C. moves the boiling boundary up the fuel channel which adds negative reactivity.

D. lowers in-shroud RPV level which assists in lowering RPV pressure.

Answer: B Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question direct asks the reason that a high pressure trip of RRC pumps is needed.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since starting RRC pumps will raise inlet subcooling. However, securing RRC pumps will lower inlet subcooling.

B. Correct. Per the bases for TS LCO 3.3.4.2, RRC pumps are automatically tripped when RPV pressure is high to add negative reactivity to offset the positive reactivity added to the core when voids collapse rising RPV pressure. This is done to prevent damage to the RPV and associated systems due to RPV overpressure.

When RRC pumps are tripped, the flow of relatively cold reactor coolant into the in-shroud region is limited to natural circulation (NC) flow. Thermal driving head (core T) must be established for NC flow. Initially (first few seconds after the transient), core T is relatively small and NC flow is negligible. At the same time, heat is still being produced in the core, especially in an ATWS condition. Core enthalpy will rise, causing additional void formation in the core, which aids in the termination of the ATWS event and, along with the safety/relief valves (SRVs), limits the peak RPV pressure to less than the ASME Section III Code Service Level C limits (1500 psig).

C. Incorrect. Tripping RRC pumps does not raise the boiling boundary. Plausible because starting RRC pumps or raising RRC flow does increase the boiling boundary.

D. Incorrect. Plausible since lowering RPV level (reducing mass in the RPV) will tend to lower RPV pressure. However, securing RRC pumps will raise RPV level due to the increase in core voiding (See attached FSAR figure 15.8-1.4).

Technical Reference(s) Attached w/ Revision # See TS B3.3.4.2, ATWS without SCRAM RRC Pump Trip Comments / Reference Final Accident Safety Report (FSAR)

Proposed references to be provided during examination: None Learning Objective: 5022 - Describe the physical and or caused and effect relationship between the RRC system and the following: (a)Core Flow (b) Reactor Power (c) NBI Question Source: Bank # LO001750 Modified Bank # (Note changes or attach parent)

New Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Question History: Last NRC Exam 2009, question #12 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

B 3.3.4.2 Revision: 92 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

FSAR Revision: Ammendment 64 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/6/2017 Tier 1 Group 1 K/A 295026.EA1.01 Level of Difficulty: 2 Importance Rating 4.1 Suppression Pool High Water Temperature: Ability to operate and/or monitor the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool cooling Question # 13 Columbia is operating in Mode 1.

A reactor scram and emergency depressurization have been initiated due to a LOCA.

Plant conditions 15 minutes after the scram:

The crew is taking actions per PPM 5.1.1, RPV Control and PPM 5.2.1, Primary Containment Control

RPV water level is -140 inches, up slow.

Wetwell temperature is 95°F, up slow.

RHR Loops B and C are injecting to recover RPV level.

Control room operators are aligning RHR A for Suppression Pool cooling due to high Wetwell temperature.

RHR-V-42A (LPCI injection valve) is OPEN

RHR-V-24A (Suppression Pool cooling valve) is CLOSED

RHR-V-48A (RHR Heat Exchanger bypass valve) is OPEN

RHR-P-2A is RUNNING

SW-P-1A is RUNNING What actions are required to complete the suppression pool cooling lineup?

A. Throttle open RHR-V-24A to establish desired flowrate, throttle RHR-V-48A as needed to maintain desired temperature.

B. Close RHR-V-42A, throttle open RHR-V-24A to establish desired flowrate, close RHR-V-48A.

C. Throttle open RHR-V-24A to establish desired flowrate, Open RHR-V-42A , open RHR-V-4A.

D. Close RHR-V-42A, throttle open RHR-V-24A to establish desired flowrate, throttle open RHR-V-48A as needed to maintain desired temperature.

Answer: B Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question requires the candidate to MONITOR the system by analyzing the current status and then OPERATE the system to change the status in response to High Suppression Pool Temperature.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since RHR-V-24A is throttled open to establish desired flowrate. However, RHR-V-42A must be opened prior to throttling RHR-V-24A. Additionally, when establishing suppression pool cooling while in the EOPs, RHR-V-48A is fully closed to maximize cooling.

B. Correct. RHR-V-42A must be closed prior to throttling open RHR-V-24A. RHR-V-24A is throttled open to establish flow between 4500 and 7000 gpm. RHR-V-48A is closed to maximize cooling.

C. Incorrect. Plausible since RHR-V-24A is throttled open. However, RHR-V-42A must be closed to allow RHR-V-24A to open. Additionally, RHR-V-4A is already open in the LPCI lineup.

D. Incorrect. Plausible since RHR-V-42A is closed and RHR-V-24A is throttled to establish desired flow. However, when establishing suppression pool cooling while in the EOPs, RHR-V-48A is fully closed to maximize cooling.

Technical Reference(s) Attached w/ Revision # See SOP-RHR-SPC-QC, Placing RHR in Suppression Pool Cooling QC Comments / Reference Proposed references to be provided during examination: None Learning Objective: RO-0218 - Initiate RHR system in suppression pool cooling.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-RHR-SPC-QC Revision: 5 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 0 Date: 12/16/2016 Tier 1 Group 1 K/A 295028.2.4.6 Level of Difficulty: 3 Importance Rating 3.7 High Drywell Temperature: Knowledge of EOP mitigation strategies.

Question # 14 Given the following parameters:

HPCS and CRD are injecting into RPV.

RPV Pressure is 550 psig and down slow.

RPV Level is -158 inches and steady.

Drywell Temperature is 285 degrees F and up slow.

Drywell Pressure is 8 psig and up slow.

Drywell conditions are within the Drywell Spray Initiation Limit (DSIL).

Wetwell Pressure is 1.5 psig and up slow.

Wetwell level is 29 feet and steady.

Which of the following should be prioritized as the next action to be taken?

A. Initiate Drywell sprays B. Raise WW level with RHR C. Initiate Wetwell Sprays D. Open 7 ADS SRVs Answer: A Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question determines if candidate knows major EOP mitigation strategies based on high drywell pressure.

SRO Only:

N/A Explanation:

Per PPM 5.2.1, when drywell temperature cannot be maintained below 135 degrees, drywell sprays should be initiated provided WW level is below 51ft and drywell temp is below DSIL.

A. Correct.

B. Incorrect. Plausible because WW level needs to be raised. However, WW level is raised using HPCS during EOPs per PPM 5.5.23 C. Incorrect. WW pressure is too low for wetwell spray. Plausible because WW spray is directed when drywell pressure is above 1.68psig (PPM 5.2.1 step P-2 to P-5)

D. Incorrect. Emergency depressurization limits have not been met. Plausible based on low RPV level and accident conditions.

Technical Reference(s) Attached w/ Revision # See PPM 5.2.1, Primary Containment Control Comments / Reference Proposed references to be provided during examination: None Learning Objective: RO-0635 - Maintain Drywell temperature below 330 degrees F Question Source: Bank # LO01890 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2009, question #14 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.2.1 Revision:

Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 1 K/A 295030.EK1.01 Level of Difficulty: 3 Importance Rating 3.8 Low Suppression Pool Water Level: Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Steam condensation Question # 15 Refer to the following conditions:

All control rods are fully inserted

Drywell Pressure is 12 psig and rising

RPV Pressure is 550 psig and stable

RPV Water Level is -50 inches

RHR A PUMP ROOM HIGH WATER LEVEL annunciator is locked in

Wetwell Level is 18 9 and lowering

Wetwell Pressure is 8 psig and rising

HPCS PUMP MOTOR OVERCURRENT annunciator is locked in.

Field operators report water coming from the RHR A pump room door.

What action should be taken:

A. Depressurize the RPV. Maintain less than a 100°F/HR cooldown rate.

B. Stop and Prevent ECCS pumps per Table 18, Vortex and NPSH Limits.

C. Emergency Depressurize using 7 ADS SRVs.

D. Spray the Wetwell using RHR-P-2A.

Answer: C Page 1 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The candidate must recognize from the plant conditions in the stem that WW level is below 192, the point at which inadequate steam condensation becomes a concern. They must then select the correct action to take (operational implications) based on that condition.

SRO Only:

N/A Explanation:

A. Incorrect. Emergency depressurization is required per PPM 5.2.1 step L-6. Plausible because with the reactor shutdown and no boron injection required, the RPV is cooled down per PPM 5.1.2 step P-6 and P-8.

B. Incorrect. Plausible since wetwell level is approaching the vortex limit for the RCIC pump (17.5 feet). However, the step that requires stopping ECCS pumps due to vortex concerns (step L-8) is performed after an emergency depressurization (step L-6).

C. Correct. PPM 5.2.1 step L-6 directs emergency depressurization when wetwell level cannot be recovered above 19 feet, 2 inches..

D. Incorrect. Plausible because wetwell spray could be directed per step P-6 of PPM 5.2.1. However, per ABN-FLOODING, step 7.1.1, RHR-P-2A should be stopped if RHR pump room A flooding is occurring.

Technical Reference(s) Attached w/ Revision # See PPM 5.2.1, Primary Containment Control Comments / Reference ABN-FLOODING, Flooding Proposed references to be provided during examination: None Learning Objective: 13567 - Given a copy of EOPs and an event, describe the basis for each variable and figure used to execute EOP strategies without error.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.2.1, WW Level Leg Revision: 23 Page 3 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.2.1, Vortex and NPSH Limits, Table Revision: 23 18 Page 4 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-FLOODING, step 4.1.1 Revision: 18, Minor 001 Page 5 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-FLOODING, step 4.6 Revision: 18, Minor 001 Page 6 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-FLOODING, step 7.1.1 Revision: 18, Minor 001 Page 7 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 1 K/A 295031.EA1.05 Level of Difficulty: 2 Importance Rating 4.3 Reactor Low Water Level: Ability to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL :

Reactor core isolation system Question # 16 CGS is operating in Mode 1.

A loss of all feedwater causes a reactor scram.

RPV level lowered to -65 inches prior to restoring feed.

Current plant conditions:

Actions of PPM 3.3.1, Reactor Scram are complete through scram reset.

The crew is controlling RPV level +13 inches to +54 inches.

Drywell pressure is 0.8 psig, up slow.

The CRS has directed resetting NS4 logic per ABN-FAZ.

What actions should the crew take to reset containment isolation logic?

Prior to depressing the Isolation Logic A&B and C&D pushbuttons on H13-P601 A. vent the Drywell per SOP-CN-CONT-VENT.

B. return both SGT trains to standby.

C. restore power to busses load shed from SM-7 and SM-8.

D. place the RCC pumps in pull-to-lock.

Answer: D Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question asks how to operate the reactor core isolation system (called NS4 system at Columbia) following an actuation on low RPV level (A signal).

SRO Only:

N/A Explanation:

Per ABN-FAZ step 4.2.19, RCC pumps should be place in PTL prior to resetting the signal in step 4.2.20.

A. Incorrect. Venting the drywell is not required for an A signal. Plausible because venting the drywell is required for an F signal (high drywell pressure).

B. Incorrect. Returning both SGT trains to standby is not required prior to reset. Plausible because SGT trains start on an A signal.

C. Incorrect. Restoring load shed busses is not required PRIOR to reset. Plausible because any FAZ signal causes non-critical loads to be shed.

D. Correct answer.

Technical Reference(s) Attached w/ Revision # See ABN-FAZ, FAZ Comments / Reference Proposed references to be provided during examination: None Learning Objective: 15764 - With the procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-FAZ Question Source: Bank #

Modified Bank # LO03416 (Note changes or attach parent)

New Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.9 55.43 Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-FAZ Revision: 17 mr 5 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Original Question LO03416 Revision: N/A A valid "A" signal has been received due to a loss of all feedwater. RPV level has been returned to a band of +13" to +54".

What is the required order for the following steps from ABN-FAZ?

(1) Reset the NS4 logic (2) Place the RCC pumps in pull-to-lock (3) Restore power to busses load shed from SM-7 and SM-8 (4) Start an RCC pump (5) Open the RCC primary containment valves A. (1)-(2)-(3)-(4)-(5)

B. (1)-(3)-(2)-(4)-(5)

C. (1)-(5)-(2)-(3)-(4)

D. (2)-(1)-(3)-(5)-(4)

Answer: D Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 4 Date: 2/7/2017 Tier 1 Group 1 K/A 295037.EK3.05 Level of Difficulty: 3 Importance Rating 3.2 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown: Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKOWN: Cold shutdown boron weight.

Question # 17 CGS is operating in Mode 1.

A transient causes an ATWS.

Reactor power is 24%.

The MSIVs are open.

SLC-P-1A and 1B are in operation and injecting normally.

Reactor level is -70", controlled with Feed and Condensate.

Which of the following is correct concerning these conditions?

Cooldown A. is not permitted to start until Cold Shutdown Boron Weight has been injected because core reactivity response for a partially borated core is unpredictable.

B. is not permitted to start until all outer ring rods are inserted because additional heat load will be imposed on the primary containment that could lead to containment failure.

C. is not permitted to start until Hot Shutdown Boron Weight has been injected because core reactivity response for a partially borated core is unpredictable.

D. is permitted immediately as long as it is secured if reactor power rises.

Answer: A Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question directly asks the REASON why cooldown is not allowed in an ATWS situation when cold shutdown boron weight has not been reached.

SRO Only:

N/A Explanation:

A. Correct. Per PPM 5.0.10, during an ATWS, cooldown is only permitted when existing control rod pattern alone can always assure reactor shutdown. This is determined with a calculation or by injecting sufficient boron into the primary. The amount of boron necessary to ensure that the reactor is shutdown under all conditions is Cold Boron Shutdown Weight.

B. Incorrect. The concern is not related to drywell parameters. Plausible since, in some ATWS conditions, SRVs would be used to supplement BPVs in cooling down the reactor and cooling down the reactor with SRVs raises containment pressure. However, the limitation on cooldown during an ATWS is based on maintaining subcriticality.

C. Incorrect. Plausible because Hot Shutdown Boron Weight exists and applies to BIIT (Boron Injection Initiation Temperature). However, per PPM 5.0.10, during an ATWS, cooldown is only permitted when existing control rod pattern alone can always assure reactor shutdown. This is determined with a calculation or by injecting sufficient boron into the primary. The amount of boron necessary to ensure that the reactor is shutdown under all conditions is Cold Boron Shutdown Weight, vice Hot Shutdown Boron Weight.

D. Incorrect. Plausible since cooldown during ATWS must be terminated if reactor power is increasing per step P-7 of PPM 5.1.2. However a cooldown can only be initiated if Cold Boron Shutdown Weight has been added or the reactor has been calculated to be shutdown in all conditions. This calculation is only performed if reactor power is below the heating range. Therefore, for the plant conditions given in the stem, a plant cooldown is not allowed.

Technical Reference(s) Attached w/ Revision # See PPM 5.0.10, Flowchart Training Manual Comments / Reference PPM 5.2.1, RPV Control - ATWS Proposed references to be provided during examination: None Learning Objective: 8184 - Given a list, identify the statement that describes plant response to conducting PPM 3.2.1, Plant Shutdown, during an ATWS if Cold Shutdown Weight of boron has not been injected.

Question Source: Bank # LO01280 Modified Bank # (Note changes or attach parent)

New Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Question History: Last NRC Exam None.

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Comments /

Reference:

5.0.10 Revision: 21 mr 1 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.1.2, RPV Control, ATWS Revision: 24 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.1.2, RPV Control, ATWS Revision: 24 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 1 K/A 295038.EK2.07 Level of Difficulty: 3 Importance Rating 3.5 High Off-Site Release Rate: Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following:

Control room ventilation Question # 18 CGS is operating in Mode 1

A High Reactor Building HVAC Exhaust Plenum Radiation Level isolation signal (Z signal) is generated.

How will the Control Room HVAC system respond?

The Control Room HVAC system will shift to A. PRESSURIZATION MODE with the Remote Air Intakes isolated.

B. RECIRCULATION MODE with the Normal Air Intake isolated.

C. PRESSURIZATION MODE with the Normal Air Intake isolated.

D. RECIRCULATION MODE with the Remote Air Intakes isolated.

Answer: C Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question asks how the Control Room Ventilation System will respond on a high exhaust plenum radiation level (Z signal).

SRO Only:

N/A Explanation:

On a Z signal, the Control Room Ventilation air intake valves isolate and suction is shifted to the remote air intakes to PRESSURIZE the control room.

A. Incorrect. The remote air intakes would not be isolated. Plausible because the remote air intakes are not normally used.

B. Incorrect. The system would be in PRESSURIZATION mode. Plausible because the normal air intakes are isolated and because RECIRCULATION implies no outside air enters the control room.

C. Correct.

D. Incorrect. The remote air intakes are not isolated. Plausible because the remote air intakes are not normally used and because RECIRCULATION implies that no outside air enters the control room.

Technical Reference(s) Attached w/ Revision # See SD000201, CR-HVAC System Description Comments / Reference Proposed references to be provided during examination: None Learning Objective: 5225 - State the automatic features associated with the following Control Room HVAC components: (f) Normal Supply Isolation, (g) Remote Intake Valves Question Source: Bank #

Modified Bank # LO02947 (Note changes or attach parent)

New Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.8 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000201, CR-HVAC System Description Revision: 15 mr1 Comments /

Reference:

Original Bank Question, LO02947 Revision: N/A An offsite release is in progress when the following alarms are received:

REMOTE INTAKE DIV 1 RAD HI-HI REMOTE INTAKE DIV 2 RAD HI-HI Considering only these alarms, how will the Control Room HVAC system respond?

A. WOA-V-51C and WOA-V-52C (Normal Outside Air Intake Isolation Valves) close WMA-AD-51A1 and WMA-AD-51B1 (Outside Air Supply Dampers) close WMA-FN-54A and WMA-FN-54B (Emergency Fltr Supply Fans) start WEA-FN-51 (Kitchen Exhaust Fan) stops B. WOA-V-51A and WOA-V-51B (Remote Air Intakes) open WOA-V-51C and WOA-V-52C (Normal Outside Air Intake Isolation Valves) close WMA-FN-54A and WMA-FN-54B (Emergency Fltr Supply Fans) start C. WMA-AD-51A1 and WMA-AD-51B1 (Outside Air Supply Dampers) close WEA-FN-51 (Kitchen Exhaust Fan) stops WEA-AD-51 (Outlet Damper) closes D. WMA-AD-51A1 and WMA-AD-51B1 (Outside Air Supply Dampers) close WMA-FN-54A and WMA-FN-54B (Emergency Fltr Supply Fans) start WEA-FN-51 (Kitchen Exhaust Fan) stops WEA-AD-51 (Outlet Damper) closes Answer: C Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 12/16/2016 Tier 1 Group 1 K/A 600000.AA2.14 Level of Difficulty: 3 Importance Rating 3.0 Plant Fire On Site: Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE: Equipment that will be affected by fire suppression activities in each zone Question # 19 CGS was operating in Mode 1 when a fire in the reactor building was reported. The crew entered ABN-FIRE, Fire. Immediate actions have been completed and the crew is performing step 4.1.4 of subsequent operator actions.

The fire brigade leader has reported the following:

The fire is out and a re-flash watch is stationed.

There is extensive smoke in Room 410. Specific equipment damage cannot be ascertained at this time.

Using the references provided, what actions are REQUIRED for post-fire safe shutdown (PFSS)?

If a reactor shutdown is required, A. use Division 2 PFSS systems. Start RHR-P-2A within 30 minutes.

B. use Division 1 PFSS systems. Open E-CB-DP/SS/IN4/A/4.

C. use Division 2 PFSS systems. Open door R408 to provide passive cooling to Room 410.

D. use Division 1 PFSS systems. Close RCC-V-130, FPC HX RCC Outlet, from the control room.

Answer: D Page 1 of 13

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question asks candidates to use plant procedures and maps to determine equipment impacted by the fire in certain fire protection zones and actions required based on this.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since actions are required for PFSS (starred steps). These actions are correct for Fire Area R-1, which is the reactor building general area. However, Room 410 (R410) is Fire Area R-18. This area has separate PFSS required actions.

B. Incorrect. Plausible since using Division 1 PFSS equipment is correct for Room 410. However, opening E-CB-DP/SS/IN4/A/4 is NOT a starred step. Therefore, it is not a REQUIRED PFSS step.

Additionally, opening E-CB-DP/SS/IN4/A/4 is to be performed for a fire in the reactor building general area, Fire Area R-1, not for Room 410 (Fire Area R-18).

C. Incorrect. Plausible since these actions are required for PFSS (starred steps). These actions are correct for Fire Area R-1, which is the reactor building general area. However, Room 410 (R410) is Fire Area R-18. This area has separate PFSS required actions.

D. Correct. Per the pre-fire plan, Room 410 is Fire Area R-18. Using Division 1 PFSS systems and closing RCC-V-130, FPC HX RCC Outlet, from the control room, are starred steps for a fire in Fire Area 18. Therefore, they are required for PFSS operability.

Technical Reference(s) Attached w/ Revision # See ABN-FIRE, Fire Comments / Reference PFB-RB-522, Reactor 522 SE Quadrant Map Proposed references to be provided during examination: PFP-RB-522, Reactor 522 SE Quadrant Map ABN-FIRE pages 39-42, 56-58 Learning Objective: 15765 - With procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-FIRE.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Page 2 of 13

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 10 CFR Part 55 Content: 55.41 41.10 55.43 Comments /

Reference:

ABN-FIRE, Bases Revision: 36 mr 1 Page 3 of 13

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-FIRE Revision: 36 mr 1 Page 4 of 13

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 5 of 13

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 6 of 13

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ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 10 of 13

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PFP-RB-522 Revision: 5 Page 11 of 13

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 12 of 13

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 13 of 13

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 1 K/A 700000.2.4.4 Level of Difficulty: 2 Importance Rating 4.5 Generator Voltage and Electric Grid Disturbances: Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Question # 20 CGS is operating in Mode 1 when the following alarms are received:

XFMR TR-S TROUBLE, P800.C4.2-7

OSCILLOGRAPH STARTED, P800.C4.4-3 The crew observes the following Startup Transformer (E-TR-S) input phase voltages:

A: 235 kv

B: 10 kv

C: 235 kv What actions should the crew take?

The crew should enter A. ABN-ELEC-GRID, Degraded Off Site Power Grid.

B. ABN-ELEC-LOOP, Loss of All Off-Site Electrical Power.

C. ABN-ELEC-SM2/SM4, SM-2, SM-4 and SL-21 Distribution System Failures.

D. ABN-ELEC-SM1/SM7, SM-1, SM-7, SM-75, SM-72, SL-71, SL-73 & SL-11 Distribution System Failures.

Answer: A Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Students must analyze phase voltages to determine an abnormal condition exists and then determine which procedure to enter based on those conditions. In this case, the indications are related to a grid disturbance/malfunction.

SRO Only:

N/A Explanation:

The conditions presented in the stem match entry conditions in ABN-ELEC-GRID which state, A loss of one phase on E-TR-S as determined by the following: Low voltage on one or more phases A. Correct. The entry conditions for ABN-ELEC-GRID include indications of a loss of phase from E-TR-S.

B. Incorrect. Plausible since TR-S is an off-site source that is lost. However, the Back-up Transformer, TR-B, is available, and entry into ABN-ELEC-LOOP is not applicable.

C. Incorrect. Power to SM-2 and SM-4 are from TR-N (Mode 1) and have not lost power. Plausible because TR-S powers SM-2 and SM-4 when the plant is shutdown.

D. Incorrect. Power to SM-1 and SM-7 are from TR-N (Mode 1) and have not lost power. Plausible because TR-S powers SM-1 and SM-7.

Technical Reference(s) Attached w/ Revision # See ABN-ELEC-GRID, Degraded Off Site Power Grid Comments / Reference Proposed references to be provided during examination: None Learning Objective: 12153 - Given plant annunciation and indications, evaluate conditions for entry into ABN-ELEC-GRID.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-ELEC-GRID Revision: 7 mr 3 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/9/2017 Tier 1 Group 2 K/A 295002.AA2.04 Level of Difficulty: 2 Importance Rating 2.8 Loss of Main Condenser Vacuum: Ability to determine and/or interpret the following as they apply to LOSS OF MAIN CONDENSER VACUUM : Offgas system flow Question # 21 CGS is operating in Mode 1.

The following annunciators are alarming:

H13.P602.A5.6-3, OFF GAS POST TREATMENT RADIATION MONITORS DOWNSCALE CRO2 reports that both Off Gas Post Treatment Radiation Monitors, OG-RIS-601A and B, are reading downscale.

How will this affect the plant and what actions should be taken to mitigate these effects?

OG-V-60, Off Gas Outlet Isolation Valve, will (1) . The crew should take actions as required by (2) .

A. (1) close (2) ABN-BACKPRESSURE, Loss of Main Condenser Backpressure B. (1) close (2) ABN-OG, Off-Gas System Trouble C. (1) remain open (2) ABN-OG, Off-Gas System Trouble D. (1) remain open (2) ABN-BACKPRESSURE, Loss of Main Condenser Backpressure Answer: A Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The candidate must understand the implications of high main condenser backpressure including the expected offgas flow change.

SRO Only:

N/A Explanation:

A. Correct. With both monitors downscale, OG-V-60 will close. This will stop off gas flow and main condenser pressure will rise. Actions should be taken per ABN-BACKPRESSURE.

B. Incorrect. Plausible since OG-V-60 will close with both monitors downscale and the alarm response for the annunciator directs operators to refer to ABN-OG. However, conditions in the question stem do not meet entry conditions for ABN-OG. With OG-V-60 closed, condenser pressure will rise and operators will take actions per ABN-BACKPRESSURE if OG-V-60 is closed since condenser vacuum will be degrading.

C. Incorrect. Plausible if it is believed that one of the radiation monitors must be in high condition to cause OG-V-60 to close. Additionally, since other OG valves are impacted by a downscale monitor, and the alarming annunciator refers the operator to ABN-OG. However, conditions in the question stem do not meet entry conditions for ABN-OG. Additionally, with both monitors downscale, OG-V-60 will close. This will stop off gas flow and main condenser pressure will rise.

Actions should be taken per ABN-BACKPRESSURE.

D. Incorrect. Plausible if it is believed that one of the radiation monitors must be in high condition to cause OG-V-60 to close. Additionally, entry into ABN-BACKPRESSURE is plausible since other OG valves are impacted by a downscale monitor. However, with both monitors downscale, OG-V-60 will close.

Technical Reference(s) Attached w/ Revision # See ABN-BACKPRESSURE, Loss of Main Condenser Backpressure Comments / Reference ABN-OG, Off-Gas System Trouble 4.602.A5.6-3, Alarm response for OFF GAS POST TREATMENT RADIATION MONITORS DOWNSCALE Proposed references to be provided during examination: None Learning Objective: 5111 Discuss the interrelationships between the Air Removal System and the following: (d) Off-Gas System Question Source: Bank # LO03484 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Comments /

Reference:

4.602.A5.6-3 Revision: 046 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-OG Revision: 004

Comments /

Reference:

ABN-BACKPRESSURE Revision: 5 mr 1 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 1 Group 2 K/A 295009.AK2.01 Level of Difficulty: 3 Importance Rating 3.9 Low Reactor Water Level: Knowledge of the interrelations between LOW REACTOR WATER LEVEL and the following:

Reactor water level indication.

Question # 22 CGS is operating in Mode 1 when an automatic reactor scram is initiated.

Current plant conditions:

Narrow Range RPV level: 0 inches

Wide Range RPV level: -149 inches and stable

Compensated Fuel Zone RPV level: -129 inches and stable

Upset Range RPV level: 0 inches

Drywell Temperature: 175° F up slow

Reactor Pressure 480 psig down slow Which of the following is correct concerning these indications?

The RO should report RPV water level as A. 0 inches.

B. -129 inches.

C. -149 inches.

D. cannot be determined.

Answer: B Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question presents a situation where a low RPV level exists. Candidates must know the requirements for when different indications are considered correct which is a direct understanding of the capabilities of the various RPV level instruments.

SRO Only:

N/A Explanation:

Per PPM 5.2.1 caution 1, Wide Range RPV level is not usable below -147. Upset level is not usable below +15 when Drywell temperature is above 161F. Therefore the correct answer is to use Fuel Zone RPV level.

A. Incorrect. Upset Range and Narrow Range indications should not be used based on PPM 5.2.1.

Plausible because they are actual RPV level indications and could be used if the situation was different.

B. Correct.

C. Incorrect. Wide Range indication should not be used because it is below -147 per PPM 5.2.1.

Plausible because WR indication would be preferred if level was higher.

D. Incorrect. Plausible since this would be the report is there were no level indications available due to being outside usable range. However, one of four ranges is still usable and the operator would report the level indicated on the operable range. With no usable RPV level indications, a transition to EOP 5.1.4, RPV Flooding, is required.

Technical Reference(s) Attached w/ Revision # See PPM 5.2.1, Primary Containment Control Comments / Reference Proposed references to be provided during examination: None Learning Objective: 11774 - Describe the operational implications of the following concepts as they apply to the Nuclear Boiler Instrumentation System: (a) Vessel level measurement Question Source: Bank # LO01558 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Comments /

Reference:

PPM 5.2.1 Revision: 23 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/8/2017 Tier 1 Group 2 K/A 295012.AK3.01 Level of Difficulty: 3 Importance Rating 3.5 High Drywell Temperature: Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL TEMPERATURE : Increased drywell cooling Question # 23 CGS is operating in Mode 1

A reactor scram has been initiated due to a steam leak in the drywell.

Drywell pressure is 1.5 psig, up slow

Drywell temperature is 137°F, up slow The crew has entered PPM 5.2.1, Primary Containment Control, and is performing the following step:

Which of the following describes the reason for this direction?

A. This action assures that the normal method of temperature control is attempted in advance of more complex actions.

B. This action assumes normal cooling is not functional and to use whatever cooling is available under the given plant conditions.

C. Other means to control temperature, such as containment spray, are not available until a LOCA signal has been received.

D. This direction is given as an initial action since drywell cooling equipment will load shed if drywell conditions degrade.

Answer: A Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question asks the reason drywell cooling is MAXIMIZED in step DT-1 to maintain drywell temperature below normal values of 135F. If drywell fans were not running or cooling was not available it would be expected that operators start the fans to maximize cooling.

SRO Only:

N/A Explanation:

A. Correct. Per PPM 5.0.10, section 8.8.5.a states that PPM 5.2.1, step DT-1 assures that the normal method of drywell temperature control is attempted in advance of initiating more complex actions to terminate increasing drywell temperature. This minimizes use of cooling methods that might have long term consequences on plant operation.

B. Incorrect. Plausible since containment temperature given in the stem implies that normal cooling systems are not functioning properly. However, containment cooling is achieved methodically, following the PPM 5.2.1 flowchart. See explanation for answer A for correct answer.

C. Incorrect. Plausible since, under most conditions, containment temperature will not reach temperatures requiring containment spray without a primary system discharging into containment.

Additionally, a LOCA signal is not required to initiate containment spray. Per PPM 5.2.1, containment spray initiation is predicated on containment temperature. However, for the plant conditions given in the stem, containment spray is not authorized since containment conditions are not within the Drywell Spray Initiation Limit (DSIL). See explanation for answer A for correct answer.

D. Incorrect. Plausible since the DW cooling fans will deenergize on a loss of power and require manual restarting. However, the fans will not load shed on high drywell pressure (F signal). See explanation for distractor A for correct answer.

Technical Reference(s)

PPM 5.0.10, Flowchart Training Manual Attached w/ Revision # See PPM 5.2.1, Primary Containment Control Comments / Reference Proposed references to be provided during examination: None Learning Objective: 8312 - Given a list, identify the statement that describes the purpose of using drywell cooling as the first method of attempting to control drywell temperature.

Question Source: Bank # LO00105 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 1999, question #65 Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.10 55.43 Comments /

Reference:

PPM 5.0.10 Revision: 21 mr 1 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.2.1, Primary Containment Control Revision: 23 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/1/2017 Tier 1 Group 2 K/A 295015.AA1.02 Level of Difficulty: 3 Importance Rating 4.0 Incomplete SCRAM: Ability to operate and/or monitor the following as they apply to INCOMPLETE SCRAM : RPS Question # 24 CGS is operating in Mode 1.

While preparing to shift RPS to Alternate power following the trip of RPS-MG-1, the following indications are observed at H13-P610:

.

Which of the following correctly describes the expected response for the given MG Set Transfer Switch operations?

Placing the MG Set Transfer Switch in...

A. ALT B would not affect any of the white indicating lights but will cause a full reactor scram.

B. ALT A would not affect any of the white indicating lights and RPS Bus A would remain de-energized.

C. ALT A would cause the white GENERATOR A FEED light to illuminate and re-energize RPS Bus A from the alternate source.

D. ALT B would cause the white GENERATOR A FEED light to illuminate and re-energize RPS Bus A from RPS-MG-2.

Page 1 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Answer: A Page 2 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

A loss of a RPS MG set causes a half scram. The question challenges the candidates ability to restore power via the alternate source and correctly interpret indications.

SRO Only:

N/A Explanation:

A. Correct. The 3 lights on the transfer switch show the power supplies that are available to power the RPS busses. The condition of the lights do not change as the switch is repositioned. The switch positions are as follows:

NORM: Both RPS buses are powered from their respective RPS motor generators

ALT A: RPS bus A is powered from the alternate power supply (MC-6B), RPS bus B is powered from its normal power supply (RPS MG B).

ALT B: RPS bus B is powered from the alternate power supply (MC-6B), RPS bus A is powered from its normal power supply (RPS MG A).

The transfer switch is break before make; the bus that is being selected will be completely disconnected from any power supply while the switch is being repositioned.

For the conditions given in the stem, RPS A is deenergized (Generator A Feed light is out with the transfer switch in NORM) and RPS B is energized by RPS MG B. When the transfer switch is taken to ALT B, RPS B will momentarily deenergize. Since RPS A is already deenergized, a reactor scram will occur.

B. Incorrect. Plausible since RPS would remain deenergized if it is believed that the power supply lights correlate to the transfer switch positions. However, taking the transfer switch to ALT A will energize RPS A from the alternate power supply. See answer A explanation for a discussion of transfer switch operation.

C. Incorrect. Plausible since taking the transfer switch to ALT A will energize RPS A from the alternate power supply. However, the GENERATOR A FEED will not light since RPS MG A is tripped. See answer A explanation for a discussion of transfer switch operation.

D. Incorrect. Plausible if it is believed that RPS busses can be supplied from either RPS MGs.

However, placing the switch in ALT B will cause a reactor scram. See answer A explanation for a discussion of transfer switch operation.

Technical Reference(s) Attached w/ Revision # See SD000161, CGS System Description, Volume 6, Chapter 8, Reactor Comments / Reference Protection System Proposed references to be provided during examination: None Page 3 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Learning Objective: 5961 - Describe the electrical alignment of RPS when the MG SET TRANSFER switch (on control room panel H13-P610) is in: NORM/ALT A/ALT B Question Source: Bank # LR00171 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.6 55.43 Comments /

Reference:

SD000161 Revision: 17 mr1 Page 4 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 5 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 6 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/8/2017 Tier 1 Group 2 K/A 295029.EK1.01 Level of Difficulty: 2 Importance Rating 3.4 High Suppression Pool Water Level: Knowledge of the operational implications of the following concepts as they apply to HIGH SUPPRESSION POOL WATER LEVEL : Containment integrity Question # 25 Following a major plant transient, primary containment water level has risen to 555'.

Which of the following is correct?

If adequate core cooling is assured, injection from sources external to the primary containment should be stopped in order to prevent...

A. unnecessary depletion of Condensate Storage Tank inventory below Technical Specification limits.

B. loss of primary containment integrity and possible substantial radioactivity release.

C. failure of low pressure Emergency Core Cooling System (ECCS) suction piping due to the static head of water.

D. overloading Emergency Core Cooling System (ECCS) pump motors due to the additional flow that would result from the higher suction head.

Answer: B Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question asks if candidates understand the consequences and operational implications if primary containment water level rises to 555. The correct answer is related to containment integrity.

SRO Only:

N/A Explanation:

Per PPM 5.0.10, the reason for PCPL limits is primary containment integrity.

A. Incorrect. Plausible due to Technical Specification Limits for CST inventory to maintain supported SSCs operable. Incorrect since the purpose of maintaining CST inventory is to respond to LOCA or accident conditions such as that described in the question stem. The water is to be used for the purpose provided.

B. Correct. Per section 7.13.1 of PPM 5.0.10, the primary containment pressure limit above 535 feet is based on the pressure capability of the girder joint of the primary containment. Failure of the primary containment girders could lead to containment failure and radioactivity release during a plant event.

C. Incorrect. Plausible because ECCS pumps draw suction from the suppression pool and primary containment pressure could be felt on the ECCS pump suctions. However, the limiting case for the conditions listed in the stem is the pressure capability of the girder joint of the primary containment.

D. Incorrect. Plausible because ECCS pumps draw suction from the suppression pool and increased primary containment will provide increased NPSH to the ECCS pumps. Incorrect because ECCS pump failure is not the primary concern. However, the limiting case for the conditions listed in the stem is the pressure capability of the girder joint of the primary containment.

Technical Reference(s) Attached w/ Revision # See PPM 5.0.10, Flowchart Training Manual Comments / Reference Proposed references to be provided during examination: None Learning Objective: 13567 - Given a copy of Emergency Operating Procedures (EOPs) and an event, describe the basis for each variable and figure used to execute the strategies of the EOPs without error.

Question Source: Bank # LR00600 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.9 55.43 Comments /

Reference:

PPM 5.0.10 Revision: 21 mr 1 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 0 Date: 12/16/2016 Tier 1 Group 2 K/A 295034.2.2.40 Level of Difficulty: 3 Importance Rating 3.4 Secondary Containment Ventilation High Radiation: Ability to apply Technical Specifications for a system.

Question # 26 The following describes different operating conditions of the reactor plant:

(1) Mode 1 (2) Mode 2 (3) Mode 3 (4) Mode 4 (5) Mode 5 (6) Operations with a potential for draining the reactor vessel (OPDRVs)

Which of the following describes all the conditions where the Reactor Building Vent Exhaust Plenum Radiation - High function of Secondary Containment Isolation Instrumentation is required by technical specifications?

A. (1), (2), (3)

B. (4), (5), (6)

C. (1), (2), (3), (6)

D. (3), (4), (5), (6)

Answer: C Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question discriminates candidates ability to correct apply technical specification applicability requirements related to secondary containment isolation ability.

SRO Only:

N/A Explanation:

Per given reference, student must analyze table 3.3.6.2.1 to determine that C is the correct answer.

A. Incorrect. Plausible because it logically follows other distractor format.

B. Incorrect. Plausible because it would be correct if note (a) did not apply.

C. Correct.

D. Incorrect. Plausible because manual initiation is not always required in technical specifications for systems and if candidate does not correctly read table 3.3.6.2.1(a).

Technical Reference(s) Attached w/ Revision # See T.S. 3.3.6.2, Secondary Containment Isolation Instrumentation LCO Comments / Reference Proposed references to be provided during examination: None Learning Objective: 5606 - Given a copy of Technical Specifications, locate and apply all of the T.S.

that are related to the NS4 system.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Tech Spec 3.3.6.2 Revision: 237 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/2/2017 Tier 1 Group 2 K/A 295036.EK1.01 Level of Difficulty: 3 Importance Rating 2.9 Secondary Containment High Sump/Area Water Level: Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL : Radiation releases Question # 27 CGS is operating in Mode 1.

A steam leak occurs in the RCIC room.

Reactor Building Ventilation exhaust radiation level is 15 mr/hr, up slow.

10 minutes later, the following annunciator alarms:

H13.P602.A13.3-1, REACTOR BUILDING EQUIPMENT SUMP HIGH LEVEL What is the condition of the Equipment Drain (EDR) system?

EDR-SUMP-R5 Pump Discharge Isolation valves, EDR-V-394/395, are (1) and EDR Sump Pump, EDR-P-5A is (2) .

A. (1) open (2) running B. (1) closed (2) secured C. (1) closed (2) running D. (1) open (2) secured Answer: B Page 1 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This question examines the candidates understanding of the response of the Plant Drain system to mitigate a uncontrolled radiation release during a high radiation level/high sump level condition.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since the sump pump would be running with a high sump level if the sump pump discharge valves were open. However, the sump pump discharge valves close when reactor building ventilation exhaust radiation levels are 13 mr/hr (Z signal), and the sump pump will not start regardless of sump level.

B. Correct. EDR-SUMP-R5 Pump Discharge Isolation valves, EDR-V-394/395, will close on a Z signal (reactor building ventilation exhaust radiation level 13 mr/hr). With the discharge isolation valves closed, the sump pump will not run in discharge mode or recirculation mode.

C. Incorrect. Plausible since EDR-SUMP-R5 Pump Discharge Isolation valves, EDR-V-394/395, will close on a Z signal (reactor building ventilation exhaust radiation level 13 mr/hr), and the sump pump should be running with a high sump level. However, the sump pump discharge valves close when reactor building ventilation exhaust radiation levels are 13 mr/hr (Z signal), and the sump pump will not start regardless of sump level.

D. Incorrect. Plausible since the equipment drain system primary containment isolation valves automatically close only on high drywell pressure (A signal) and RPV low Level 2 (F signal). Both of these signals are not present. However, EDR-SUMP-R5 Pump Discharge Isolation valves, EDR-V-394/395, will close on a Z signal (reactor building ventilation exhaust radiation level 13 mr/hr). With the discharge isolation valves closed, the sump pump will not run in discharge mode or recirculation mode.

Technical Reference(s)

SD000130, CGS System Description, Volume 9, Chapter 7, Plant Attached w/ Revision # See Drains Comments / Reference PPM 4.602.A13, 602.A13 Annunciator Panel Alarms Proposed references to be provided during examination: None Learning Objective: 5333 - List the isolation signals and setpoints for the following valves: (e) EDR-V-394 (f) EDR-V-395 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Page 2 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 10 CFR Part 55 Content: 55.41 41.13 55.43 Comments /

Reference:

SD000130, section V.B.3.1, EDR sump Revision: Major 12 pump starting logic Page 3 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000130, section V.C.4, EDR sump Revision: Major 12 discharge isolation valve closing logic Page 4 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Ref. B, Annunciator 3-1, REACTOR Revision: Major 024 BUILDING EQUIPMENT SUMP HIGH LEVEL, response.

Page 5 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Ref. B, Annunciator 4-1, REACTOR Revision: Major 024 BUILDING EQUIPMENT SUMP TEMPERATURE HIGH, response.

Page 6 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 2 Group 1 K/A 203000.K5.02 Level of Difficulty: 2 Importance Rating 3.5 RHR/LPCI: Injection Mode: Knowledge of the operational implications of the following concepts as they apply to RHR/LPCI: INJECTION MODE: Core cooling methods Question # 28 Select the valve that has an automatic open signal only for the first ten minutes after an emergency core cooling system (ECCS) initiation signal.

A. RHR-V-42A (injection valve).

B. RHR-V-48A (heat exchanger bypass).

C. RHR-V-53A (shutdown cooling return isolation).

D. RHR-V-24A (suppression pool cooling return).

Answer: B Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question asks for candidate knowledge of ECCS interlocks related to core cooling (HTXR bypass valve).

SRO Only:

Type an explanation as to how this question clearly meets the NRC guidance for SRO only questions.

If this cannot be done without a lengthy explanation the question may be RO level.

Explanation:

A. Incorrect. Plausible since RHR-V-42A receives an automatic open signal on a LPCI initiation signal. However, RPV pressure must be below 470 psig concurrently with the Level 1 signal, and the automatic open signal does not automatically clear after 10 minutes.

B. Correct. Upon receipt of an LPCI initiation signal, RHR-V-48A will receive a signal to go full open to ensure maximum LPCI flow is available to the RPV if needed. 10 minutes after the ECCS initiation signal, the automatic open signal is removed to allow the valve to be throttled as necessary to provide necessary cooling to the RPV.

C. Incorrect. Plausible since RHR-V-53A receives a closed signal when RPV level reaches +13 inches (Level 3) to isolate shutdown cooling from the RPV during a LOCA. Additionally, the valve cannot be opened for 10 minutes after a LPCI initiation signal. However, it does not receive an open signal.

D. Incorrect. Plausible since RHR-V-24A receives an automatic close signal when LPCI initiates.

does not receive an open signal. Plausible because it is an RHR valve related to cooling.

However, it does not receive an open signal and the signal does not clear 10 minutes after the Level 3 signal is received.

Technical Reference(s) Attached w/ Revision # See SD000198, CGS System Description, Volume 7, Chapter 4, Comments / Reference Residual Heat Removal Proposed references to be provided during examination: None Learning Objective: 11801 - Describe the function, purpose and design features of the following RHR components: (p) RHR-V-48A Question Source: Bank # LO00834 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 10 CFR Part 55 Content: 55.41 41.5 55.43 Comments /

Reference:

SD000198 Revision: 16 mr 0 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO X SRO Rev.3 Date: 2/2/2017 Tier 2 Group 1 K/A 203000.A1.03 Level of Difficulty: 3 Importance Rating 3.8 RHR/LPCI: Injection Mode: Ability to predict and/or monitor changes in parameters associated with operating the RHR/LPCI:

INJECTION MODE controls including: System flow.

Question # 29 CGS is operating in Mode1.

A LOCA has occurred, requiring the crew to scram the reactor and enter the EOPs.

Current plant conditions:

RPV level is -135 inches, up slow.

RPV pressure is 320 psig, down slow.

Drywell pressure peaked at 1.60 psig. Current pressure is 1.55 psig, down slow.

RHR-P-2B failed to automatically start.

All other automatic actuations were successful.

What is the status of RHR Loop B if the operator ARMS and DEPRESSES RHR-RMS-S60 (RHR B and C MANUAL INITIATION pushbutton)?

RHR-V-64B (RHR Loop B Minimum Flow Bypass valve) is (1) and RHR Loop B is (2) into the RPV.

A. (1) open (2) injecting B. (1) closed (2) not injecting C. (1) closed (2) injecting D. (1) open (2) not injecting Answer: D Page 1 of 4

Justification for Modifying Question RO-29 During the Examination.

Modified Question:

CGS is operating in Mode1.

A LOCA has occurred, requiring the crew to scram the reactor and enter the EOPs.

Current plant conditions:

RPV level is -135 inches, up slow.

RPV pressure is 320 psig, down slow.

Drywell pressure peaked at 1.60 psig. Current pressure is 1.55 psig, down slow.

RHR-P-2B failed to automatically start.

All other automatic actuations were successful.

What is the status of RHR Loop B if one minute after the operator ARMS and DEPRESSES RHR-RMS-S60 (RHR B and C MANUAL INITIATION pushbutton)?

RHR-V-64B (RHR Loop B Minimum Flow Bypass valve) is (1) and RHR Loop B is (2) into the RPV.

A. (1) open (2) injecting B. (1) closed (2) not injecting C. (1) closed (2) injecting D. (1) open (2) not injecting As-Given Question. See reason on next page.

ATTACHMENT 10 (Page 4 of 5)

Justification for Modifying Question RO-29 During the Examination.

Question 55-42915: After Arm & Depress What is the status? Does this mean immediately, 1 second later, 15 seconds later?

Station Response RHR-V-64B has an eight second delay prior to opening on an automatic open signal.

The original question stem did not specify a time for checking system status after the automatic initiation signal from operating the RHR A and B Manual Initiation pushbutton.

After consulting with the chief examiner, the question stem was modified, as described below.

The change was communicated to all applicants during the examination.

ATTACHMENT 10 (Page 3 of 5)

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This question requires the candidate to demonstrate knowledge of how the LPCI system responds when manually started after a LOCA.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since a valid A signal is present with RPV level at -135 inches and RHR-P-2B will start. Additionally, if it is believed that RHR pump shutoff head is similar to the LPCS pump shutoff head ( 320 psig) and the pump is injecting some flow into the core, but not full flow since RPV pressure is close to the pump shutoff head, the min. flow bypass valve would be open.

However, RHR pump shutoff head is approximately 220 psig and the pump will not be injecting at the RPV pressure given.

B. Incorrect. Plausible if it is believed that the issue that caused the RHR pump not to start in manual affects the manual initiation circuitry. Additionally, the min. flow bypass valve position is consistent with the pump not running. However, the manual initiation signal is separate from the automatic initiation signal, so the pump will start on manual initiation. Additionally, the min. flow bypass valve will be open since RPV pressure is above RHR pump shutoff head and the RHR pump will not be injecting into the RPV.

C. Incorrect. Plausible if it is believed that the RHR pump shutoff head is above the RPV pressure given in the stem. However, RHR pump shutoff head is 220 psig, which is 100 psig below the RPV pressure given in the stem. The RHR pump will be running, but not injecting into the RPV.

The min. flow bypass valve will be open since the RHR pump breaker is closed and flow is LT 800 gpm.

D. Correct. With a valid A initiation signal, RHR-P-2B will start. Since RPV pressure is greater than the pump shutoff head of 220 psig, there will be no injection into the RPV and the minimum flow bypass valve will be open.

Technical Reference(s) Attached w/ Revision # See SOP-RHR-INJECTION, RHR RPV Injection Comments / Reference Proposed references to be provided during examination: None Learning Objective: 5779 - Describe the expected system response for any routine lineup, when the initiation logic for the LPCI mode of the RHR system is satisfied.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.8 55.43 Comments /

Reference:

SOP-RHRA-INJECTION Revision: Major 003 Minor 002 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/1/2017 Tier 2 Group 1 K/A 205000.K6.01 Level of Difficulty: 3 Importance Rating 3.3 Shutdown Cooling System (RHR Shutdown Cooling Mode): Knowledge of the effect that a loss or malfunction of the following will have on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) : A.C. electrical power Question # 30 CGS is in Mode 4.

RHR Loop A is in Shutdown Cooling (SDC) mode.

DG-2 is inoperable and unavailable.

RRC pumps are secured.

A Loss of Offsite Power (LOOP) occurs.

With no operator action, what is the status of the following SDC components 1 minute after the LOOP?

(1) RHR-V-8, RHR Shutdown Cooling Suction Outboard Isolation (2) RHR-V-9, RHR Shutdown Cooling Suction Inboard Isolation A. (1) Closed (2) Closed B. (1) Open (2) Open C. (1) Closed (2) Open D. (1) Open (2) Closed Answer: C Page 1 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This question requires the candidate to demonstrate knowledge of the effect of a loss of an AC bus on the shutdown cooling lineup.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since on a loss of offsite power where both SM-7 and SM-8 are restored, this would be the expected lineup due to both RPS busses deenergizing (see explanation for answer C). However, since SM-8 is de-energized, RHR-V-9 will not close even though it has a signal to do so.

B. Incorrect. Plausible since both valves would be open if the RPS motor generator flywheels sustained RPS busses until the diesel generators could pick up the busses DG-2 was available to energize SM-8. However, both RPS busses are lost following the LOOP (see explanation for answer C) and an isolation signal is generated for all NS4 inboard and outboard valves. Since SM-8 is de-energized (DG-2 unavailable), RHR-V-9 will not close.

C. Correct. With the reactor shutdown, SM-7 and SM-8 are powered from TR-S. With a LOOP, both RPS busses are lost since the RPS motor generator flywheels will maintain power for approximately 4 seconds and the shift to the diesel generators takes approximately 10 seconds.

Since both RPS buses are lost, an isolation signal is generated for all NS4 inboard and outboard valves. Since SM-8 is de-energized (DG-2 unavailable), RHR-V-9 will not close.

D. Incorrect. Plausible since. valve positions are correct if it is believed that loss of RPS B only generates a NS4 inboard isolation signal and RPS A was not lost. However, both RPS A and B are lost (see discussion for answer B), generating an isolation signal for both inboard and outboard valves.

Technical Reference(s) Attached w/ Revision # See SD000182, CGS System Description, Vol. 1 Chap. 2, AC Distribution Comments / Reference SD000161, CGS System Description, Vol. 6 Chap. 8, Reactor Protection System Proposed references to be provided during examination: None Learning Objective: 5781 - List the interlocks and trips associated with the following RHR system components: a. RHR pumps, d. RHR-V-8 & RHR-V-9 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Page 2 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Comments /

Reference:

AC Distribution System Description Revision: 19 Page 3 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 4 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

RPS System Description Revision: Major 17 Minor 001 Page 5 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 6 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 1 K/A 209001.A1.01 Level of Difficulty: 2 Importance Rating 3.4 Low Pressure Core Spray System: Ability to predict and/or monitor changes in parameters associated with operating the LOW PRESSURE CORE SPRAY SYSTEM controls including: Core spray flow Question # 31 Given the following:

LPCS-P-1 automatically initiated on low RPV water level.

LPCS flow is observed to be 7200gpm.

3 minutes later, RPV pressure is 290 psig and the following is observed:

Which of the following describes the reason for these indications?

A. RPV Pressure is above LPCS-P-1 shutoff head.

B. LPCS-P-1 shaft seizure.

C. LPCS-P-1 shaft shear D. LPCS-P-1 supply breaker opened on overcurrent.

Answer: C Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Candidate is using LPCS indications to monitor a change in the system status and describe the cause of that change.

SRO Only:

N/A Explanation:

The red indicating light for LPCS-P-1 means the breaker is still closed and the motor is running.

Current is abnormally low and both flow and discharge pressure have lowered to zero. This indicates a sheared shaft.

A. Incorrect. RPV pressure of 290 psig is below the shutoff head for LPCS-P-1 ( 320 psig). Plausible because LPCS flow would lower to zero if below shutoff head and the red indicating light for LPCS-P-1 would still be on as it is in the pictures.

B. Incorrect. Shaft seizure would result in a breaker trip and a green light on LPCS-P-1. Plausible because a shaft seizure would lower LPCS pump flow and discharge pressure.

C. Correct.

D. Incorrect. Supply breaker opening would result in a green light on LPCS-P-1. Plausible because the supply breaker opening would result in 0 flow and low LPCS discharge pressure.

Technical Reference(s) Attached w/ Revision # See GFES BC05Sr4_Motors Comments / Reference PPM 5.0.10, Flowchart Training Manual PPM 5.1.1, RPV Control Proposed references to be provided during examination: None Learning Objective: 11586 Describe the function, purpose and design features of the following Low Pressure Core Spray System components: b. LPCS pump LPCS-P-1 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

GFES BC05Sr4_Motors Revision: May 2011 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.0.10 Revision: 21 minor 001 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.1.1 Revision: 21 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 1 K/A 209002.K4.07 Level of Difficulty: 2 Importance Rating 3.5 High Pressure Core Spray System (HPCS): Knowledge of HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) design feature(s) and/or interlocks which provide for the following: Override of reactor water level interlock Question # 32 When the HPCS RPV INJ VALVE INTERLOCK OVERRIDE keylock switch (HPCS-RMS-S25) is placed in the "OVERRIDE" position, HPCS-V-4 (HPCS Injection valve),

(1) be throttled and (2) open on a HPCS initiation signal.

A. (1) cannot (2) will not B. (1) can (2) will not C. (1) cannot (2) will D. (1) can (2) will Answer: B Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Questions asks how overriding the HPCS Injection Valve low RPV water level interlock will impact its behavior.

SRO Only:

N/A Explanation:

Overriding the switch allows HPCS-V-4 to be throttled. The valve will no longer open on low RPV water level signals.

A. Incorrect. HPCS-V-4 can be throttled once the switch is in the OVERRIDE positon.

B. Correct.

C. Incorrect. HPCS-V-4 can be throttled once the switch is in the OVERRIDE positon.

D. Incorrect. HPCS-V-4 will no longer open on low RPV level.

Technical Reference(s) Attached w/ Revision # See SD000174, CGS System Description, Volume 7, Chapter 2, High Comments / Reference Pressure Core Spray Proposed references to be provided during examination: None Learning Objective: 7661 - State the function of the HPCS RPV Injection Valve Interlock Override key switch Question Source: Bank # LO02256 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000174 Revision: r13 mr0 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 1 K/A 211000.A4.08 Level of Difficulty: 3 Importance Rating 4.2 Standby Liquid Control System: Ability to manually operate and/or monitor in the control room: System initiation Question # 33 While placing the control switches for SLC-P-1A and SLC-P-1B to OPERATE, SLC-V-1A (Storage tank outlet valve) fails to open (all other components functioned as designed).

Which of the following correctly describes the SLC system response to this condition?

A. SLC-P-1A and SLC-P-1B will start and continue running.

B. SLC-P-1A and SLC-P-1B will not start.

C. SLC-P-1A will not start, SLC-P-1B will start and continue running.

D. SLC-P-1A will start and trip, SLC-P-1B will start and continue running.

Answer: A Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question determines if a candidate is able to understand system initiation response given indications available in the control room.

SRO Only:

N/A Explanation:

SLC pumps will start provided ONE of the tank outlet valves SLC-V-1A or SLC-V-1B is open after the switch for SLC-P-1A and SLC-P-1B are placed in the OPERATE POSTION. The suction to the pumps is connected downstream of the two suction valves.

A. Correct Answer.

B. Incorrect because SLC-P-1B WILL start. Plausible because it is common for pumps to have start interlocks based on their specific suction valve.

C. Incorrect because SLC-P-1A will NOT trip. Plausible because pumps often have low suction pressure trips that would trip the pump if the suction valve were closed.

D. Incorrect because both pumps WILL start. Plausible because some systems have multiple suction valves in series before the pumps that must be open before the system starts.

Technical Reference(s) Attached w/ Revision # See SD000172, CGS System Description, Volume 5, Chapter 3, Standby Comments / Reference Liquid Control Proposed references to be provided during examination: None Learning Objective: 5925 - Describe the expected response to placing the SLC SYSTEM A or B keylock switch in the OPERATE position.

Question Source: Bank # LR00922 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000172 Revision: 13 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 1 K/A 212000.A3.01 Level of Difficulty: 3 Importance Rating 4.4 Reactor Protection System: Ability to monitor automatic operations of the REACTOR PROTECTION SYSTEM including:

Reactor Power Question # 34 CGS is operating in Mode 2.

The following indications are observed:

What is the status of RPS A?

A. IRM ACEG UPSCL TRIP OR INOP annunciator, Rod Block, 1/2 SCRAM B. IRM MONITOR UPSCALE annunciator, NO Rod Block, NO 1/2 SCRAM C. IRM ACEG UPSCL TRIP OR INOP annunciator, NO Rod Block, NO 1/2 SCRAM D. IRM MONITOR UPSCALE annunciator, Rod Block, NO 1/2 SCRAM Answer: D Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Student must interpret/monitor system indications for IRM and determine whether automatic RPS actions have occurred. Indications are based on reactor power and IRM readings.

SRO Only:

N/A Explanation:

Any IRM channel above 120% will cause a 1/2 SCRAM.

Any IRM channel above 108% will cause a ROD BLOCK.

Rod Blocks, 1/2 SCRAMs, and Annunciators are disabled if an IRM channel is bypassed for that channel only.

Because IRM Channel G is bypassed, there is no 1/2 SCRAM. IRM A, however, has a rod block signal.

A. Incorrect. Plausible since a rod block occurs. Incorrect because an IRM ACEG Upscale Trip or Inop would not alarm (channel GT 120%) and there is no 1/2 SCRAM since channel G is bypassed.

B. Incorrect. Plausible since there is no 1/2 SCRAM and the IRM Monitor Upscale annunciator is in alarm (108%). Incorrect because a rod block condition exists.

C. Incorrect. Plausible because IRM ACEG UPSCL TRIP OR INOP annunciator sounds similar to the IRM MONITOR UPSCALE alarm and have similar setpoints (120% vice 108%). Incorrect, however, IRM UPSCL TRIP OR INOP would not be in alarm AND a rod block exists.

D. Correct. With IRM channel A reading 112.5%, the IRM MONITOR UPSCALE annunciator is in alarm and a rod block signal exists. Since channel G is bypassed, no 1/2 SCRAM signal is processed.

Technical Reference(s) Attached w/ Revision # See SD000138, CGS System Description, Volume 6, Chapter 2, Comments / Reference Intermediate Range Monitor PPM 4.603.A7, Annunciator Response Procedure Proposed references to be provided during examination: None Learning Objective: 11794 - Describe the physical connection and/or cause-and-effect relationship between the Intermediate Range Monitoring System and the following: (a) RPS Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Comments /

Reference:

4.603.A7 Revision: 51 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000138 Revision: 10 mr 1 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/8/2017 Tier 2 Group 1 K/A 215003.2.1.31 Level of Difficulty: 3 Importance Rating 4.6 Intermediate Range Monitor (IRM) System: Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

Question # 35 A cold reactor startup is in progress per PPM 3.1.2, Reactor Startup.

All IRMs were operable and fully inserted prior to commencing the startup.

You observe the following indications:

H Which of the following is correct regarding system lineup?

A. IRM H needs to be withdrawn from the core.

B. IRM B, D, and F range switches need to be raised.

C. IRM/APRM overlap requirements are not met on IRM H.

D. IRM B, D, and F need to be inserted for IRM/APRM overlap verification.

Answer: A Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The K/A requires the ability to determine if IRM system are in the correct lineup for plant conditions.

This question requires candidates to understand the required IRM lineup during a reactor startup, to determine that the current IRM lineup is not in the required lineup, and to identify the problem.

SRO Only:

N/A Explanation:

Following APRM/IRM overlap at 5% power, all IRMs are supposed to be withdrawn per PPM 3.1.2 step Q42. In the stem, APRM power indicates close to 10%, all other IRMs are withdrawn, but IRM H has not been withdrawn. The K/A requires the candidate to identify and determine if plant components are in the correct lineup. The question determines if the candidate can identify that IRM H SHOULD be withdrawn and is NOT withdrawn.

A. Correct.

B. Incorrect. IRM range switches should be on Range 1 per step Q42 of PPM 3.1.2. Plausible since during a reactor startup, the IRM range switches must be raised as power increases to keep power on scale.

C. Incorrect. IRM/APRM overlap requirements are met (APRM readings greater than 5%) and have already been verified as indicated by the withdrawal of all other IRMs. Plausible because if the range switch for IRM H was on another range, IRM overlap requirements would NOT be met.

D. Incorrect. IRM/APRM overlap requirements are met (APRM readings greater than 5%) and have already been verified. Plausible because IRMs are inserted during shutdown for overlap verification per PPM 3.2.1, Reactor Plant Shutdown.

Technical Reference(s)

PPM 3.1.2, Reactor Startup, Attachment 7.3, Startup Flowchart Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: 11796 - Describe the operational implications of the following concepts as they apply to the IRM system: (a) Detector operation.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 10 CFR Part 55 Content: 55.41 41.10 55.43 Comments /

Reference:

Revision:

(Ref. B) PPM 3.1.2, Reactor Startup, Attachment 7.3, Startup Major Rev. 081 Flowchart, steps Q40 to Q44 Minor Rev. 001 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Revision:

(Ref. B) PPM 3.1.2, Reactor Startup, Attachment 7.3, Startup Major Rev. 081 Flowchart, steps Q37 to Q39 Minor Rev. 001 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/2/2017 Tier 2 Group 1 K/A 215003.A2.06 Level of Difficulty: 3 Importance Rating 3.0 Intermediate Range Monitor (IRM) System: Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Faulty range switch Question # 36 CGS is operating in Mode 2.

A plant startup is in progress with power increasing.

Power indicates 37/40 with IRM G on range 5.

All other IRM channels are on range 5 and indicate between 20/40 and 30/40.

The range switch for IRM G fails from range 5 to range 4.

IRM G range can no longer be manually changed.

Which of the following is the next action that should be taken, with CRS direction?

A. Shutdown the reactor per PPM 3.2.1, Normal Reactor Shutdown.

B. Reset the 1/2 SCRAM using SOP-RPS-OPS.

C. Enter PPM 3.3.1, Reactor SCRAM.

D. Bypass IRM G per ARP 4.603.A7.

Answer: D Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The candidate must determine the impact to the system based on the failure such as whether or not a SCRAM will occur. They must also know the correct action to take per procedures.

SRO Only:

N/A Explanation:

The failure will cause a rod block and half scram on RPS A. Startup can continue if the IRM channel is bypassed.

A. Incorrect. Shutting down is not necessary. Plausible because 1/2 SCRAM occurs and the IRM system is not functioning as designed.

B. Incorrect. The stem asked for the NEXT action to be taken. Resetting the 1/2 SCRAM will not work until the IRM channel is bypassed. Plausible because a 1/2 SCRAM occurs.

C. Incorrect. An automatic SCRAM does not occur and a manual SCRAM is not required. Plausible because a 1/2 SCRAM occurs.

D. Correct.

Technical Reference(s) Attached w/ Revision # See PPM 4.603.A7, Annunciator Response Procedure Comments / Reference Proposed references to be provided during examination: None Learning Objective: 11798 - Predict the impacts of the following on the Intermediate Range Monitoring System: (f) Faulty Range Switch Question Source: Bank #

Modified Bank # LO01128 (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.6 55.43 Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 4.603.A7 Revision: 51 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LO01128, Original Question Revision: N/A A plant startup is in progress with power at 0.2% and increasing. Maintenance activities has caused RPS B to have a 1/2 scram in on it. A range switch has failed causing IRM G to remain on Range 4.

The failure goes unnoticed. All other plant systems operate as designed.

Which of the following is correct for these conditions?

A. Enter PPM 3.3.1 Reactor Scram B. The startup continues after the rod block is reset.

C. The power increase stops when the rod block is received.

D. Enter PPM 5.1.2 RPV Control ATWS.

Answer: A Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/2/2017 Tier 2 Group 1 K/A 215004.K1.01 Level of Difficulty: 3 Importance Rating 3.6 Source Range Monitor (SRM) System: Knowledge of the physical connections and/or cause-effect relationships between SOURCE RANGE MONITOR (SRM) SYSTEM and the following: Reactor protection system Question # 37 Given the following:

The reactor is in Mode 4.

Reactor Protection System (RPS) shorting links have been removed as part of an approved special test.

As the crew is withdrawing control rods for the special test, the following source range indications are observed:

What is the condition of RPS?

RPS A is (1) and RPS B is (2) .

A. (1) not tripped (2) not tripped B. (1) tripped (2) not tripped C. (1) not tripped (2) tripped Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 D. (1) tripped (2) tripped Answer: D Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question determines if candidate understands when an RPS scram trip would occur based on signals from the SRM system.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since this is the condition that would normally be present. However, with the shorting links removed, any SRM that reaches 2.0x105 cps will cause both RPS channels to trip.

B. Incorrect. Plausible if it is believed that a SRM that reaches 2.0x105 cps will cause a trip of its associated RPS channel. However, one channel above the setpoint will cause both RPS channels to trip.

C. Incorrect. Plausible if it is believed that both channels of SRMs in a single RPS channel will cause a trip of the associated RPS channel and if the setpoint is believed to be 2.0x104 cps. However, with the shorting links removed, any SRM that reaches 2.0x105 cps will cause both RPS channels to trip.

D. Correct. With the shorting links removed, any SRM that reaches 2.0x105 cps will cause both RPS channels to trip.

Technical Reference(s) Attached w/ Revision # See SD000161, CGS System Description, Volume 6, Chapter 8, Reactor Comments / Reference Protection System SD000132, CGS System Description, Volume 6, Chapter 1, Source Range Monitor Proposed references to be provided during examination: None Learning Objective: 11997 - Describe the function, purpose, and design features of the following major SRM system components: (j) trip circuits Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.2 55.43 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000132 Revision: 12 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000161 Revision: 17 mr 1 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 0 Date: 12/16/2016 Tier 2 Group 1 K/A 215005.K2.02 Level of Difficulty: 3 Importance Rating 2.6 Average Power Range Monitor/Local Power Range Monitor System: Knowledge of electrical power supplies to the following: APRM channels Question # 38 What is the power supply to Average Power Range Monitor (APRM) channel "3"?

A. Critical Instrument Power Inverter IN-3.

B. 125 VDC Distribution Panel DP-S1-1A.

C. 24 VDC Distribution Panel DP-SO-A.

D. RPS A or RPS B Answer: D Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question specifically asks the power supply to APRMs.

SRO Only:

N/A Explanation:

APRMs are powered by both RPS A and B via an auctioneered system.

A. Incorrect. Plausible because IN-3 is called the Critical Instrument Power Inverter B. Incorrect. Plausible because DP-S-1A powers safety related DC components.

C. Incorrect. Plausible because DP-SO-A powers safety related DC components.

D. Correct.

Technical Reference(s) Attached w/ Revision # See SD0001819, CGS System Description, Vol. 6 Chap. 10, Power Range Comments / Reference Neutron Monitor Proposed references to be provided during examination: None Learning Objective: 13709 - Identify the normal, alternate, and or emergency power supplies for major PRNM system components.

Question Source: Bank # LO03310 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD001819 Revision: 2 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/1/2017 Tier 2 Group 1 K/A 217000.K3.04 Level of Difficulty: 3 Importance Rating 3.6 Reactor Core Isolation Cooling System (RCIC): Knowledge of the effect that a loss or malfunction of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) will have on following: Adequate core cooling Question # 39 Current plant conditions:

A reactor scram was initiated due to a LOCA

Reactor Core Isolation Cooling (RCIC) is the only system available for RPV injection.

RCIC is injecting into the RPV.

RPV pressure is 850 psig.

Emergency Depressurization (ED) has been initiated.

When should emergency depressurization be terminated?

ED should be terminated when RPV pressure is A. approximately 10 psig below DW pressure to minimize the radiological release from primary containment to the environment.

B. within 40 psig of WW pressure to ensure continued rate of energy addition is within capacity of primary containment vent path.

C. LT 125 psig to maintain core cooling while transitioning to shutdown cooling using the RHR system.

D. between 175 psig and 300 psig until RCIC operation is no longer needed to maintain adequate core cooling.

Answer: D Page 1 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This question requires that the candidate demonstrate the understanding that maintaining adequate core cooling by maintaining RCIC injection, when RCIC is the only RPV injection source, during a situation that requires emergency depressurization.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible because RPV pressure below drywell pressure would minimize release.

Incorrect because this is not the reason ED is terminated.

B. Incorrect. Plausible because WW pressure can be a concern. Incorrect because this is not the reason ED is terminated.

C. Incorrect. Plausible because 125 psig is the highest pressure in which SDC can be placed in service. Incorrect because this is not the reason ED is terminated.

D. Correct. As stated in Ref. B, section 4.2.4, when RCIC is the only injection source and emergency depressurization is required, the depressurization is stopped when RPV pressure is between 175 psig and 300 psig. This is accomplished to maintain RCIC flow and restore RPV level to greater than TOP of Active Fuel (TAF), which is -161 inches. If RCIC flow is not maintained, loss of adequate core cooling will occur (see Ref. A, 4.2.1 for definition of adequate core cooling).

Technical Reference(s)

PPM 5.0.10, Flowchart Training Manual Attached w/ Revision # See OI-15, EOP and EAL Clarifications Comments / Reference Proposed references to be provided during examination: None Learning Objective: 11229 - Analyze plant conditions and determine the bases for prioritizing emergency procedure implementation during emergency operations.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.0.10, section 4.2.1, definition of Revision: Major 021 Minor 001 Adequate Core Cooling Page 3 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.0.10, section 8.4.1, Purpose of Revision: Major 021 Minor 001 emergency depressurization Page 4 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.0.10, section 8.4.2, Overview of Revision: Major 021 Minor 001 emergency depressurization strategy Page 5 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.0.10, section 8.4.3.g.1, Definition Revision: Major 021 Minor 001 of Decay Heat Removal Pressure (DHRP)

Page 6 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

OI-15, section 4.2.4, RPV Pressure Control Revision: Major 025 Page 7 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/7/2017 Tier 2 Group 1 K/A 217000.A2.02 Level of Difficulty: 3 Importance Rating 3.8 Reactor Core Isolation Cooling System (RCIC): Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Turbine Trips Question # 40 CGS is operating in Mode 1.

A reactor scram was initiated due to a LOCA.

The HPCS system is out of service.

RCIC is in operation and injecting into the RPV.

RPV pressure is 960 psig, down slow.

MSIVs are closed.

RPV level is -155 inches, up slow.

The following annunciator alarm is received:

H13.P601.A4.2-8, RCIC TURBINE EXHAUST PRESS HIGH What is the impact on the RCIC system?

The RCIC-P-1 turbine is A. in operation. The RCIC turbine must be tripped.

B. tripped. The RCIC turbine must remain tripped.

C. in operation. The RCIC system should remain in service.

D. tripped. The RCIC system should be restored to service.

Answer: D Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question requires the candidate to demonstrate knowledge of the effects of a RCIC turbine trip due to high backpressure, and the procedural guidance to bypass the trip to maintain the RCIC system in operation for the conditions given in the stem.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible if it is believed that it one RCIC exhaust pressure switch will give the alarm while both switches are required to automatically trip the RCIC turbine; and that the RCIC turbine should be manually tripped since one pressure switch is above the limit. However, either pressure switch (RCIC-PS-9A/B) will actuate the alarm and cause the RCIC turbine to trip.

B. Incorrect. Plausible since the RCIC turbine will trip for the given conditions. However, for the conditions given in the stem, the RCIC high exhaust pressure trip should be overridden using PPM 5.5.5, since RCIC is currently the only system capable of maintaining RPV level.

C. Incorrect. Plausible if it is believed that it one pressure switch will give the alarm while both switches are required to automatically trip the RCIC turbine. Plausibility is enhanced since, for the conditions given in the stem, the RCIC system should be restored to service. However, either RCIC exhaust pressure switch (RCIC-PS-9A/B) will actuate the alarm and cause the RCIC turbine to trip.

D. Correct. Either RCIC exhaust pressure switch (RCIC-PS-9A/B) will actuate the alarm and cause the RCIC turbine to trip at a pressure of 25 psig. With the conditions given in the stem, PPM 5.1.1, table 1 allows the RCIC turbine exhaust pressure trip to be overridden and the RCIC system restored to service when needed. The stem conditions establish that RCIC is the only available system capable of restoring RPV level.

Technical Reference(s) 4.601.A4, 601.A4 Annunciator Panel Alarms Attached w/ Revision # See PPM 5.1.1 RPV Control Comments / Reference PPM 5.5.5, Overriding RCIC Low RPV Pressure Isolation and High Exhaust Pressure Isolation Proposed references to be provided during examination: None Learning Objective: 11671 - From memory, draw a simplified diagram of the RCIC system showing all major components and flow paths. (g) rupture discs Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.5 55.43 Comments /

Reference:

4.601.A4, 601.A4 Annunciator Panel Revision: 39 mr 2 Alarms, alarm 2-8 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.5.5 Revision: 007 Comments /

Reference:

PPM 5.1.1 Revision: 21 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 2 Group 1 K/A 218000.K4.03 Level of Difficulty: 3 Importance Rating 3.8 Automatic Depressurization System: Knowledge of AUTOMATIC DEPRESSURIZATION SYSTEM design feature(s) and/or interlocks which provide for the following: ADS logic control Question # 41 Given the following:

A loss of coolant accident (LOCA) has occurred concurrent with a loss of offsite power.

Both diesel generator #1 (DG1) and diesel generator #2 (DG2) have tripped and cannot be re-started.

RPV level has been lowering at a constant rate of 30 inches/minute since the start of the LOCA.

Current RPV level is -120 inches.

Reactor core isolation cooling (RCIC) has tripped.

High pressure core spray (HPCS) is operating as expected and injecting into the vessel.

Which of the following describes the response of the automatic depressurization system (ADS) when vessel level reaches Level 1?

A. ADS timers will start and time out. ADS will then initiate.

B. ADS timers will not start. ADS will not initiate.

C. ADS timers will start and time out. ADS will not initiate.

D. ADS timers will already be timed-out when Level 1 is reached. ADS will initiate immediately upon reaching Level 1.

Answer: C Page 1 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question provides plant conditions and discriminates whether a candidate knows the ADS logic behavior based on those conditions.

SRO Only:

N/A Explanation:

With DG1 and DG2 out of service, no low pressure ECCS pumps are available so ADS initiation will not occur. The ADS timer, however, will still start when RPV level reaches level 1 (-129).

A. Incorrect. ADS will not initiation. Plausible because if low pressure ECCS pumps were available, ADS would initiate.

B. Incorrect. ADS timers WILL start and timeout. Plausible because ADS does not initiate when no low pressure ECCS pumps are running.

C. Correct.

D. Incorrect. ADS timers start when Level 1 is reached, not before. Plausible because ADS senses when water level is below BOTH level 3 and level 1 and the stem states that RPV level is below level 3.

Technical Reference(s) Attached w/ Revision # See SD000186, CGS System Description, Volume 7, Chapter 5, Automatic Comments / Reference Depressurization Proposed references to be provided during examination: None Learning Objective: 5070 - State the interlocks (conditions) that must be satisfied prior to automatic or manual initiation of ADS.

Question Source: Bank # LR00641 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000186 Revision: 12 Page 3 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 1 K/A 223002.K6.04 Level of Difficulty: 2 Importance Rating 3.3 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off: Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF:

Nuclear boiler instrumentation Question # 42 CGS is operating in Mode 1.

MS-LT-61A, WR RPV Level Transmitter, failed low.

P601.A12.2-4, NS4 GROUP 1 ISOLATION RPV LEVEL LOW (-129"), is LIT.

Subsequently, a LOCA occurred. The crew scrammed the reactor and entered PPM 3.3.1, Reactor Scram and PPM 5.1.1, RPV Control. Current plant conditions:

MSIV ISOL SYS A/B LOW RPV LVL / HI STM TUNNEL TEMP BYPASS switches are NOT in BYPASS

Wide Range RPV Level indication:

MS-LIS-200A: Downscale MS-LIS-200B: -115 inches MS-LIS-200C: -132 inches MS-LIS-200D: -125 inches What is the condition of the MSIV inboard and outboard isolation valves?

MSIV Inboard Isolation valves, MS-V-22A-D, are (1) and MSIV Outboard Isolation valves, MS-V-28A-D, are (2) .

A. (1) OPEN (2) OPEN B. (1) CLOSED (2) OPEN C. (1) OPEN (2) CLOSED D. (1) CLOSED (2) CLOSED Answer: A Page 1 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This question evaluates the candidates knowledge of the effect on the NS4 system when one RPV level transmitter fails and an additional level transmitter actuates a NS4 isolation signal.

SRO Only:

N/A Explanation:

A. Correct. Low RPV level (-129 inches) on MS-LIS-200A/B/C/D results in a group 1 isolation (MSIVs). Group 1 isolation has a one out of two taken twice logic meaning that one transmitter in the A/C channels must be low and one transmitter in the B/D channels must be low for an actuation to occur. In this case, BOTH A and C channels are below the trip setpoint, but neither B or D are below the setpoint, so no actuation would occur.

B. Incorrect. Plausible since isolation logic for other groups employ a 2 of 2 taken once scheme. With both A & C channels below the trip setpoint (-129 inches), isolation valves would close.

Additionally, it is plausible to incorrectly believe that MSIVs are division dependent like other isolation valves. Examples include Group 6: RHR-V-9 (SDC Supply Inboard Isolation) and RHR-V-8 (SDC Outboard Isolation); Group 7: RWCU-V-1 (RWCU Suction Inboard Isolation) and RWCU-V-4 (RWCU Suction Outboard Isolation). However, MSIVs are in isolation group 1, which employs a 1 of 2 taken twice logic scheme. Additionally, MSIVs are not division dependent. See distractor A explanation.

C. Incorrect. Plausible since isolation logic for other groups employ a 2 of 2 taken once scheme. With both A & C channels below the trip setpoint (-129 inches), isolation valves would close.

Additionally, it is plausible to incorrectly believe that MSIVs are division dependent like other isolation valves. Examples include Group 6: RHR-V-9 (SDC Supply Inboard Isolation) and RHR-V-8 (SDC Outboard Isolation); Group 7: RWCU-V-1 (RWCU Suction Inboard Isolation) and RWCU-V-4 (RWCU Suction Outboard Isolation). However, MSIVs are in isolation group 1, which employs a 1 of 2 taken twice logic scheme. Additionally, MSIVs are not division dependent. See distractor A explanation.

D. Incorrect. Plausible since isolation logic for other groups employ a 2 of 2 taken once scheme. With both A & C channels below the trip setpoint (-129 inches), an isolation valve would close.

Additionally, since all MSIVs close together, its plausible to believe that both valves would be closed. However, MSIVs are in isolation group 1, which employs a 1 of 2 taken twice logic scheme.

See distractor A explanation.

Technical Reference(s)

SD000173, CGS System Description, Volume 8, Chapter 5, Nuclear Steam Supply Shutoff System (NS4)

Attached w/ Revision # See SD000126, CGS System Description, Volume 4, Chapter 2, Nuclear Comments / Reference Boiler Instrumentation (NBI)

PPM 4.601.A12, Annunciator Panel Alarms, Annunciator 2-4

Response

Proposed references to be provided during examination: None Learning Objective: 5596 - Describe the isolation logic used by the NS4 system for MSIV isolation Page 2 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 and Group 3 and 4.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.9 55.43 Comments /

Reference:

SD000173 Revision: 14 mr2 Page 3 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 4 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000126 Revision: 13 mr 1 Page 5 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

P601.A12.2-4 Annunciator Response Revision: Major 026 Minor 001 Page 6 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 1 K/A 239002.K5.05 Level of Difficulty: 3 Importance Rating 2.6 Relief/Safety Valves: Knowledge of the operational implications of the following concepts as they apply to RELIEF/SAFETY VALVES : Discharge line quencher operation Question # 43 Which of the following describes the reason for Emergency Depressurization with a high Suppression Pool Level?

A. SRV discharge would result in exceeding code allowable stresses which could cause a failure of the SRV tailpipes.

B. A large LOCA would exceed the Heat Capacity Temperature Limit resulting in the failure of the wetwell/drywell interface.

C. A large LOCA would result in exceeding the SRV Tail Pipe Level limit and exceed the Primary Containment Pressure Limit.

D. SRV discharge would cause excessive containment pressure on the drywell floor and exceed the Primary Containment Pressure Limit.

Answer: A Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question determines if candidate understands the reason SRVs are opened on high WW level which directly ties to the design and capabilities of the quenchers.

SRO Only:

N/A Explanation:

SRVs are opened to Emergency Depressurize the reactor prior to wetwell level being too high (SRVTPLL) and challenging the quenchers/tailpipes.

A. Correct.

B. Incorrect. The reason/concern related to high wetwell level is NOT exceeding HCTL. Plausible since HCTL is related to wetwell temperature.

C. Incorrect. This condition does not challenge PCPL. Plausible since distractor includes SRVTPLL.

D. Incorrect. Pressure on the drywell floor is not the concern. Plausible since the wetwell performs the quenching function to minimize the impact on primary containment.

Technical Reference(s) Attached w/ Revision # See PPM 5.0.10, Flowchart Training Manual Comments / Reference Proposed references to be provided during examination: None Learning Objective: 8387 - Given a list, identify the statement that describes the reason for emergency depressurizing the RPV if wetwell level and reactor pressure cannot be restored and maintained below SRVTPLL.

Question Source: Bank # LO00144 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

5.0.10 Revision: 21 mr 1 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 1 K/A 259002.A1.04 Level of Difficulty: 2 Importance Rating 3.6 Reactor Water Level Control System: Ability to predict and/or monitor changes in parameters associated with operating the REACTOR WATER LEVEL CONTROL SYSTEM controls including: Reactor water level control controller indications Question # 44 RFW-P-1B is being placed in service as the second Reactor Feed Pump per SOP-RFT-START, following a pump trip. Feedwater Level Control is aligned as follows:

RFW-P-1A is in AUTO.

RFW-P-1B is in MDEM with RFW-V-102B (Pump Discharge Valve) open. RFW-P-1B discharge pressure is at shutoff head.

RPV level is being maintained by RFW-LIC-600 (RPV Master Level Controller) in AUTO If CRO1 depresses the UP arrow for RFW-P-1B on RFT-COMP-1, RFW-P-1B speed will A. rise and RFW-P-1A speed will remain the same.

B. rise and RFW-P-1A speed will rise.

C. rise and RFW-P-1A speed will lower.

D. remain the same and RFW-P-1A speed will remain the same.

Answer: C Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question determines if candidate can predict changes in the feedwater level control system controllers following a feedwater pump speed adjustment.

SRO Only:

N/A Explanation:

A. Incorrect: If RFW-V-102B were not open, this distractor would be correct. RFW-V-102B is opened when RFW-P-1B discharge pressure is within 20 to 30 psi of RFW-P-1A discharge pressure. As a result, once speed is raised, RFW-P-1B will feed the reactor and raising the speed of one pump will cause the speed of the other pump to lower.

B. Incorrect: See A. The UP arrows on RFT-COMP-1 control feed turbine speeds individually. The response in this distractor would be expected if the INC button on RFW-LIC-600 was depressed with both pumps in AUTO.

C. Correct: With the conditions provided in the stem, depressing the UP arrow for RFW-P-1B will cause the B feed turbine speed to rise. This will result in an increase in feed flow, which the FWLC system will detect and respond to by lowering the speed of RFW-P-1A.

D. Incorrect: See A. The UP arrow will not raise the speed of a feed turbine if the turbine has not been reset following a trip. The speed of the other feed turbine would also remain unchanged in that situation.

Technical Reference(s) Attached w/ Revision # See SD000157, CGS System Description, Volume 4, Chapter 3, Feedwater Comments / Reference Level Control Proposed references to be provided during examination: None Learning Objective: 5394 - Describe the function of each of the following controls and how they relate to each other: (a)Turbine Speed Controllers Question Source: Bank # LO02781 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.5 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000157 Revision: 16 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 1 K/A 261000.A2.03 Level of Difficulty: 3 Importance Rating 2.9 Standby Gas Treatment System: Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High train temperature Question # 45 CGS is operating in Mode 1

A LOCA occurred.

Both Standby Gas Treatment (SGT) trains automatically started during the event.

SGT System A has been placed in standby, SGT System B is running.

The following annunciators have just alarmed:

H13-P811.K2-2.4, CHARCOAL FLTR B-1 OUTLET TEMP HIGH H13-P811.K2-2.5, CHARCOAL FLTR B-2 OUTLET TEMP HIGH What actions should the crew take to address these alarms per ABN-SGT-TEMP/RAD, Standby Gas Treatment Charcoal High Temperature/Radiation?

The crew should A. start SGT System A, place SGT System B in standby, and initiate emergency deluge on SGT System B.

B. start SGT System A, place SGT System B in recirculation, and visually check for fire on SGT System B.

C. maintain SGT System A in standby, place SGT System B in standby and initiate emergency deluge on SGT System B.

D. maintain SGT System A in standby, place SGT System B in recirculation, and visually check for fire on SGT System B.

Answer: B Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question determines if candidates know the correct actions to take in the event of high temperature condition in the SGT system.

SRO Only:

N/A Explanation:

Per ABN-SGT-TEMP/RAD the major actions/strategy that should be taken is to place the running (high temp) train in recirculation and to start the standby train. ABN-SGT-TEMP/RAD describes that high temperatures can occur from decay of particulates following a LOCA and that operating the associated SGT fan will cool the charcoal bed. This is the reason that the effected train should remain in recirculation and not secured.

A. Incorrect. SGT Train B should be placed in recirculation, not standby. Plausible because all other actions in the distractor are appropriate and shutting down a train that is experiencing high temperatures would be appropriate in other plant systems and conditions.

B. Correct.

C. Incorrect. SGT Train B should be placed in recirculation, not standby. Plausible because securing systems that might be impacted by high temperatures is a correct action in other plant systems and conditions. Also, initiating the emergency deluge system would reduce temperatures, but is only allowed in a fire situation.

D. Incorrect. SGT Train A should be placed in service. Plausible because other portions of the distractor are correct.

Technical Reference(s) Attached w/ Revision # See ABN-SGT-TEMP/RAD, Standby Gas Treatment Charcoal High Comments / Reference Temperature/Radiation Proposed references to be provided during examination: None Learning Objective: 15809 - With procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-SGT-TEMP/RAD Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 10 CFR Part 55 Content: 55.41 41.5 55.43 Comments /

Reference:

ABN-SGT-TEMP/RAD Revision: 3 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-SGT-TEMP/RAD Revision: 3 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 2 Group 1 K/A 262001.A3.01 Level of Difficulty: 3 Importance Rating 3.1 A.C. Electrical Distribution: Ability to monitor automatic operations of the A.C. ELECTRICAL DISTRIBUTION including:

Breaker tripping Question # 46 CGS is operating in Mode 1.

A switchyard fault caused the concurrent loss of the main transformers and the startup transformer.

Which of the following is the expected lineup for bus SM-7 supply breakers?

A.

B.

C.

D.

Answer: A Page 1 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question determines if candidates can select the correct status of the electrical distribution system following a trip and loss of the main and startup transformer.

SRO Only:

N/A Explanation:

Bus SM-7 will experience an undervoltage condition which will cause an automatic transfer of power to the backup transformer (TR-B). In addition, the EDG will start, but will not connect to the bus due to the availability of TR-B.

A. Correct.

B. Incorrect. CB-DG1/7 would not close. Plausible because if TR-B was, unavailable this distractor would be correct.

C. Incorrect. DG 1 should start. Plausible because CB-B7 will close and supply power to SM-7.

D. Incorrect. Plausible if it is believed that the normal transformer (TR-N1/2) is powered from offsite sources. However, TR-N1/2 is lost when the main transformer is lost. With a concurrent loss of TR-S, SM-7 will be powered from TR-B.

Technical Reference(s) Attached w/ Revision # See SD000182, CGS System Description, Volume 1, Chapter 2, AC Comments / Reference Distribution Proposed references to be provided during examination: None Learning Objective: 5051 - Explain the design features and/or system interlocks or response which provide for the following: (e) SM-7(8) response to undervoltage, including load shedding.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000182 Revision: r19 mr0 Page 3 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/1/2017 Tier 2 Group 1 K/A 262002.A4.01 Level of Difficulty: 3 Importance Rating 2.8 Uninterruptable Power Supply (A.C./D.C.): Ability to manually operate and/or monitor in the control room: Transfer from alternative source to preferred source Question # 47 CGS is operating in Mode 1.

Efforts are underway to transfer IN-1 loads from the Alternate AC Source through the Maintenance Bypass Switch to the UPS Inverter through the Static Switch.

Concerning the transfer, which of the following is correct?

Power is being transferred to the UPS Inverter from A. MC-7F. It will be a make-before-break transfer.

B. MC-7A. It will be a make-before-break transfer.

C. MC-7F. It will be a break-before-make transfer.

D. MC-7A. It will be a break-before-make transfer.

Answer: A Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question directly asks the expected conditions/response when transferring IN-1 from the maintenance source back to the normal source.

SRO Only:

N/A Explanation:

A. Correct. The alternate source is powered by MC-7F. The maintenance source goes through the static switch which is a make-before-break connection.

B. Incorrect. Plausible since this transfer will be make-before-break. However, the alternate source of power to IN-1 is MC-7F C. Incorrect. Plausible since the normal source of power is MC-F. However, this will be a make-before-break transfer.

D. Incorrect. Plausible since the Bypass source to IN-1 is MC-7A and transferring to the Bypass source is a break-before-make transfer. However, the alternate power supply is MC-7F and this will be a make-before-break transfer.

Technical Reference(s) Attached w/ Revision # See SD000194, CGS System Description, Volume 1, Chapter 3, Comments / Reference Uninterruptible Power Supply System Proposed references to be provided during examination: None Learning Objective: 5896 - List the power supplies to each inverter: (a) E-IN-1 5891 - State the purpose and various functions of the following with respect to E-IN-1: (c) static switch, bypass switch Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000194 Revision: 13 mr0 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000194 Revision: 13 mr 0 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 3 Date: 2/10/2017 Tier 2 Group 1 K/A 263000.2.2.44 Level of Difficulty: 3 Importance Rating 4.0 D.C. Electrical Distribution: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Question # 48 CGS is operating in Mode 4.

RHR A is operating in Shutdown Cooling mode per SOP-RHR-SDC, RHR Shutdown Cooling.

DLO-P-2B1, DC Soakback Lube Oil Pump is currently running to collect vibration data.

Post maintenance testing is being conducted on TG-EOP-1, Main Turbine Emergency Oil Pump and the pump is currently running.

The following indications are observed:

Based on these indications, what action, if any, should be taken?

A. Secure pump TG-EOP-1.

B. Direct an EO to secure DLO-P-2B1.

C. Place RHR B in service and secure RHR A.

D. No actions are required based on these indications.

Answer: B Page 1 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question asks the candidate to evaluate control room indications related to the DC Distribution system, identify a ground condition exists on bus S1-2, related this condition to current plant operations and determine what actions to take to troubleshoot the ground.

SRO Only:

N/A Explanation:

In accordance with PPM 4.800.C5-8-2, Annunciator Panel Alarms for 125 VDC BATTERY B1-2 GND, a less than 25K can be considered a severe ground. Step 4 of the ARP directs that performance of ground isolation should occur by isolating selected loads. Based on plant conditions, the only plant activity that impacts 125VDC Bus S1-2 is the testing of DLO-P-2B1 and the pump should be secured.

A. Incorrect. Plausible because TG-EOP-1 is a DC pump and is powered by 250VDC, but securing this pump would not resolve the ground condition on 125VDC Bus S1-2.

B. Correct. DLO-P-2B1 is powered by S1-2. This is the only potential source of the ground from the conditions given in the stem.

C. Incorrect. Plausible because RHR-P-2A control power is from 125VDC Bus S1-1, however the ground condition is related to bus S1-2. Swapping to RHR-P-2B would not resolve the ground condition.

D. Incorrect. Plausible because no action would be taken if the ground were greater than 25K or if the candidate believes the indications given are normal. Actions should be taken, however, to isolate the ground.

Technical Reference(s)

A PPM 4.800.C5-8-2, Annunciator Response Procedure Attached w/ Revision # See SOP-ELEC-DC-LU, DC Electrical Distribution System Breaker B Comments / Reference Lineup Proposed references to be provided during examination: None Learning Objective:

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Page 2 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 10 CFR Part 55 Content: 55.41 41.5 55.43 PPM 4.800.C5-8-2, Annunciator Response Procedure Revision: 31 Page 3 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 SOP-ELEC-DC-LU Revision: 5 Page 4 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 SOP-ELEC-DC-LU Revision: 5 Page 5 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 SOP-ELEC-DC-LU Revision: 5 Page 6 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 SOP-ELEC-DC-LU Revision: 5 Page 7 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/7/2017 Tier 2 Group 1 K/A 263000.K5.01 Level of Difficulty: 3 Importance Rating 2.6 D.C. Electrical Distribution: Knowledge of the operational implications of the following concepts as they apply to D.C.

ELECTRICAL DISTRIBUTION: Hydrogen generation during battery charging Question # 49 CGS is operating in Mode 1. An equalizing battery charge is in progress on 250 vdc battery B2-1 when the following conditions are noted:

Annunciator 826.P1.3-4, BATTERY ROOM FAN 53A DIFFERENTIAL PRESSURE LOW, is in alarm

Battery Room No. 1 Exhaust Fan, WEA-FN-53A, is OFF.

Field operators report that the breaker for WEA-FN-53A is tripped and will not reset.

Battery Room No. 1 temperature is 82°F and steady.

What actions should the crew take and why?

The crew should secure the equalizing charge on battery B2-1 and A. install temporary ventilation in Battery Room No. 1 in accordance with ABN-HVAC, HVAC Trouble, to reduce hydrogen buildup.

B. remove battery B2-1 from service in accordance with SOP-ELEC-250VDC-SHUTDOWN, 250 vdc System Shutdown, to reduce hydrogen buildup.

C. install temporary ventilation in Battery Room No. 1 in accordance with ABN-HVAC, HVAC Trouble, to prevent battery electrolyte temperature from exceeding tech spec limits.

D. remove battery B2-1 from service in accordance with SOP-ELEC-250VDC-SHUTDOWN, 250 vdc System Shutdown, to prevent battery electrolyte temperature from affecting battery performance.

Answer: A Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This questions requires the candidates to demonstrate understanding of the method used to prevent hydrogen buildup during a battery charge and the operational requirements if ventilation is lost.

SRO Only:

N/A Explanation:

Per ABN-HVAC, if WMA-FN-53A is lost, portable fans are installed to direct ventilation to battery room 1.

A. Correct.

B. Incorrect. The battery should not be removed from service. Plausible because removing the battery from service would stop hydrogen generation.

C. Incorrect. Plausible since maintaining battery room ventilation will assist in lowering temperatures.

However, there is no high temperature specification for battery electrolyte. There is a low spec. of 60°F.

D. Incorrect. The battery should not be removed from service. Plausible because battery electrolyte temperatures are related to battery operation.

Technical Reference(s)

SOP-ELEC-250V-OPS, 250 vdc System Operations Attached w/ Revision # See Comments / Reference ABN-HVAC, HVAC Trouble TS Bases, LCO 3.8.6 Proposed references to be provided during examination: None Learning Objective: 15777 - With the procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-HVAC.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.5 55.43 Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-ELEC-250V-OPS Revision: 2 Comments /

Reference:

ABN-HVAC Revision: 13 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

TS Bases, LCO 3.8.6 Revision: 92 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 0 Date: 12/16/2016 Tier 2 Group 1 K/A 264000.K1.05 Level of Difficulty: 2 Importance Rating 3.2 Emergency Generators (Diesel/Jet): Knowledge of the physical connections and/or cause-effect relationships between EMERGENCY GENERATORS (DIESEL/JET) and the following: Emergency generator fuel oil supply system Question # 50 How is diesel fuel delivered to the Emergency Diesel Generator DG-1 and DG-2 fuel injectors during diesel generator operation?

A. An engine-driven main supply pump with a DC-powered backup pump.

B. An AC-powered main supply pump with a DC-powered backup pump.

C. A DC-powered main supply pump with an engine-driven backup pump.

D. An engine-driven main supply pump with an AC-powered backup pump.

Answer: A Page 1 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question determines whether a candidate understands what type of pump supplies fuel to the engine and how it is powered.

SRO Only:

N/A Explanation:

Per SD000200, the fuel is supplied by an engine driven pump during normal operation with a DC powered backup pump available as a backup.

A. Correct.

B. Incorrect. The primary supply pump is NOT AC powered.

C. Incorrect. The primary supply pump is NOT DC powered.

D. Incorrect. The backup pump is NOT AC powered.

Distractors are plausible because the Fuel Oil Transfer Pumps are AC powered and distractors are variations of the possible combinations.

Technical Reference(s) Attached w/ Revision # See SD000200, CGS System Description, Volume 7, Chapter 8, Diesel Comments / Reference Generator Proposed references to be provided during examination: None Learning Objective: 12403 - Explain the function and operation of the following components, including any interlocks or automatic features associated with them: (f) Fuel Oil Storage and Transfer System Question Source: Bank # EO00573 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.2 55.43 Page 2 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000200 (pages 12 and 13 Revision: 12 mr 1 Page 3 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 1 K/A 300000.K2.01 Level of Difficulty: 2 Importance Rating 2.8 Instrument Air System (IAS): Knowledge of electrical power supplies to the following: Instrument air compressor Question # 51 What is the power supply to air compressor CAS-C-1B?

A. MC-7A B. MC-8A C. MC-6C D. MC-6B Answer: B Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The candidate must recall the power supply to the CAS compressors.

SRO Only:

N/A Explanation:

CAS-C-1B is powered by MC-8A A. Incorrect. CAS-C-1B is powered by MC-8A. Plausible because CAS-C-1A is powered by MC-7A.

B. Correct.

C. Incorrect. CAS-C-1B is powered by MC-8A. Plausible because service air compressor SA-C-1 is powered by MC-6C.

D. Incorrect. CAS-C-1B is powered by MC-8A. Plausible because service air dryer SA-DY-1 is powered by MC-6B.

Technical Reference(s) Attached w/ Revision # See SD000205, CGS System Description, Volume 2, Chapter 11, Comments / Reference Control and Service Air Cas System Description Proposed references to be provided during examination: None Learning Objective: 5881 - Given a list of various plant systems, describe their interrelationship with the Control Air and Service Air systems. (a) AC Power Distribution Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000205 Revision: 11 mr 1 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 0 Date: 12/16/2016 Tier 2 Group 1 K/A 300000.K6.07 Level of Difficulty: 3 Importance Rating 2.5 Instrument Air System (IAS): Knowledge of the effect that a loss or malfunction of the following will have on the INSTRUMENT AIR SYSTEM: Valves Question # 52 CGS is operating in Mode 1.

During a valve lineup, field personnel erroneously operated the CN to CIA crosstie isolation valve, CIA-V-728, in the close direction.

The error was detected prior to fully closing the valve and CIA-V-728 was restored to the full open position.

The pressure transient on the Containment Instrument Air (CIA) header is denoted in the following table:

Time Pressure (minutes) (psig) 0 185 1 150 2 100 3 90 4 145 5 185 What is the current status of the CIA system?

Inboard MSIVs are (1) , ADS Header Supply valves, CIA-V-39A & B (2) .

A. (1) closed (2) remained open B. (1) closed (2) closed and reopened C. (1) open (2) remained open D. (1) open (2) closed and reopened Answer: D Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This question requires the candidates to understand the effects on the instrument air system when the nitrogen supply valve is shut.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since MSIVs will close when CIA pressure drops to 50-80 psig. However, CIA pressure only reached 90 psig. Additionally, ADS Header Supply valves, CIA-V-39A & B, will close if CIA pressure drops below 160 psig for 3 minutes, and then reopen when CIA header pressure is

> 160 psig.

B. Incorrect. Plausible since the ADS Header Supply valves, CIA-V-39A & B, will close if CIA pressure drops below 160 psig for 3 minutes, and then reopen when CIA header pressure is > 160 psig. However, MSIVs will close when CIA pressure drops to 50-80 psig. CIA pressure only reached 90 psig and the MSIVs will remain open.

C. Incorrect. Plausible since MSIVs will remain open if CIA pressure remains above 50-80 psig.

However, ADS Header Supply valves, CIA-V-39A & B, will close if CIA pressure drops below 160 psig for 3 minutes, and then reopen when CIA header pressure is > 160 psig.

D. Correct. MSIVs will remain open if CIA pressure remains above 50-80 psig. ADS Header Supply valves, CIA-V-39A & B, will close if CIA pressure drops below 160 psig for 3 minutes, and then reopen when CIA header pressure is > 160 psig.

Technical Reference(s) Attached w/ Revision # See SD000156, CGS System Description, Vol. 8, Chap. 4, Containment Comments / Reference Instrument Air (CIA)

ABN-CIA, Containment Instrument Air System Failure Proposed references to be provided during examination: None Learning Objective: 11755 - Describe the function, purpose and design features of the following Containment Instrument Air System: (e) Nitrogen Bank Bottles.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

CGS System Description, CIA Revision: 11 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-CIA Revision: Major 006 Minor 004 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/2/2017 Tier 2 Group 1 K/A 400000.K3.01 Level of Difficulty: 2 Importance Rating 2.9 Component Cooling Water System (CCWS): Knowledge of the effect that a loss or malfunction of the CCWS will have on the following: Loads cooled by CCWS Question # 53 Columbia is operating in MODE 1.

Annunciator 825.N.3-4, REACTOR CLOSED COOLING SURGE TANK LEVEL HIGH, is in alarm.

RCC surge tank level is 4.6 feet, up slow.

RCC radiation level is 65 cps, up slow.

What is the most likely cause for the change in these parameters?

There is a leak in A. the Reactor Building Equipment Drain Heat Exchanger, EDR-HX-2.

B. a Control Rod Drive pump motor cooler.

C. a Reactor Water Cleanup pump motor cooler.

D. a Reactor Recirculation pump shaft seal jacket cooler.

Answer: D Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question presents a malfunction in the RCC (CCWS) system and determines if candidate can properly evaluate the expected system response.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since the EDR system is radioactive and a leak into RCC from this system could cause RCC radiation levels to rise. However, the cooling medium on the shell side of the HX is low pressure refrigerant. Therefore, a leak in this component would be out of the RCC system.

B. Incorrect. Plausible since a leak into RCC from the CRD pumps would cause RCC radiation levels to rise. However, CRD system pressure is below RCC pressure and a leak would be out of the RCC system.

C. Incorrect. Plausible since a leak into the RCC from the RWCU pump coolers would cause RCC radiation levels and surge tank level to rise. However, since the RWCU pump motor coolers have high-pressure primary water on their shell-sides and RCC on their tube- sides, these components are less likely to leak water into the RCC system.

D. Correct. The RRC pump seal has two coolers: a jacket cooler and a cooler for the water recirculated by the auxiliary impeller on the RRC pump shaft. This water is at reactor pressure and could leak into the RCC system. Additionally, a leak from the RRC shaft seal cooler would cause RCC radiation levels to rise.

Technical Reference(s) Attached w/ Revision # See SD000196, CGS System Description, Volume 3, Chapter 1, Reactor Comments / Reference Closed Cooling Water Proposed references to be provided during examination: None Learning Objective: 7669 - Predict the possible sources of leakage into the RCC system and how the sources may be identified.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 10 CFR Part 55 Content: 55.41 41.7 55.43 Comments /

Reference:

SD000196 Revision: 14 mr 1 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 2 K/A 201002.A1.03 Level of Difficulty: 3 Importance Rating 3.0 Reactor Manual Control System: Ability to predict and/or monitor changes in parameters associated with operating the REACTOR MANUAL CONTROL SYSTEM controls including: Rod movement sequence lights.

Question # 54 Given the following:

A control rod is inserted using the Continuous Insert Pushbutton.

The ROD DRIFT annunciator on H13-P603 alarms after releasing the Continuous Insert Pushbutton.

The red DRIFT light associated with the selected control rod is lit.

Which of the following caused these indications?

A. RMCS has detected a reed switch that failed to close at some point during the requested control rod motion.

B. Control Rod Drive Hydraulic Cooling Water flow is set too low allowing the selected control rod to insert past the requested position.

C. Control Rod Drive Hydraulic Drive water d/p is set too high causing the control rod to settle after the RMCS rod motion timer has timed out.

D. The Continuous Insert Pushbutton bypasses the settle function and an odd reed switch was made up before the rod settled.

Answer: D Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question presents a situation where rods were inserted and provides indications. The question determines if the candidate understands the associated indications (ability to monitor).

SRO Only:

N/A Explanation:

The rod settle function is bypassed when using the Continuous Insert Pushbutton. If the pushbutton is released when the rod is near an odd reed switch, the drift light will be activated.

A. Incorrect. A reed switch position failure would not result in a rod drift alarm. Plausible because a rod drift alarm is caused by a rod passing an odd reed switch when the rod is not in the driving cycle.

B. Incorrect. CRD flow being too low does not cause a rod drift condition. Plausible because CRD flow being too high will cause a rod drift condition.

C. Incorrect. High water D/P does not cause a rod drift condition. Plausible since CRD water is kept at a higher pressure then RPV pressure.

D. Correct.

Technical Reference(s) Attached w/ Revision # See PPM 4.603.A7, Annunciator Response Comments / Reference SD000148, CGS System Description, Volume 5, Chapter 6, Reactor Manual Control Proposed references to be provided during examination: None Learning Objective: 5792 - State the functions and interrelationships of these P603 controls: (a)

Insert Pushbutton (b) Continuous Insert Pushbutton Question Source: Bank # LO02183 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.6 55.43 Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

4.603.A7 Revision: 51 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000148 pages 11 and 12 Revision: 14 mr 1 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000148 page 17 Revision: 14 mr 1 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 2 K/A 201003.K5.08 Level of Difficulty: 3 Importance Rating 3.1 Control Rod and Drive Mechanism: Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD AND DRIVE MECHANISM: How control rods affect shutdown margin.

Question # 55 Which of the following conditions satisfies the Maximum Subcritical Banked Withdrawal Position, to ensure that there is sufficient SDM to keep the reactor shutdown under all conditions?

A. Two control rods at 08, all other control rods at 00.

B. One control rod at 08, one control rod at 04, all other control rods at 02.

C. One control rod at 48, one control rod at 04, all other control rods at 00.

D. One control rod at 48, two control rods at 02, all other control rods at 00.

Answer: D Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question presents a situation where not all rods are inserted fully following a SCRAM and determines if candidate can identify which conditions would result in entry into the ATWS EOP (due to inadequate shutdown margin).

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since most control rods are below position 2. However, two control rods are greater than position 2 and Maximum Subcritical Bank Withdrawal Position (MSBWP) is not met.

B. Incorrect. Plausible since most control rods are at position 2. However, two control rods are greater than position 2 and Maximum Subcritical Bank Withdrawal Position (MSBWP) is not met.

C. Incorrect. Plausible since one control rod is full out. However, multiple rods are greater than position 2 and Maximum Subcritical Bank Withdrawal Position (MSBWP) is not met.

D. Correct. This meets the criteria for assuring the reactor is shutdown with control rod position alone:

Maximum Subcritical Banked Withdrawal Position (MSBWP): No more than one control rod greater than Notch 02 and all other rods at Notch 02 or less.

Technical Reference(s) Attached w/ Revision # See PPM 5.0.10, Flowchart Training Manual Comments / Reference Proposed references to be provided during examination: None Learning Objective: 8182: Given a list, identify the criteria that must be met to ensure that the reactor is shutdown with no boron injected.

Question Source: Bank # LO01786 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.6 55.43 Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.0.10, Flowchart Training Manual Revision: Major 021 Minor 001 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 2 K/A 202002.K6.05 Level of Difficulty: 3 Importance Rating 3.1 Recirculation Flow Control System: Knowledge of the effect that a loss or malfunction of the following will have on the RECIRCULATION FLOW CONTROL SYSTEM: Reactor water level Question # 56 CGS is operating in Mode 1 with both Reactor Recirculation (RRC) pumps at 60 Hz when the following alarm was received:

P602.A6.5-1, LOOP A ASD CHANNEL FAILURE LIMIT Approximately 1 minute later, Reactor Feed Pump Turbine (RFPT) 1A tripped. Plant conditions 1 minute after the RFPT trip include:

Annunciator P603.A8.3-7, REACTOR PRESSURE VESSEL LEVEL HIGH/LOW ALERT, in alarm.

RPV level: 30.5 inches up slow.

Lowest RPV level reached: 26.0 inches.

What is the current operating frequency for both RRC pumps?

A. RRC-P-1A: 15 Hz RRC-P-1B: 15 Hz B. RRC-P-1A: 30 Hz RRC-P-1B: 30 Hz C. RRC-P-1A: 51 Hz RRC-P-1B: 15 Hz D. RRC-P-1A: 51 Hz RRC-P-1B: 30 Hz Answer: B Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question requires the candidate to understand the loss of actual reactor water level affects the Reactor Recirculation (RRC) system. A malfunction of a single RPV level channel would not affect the RRC system since the output of the failed channel would be substituted with a functioning channels output automatically.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since the Low RPV level runback (-13 inches, Ref. A, section V.C.1.e.(3)) will reduce both RRC pumps to 15 Hz. This could occur during a reactor scram. Information in the stem indicates that a reactor scram did not occur since RPV level on a reactor scram goes below 0 inches prior to recovery.

B. Correct. A Reactor Feed Pump Turbine (RFPT) trip along with RPV level below the Low Level Alert setpoint (31.5 inches, Ref. C) will cause both Reactor Recirculation (RRC) pumps to runback to 30 Hz (see Ref. A, section V.C.1.e.(2)). Although RRC-P-1A was operating in Manual at a reduced frequency of 51Hz due to an earlier loss of ASD Channel runback, the pump will runback to 30 Hz when the Loss of RFPT runback initiates.

C. Incorrect. Plausible since RRC-P-1A did initially runback to 51Hz due to the loss of an ASD channel (Ref. B). See distractor A plausibility statement above for further information. However, the RFPT runback will reduce BOTH RRC pump operating frequencies to 30 Hz (see Ref. A, section V.C.1.e.(2)).

D. Incorrect. Plausible since RRC-P-1A did initially runback to 51Hz due to the loss of an ASD channel (Ref. B). However, the RFPT runback will reduce BOTH RRC pump operating frequencies to 30 Hz (see Ref. A, section V.C.1.e.(2)).

Technical Reference(s)

SD000184, CGS System Description, Volume 5, Chapter 5, Reactor Recirculation Flow Control (ASD)

Attached w/ Revision # See PPM 4.602.A6, Annunciator Panel Alarms, Window 5-1, LOOP A ASD Comments / Reference CHANNEL FAILURE LIMIT PPM 4.603.A8, Annunciator Panel Alarms, Window 3-7, REACTOR PRESSURE VESSEL LEVEL HIGH/LOW ALERT Proposed references to be provided during examination: None Learning Objective: 5022 - Describe the physical and/or cause-and-effect relationship between the RRC system and the following: (k) Reactor water level Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Comments /

Reference:

SD000184,Section V.C.1, Runbacks Revision: Major 19, Minor 001 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 4.602.A6, Annunciator Panel Alarms, Revision: 32 Window 5-1, LOOP A ASD CHANNEL FAILURE LIMIT Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 4.603.A8, Annunciator Panel Alarms, Revision: 36 Window 3-7, REACTOR PRESSURE VESSEL LEVEL HIGH/LOW ALERT Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/7/2017 Tier 2 Group 2 K/A 204000.K4.02 Level of Difficulty: 3 Importance Rating 2.7 Reactor Water Cleanup System: Knowledge of REACTOR WATER CLEANUP SYSTEM design feature(s) and/or interlocks which provide for the following: Piping over-pressurization protection Question # 57 During a plant shutdown and cooldown per PPM 3.2.1, under which of the following operating conditions may RWCU-V-31, Blowdown Orifice Bypass Valve FIRST be opened?

A. When MODE 4 is entered.

B. After RHR is placed in shutdown cooling.

C. At any reactor pressure when using RWCU-MOV-34 (Discharge to Main Condenser)

D. When RPV pressure is LE 125 psig.

Answer: D Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question ensures candidate understands the design limitation of the RWCU system with regard to RWCU-V-31 to prevent piping over pressurization.

SRO Only:

N/A Explanation:

Per PPM 3.1.2 and the system description, RWCU-V-31 should not be open above 125psig RPV pressure to avoid over pressurization.

A. Incorrect. RPV pressure must be below 125psig. Plausible because MODE 4 is a shutdown condition where temperature is below 200F.

B. Incorrect. Plausible since the RHR SDC permissive is satisfied at approximately 125 psig.

However, RHR cannot be placed in SDC until RPV pressure is LE 48 psig, which is below the 125 psig limit for opening RWCU-V-31.

C. Incorrect. Plausible because RWCU-V-34 is in the same flowpath as RWCU-V-31 and would impact flow rates and discharge location.

D. Correct.

Technical Reference(s)

PPM 3.1.2, Normal Plant Startup Attached w/ Revision # See SD000190, CGS System Description, Volume 4, Chapter 4, Reactor Comments / Reference Water Cleanup PPM 3.2.1, Normal Plant Shutdown Proposed references to be provided during examination: None Learning Objective: 5034 - State the function of the following components: (c) Orifice Bypass Valve (RWCU-V-31)

Question Source: Bank # LO03309 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 3.1.2 Revision: 81 mr 2 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000190 (page 5) Revision: 14 mr 1 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 3.2.1, Normal Plant Shutdown Revision: 83 mr 1 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 0 Date: 12/16/2016 Tier 2 Group 2 K/A 215001.A2.01 Level of Difficulty: 3 Importance Rating 2.7 Traversing In-Core Probe: Ability to (a) predict the impacts of the following on the TRAVERSING IN-CORE PROBE ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low reactor water level: Mark-I&II(Not-BWR1)

Question # 58 CGS is operating in Mode 1 with a Traversing In-Core Probe (TIP) trace in progress.

A transient occurs.

RPV level: -65 inches, up slow.

Drywell pressure: 1.45 psig, down slow.

Highest drywell pressure: 1.55 psig.

Current condition of the TIP system:

How should the crew respond to these indications?

A. Place the MODE switch in MAN and retract the TIP probe in manual mode. TIP ball valve, TIP-V-1, will automatically close.

B. Press the AUTO START pushbutton to automatically withdraw the TIP probe and take the Shear Valve Control keyswitch on Valve Control Channel A to Fire.

Page 1 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 C. Verify that the TIP probe automatically retracted and take the Shear Valve Control keyswitch on Valve Control Channel A to Fire.

D. Use the manual hand crank to retract the TIP probe. TIP ball valve, TIP-V-1, will automatically close.

Answer: C K/A Match:

This question requires the candidate to demonstrate understanding of the expected response of the Traversing In-Core Probe system (TIPS) to a low RPV level and the actions required if TIPS does not respond as expected.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since this is a possible action if the probe did not automatically retract.

However, the IN-SHIELD light lit on Control Drawer A indicates that all probes are in the in-shield position and the ball valve, TIP-V-1, failed to automatically shut.

B. Incorrect. Plausible since it is necessary to activate the shear valve, TIP-V-7, to isolate the TIP system. However, the IN-SHIELD light lit on Control Drawer A indicates that all probes are in the in-shield position and no action is necessary to withdraw the probe.

C. Correct. The IN-SHIELD light lit on Control Drawer A indicates that all probes are in the in-shield position, but the ball valve, TIP-V-1, failed to automatically shut. It is necessary to activate the shear valve, TIP-V-7, to isolate the TIP system.

D. Incorrect. Plausible since this is a possible action if the probe did not automatically retract or could not be retracted electrically. However, the IN-SHIELD light lit on Control Drawer A indicates that all probes are in the in-shield position and the ball valve, TIP-V-1, failed to automatically shut.

Technical Reference(s) Attached w/ Revision # See ABN-TIPS, TIP System Failure to Isolate Comments / Reference SD000155, CGS System Description, Vol 6, Chapter 7, Traversing In-Core Probe (TIP).

Proposed references to be provided during examination: None Learning Objective: 6989 - Explain the TIP system response to an FA (LOCA) signal.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Page 2 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.5 55.43 Comments /

Reference:

TIP System Description Revision: 13 Page 3 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 4 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 5 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-TIPS Revision: Major 002 Minor 003 Page 6 of 6

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/7/2017 Tier 2 Group 2 K/A 215002.A3.04 Level of Difficulty: 3 Importance Rating 3.6 Rod Block Monitor System: Ability to monitor automatic operations of the ROD BLOCK MONITOR SYSTEM including:

Verification of proper functioning/ operability: BWR-3,4,5 Question # 59 The reactor is operating in Mode 1 with the following conditions:

APRM power levels:

APRM 1: 24%

APRM 2: Bypassed APRM 3: 31%

APRM 4: 30%

A center control rod is selected on the Rod Select Matrix. Selected rod position is 32.

What is the status of the Rod Block Monitors (RBM)?

A. RBM A is NORMAL and RBM B is BYPASSED.

B. RBM A is NORMAL and RBM B is NORMAL.

C. RBM A is BYPASSED and RBM B is BYPASSED.

D. RBM A is BYPASSED and RBM B is NORMAL.

Answer: D Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question requires demonstrating knowledge of Rod Block Monitor (RBM) automatic actions when Average Power Range Monitor (APRM) inputs are below the Low Power Setpoint (LPSP) or bypassed.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible if student does not understand where the LPSP is derived.

B. Incorrect. Plausible if student does not understand where the LPSP is derived or the LPSP value.

C. Incorrect. Plausible if student does not understand where the LPSP is derived or that RBM is not bypassed when a center rod is selected.

D. Correct. RBM A derives Simulate Thermal Power (STP) from APRM 1. Since APRM 1 is below the Low Power Setpoint (LPSP) of 26%, RBM A is bypassed. RBM B derives Simulate Thermal Power (STP) from APRM 2. Since APRM 2 is above the Low Power Setpoint (LPSP) of 26%, RBM B is in normal operation.

Technical Reference(s)

SD001819, CGS System Description, Vol. 6, Chap. 10, Power Range Neutron Monitor Attached w/ Revision # See Comments / Reference CGS Licensee Controlled Specifications (LCS)

ISP-RBM-B301, RBM-CHS-A Calibration Proposed references to be provided during examination: None Learning Objective: 5699 Explain how the following systems interrelate with the RBM:

a. APRM System Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD001819 Revision: 2 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LCS 1.3.2.1, Trip Setpoints Revision: 92 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ISP-RBM-B301 Revision: 001 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/2/2017 Tier 2 Group 2 K/A 226001 A4.19 Level of Difficulty: 3 Importance Rating 3.4 RHR/LPCI: Containment Spray System Mode: Ability to manually operate and/or monitor in the control room: Drywell Temperature Question # 60 Columbia is operating at 100% power.

A large LOCA occurs.

RPV level lowers to -159 and is rising slowly.

RPV pressure is 200 psig and lowering.

Drywell temperature is 285°F, up slow.

The CRS directs lowering drywell temperature by spraying the drywell with RHR B.

What action needs to be taken before opening RHR-V-17B (Upper Drywell Spray Inboard Isolation Valve) and RHR-V-16B (Upper Drywell Spray Outboard Isolation Valve)?

A. STOP RRC Pumps, CLOSE RHR-V-42B (LPCI Injection).

B. STOP RRC Pumps, CLOSE RHR-V-24B (Suppression Pool Cooling/Test Return).

C. STOP ALL Drywell Cooling Fans, CLOSE RHR-V-42B (LPCI Injection).

D. STOP ALL Drywell Cooling Fans, CLOSE RHR-V-24B (Suppression Pool Cooling/Test Return).

Answer: C Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question ensures the candidate knows the actions needed to align RHR B for drywell spray which will lower and control drywell temperature.

SRO Only:

N/A Explanation:

Due to the low RPV level, RRC pumps have tripped and do not need to be secured. Drywell cooling flow (RCC) is lost as well, but drywell fans are still running and need to be secured. Based on conditions, RHR B would be injecting and RHR-V-42B will need to be closed.

A. Incorrect. RRC pumps are already secured. Plausible because PPM 5.2.1 directs securing RRC pumps prior to initiating spray. Plausible because RHR-V-42B needs to be closed.

B. Incorrect. RRC pumps are already secured. Plausible because PPM 5.2.1 directs securing RRC pumps prior to initiating spray. Plausible because RHR-V-24B would need to be closed if OPEN, but isnt OPEN in this condition.

C. Correct Answer.

D. Incorrect. RHR-V-24B does not need to be closed. Plausible because RHR-V-24B would need to be closed if it was open.

Technical Reference(s) Attached w/ Revision # See PPM 5.2.1, Primary Containment Control Comments / Reference SOP-RHR-SPRAY-DW-QC Proposed references to be provided during examination: None Learning Objective: 5774 - Describe the flow path within the appropriate RHR system for each of the following: (e) Drywell spray Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-RHR-SPRAY-DW-QC Revision: 3 mr 7 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.2.1 Revision: 23 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 2 Group 2 K/A 239001.2.2.42 Level of Difficulty: 2 Importance Rating 3.9 Main and Reheat Steam: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Question # 61 CGS is operating in Mode 1. Reactor power is 25%.

With no other actions, how many Main Turbine Bypass Valves (BPVs) must be operable to continue unrestricted operation with the current plant conditions.

A. 0 B. 2 C. 3 D. 4 Answer: D Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This question requires the candidate to demonstrate knowledge of the entry conditions for the Main/Reheat Steam-related Technical Specification LCO 3.7.6, Main Turbine Bypass System.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since operation at >25% reactor power may continue with the Main Turbine Bypass system inoperable if modifications to the MCPR limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") are made. However, with no additional actions, all 4 Bypass Valves (BPVs) must be operable B. Incorrect. Plausible if it believed that the Main Turbine Bypass system is rated for a load reduction of 50%. However, the system is rated for 50% load reduction, and all BPVs must be operable when 25%.

C. Incorrect. Plausible if it believed that the Main Turbine Bypass system is rated for a load reduction of 100%. However, the system is rated for 50% load reduction, and all BPVs must be operable when 25%.

D. Correct. When reactor power is 25%, and no actions have been taken to modify the MCPR, then all 4 BPVs must be operable to allow continued operation.

Technical Reference(s) Attached w/ Revision # See TS LCO 3.7.6, Main Turbine Bypass System Comments / Reference TS Bases, LCO 3.7.6, Main Turbine Bypass System Proposed references to be provided during examination: None Learning Objective: 5545 - Referencing Columbia Generating Station Technical Specifications (section 3 only for initial license candidates) associated with the Main Steam System and a set of plant conditions, determine as applicable the LSSS, the LCO, the action statement, and the appropriate bases.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

LCO 3.7.6 Revision: 237 Comments /

Reference:

TS Bases, LCO 3.7.6 Revision: 92 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/2/2017 Tier 2 Group 2 K/A 259001.K1.11 Level of Difficulty: 3 Importance Rating 2.7 Reactor Feedwater System: Knowledge of the physical connections and/or cause-effect relationships between REACTOR FEEDWATER SYSTEM and the following: RFP lube oil system Question # 62 CGS is operating in Mode 1.

The following RFW-P-1A annunciators are alarming:

P840.A1.7-1, TURBINE A VIBRATION HIGH

P840.A1.8-1, TURBINE A THRUST BEARING WEAR HIGH If turbine thrust bearing wear rises to 15 mils, the turbine trip oil header for RFW-P-1A will...

A. direct oil to the tops of the high pressure and low pressure stop valves causing them to close.

B. pressurize to align the high pressure and low pressure stop valve drains to the oil reservoir.

C. depressurize allowing the high pressure and low pressure stop valves to close.

D. remain pressurized. Thrust bearing wear does not cause a feed turbine trip.

Answer: C Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This question requires the candidate to demonstrate knowledge of the effect of high turbine thrust bearing wear on the reactor feedwater system.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since pressure on the top of the valves would cause them to close. However, the stop valves are hydraulically open, spring shut. They are tripped by venting control oil pressure from the bottom of the valves.

B. Incorrect. Plausible since the stop valves are tripped by venting control oil pressure from the bottom of the valves. However, this is accomplished by depressurizing the control oil header.

C. Correct. When Turbine Thrust Bearing wear reaches 15 mils, the turbine trip oil system depressurizes the oil lines leading to the high and low pressure stop valves. These valves close by spring force within 0.5 seconds.

D. Incorrect. Plausible since pump thrust bearing will actuate the Turbine A Vibration High alarm.

However, high wear on the pump thrust bearing will not cause a turbine trip. There are two thrust bearings, the turbine thrust bearing and the pump thrust bearing. Both bearings will cause a Turbine A Vibration High alarm. However, only the turbine thrust bearing will cause a turbine trip.

Technical Reference(s) Attached w/ Revision # See SD000151, CGS System Description, Vol. 2 Chap. 3, Reactor Comments / Reference Feedwater P840.A1.7-1, TURBINE A VIBRATION HIGH, Annunciator Response P840.A1.8-1, TURBINE A THRUST BEARING WEAR HIGH, Annunciator Response Proposed references to be provided during examination: None Learning Objective: 5750 - Explain the following pertaining to the Turbine Oil System:

c. Basic relationship between trip oil pressure and steam stops/ hydraulic trip valve during normal and tripped conditions.

d. How RFP trip oil responds to a turbine trip and turbine reset.

Question Source: Bank # LO01945 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.7 55.43 Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000151 Revision: Major 13 Minor 003 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

P840.A1.7-1, Annunciator Response Revision: Major 022 Minor 002 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

P840.A1.8-1, Annunciator Response Revision: Major 022 Minor 002 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 0 Date: 12/16/2016 Tier 2 Group 2 K/A 268000.K3.04 Level of Difficulty: 3 Importance Rating 2.7 Radwaste: Knowledge of the effect that a loss or malfunction of the RADWASTE will have on following: Drain sumps Question # 63 On a loss of control air to the Reactor Building, Reactor Building radioactive floor drain (FDR) sumps...

A. cannot be pumped down.

B. may be pumped to the Waste Collector Tank only.

C. may be pumped to the Floor Drain Collector Tank only.

D. may be pumped to either the Floor Drain Collector Tank or the Waste Collector Tank.

Answer: A Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question determines if candidates understand the impact of a malfunction (Loss of air) on the Radwaste system with regard to floor drains.

SRO Only:

N/A Explanation:

A. Correct. When control air is lost, sump pump discharge isolation valves FDR-V-219/220/221/222 fail closed. Reactor building sumps cannot be pumped down.

B. Incorrect. Plausible since the Floor Drain Collector Tank to Waste Collector Tank crosstie, FDR-V-33, fails as-is on a loss of control air. However, all reactor building sump pump discharge valves fail close and the sumps cannot be pumped down.

C. Incorrect. Plausible if it is believed that the sump pump discharge isolation valves FDR-V-219/220/221/222 fail as-is or open and the Floor Drain Collector Tank to Waste Collector Tank crosstie, FDR-V-33, fails closed. However, when control air is lost, sump pump discharge isolation valves FDR-V-219/220/221/222 fail closed. Reactor building sumps cannot be pumped down.

D. Incorrect. Plausible if it is believed that the sump pump discharge isolation valves FDR-V-219/220/221/222 fail as-is or open. However, when control air is lost, sump pump discharge isolation valves FDR-V-219/220/221/222 fail closed. Reactor building sumps cannot be pumped down.

Technical Reference(s) Attached w/ Revision # See ABN-CAS, Control Air System Failure Comments / Reference SD000130, CGS System Description, Vol. 9, Chap. 7, Plant Drains Proposed references to be provided during examination: None Learning Objective: 12507 - Explain he interrelationships between the Plant Drains system and the following systems: (c) Control and Service Air Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.5 55.43 Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-CAS Revision: 009 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Plant Drains System Description Revision:

Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/7/2017 Tier 2 Group 2 K/A 272000.K4.02 Level of Difficulty: 3 Importance Rating 3.7 Radiation Monitoring System: Knowledge of RADIATION MONITORING System design feature(s) and/or interlocks which provide for the following: Automatic actions to contain the radioactive release in the event that the predetermined release rates are exceeded.

Question # 64 CGS is operating in Mode 1.

The following annunciators are alarming:

P602.A5.1-5: REACTOR BLDG EXH PLENUM RAD HIGH

P602.A5.1-4: REACTOR BLDG EXH PLENUM RAD HI-HI Reactor building exhaust plenum radiation levels read:

REA-RIS-609A: 15.5Mr/hr

REA-RIS-609B: 16.1Mr/hr

REA-RIS-609C: 14.8Mr/hr

REA-RIS-609D: 15.6Mr/hr Which of the following will occur in response to these conditions?

A. Control room ventilation fans trip.

B. Offgas system discharge isolates.

C. Reactor water sample valves close.

D. Containment Nitrogen Makeup isolates.

Answer: D Page 1 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question evaluates whether the candidate understands automatic actions that will occur as a result on high radiation detected in the reactor building exhaust plenum SRO Only:

N/A Explanation:

High reactor building exhaust plenum levels cause a Z signal when above 13mr/hr. Per ABN-FAZ-QC, Containment Nitrogen makeup isolates on a Z signal.

A. Incorrect. Control room ventilation fans do not trip. Instead the CR emergency filtration systems start and align air to the remote intakes. Plausible because automatic actions are associated with the control room ventilation system.

B. Incorrect. Offgas does not isolate on a Z signal. Plausible because the offgas system processes radiative non-condensable gases from the condenser during normal operations.

C. Incorrect. Reactor water sample valves do not isolate on a Z signal. Plausible because they do isolate on an F or A signal.

D. Correct.

Technical Reference(s) Attached w/ Revision # See ABN-FAZ-QC FAZ Automatic Actions - Quick Card Comments / Reference PPM 4.602.A5, Annunciator Response Procedure Proposed references to be provided during examination: None Learning Objective: 11937 - Describe the effect that a loss or malfunction of each of the following would have on the NS4 system. (b) RB Exhaust Plenum Radiation Monitor Question Source: Bank # LX00825 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.9 55.43 Page 2 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

4.602.A5 Revision: 46 Page 3 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-FAZ-QC Revision: 3 mr 1 Page 4 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-FAZ-QC Revision: 3 mr 1 Page 5 of 5

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 2 Group 2 K/A 286000.K2.02 Level of Difficulty: 2 Importance Rating 2.9 Fire Protection System: Knowledge of electrical power supplies to the following: Pumps Question # 65 What is the power supply to FP-P-2A, Fire Protection Pump 2A?

A. MC-5N B. MC-3C C. MC-7A D. Diesel Engine Answer: A Page 1 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question determines if candidates know the power supply to fire protection pumps.

SRO Only:

N/A Explanation:

FP-P-2A is powered by MC-5N A. Correct.

B. Incorrect. Plausible because FP-TK-1 (cardox unit) is powered by MC-3C.

C. Incorrect. Plausible because FP-P-2B is powered by MC-6N.

D. Incorrect. Plausible because FP-P-110 is a diesel powered pump.

Technical Reference(s) Attached w/ Revision # See A SD000177, CGS System Description, Volume 3, Chapter 2, Comments / Reference Fire Protection Proposed references to be provided during examination: None Learning Objective: 12271 Explain the function and operation of the following Fire Protection System components, including any automatic features or interlocks: Fire Pumps Question Source: Bank # LO03268 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.4 55.43 Page 2 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000177 Revision: 16 mr 1 Page 3 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/9/2017 Tier 3 Group K/A 2.1.2 Level of Difficulty: 2 Importance Rating 4.1 Knowledge of operator responsibilities during all modes of plant operation.

Question # 66 CGS is operating in Mode 4.

Maintenance has requested that the following valve be cycled to consolidate new packing that has been installed:

Per PPM 1.3.1, Operating Policies, Programs and Practices, what are the required limits for cycling this valve?

During valve cycling, the duty cycle should not exceed (1) minutes, or (2) starts, whichever comes first, without meeting cooldown time requirements.

A. (1) 5 (2) 8 B. (1) 5 (2) 10 C. (1) 15 (2) 10 D. (1) 15 (2) 8 Page 1 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Answer: A K/A Match:

Question determines if operators understand operator responsibilities for limiting duty cycle on MOVs operated from the control room.

SRO Only:

N/A Explanation:

A. Correct. RWCU-V-4 is powered from MC-S2-1A, which is DC. Per PPM 1.3.1, section 4.14.3.l, DC motor operator cycle time requirements are 5 minutes or 8 starts, whichever is first.

B. Incorrect. Plausible since DC motor duty cycle should not exceed 5 minutes. However, DC actuators are limited to 8 starts prior to a cooldown. AC motor operators are allowed 10 starts.

C. Incorrect. Plausible since AC motor operators are limited to 15 minutes and 10 starts without a cooldown. However, RWCU-V-4 is DC powered. DC motor operators are limited to 5 minutes and 8 starts.

D. Incorrect. Plausible since DC motor duty cycle is limited to 8 starts. However, DC actuators are limited to 5 minutes prior to a cooldown. AC motor operators are allowed 15 minutes.

Technical Reference(s) Attached w/ Revision # See PPM 1.3.1 Operating Policies, Programs, and Practices. Comments / Reference Proposed references to be provided during examination: None Learning Objective:

Question Source: Bank # LO01463 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 1.3.1 Revision: Major 120 Minor 005 Page 3 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 3 Group K/A 2.1.20 Level of Difficulty: 3 Importance Rating 4.6 Ability to interpret and execute procedure steps.

Question # 67 Control room operators are performing a continuous use procedure when they come to a conditional step that does not apply to the current plant conditions. The operators mark the step N/A.

What is required to mark this step N/A?

A. Document the reason next to the step, and obtain supervisor initials for approval.

B. Document the reason next to the step; supervisor initials are not required for approval.

C. Obtain supervisor initials next to the step for approval; a documented reason is not required.

D. No additional documentation or supervisor approval is required.

Answer: D Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question provides a reference and plant conditions and determines if the candidate can correctly interpret and execute procedural steps.

SRO Only:

N/A Explanation:

Student must interpret PPM 3.1.2 Attachment 7.1.

A. Incorrect. Plausible since documenting a reason and supervisor initials are required to N/A non-conditional steps. However, for conditional steps that do not apply to the current plant conditions, documenting a reason and supervisor initials are not required.

B. Incorrect. Plausible since documenting a reason is required to N/A non-conditional steps.

However, for conditional steps that do not apply to the current plant conditions, documenting reasons and supervisor approvals are not required.

C. Incorrect. Plausible since supervisor initials are required to N/A non-conditional steps. However, for conditional steps that do not apply to the current plant conditions, no other approval is required.

D. Correct. When a conditional step does not apply to the current plant conditions, it may be marked N/A and no additional justification or approval is required.

Technical Reference(s) Attached w/ Revision # See SWP-PRO-01, Procedure and Work Instruction Use and Adherence Comments / Reference Proposed references to be provided during examination: None Learning Objective: 10774 - Using the appropriate procedures, discuss procedure use and adherence at Columbia.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SWP-PRO-01, section 4.7 Revision: 30 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 3 Group K/A 2.1.23 Level of Difficulty: 3 Importance Rating 4.3 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Question # 68 CGS is operating in Mode 2.

A reactor startup is in progress in accordance with PPM 3.1.2, Start Up Flow Chart.

The following step is being performed:

What indications are used to determine criticality?

Criticality shall be identified by increasing neutron level, A. a constant positive period and no change in reactor coolant temperature.

B. no simultaneous control rod motion and no change in reactor coolant temperature.

C. a constant positive period and no simultaneous control rod motion.

D. positive reactor period getting longer and no simultaneous control rod motion.

Answer: C Page 1 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question determines if candidates can correctly interpret and execute general (not overly specific) procedural requirements. Procedural step is from PPM 3.1.2, Plant Startup, which is an integrated plant procedure.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since a constant positive period is used to determine criticality. However, no change in reactor coolant temperature is not used.

B. Incorrect. Plausible since a no simultaneous control rod motion is used to determine criticality.

However, no change in reactor coolant temperature is not used.

C. Correct. Per note N6 of PPM 3.1.2, Startup Flowchart, attachment 7.3, criticality is determined by increasing neutron level, a constant steady period, and no simultaneous control rod motion.

D. Incorrect. Reactor period getting longer would indicate a subcritical reactor. Plausible because reactor period is observed and no rod motion is verified as part of the check for criticality.

Technical Reference(s) Attached w/ Revision # See PPM 3.1.2, Startup Flowchart, Attachment 7.3 Comments / Reference Proposed references to be provided during examination: None Learning Objective: 6651 - With procedures available, determine how criticality is identified.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.10 55.43 Comments /

Reference:

PPM 3.1.2, Attachment 7.3 Revision: Major 81 Minor 001 Page 2 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Page 3 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 3 Group K/A 2.2.6 Level of Difficulty: 3 Importance Rating 3.0 Knowledge of the process for making changes to procedures.

Question # 69 CGS is operating in Mode 3.

A reactor startup is in progress.

CRO2 is preparing to start RCC-P-1B per SOP-RRC-START.

Prior to giving direction for OPS2 to close RCC-V-2B (pump discharge valve), CRO2 notes that the procedure contains an error. RCC-V-2B is incorrectly listed as RCC-V-2A in the procedure.

How should this error be resolved?

RCC startup may continue A. after receiving a peer check from CR01 on the Equipment Part Number (EPN), revising the erroneous EPN using pen and ink. No supervisory review is required.

B. after an electronic Procedure Change Notification (PCN) has been initiated and approved by the Shift Manager and the Operations Manager.

C. after a Minor Revision Procedure Change Notification (PCN) has been initiated electronically and approved by the Plant Operations Committee (POC).

D. after receiving verbal approval for the procedure change (PCN) from the CRS or Shift Manager. Document the verbal approval in the procedure.

Answer: D Page 1 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

Question presents an operationally valid situation where a minor error in the procedure needs to be resolved and determines if the candidates have knowledge of the process to do so SRO Only:

N/A Explanation:

Per SWP-PRO-02, once a verbal PCN has been completed per section 5.5.1, work may continue.

A. Incorrect. The PCN process should be used. Plausible because the procedural error is minor. Also, when a procedure is not currently in use and an error is identified, it is appropriate to write a CR.

B. Incorrect. Operations manager approval is not required for PCNs. Plausible because a PCN should be initiated to correct the procedural error.

C. Incorrect. POC review and approval is NOT required prior to continuing work provided the verbal PCN process is used. Plausible because POC approval may be required when submitting the procedure for revision depending on the significance of the change.

D. Correct.

Technical Reference(s) Attached w/ Revision # See SWP-PRO-02 Preperation, Review, Approval and Distribution of Comments / Reference Procedures.

Proposed references to be provided during examination: None Learning Objective: 6064 - State when a verbal temporary change to a procedure can be used.

Question Source: Bank # LO02784 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SWP-PRO-02 Revision: 45 Page 3 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/2/2017 Tier 3 Group 2 K/A 2.2.35 Level of Difficulty: 2 Importance Rating 3.6 Ability to determine Technical Specification Mode of Operation Question # 70 Given the following:

All reactor vessel head closure bolts are fully tensioned.

The mode switch is in shutdown.

Average reactor coolant temperature is 200°F.

The reactor is in Mode A. 2 B. 3 C. 4 D. 5 Answer: C Page 1 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question provides plant conditions and determines if the candidate can select the correct MODE of operation based on those conditions.

SRO Only:

N/A Explanation:

With temperature less than or equal to 200 degrees F and all vessel head closure bolts tensioned, the reactor vessel is in Mode 4.

A. Incorrect. Plausible if the mode switch was in Refuel.

B. Incorrect. Plausible if RCS temperature were above 200 degrees F.

C. Correct.

D. Incorrect. Plausible if one or more reactor vessel head closure bolts was less than fully tensioned.

Technical Reference(s) Attached w/ Revision # See CGS Technical Specifications, Table 1.1-1, Modes Comments / Reference Proposed references to be provided during examination: None Learning Objective: 10297: Define: Mode 1, Mode 2, Mode 3, Mode 4, and Mode 5.

Question Source: Bank #

Modified Bank # LO01576 (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

CGS TS table 1.1-1, Modes Revision: 237 Comments /

Reference:

Question LO01576 Revision:

With the mode switch in shutdown and reactor pressure at 135 psig, the reactor would be in Mode:

A. 2 B. 3 C. 4 D. 5 Answer: B Page 3 of 3 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 0 Date: 12/16/2016 Tier 3 Group 2 K/A 2.2.12 Level of Difficulty: 2 Importance Rating 3.7 Knowledge of surveillance procedures.

Question # 71 When reading surveillance procedures, which symbol indicates steps that contain technical specification surveillance acceptance criteria?

A. #

B.

C. $

D.

Answer: A Page 1 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The question determines whether the candidate knows about important symbols contained throughout surveillance procedures. Since the symbol applies to all surveillance procedures, it is generic in nature and meets the K/A.

SRO Only:

N/A Explanation:

SWP-PRO-03 states the following:

IDENTIFY steps that contain or satisfy Technical Specification Surveillance acceptance criteria with a

sign to the left of the step number.

A. Correct.

B. Incorrect. Plausible because it is a commonly used symbol in technical writing.

C. Incorrect. Plausible because it is a commonly used symbol (indicates ODCM and LCS criteria).

D. Incorrect. Plausible because it is a commonly used symbol in technical writing.

Technical Reference(s) Attached w/ Revision # See SWP-PRO-03, Writers Manual Comments / Reference Proposed references to be provided during examination: None Learning Objective: 13559 - Given copies of plant procedures, locate and demonstrate an understanding of the procedural steps that apply to performing system tasks, including precautions and limitations, and Operations responsibilities.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SWP-PRO-3 Revision: 22 Page 3 of 3

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 0 Date: 12/16/2016 Tier 3 Group 3 K/A 2.3.12 Level of Difficulty: 3 Importance Rating 3.2 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Question # 72 CGS is in Mode 2, performing a reactor startup following a refueling outage.

Reactor power is 5% and steady.

An equipment configuration issue is identified which requires a containment entry to realign equipment.

Two operators will enter to complete the equipment lineup.

Who must give permission for this containment entry?

Permission for containment entry is required from the Shift Manager (SM)

A. only.

B. and Plant General Manager (PGM) only.

C. and Radiation Protection Manager (RPM) only.

D. and Radiation Protection Manager (RPM) and Plant General Manager (PGM) only.

Answer: D Page 1 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

The K/A asks for knowledge of radiation principles, such as containment entry requirements. This question requires the candidate to differentiate between the required approvals for various containment entries.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible if the containment entry was not the initial containment entry. However, since this the first entry since de-inerting containment following a reactor startup (see Ref. A, section 2.4 below), this entry is classified as an initial entry and additional permissions are required.

B. Incorrect. Plausible if this containment entry is being performed at power. However, plant conditions given show the reactor is shutdown. Therefore, plant general manager permission to enter containment is not required.

C. Incorrect. Per Ref. A, step 6.1.5, initial containment entry with the reactor shutdown requires Shift Manager and Radiation Protection Manager permission only.

D. Correct. Step 4.1.2 or Ref. A requires Plant General Managers permission to enter containment at power.

Technical Reference(s)

SOP-ENTRY-DW, Personnel Entry Into Drywell Attached w/ Revision # See PPM 11.2.7.3, High Radiation Area, Locked High Radiation Area, and Comments / Reference Very High Radiation Area Controls Proposed references to be provided during examination: None Learning Objective: 13262 - Identify the requirements necessary to coordinate an initial DW entry.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.12 55.43 Page 2 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-ENTRY-DW, section 6.1, Drywell Revision: Major: 023, Minor 001 Entry Preparation, step 6.1.5 Page 3 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

A, section 6.3, Subsequent Drywell Revision: Major: 023, Minor 001 Entry, step 6.3.3 Page 4 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-ENTRY-DW, section 2.4 Revision: Major: 023, Minor: 001 Page 5 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 11.2.7.3, section 6.6, steps 6.6.1 Revision: Major: 041, Minor: 001

& 6.6.2 Page 6 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 11.2.7.3, section 6.6, step 6.6.3 Revision: Major: 041, Minor: 001 Page 7 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 3 Group 3 K/A 2.3.14 Level of Difficulty: 2 Importance Rating 3.4 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Question # 73 CGS is in Mode 4.

An event requires operators to enter the Reactor Building to perform a system isolation that will protect some valuable equipment.

An individual has volunteered to perform the task.

At what classification are Energy Northwests administrative exposure hold points first waived and, what is the maximum dose that the Emergency Director may authorize the individual to receive?

A. Alert; 10 rem TEDE B. Site Area Emergency; 10 rem TEDE C. Alert; 25 rem TEDE D. Site Area Emergency; 25 rem TEDE Answer: A Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This question requires the candidate to demonstrate knowledge of the emergency allowable dose limits and when the Energy Northwest dose hold points are waived.

SRO Only:

N/A Explanation:

A. Correct. A total dose of 10 rem TEDE may be authorized to save valuable equipment. Energy Northwest administrative exposure hold points are automatically waived at an Alert classification.

B. Incorrect. Plausible since a total dose of 10 rem TEDE may be authorized to save valuable equipment. However, Energy Northwest administrative exposure hold points are automatically waived at an Alert classification.

C. Incorrect. Plausible since Energy Northwest administrative exposure hold points are automatically waived at an Alert classification. However, a total dose of 10 rem TEDE may be authorized to save valuable equipment.

D. Incorrect. Plausible since 25 rem is the emergency exposure limit for life saving. However, a total dose of 10 rem TEDE may be authorized to save valuable equipment. Additionally, Energy Northwest administrative exposure hold points are automatically waived at an Alert classification.

Technical Reference(s) Attached w/ Revision # See PPM 13.2.1, Emergency Exposure Levels / Protective Action Guides Comments / Reference Proposed references to be provided during examination: None Learning Objective: 11258 - Knowledge of the facility ALARA program Question Source: Bank # LO01923 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2009, #23 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 41.12 55.43 Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 13.2.1, section 2.2, Emergency Revision: 22 Exposure Controls Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 13.2.1, EPA 400 Protective Action Revision: 22 Guides for Emergency Workers Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 3 Group 4 K/A 2.4.4 Level of Difficulty: 3 Importance Rating 4.5 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Question # 74 The plant is in MODE 1.

The crew has identified a slow increase in drywell unidentified leakage.

Drywell pressure: 0.7 psig, up slow.

A loss of IN-1 occurs, resulting in a loss of power to US-PP.

At a Drywell pressure of 1.0 psig, the CRS orders a reactor scram.

RPV level: -20 inches, up slow.

Drywell pressure: 1.5 psig, up slow

Drywell temperature: 125°F, up slow

Reactor Building exhaust plenum radiation level: 2.5 mr/hr, up slow.

Which EOP(s) should be entered?

A. Enter PPM 5.1.1 RPV Control only.

B. Enter PPM 5.1.1 RPV Control and PPM 5.2.1 Primary Containment Control.

C. Enter PPM 5.1.1 RPV Control and transition to PPM 5.1.2 RPV Control-ATWS.

D. Enter PPM 5.1.1 RPV Control and PPM 5.3.1 Secondary Containment Control.

Answer: A Page 1 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This question requires the candidate to demonstrate knowledge of EOP entry conditions.

SRO Only:

N/A.

Explanation:

A. Correct. Entry into PPM 5.1.1 is required due to RPV level below +13 inches.

B. Incorrect. Plausible since Drywell pressure and temperature are rising. However, both parameters do not meet the requirements to enter PPM 5.2.1.

C. Incorrect. Plausible since the loss of IN-1 will cause a loss of the full-core display. However, all rod positions may be verified with the Rod Worth Minimizer (RWM) and rods may be verified to be fully inserted. Therefore, entry into PPM 5.1.2 is not required.

D. Incorrect. Plausible since RB exhaust plenum radiation level is rising. However, the level does not meet the requirement to enter PPM 5.3.1.

Technical Reference(s) Attached w/ Revision # See ABN-ELEC-INV, 120 VAC Critical Distribution System Failures Comments / Reference PPM 5.1.1, RPV Control Proposed references to be provided during examination: None Learning Objective: 8017: Given plant conditions, recognize an EOP entry condition(s) and enter the appropriate flow chart.

Question Source: Bank # LR00114 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 2 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-ELEC-INV Revision: Major 13 Minor 001 Page 3 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.1.1 Entry Requirements Revision: 021 Page 4 of 4

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 1/31/2017 Tier 3 Group 4 K/A 2.4.5 Level of Difficulty: 3 Importance Rating 3.7 Type the K/A System or Condition Here: Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.

Question # 75 CGS was operating in Mode 1 when a condition caused the crew to enter an abnormal procedure (ABN). Conditions degraded such that the crew initiated a manual reactor scram and entered the reactor scram procedure and an emergency operating procedure (EOP) prior to completing all steps in the ABN.

Which of the following is correct concerning continuing execution of ABN steps?

Steps in the ABN may be executed A. concurrently with the EOP ONLY IF specifically called out by the EOP.

B. concurrently with the EOP ONLY IF ABN actions do not conflict with EOP actions.

C. ONLY AFTER the shift manager has determined that an emergency no longer exists.

D. ONLY AFTER applicable steps of the reactor scram procedure have been completed.

Answer: B Page 1 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 K/A Match:

This questions tests the candidates knowledge of the hierarchy of use for operating procedures, including emergency, abnormal and normal procedures.

SRO Only:

N/A Explanation:

A. Incorrect. Plausible since Ref. A, section 4.8.3.a states that All required concurrent execution of any Volume 2, 3, or 4 Procedure are specifically called out by the EOPs. However, the section continues by stating This does not mean that a Volume 4, Abnormal Procedure, cannot otherwise be concurrently executed with the EOPs so long as its specified actions do not conflict with the direction given by the EOPs. Therefore, ABN actions may be performed concurrently with EOP actions without being specifically referenced in the EOP.

B. Correct. As delineated in Ref. A, section 4.8.3.a, Abnormal procedure (ABN) actions may be executed concurrently with Emergency procedure (EOP) actions as long as the ABN actions do not conflict with the direction given by the EOPs.

C. Incorrect. Plausible since Ref. A, section 4.8.1.a states The Volume 5 Emergency Operating Procedures (EOPs) and the actions specified therein have priority/precedence over all Volume 2, 3, and 4 Procedures when an emergency exists (EOP entry condition(s) is/are met). This statement implies that once an EOP is entered, actions from lower tier procedures are stopped.

Additionally, Ref. A, section 4.8.2.e states EOPs are exited only if the Shift Manager determines that an emergency no longer exists and directs EOP exit or the EOPs direct exit to appropriate plant procedures. This distractor infers that the EOPs must be exited prior to completing ABN actions. This is not the case, however. See explanation for answer B above.

D. Incorrect. Plausible since Ref. A, section 4.8.1.b states During a transient, and as plant conditions continue to degrade, the flow path of procedure usage is from the Volume 2 and 3 Procedures to the Volume 4, Abnormal Procedures, and then to the EOPs. This infers that actions from a Volume 3, General Operating Procedure, such as PPM 3.3.1, Reactor Scram, should be completed prior to completing actions from a Volume 4 procedure, such as ABNs. This is not the case. See the explanation for answers B and C above.

Technical Reference(s)

PPM 1.3.1, Operating Policies, Programs and Practices Attached w/ Revision # See Comments / Reference Page 2 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Proposed references to be provided during examination: None Learning Objective: 6105 - State which procedures have priority/precedence over all other operating procedures when an emergency exists.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 41.10 55.43 Page 3 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 1.3.1, section 4.8.3, Procedure Use Revision: 120, Minor 003 During an Emergency Page 4 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 1.3.1, section 4.8.4, Abnormal Revision: 120, Minor 003 Procedures Page 5 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 1.3.1, section 4.8.1, Procedure Revision: 120, Minor 005 Hierarchy Page 6 of 7

ES-401 CGS NRC 2017 RO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 1.3.1, section 4.8.2, Volume 5 Revision: 120, Minor 005 Emergency Operating Procedures (EOPs)

Page 7 of 7

ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet U. S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:

Date: March 9, 2017 Facility / Unit: CGS Region: I II III IV Reactor Type: W CE BW GE Start Time: 0900 Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80 percent overall, with 70 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80 percent to pass. You have 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to complete the combined examination, and 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months if you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

__________________________________

Applicant's Signature Results RO/SRO-Only/Total Examination Values 75 / 25 / 100 Points 74 99 Applicant's Score ______ / ______ / ______

Points Applicant's Grade ______ / ______ / ______

Percent

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 1 Group 1 K/A 295001.AA2.03 Level of Difficulty: 3 Importance Rating 3.3 PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Actual core flow Question # 76 CGS is in Mode 1. Reactor power is 100%.

An event causes both RRC pumps to run back.

Current conditions:

Reactor power: 80%

Core flow:

What is the earliest action required to satisfy Technical Specifications?

A. Declare RRC Loop A inoperable in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

B. Shutdown the reactor to Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

C. Reduce reactor power to less than 70% in 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months .

D. Secure RRC-P-1A and enter single loop operation within 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

Answer: A Page 1 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 K/A Match:

Requires the candidate to interpret core flow indications during a partial loss of core flow and determine actions necessary to mitigate the casualty.

SRO Only:

K/A is an A2 Statement.

Page 2 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Total core flow is below 70% of rated core flow (75.95 Mlb/hr). In accordance with TS bases (Ref. B), and the daily surveillance (Ref. C), Jet Loop Flow mismatch must be LE 4172 gpm.

From indications in the stem, the mismatch is 4310 gpm, which does not meet the criterion established in surveillance 3.4.1.1. Condition A of LCO 3.4.1 (Ref. A) applies, and the Loop with the lowest flow (Loop A) should be declared inoperable within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

B. Incorrect. In accordance with LCO 3.4.1, condition C (Ref. A), placing the reactor in Mode 3 is a correct action if conditions A & B are not complete. However, the stem asked for the earliest action to meet TS, which is answer A.

C. Incorrect. Plausible if candidate believes that loop flow deviation limit of LE 10% is applicable >

70% power vice 70% rated flow.

D. Incorrect. Plausible since securing RRC-P-1A and entering single loop operations would meet the requirements of LCO 3.4.1. However, the stem asked for the earliest action to meet TS, which is answer A.

Technical Reference(s)

A Technical Specification 3.4.1, Recirculation Loops Operating Attached w/ Revision # See B Technical Specifications Bases for TS 3.4.1 Comments / Reference OSP-RRC-D701, Jet Pump Operability and Recirculation Loop C

Flow Mismatch, Two Loop Operation Proposed references to be provided during examination: None Learning Objective: 5031 - Referencing Columbia Generating Station Technical Specifications associated with the Reactor Recirculation System and a set of plant conditions, determine as applicable the LSSS, the LCO, the action statement, and the appropriate bases.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 3 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

TS 3.4.1, Page 1 Revision: Amendment 237 Page 4 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

TS 3.4.1, Page 2 Revision: Amendment 237 Page 5 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

TS Bases for TS 3.4.1 Revision: 92 Page 6 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

OSP-RRC-D701, Daily Surveillance Revision: Major 018, Minor 002 Page 7 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 1 Group 1 K/A 295003.2.2.36 Level of Difficulty: 3 Importance Rating 4.2 Partial or Complete Loss of AC: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Question # 77 CGS is in Mode 1. DG-2 is in a maintenance window.

DG-2 was declared inoperable on Jan. 1, 2017, at 1600.

A risk management action plan for the use of the alternate AC source, DG-4, has been established prior to the DG-2 maintenance window.

On Jan. 8, 2017, at 2100, a fault caused a lock-out condition on the Backup Transformer, TR-B, and TR-B was declared inoperable.

Using the references provided, which is the earliest required action that will preclude a required shutdown?

Declare A. TR-B operable prior to 2100 on Jan. 9, 2017.

B. DG-2 operable prior to 0900 on Jan. 9, 2017.

C. TR-B operable prior to 2100 on Jan. 11, 2017.

D. DG-2 operable prior to 1600 on Jan. 15, 2017.

Answer: B K/A Match:

The question requires the candidate to understand how a diesel generator maintenance window affects LCO 3.8.1 with a concurrent loss of one qualified offsite circuit.

Page 1 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a G statement linked to 10CFR 55.43.5 Page 2 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since this action and time would be applicable if two offsite sources were lost (LCO 3.8.1, condition C). However, with one offsite source and one diesel generator inoperable, either one must be declared operable within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to meet the requirements of LCO 3.8.1 condition D.

B. Correct. Since one DG and one qualified offsite circuit is inoperable, LCO 3.8.1 condition D is applicable.

C. Incorrect. One inoperable qualified offsite circuit is normally required to be restored to operable within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months per LCO 3.8.1, condition A. However, since one DG and one qualified offsite circuit is inoperable, LCO 3.8.1 condition D is applicable.

D. Incorrect. With a risk management plan for the use of the alternate AC source, DG may be inoperable for 14 days per LCO 3.8.1, condition B. However, since one DG and one qualified offsite circuit is inoperable, LCO 3.8.1 condition D is applicable.

Technical Reference(s)

Technical Specification 3.8.1, AC Sources-Operating Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: LCO 3.8.1, Action Table Learning Objective: 5059 - Referencing Technical Specifications associated with the AC Distribution System and a set of plant conditions, determine as applicable the LCO, the action statement, and the appropriate bases.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.2 Page 3 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

TS 3.8.1, AC Sources-Operating Revision: Amendment 237 Page 4 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 5 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 6 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 7 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 8 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/7/2017 Tier 1 Group 1 K/A 295004 AA2.01 Level of Difficulty: 3 Importance Rating 3.6 Partial or Total Loss of DC Pwr: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Cause of partial or complete loss of D.C.power Question # 78 CGS is operating in Mode 1.

Battery charger E-C1-1B is out of service for corrective maintenance A reactor scram is initiated due to a LOCA. Current plant conditions:

RPV pressure: 800 psig and stable. SRVs are being controlled in manual

RPV level: -160 inches, up slow. RCIC is being used to control level.

Drywell pressure: 3.52 psig, down slow. RHR-P-2B is spraying the wetwell.

HPCS is running on minimum flow.

The following occurs:

Annunciator P.800.C5.9-2, 125 VDC CHARGER C1-1A/1B TROUBLE, alarms.

Field operators report that the AC input breaker to battery charger E-CB-1-1A is tripped open and cannot be closed.

What actions should the CRS direct to control the plant?

The CRS should direct operators to A. insert keys and operate SRVs from panel H13-P631 to control RPV pressure.

B. secure the RCIC turbine and operate HPCS as necessary to control RPV level.

C. reopen MSIVs and control RPV pressure using DEH in automatic.

D. continue to operate RCIC to control level until RHR is in shutdown cooling.

Answer: B K/A Match:

This question requires candidates to demonstrate an understanding that a partial loss of 125 vdc has occurred and the EOP actions required to mitigate the event.

Page 1 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is an A2 statement and Page 2 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since ADS SRVs are operated from H13-P631 on a complete loss of division 1 DC. However, with both battery chargers inoperable, the battery will maintain the DC bus and SRVs will be controlled from the front panel H13-P601.

B. Correct. Since division 1 DC is not lost, the RCIC system will continue to operate. However, ABN-ELEC-125VDC, step 4.2.5 directs placing an alternate water source in service when both division 1 battery chargers are lost in anticipation of losing the RCIC turbine on a loss of the division 1 DC bus.

C. Incorrect. Plausible since reopening MSIV is allowed in step PPM 5.1.1, step 3 and a complete loss of division 1 DC would render SRVs inoperable from the front panel. However, RPV low level interlocks cannot be bypassed to open MSIVs and SRVs are available.

D. Incorrect. Plausible since, division 1 DC is still available and the RCIC system will continue to operate. However, ABN-ELEC-125VDC, step 4.2.5 directs placing an alternate water source in service when both division 1 battery chargers are lost in anticipation of losing the RCIC turbine on a loss of the division 1 DC bus. Additionally, the division 1 battery is rated to maintain DC voltage on the bus for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Brining the primary to the point of SDC will take much longer, and the RCIC turbine will trip once battery bus voltage is lost.

Technical Reference(s)

ABN-ELEC-125VDC, Plant BOP, DIV 1,2 & 3 125 VDC Distribution System Failures Attached w/ Revision # See Comments / Reference PPM 5.1.1, RPV Control Proposed references to be provided during examination: None Learning Objective: 7652: Predict the effects that a loss of DC bus S1-1 will have on b. RCIC Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 3 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-ELEC-125VDC, Plant BOP, DIV 1,2 Revision: 014

& 3 125 VDC Distribution System Failures Page 4 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 5 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.1.1, RPV Control Revision: 21 Page 6 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 7 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 1 Group 1 K/A 295016.2.4.35 Level of Difficulty: 3 Importance Rating 4.0 Control Room Abandonment: Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

Question # 79 CGS is operating in Mode 1.

The shift manager directs a control room evacuation due to a chemical spill that makes the control room uninhabitable.

All immediate actions were completed before the control room crew evacuated the control room.

Supplemental actions are in progress outside the control room.

Using the reference provided, what actions should the CRS direct the equipment operators to perform?

The CRS should direct the equipment operators to A. trip RPS breakers to ensure that the reactor is scrammed, all MSIVs are closed, and DG-2 remains available.

B. open circuit breakers and remove control power fuses for condensate and condensate booster pumps to prevent uncontrolled RPV injection.

C. trip and then restart DG-2 to ensure that it is started in the correct sequence for control room evacuation due to a control room fire.

D. manually close breaker CB-8/DG2 to ensure DG-2 remains available while the control room is evacuated.

Answer: B K/A Match:

This question requires the candidate to demonstrate knowledge of equipment operator actions during control room evacuation and the effects that these field actions have on mitigation strategies.

Page 1 of 6 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is G statement and Page 2 of 6 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since these actions are performed during control room evacuation due to a fire, and other breakers are opened by equipment operators in attachment 7.7. However, a control room fire is not occurring and these actions are performed by a control room operator.

B. Correct. In accordance with ABN-CR-EVAC, attachment 7.7, when the control room is evacuated, and equipment operator will open condensate and condensate booster pump breakers to prevent RPV overfill following emergency depressurization.

C. Incorrect. Plausible since these actions are performed by an equipment operator during control room evacuation in accordance with ABN-CR-EVAC, attachment 7.5. However, they are only performed if the evacuation stems from a control room fire.

D. Incorrect. Plausible since these actions are performed following a control room evacuation in accordance with ABN-CR-EVAC, attachment 7.3, and other DG-2 actions are performed by equipment operators in attachment 7.5. However, they are performed by a control room operator only if the evacuation stems from a control room fire.

Technical Reference(s) Attached w/ Revision # See ABN-CR-EVAC, control Room Evacuation and Remote Cooldown Comments / Reference Proposed references to be provided during examination: ABN-CR-EVAC Flow Chart Learning Objective: 11272 - Knowledge of RO tasks performed outside the main control room during emergency operations including system geography and system implications.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 3 of 6 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-CR-EVAC, Flow Chart (Reference) Revision: 035 Page 4 of 6 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 5 of 6 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-CR-EVAC, Bases Revision: 035 Page 6 of 6 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 1 Group 1 K/A 295019 AA2.02 Level of Difficulty: 3 Importance Rating 3.7 Partial or Total Loss of Inst. Air: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INTRUMENT AIR: Status of safety-related instrument air system loads.

Question # 80 CGS is operating in Mode 1.

RHR-P-2A was declared inoperable due to an overcurrent trip at 1200, on 1/1/2017.

The crew entered LCO 3.5.1, condition A, when RHR-P-2A tripped.

At 1800, on 1/4/2017, an operator reported that ADS Division 1 Nitrogen Bank average cylinder pressure was 2050 psig and steady.

With no other operator actions, using the reference provided, what is the latest time that the reactor must be in Mode 3 according to technical specifications?

A. 0100 on 1/5/2017.

B. 0600 on 1/5/2017.

C. 0600 on 1/8/2017.

D. 0000 on 1/9/2017.

Answer: B K/A Match:

The question requires the candidate to demonstrate knowledge of the relationship between the ADS nitrogen banks (part of the Containment Instrument Air system), and operability of Safety Relief Valves (SRVs).

Page 1 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a G statement tied to 10CFR.55.43 and Page 2 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible for LCO 3.5.1, condition H (ie - two additional ECCS subsystems inoperable, OR HPCS and LPCS inoperable, OR HPCS and ADS valve inoperable, etc.). aHowever, only one non-HPCS ECCS subsystem is inoperable.

B. Correct. In accordance with TS 3.5.1 bases, an average pressure of 2200 psig is required in each ADS nitrogen bank to maintain operability. The division 1 nitrogen bank supplies 3 SRVs.

Therefore, 2 of the required 6 ADS SRVs is inoperable. LCO 3.5.1 action G.1 applies and the reactor must be in Mode 3 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from the discovery that the LCO is not met.

C. Incorrect. Plausible for LCO 3.5.1 action statements F and G.

D. Incorrect. Plausible for LCO 3.5.1, condition A and B, if it is believed that ADS SRVs are still operable with Containment Instrument Air (CIA) only. See distractor A and B explanations.

Technical Reference(s)

TS 3.5.1, ECCS - Operating Attached w/ Revision # See TS 3.5.1, ECCS - Operating, Bases Comments / Reference Proposed references to be provided during examination: TS 3.5.1, Actions table.

Learning Objective: LO7748 - Predict the plant impact that a loss or malfunction of the Containment Instrument Air System will have on the following: SRVs Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 3 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

TS 3.5.1, ECCS - Operating (Reference) Revision: 237 Page 4 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 5 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 6 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

TS 3.5.1, ECCS - Operating Bases Revision: 92 Page 7 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 8 of 8

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 1 Group 1 K/A 295021.2.1.19 Level of Difficulty: 3 Importance Rating 3.8 Loss of Shutdown Cooling: Ability to use plant computers to evaluate system or component status.

Question # 81 Page 1 of 10

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 CGS is in Mode 3.

RHR Loop B is in Shutdown Cooling (SDC).

RHR-P-2B is inadvertently secured.

3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months later, the crew is ready to restore SDC Using the Plant Computer display below, what is the procedurally-preferred method to restore SDC?

88 . 9 2.5 111.5 0.

The CRS should direct the crew to A. Place RHR Loop B in SDC in accordance with SOP-RHR-SDC, section 5.7, RHR Loop B Shutdown Cooling Quick Restart.

B. Place RHR Loop A in SDC in accordance with SOP-RHR-SDC, section 5.1, RHR Loop A Shutdown Cooling Initiation.

C. Place RHR Loop B in SDC in accordance with SOP-RHR-SDC, section 5.2, RHR Loop B Shutdown Cooling Initiation.

D. Establish SDC in accordance with ABN-RHR-SDC-ALT, section 7.1, Discharge Steam to the Main Condenser.

Answer: C Page 2 of 10

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 K/A Match:

This question requires the candidate to use information from the plant computer to evaluate the appropriate method to restore shutdown cooling.

SRO Only:

K/A is a G statement and Page 3 of 10

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

Page 4 of 10

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 A. Incorrect. Plausible since the quick start procedure could be used to restore SDC if the T between RRC inlet temperature and RHR-P-2B outlet temperature were < 80°F.

B. Incorrect. Plausible since RHR Loop A was not in SDC and section 5.1 of SOP-RHR-SDC would need to be completed to use this loop. However, step 4.1.6, states that it is preferred to start the SDC loop that was previously running.

C. Correct. Since the T between RRC inlet temperature and RHR-P-2B outlet temperature is > 80°F, SDC must be restored using the normal section of the procedure to warm up the RHR system.

D. Incorrect. Plausible if RPV pressure was > 30 psig or if the normal SDC suction valves are inoperable.

Technical Reference(s) Attached w/ Revision # See SOP-RHR-SDC, RHR Shutdown Cooling Comments / Reference ABN-RHR-SDC-ALT, Residual Heat Removal Alternate Shutdown Cooling ABN-RHR-SDC-LOSS, Loss of Shutdown Cooling Proposed references to be provided during examination: None Learning Objective: 11814 - Predict the impact of the following on the Residual Heat Removal System: b. Pump trips Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.5 Comments /

Reference:

SOP-RHR-SDC, RHR Shutdown Cooling Revision: Major 025 Minor 002 Page 5 of 10

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 6 of 10

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 7 of 10

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 8 of 10

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-RHR-SDC-ALT, Residual Heat Revision: 013 Removal Alternate Shutdown Cooling Comments /

Reference:

ABN-RHR-SDC-LOSS Revision: 006 Page 9 of 10

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 10 of 10

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 1 Group 1 K/A 700000.AA2.08 Level of Difficulty: 3 Importance Rating 4.4 Generator Voltage and Electric Grid Disturbances: Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Criteria to trip the turbine or reactor.

Question # 82 CGS is operating in Mode 1.

Operators notice main generator parameters are fluctuating.

Dittmer reports that there are no transmission system disturbances in progress.

The Power System Stabilizer (PSS) has been turned off.

The main generator voltage regulator is in TEST.

What actions should be taken?

The CRS will direct the operators to A. restore the PSS within 30 minutes if main generator fluctuations have not subsided.

B. scram the reactor if main generator power fluctuations reach 200 MWe peak to peak.

C. scram the reactor 3 minutes after main generator output voltage fluctuations reach 5 kv peak to peak.

D. restore the main generator voltage regulator to AUTO once parameters are within the Generator Capability Curve.

Answer: B K/A Match:

The question requires the candidate to determine the reactor scram requirements during main generator output fluctuations.

Page 1 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a G statement tied to 10CFR.55.43 and This question requires the candidate to identify that the actions of the abnormal procedure to reduce electric system fluctuations have been completed and were not successful, and that system Page 2 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 fluctuations have risen to the point where a turbine trip is required. Additionally, the candidate must know that current plant conditions require that the reactor is tripped prior to tripping the turbine.

Explanation:

A. Incorrect. Plausible since removing the PSS from service requires a report to Dittmer within 30 minutes. However, there is no requirement to restore the PSS to service with main generator fluctuations in progress.

B. Correct. In accordance with ABN-GENERATOR, step 4.6.1.c.6), the reactor should be scrammed and the main generator tripped if main generator output power fluctuations reach 200 MWe peak to peak. This is not an immediate action. The reactor is scrammed and the turbine is tripped after the PSS has been turned off and the voltage regulator is taken to test and the fluctuations have not subsided.

C. Incorrect. Plausible since there is a requirement to scram the reactor if main generator output power fluctuations are 100 MWe peak to peak sustained for GT 3 minutes and main generator output voltage may be fluctuating. However, there is no requirement to scram the reactor on high output voltage fluctuations.

D. Incorrect. Plausible since the voltage regulator is taken out of AUTO in an attempt to stabilize main generator output parameters. However, there is no direction to restore the voltage regulator to AUTO based on status of the Generator Capability curve.

Technical Reference(s) Attached w/ Revision # See ABN-GENERATOR, Main Generator Trouble Comments / Reference Proposed references to be provided during examination: None Learning Objective: LO5531: Describe the effect that a loss or malfunction of the following will have on the Main Generator - Voltage regulation.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 3 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-GENERATOR, section 4.6, Main Revision: Major 015 Minor 003 Generator Power Oscillations Page 4 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 5 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 6 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-GENERATOR, section 5.0, Bases for Revision: Major 015 Minor 003 section 4.6 Page 7 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/6/2017 Tier 1 Group 2 K/A 295007.2.4.21 Level of Difficulty: 2 Importance Rating 4.6 High Reactor Pressure: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Question # 83 Why is reactor steam dome pressure limited to 1325 psig?

This limit is established to specifically ensure that RCS pressure will not exceed (1) of design pressure (2) .

A. (1) 110%

(2) at the lowest elevation of the RCS B. (1) 110%

(2) in the steam dome C. (1) 125%

(2) at the lowest elevation of the RCS D. (1) 125%

(2) in the steam dome Answer: A K/A Match:

The question requires the candidate to understand the logic used to establish the steam dome pressure safety limit, which maintains RCS integrity.

Page 1 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a G statement tied to 10CFR.55.43 and Page 2 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. In accordance with TS bases for SL 2.1.2, the limit for reactor steam dome pressure ensures that pressure at the lowest point of the RCS is maintained below 110% of the design pressure of 1250 psig.

B. Incorrect. Plausible since the limit for reactor steam dome pressure is to ensures that pressure the highest RCS pressure point is maintained below 110% of the design pressure of 1250 psig.

However, the highest RCS pressure is at the lowest point in the RCS.

C. Incorrect. Plausible since the steam dome safety limit will limit pressure below design pressure at the lowest point in the RCS. However, pressure is limited to 110% of design pressure at this point.

The transient pressure limit for RCS piping and valves is 125% of design pressure. The safety limit is based on the most conservative limit, or 110% of design pressure.

D. Incorrect. Plausible The transient pressure limit for RCS piping and valves is 125% of design pressure. However the safety limit is based on the most limiting allowance, which is 110% of design pressure for the pressure vessel. Additionally, the limiting area is the lowest point in the RCS.

Technical Reference(s) Attached w/ Revision # See Technical Specification Bases, 2.1.2, Reactor Coolant System (RCS) Comments / Reference Pressure SL Proposed references to be provided during examination: None Learning Objective: 13427 Describe the bases for the Reactor Steam Dome Pressure Safety Limit.

[TS Bases] (SRO-only)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 3 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Technical Specification Bases, 2.1.2, Revision: 92 Reactor Coolant System (RCS) Pressure SL Page 4 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 5 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 1 Group 2 K/A 295010.AA2.03 Level of Difficulty: 3 Importance Rating 3.6 High Drywell Pressure: Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:

Drywell radiation levels.

Question # 84 CGS is operating in Mode 1.

An event occurs that requires entry into the Emergency Operating Procedures (EOPs).

Primary containment radiological conditions are greater than the ODCM RFO Offsite Radioactivity Release Limits.

Drywell pressure is 1.75 psig, up slow.

Drywell temperature is 200°F, up slow.

PC pressure is below the Primary Containment Pressure Limit (PCPL).

Under what conditions should the CRS consider venting Primary Containment (PC) through the Standby Gas Treatment (SGT) system?

The CRS should consider venting the PC A. if PC hydrogen is detected, to prevent an explosive gas mixture from developing in containment.

B. if emergency depressurization (ED) is required and less than 7 SRVs are available, to ensure RPV pressure is below Decay Heat Removal Pressure (DHRP).

C. if Wetwell level exceeds the SRV Tail Pipe Level Limit (SRVTPL), to minimize potential for Drywell floor failure.

D. if significant fuel damage is anticipated, to reduce the overall radiation release by venting before containment atmosphere becomes more contaminated.

Answer: D K/A Match:

This question requires the candidate to demonstrate understanding of the relationship between high drywell pressure and drywell radiological conditions and the strategies available to minimize off site radiation release.

Page 1 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is an A2 statement and Page 2 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since primary containment venting is allowed if hydrogen gas concentration is detected (> 0.6%). However, in this event, venting is not allowed if primary containment radiological conditions exceed the ODCM RFO Offsite Radioactivity Release Limits.

B. Incorrect. Plausible since this condition requires rapid RPV depressurization to reduce RPV to drywell differential pressure to allow RHR shutdown cooling. However, venting containment will exacerbate the d/p.

C. Incorrect. Plausible since this event requires the RPV to be depressurized via emergency depressurization. However, venting the primary containment will not mitigate this event.

D. Correct. In accordance with PPM 5.0.10, venting below the Primary Containment Pressure Limit may be appropriate to limit radioactivity release if significant fuel damage is anticipated. Reducing primary containment pressure while the primary containment atmosphere is still relatively clean increases the capacity of the containment to retain fission products. Later releases, after core damage has progressed, may thereby be avoided.

Technical Reference(s) Attached w/ Revision # See PPM 5.0.10, Flowchart Training Manual Comments / Reference PPM 5.2.1, Primary Containment Control Proposed references to be provided during examination: None Learning Objective:

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.4 Page 3 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.0.10 - Flowchart Training Manual Revision: Major 21 Minor 001 Page 4 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 5 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.2.1, Primary Containment Control, Revision: 23 Override P-4 Page 6 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.2.1, Primary Containment Control, Revision: 23 PC Gas Leg Page 7 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.2.1, Primary Containment Control, Revision: 23 Wetwell Level High Leg Page 8 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.1.3, Emergency RPV Revision: 20 Depressurization Page 9 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/9/2017 Tier 1 Group 2 K/A 295033.2.2.37 Level of Difficulty: 3 Importance Rating 4.6 High Secondary Containment Area Radiation Levels: Ability to determine operability and/or availability of safety related equipment.

Question # 85 What is the basis for the Maximum Safe Operating Values (MSOVs) for Reactor Building Area Radiation Monitors listed in Table 24 of PPM 5.3.1, Secondary Containment Control?

The max safe value is set A. low enough to require emergency depressurization prior to exceeding the SITE AREA EMERGENCY (SAE) emergency action level.

B. high enough to confirm that an emergency depressurization is required to prevent core damage.

C. low enough to allow time for mitigating actions without damaging safety related equipment due to high radiation exposure.

D. high enough to confirm primary system leakage into secondary containment, requiring a normal reactor shutdown.

Page 1 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Answer: C K/A Match:

This question requires the candidate to demonstrate an understanding of how equipment operability is determined during a high secondary containment event.

Page 2 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is an A2 statement and Page 3 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since a RB ARM greater than the max safe operating value (MSOV) is a threshold for a Site Area Emergency EAL. However, the timing of the EAL call is determined by the severity of the event, and when the crew recognizes that an EAL call is necessary. The MSOV setpoint is not determined to ensure that emergency depressurization is initiated prior to a SAE determination.

B. Incorrect. Plausible since emergency depressurization (ED) may be performed due to a RB ARM exceeding MSOV in two areas with a primary system discharging into the secondary containment.

However, exceeding the MSOV setpoint is indicative of a threat to one or more safety function, and requires that the parameter be controlled by ensuring the reactor is scrammed and PPM 5.1.1, RPV Control is entered to reduce the energy entering secondary containment. The MSOV setpoint itself was not selected to require an ED. For instance, if multiple locations are exceeding a MSOV in the same parameter but there were no primary system discharging into secondary containment, an ED would not be required (PPM 5.3.1, steps SC-13 through SC-15). Additionally, multiple parameters exceeding their MSOV in one location does not require ED. Another aspect that makes this distractor incorrect is that the distractor states that the MSOV setpoint is set high enough Setting the MSOV higher will bring the plant closer to core damage prior to an ED, which is less conservative.

C. Correct. The MSOV for RB ARMs is low enough to allow time for shutdown or isolation of a leak without exceeding the total integrated dose allowable for even the most sensitive safety related equipment.

D. Incorrect. Plausible since a normal reactor shutdown is performed based on parameters exceeding MSOV. However, normal shutdown is only performed when a parameter exceeds MSOV in two areas AND a primary system IS NOT discharging into secondary containment. See the discussion for distractor B.

Technical Reference(s) Attached w/ Revision # See PPM 5.0.10, Flowchart Training Manual Comments / Reference PPM 5.3.1, Secondary Containment Control Proposed references to be provided during examination: None Learning Objective: 8456 - Define Maximum Safe Operating Value for the following secondary containment parameters: c.Area radiation levels (PPM 5.3.1)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 Page 4 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 55.43 43.5 Comments /

Reference:

PPM 5.0.10 Revision: Major 21 Minor 001 Page 5 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.0.10 Revision: Major 21 Minor 001 Page 6 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.3.1 Revision: 20 Page 7 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/6/2017 Tier 2 Group 1 K/A 205000 A2.06 Level of Difficulty: 4 Importance Rating 3.5 Type the K/A System or Condition Here: Shutdown Cooling, Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: SDC/RHR pump trips.

Question # 86 CGS is operating in Mode 5.

The reactor was shutdown 10 days ago for a refueling outage.

Refueling operations are in progress. 15 fuel bundles have been removed from the core.

RHR Loop A is in shutdown cooling (SDC).

Reactor recirculation pumps are OOS for maintenance.

The main condenser is OOS and open for tube cleaning.

SW-P-1B has been placed OOS for corrective maintenance.

RHR-P-2A trips and cannot be restarted.

What actions should be taken?

The CRS should direct starting A. RHR-P-2B, with suction from the Spent Fuel Pool/FPC, and discharging through RHR-V-42B per SOP-FPC-ASSIST-ALT.

B. RHR-P-2C, with suction from the suppression pool and rejecting RPV water to the suppression pool per SOP-RHR-INJECTION.

C. RHR-P-2B, with suction from the Skimmer Surge Tanks and returning to the Fuel Pool Cooling system per SOP-FPC-ASSIST-ALT.

D. a condensate pump with suction from the condensate storage tank and rejecting RPV water to the suppression pool per ABN-RHR-SDC-ALT.

Answer: C K/A Match:

This question requires the candidate to demonstrate an understanding of the impact on the shutdown cooling system on a loss of RHR heat exchanger cooling, and the method used to mitigate this loss by validating an alternate shutdown cooling method is available.

Page 1 of 4

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is an A2 statement and Page 2 of 4

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since this is an allowed method of alternate shutdown cooling listed in ABN-RHR-SDC-LOSS. However, this method can only be used if the reactor has been shutdown > 13 days and should not be used if fuel is removed from the core to preclude possible damage to core instrumentation.

B. Incorrect. Plausible since this is an allowed method of alternate shutdown cooling listed in ABN-RHR-SDC-LOSS. However, this method requires a functional suppression pool cooling system in service, which is not available.

C. Correct. With the given plant conditions, alternate shutdown cooling is provided by circulating water between the RPV and spent fuel pool via RHR-P-2B.

D. Incorrect. Plausible since this is an allowed method of alternate shutdown cooling listed in ABN-RHR-SDC-LOSS. However, this method requires a functional suppression pool cooling system in service, which is not available.

Technical Reference(s)

ABN-RHR-SDC-LOSS, Loss of Shutdown Cooling Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: 11814 - Predict the impact of the following on the Residual Heat Removal System: b. Pump trips Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.7 Page 3 of 4

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-RHR-SDC-LOSS, section 4.8 Revision: 006 Page 4 of 4

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/6/2017 Tier 2 Group 1 K/A 209001.2.4.9 Level of Difficulty: 3 Importance Rating 4.2 LPCS: Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

Question # 87 CGS is operating in Mode 3.

A LOCA occurred requiring entry into the EOPs.

An Emergency Depressurization (ED) has been initiated due to low RPV level.

Current plant conditions:

RPV level: -200 inches, down slow

RPV pressure: 40 psig, down slow How should adequate core cooling be achieved?

A. Inject with RCIC to raise RPV level > -161 inches.

B. Establish RPV injection using LPCS at 6000 gpm.

C. Flood containment using all available injection sources.

D. Stabilize RPV pressure to allow Steam Cooling without injection.

Answer: B K/A Match:

This question requires the candidate to demonstrate knowledge of the conditions when LPCS is required to mitigate effects of a Loss of Coolant Accident from a shutdown condition.

Page 1 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a G statement and Page 2 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since restoring RPV level above above -161 inches, which is Top of Active Fuel (TAF), will ensure adequate core cooling is maintained. However, with the given conditions in the question stem, there is insufficient steam pressure to operate the RCIC turbine.

B. Correct. With RPV level below -198 inches but above -210 inches, injecting with HPCS OR LPCS with a flow rate 6000 gpm will maintain adequate core cooling.

C. Incorrect. Plausible since flooding containment to provide core cooling is the strategy used in the Severe Accident Guidelines (SAGs). However, this strategy does not ensure adequate core cooling. Additionally, the SAGs are not entered until ALL EOP mitigation strategies are ineffective in maintaining core cooling.

D. Incorrect. Plausible since Steam Cooling is an effective method for adequate core cooling.

However, Steam Cooling is only effective when RPV level is -198 inches.

Technical Reference(s) Attached w/ Revision # See PPM 5.0.10, Flowchart Training Manual Comments / Reference PPM 5.1.1, RPV Control, RPV Level Leg Proposed references to be provided during examination: None Learning Objective: LO8041 - List the four methods used to provide adequate core cooling in the EOPs.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 3 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.0.10, Flowchart Training Manual Revision: 020 Page 4 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 5 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.7.1, RPV and Primary Revision: Major 006 Minor 001 Containment Flooding SAG Page 6 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.1.1, RPV Control, RPV Level Leg Revision: 21 (Post Emergency Depressurization)

Page 7 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 0 Date: Tier 2 Group 1 K/A 215003.A2.05 Level of Difficulty: 3 Importance Rating 3.5 IRM: Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Faulty or erratic operation of detectors/system.

Question # 88 CGS is operating in Mode 2.

A reactor startup is in progress in accordance with PPM 3.1.2, Startup Flow Chart.

Current Conditions:

IRM-B is bypassed due to a failed power supply. Repair parts are on order.

All other IRM channels are on range 10 The crew is verifying IRM/APRM overlap when IRM channel D fails.

Maintenance estimates 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months to repair IRM channel D.

Using the provided reference, what is the earliest permitted action that will satisfy all technical specifications requirements (Assume the action is taken at the time listed below)?

A. Place the reactor mode switch in Run at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

B. Place IRM-D mode switch in Standby at 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

C. Repair IRM-D and return it to operable at within 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months .

D. Place the reactor mode switch in Shutdown at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Answer: A K/A Match:

The question requires the candidate to understand that the IRM failure cause entry into a technical specification Limiting Condition for Operation, and identify the actions required to meet technical specification requirements.

Page 1 of 9 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is an A2 statement and Page 2 of 9 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Placing the reactor mode switch in Run puts the reactor in Mode 1, which is outside the mode of applicability of TS 3.3.1.1 for IRMs. Even though IRM/APRM overlap cannot be completed with two inoperable IRM channels on a single RPS system, this overlap is ONLY required to be met during entry into Mode 2 from Mode 1 (reactor shutdown).

B. Incorrect. Plausible since placing the IRM mode switch in Standby inserts a trip signal and meets LCO 3.3.1.1, condition A requirements to trip the failed channel. However, it is not the earliest action listed as required by the question stem.

C. Incorrect. Plausible since returning IRM-D to operable within 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months places the plant outside the mode of applicability for LCO 3.3.1.1 prior to a required plant shutdown per LCO 3.3.1.1, condition G. However, it is not the earliest action listed as required by the question stem.

D. Incorrect. Placing the reactor mode switch in Shutdown puts the reactor in Mode 3, which is outside the mode of applicability of TS 3.3.1.1 for IRMs. However, it is not the earliest action listed as required by the question stem.

Technical Reference(s) Attached w/ Revision # See CGS Technical Specifications Comments / Reference CGS Technical Specifications Bases Proposed references to be provided during examination: TS 3.3.1.1, Actions Table and table 3.3.1.1-1 (Intermediate Range portion without allowable values)

Learning Objective: 7637 - Predict the effects that a failure of the IRM system will have on the following: RPS Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 43.2 Page 3 of 9 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

CGS Technical Specifications (Proposed Revision: 237 Reference)

Page 4 of 9 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 5 of 9 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 6 of 9 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 7 of 9 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 8 of 9 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

CGS Technical Specifications Bases Revision: 92 Page 9 of 9 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 2 Group 1 K/A 218000 2.4.6 Level of Difficulty: 3 Importance Rating 4.7 ADS: Knowledge of EOP mitigation strategies.

Question # 89 CGS is operating in Mode 1.

A LOCA occurs requiring the crew to enter the EOPs.

The reactor mode switch is in shutdown.

Reactor power is 10% and steady.

RPV level: -127 inches, down slow.

MSIVs are open.

Why is ADS inhibited?

ADS is inhibited to prevent A. adding positive reactivity from LPCI and LPCS injection.

B. loss of adequate core cooling by depressurizing the RPV.

C. a severe thermal transient from increased HPCS injection rate.

D. impacting RPV level recovery by degrading RCIC operation.

Answer: A K/A Match:

The question requires the candidate to demonstrate knowledge of the reasons for inhibiting ADS in the EOPs, specifically PPM 5.1.2, RPV Control - ATWS.

Page 1 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a G statement and Page 2 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. In an ATWS condition, ADS is inhibited to prevent positive reactivity addition via an uncontrolled addition of cold, unborated water from low pressure injection sources (LPCI or LPCS).

B. Incorrect. Plausible since depressurization with RPV level below Top of Active Fuel (TAF) could cause a loss of adequate core cooling if RCIC were the only available injection source. However, the conditions given in the stem do not support this.

C. Incorrect. Plausible since ADS may cause a severe thermal transient. However, the transient is not from an increase in HPCS injection, but rather from uncontrolled low pressure injection (LPCI/LPCS). Additionally, during an ATWS, HPCS is manually secured.

D. Incorrect. Plausible since RCIC operation is degraded if RPV pressure is reduced by ADS.

However, in this situation, RPV level is intentionally lowered to assist in reducing reactor power.

Technical Reference(s)

PPM 5.0.10, Flowchart Training Manual Attached w/ Revision # See Comments / Reference PPM 5.1.2, RPV Control - ATWS Proposed references to be provided during examination: None Learning Objective: LO8089 - Given a list, identify the statement that describes the two reasons that ADS is inhibited during an ATWS Question Source: Bank #

Modified Bank # LO00098 (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 3 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.0.10, Flowchart Training Manual Revision: Major 21 Minor 001 Page 4 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.1.2, RPV Control - ATWS Revision: 24 Comments /

Reference:

Bank Question LO00098 Revision: N/A The first step in PPM 5.1.2, RPV Control - ATWS is to Inhibit ADS Which of the following is the basis for this action?

A. This action prevents unwanted depressurization of the RPV which could result in uncontrolled injection from unborated water causing a power excursion.

B. The ADS system is not analyzed for ATWS conditions, more controlled pressure reductions are performed manually when conditions in the EOPs direct it.

C. This action prevents unwanted depressurization of the RPV which could result in uncontrolled reactivity addition due to addition of voids in the core.

D. The ADS system is not analyzed for ATWS conditions. Auto actuation at the ADS setpoint would be premature for ATWS conditions.

Correct Answer: A Page 5 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/9/2017 Tier 2 Group 1 K/A 400000.A2.01 Level of Difficulty: Importance Rating 3.4 Component Cooling Water: Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Loss of CCW pump.

Question # 90 CGS is operating in Mode 1.

RCC-P-1A and RCC-P-1B are running.

RCC-P-1B tripped.

RCC-P-1B and RCC-P-1C cannot be started.

Troubleshooting activities are in progress on both pumps.

Current plant conditions:

RRC-P-1A stator temperature is 190F, up slow.

Fuel Pool Cooling (FPC) HX outlet temperature is 100F, up slow.

Drywell pressure is 0.26 psig, up slow.

Field operators report that CRD-P-1A (running CRD pump), lube oil housing is hot to the touch.

What is the next action that should be taken?

The CRS should direct the crew to A. trip RRC-P-1A and enter ABN-RRC, Loss of Reactor Recirculation Flow.

B. start CRD-P-1B and secure CRD-P-1A per SOP-CRD-PUMPS, Control Rod Drive System Pump Operations.

C. transfer the FPC system cooling to Service Water per SOP-FPC-OPS, Fuel Pool Cooling and Cleanup Operations.

D. vent the drywell per SOP-CN-CONT-VENT, Containment Vent, Deinert, Purge, and Ventilating.

Answer: B K/A Match:

This question requires the candidate to demonstrate knowledge of procedures and actions that should be implemented to address abnormal plant parameters following a partial loss of RCC.

Page 1 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is an A2 statement and Page 2 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since ABN-RCC, step 4.2.4 directs operators to monitor RRC pump parameters, RRC-P-1A stator temperature has risen approximately 50F, and if the temperature continues to rise, the pump should be tripped per Alarm Response Procedure 602.A6.6-6.

However, the pump is not required to be tripped until stator temperature reaches 266F, and the CRD pump is a more immediate issue.

B. Correct. With reduced RCC flow, ABN-RCC, step 4.2.6 directs CRD pump cooling be transferred to the Condensate Transfer header OR CRD pumps be periodically shifted if CRD pump lube oil housing temperature is hot to the touch.

C. Incorrect. Plausible since a FP temperature has risen approximately 10F. However, ABN-RCC, step 4.2.11 directs transferring FPC cooling to SW is FP temperature cannot be maintained 125F, and the CRD pump is a more immediate issue.

D. Incorrect. Plausible since ABN-RCC, step 4.2.8 directs the drywell to be vented if pressure is rising due to inadequate RCC cooling. However, ABN subsequent operator actions should be performed in order and addressing the CRD pumps is in step 4.2.6. Additionally, drywell pressure is still within the normal band of 0.25 psig to 0.75 psig (see SOP-CN-CONT-VENT, 4.16), with pressure slowly rising. The CRD pump is a more immediate concern.

Technical Reference(s)

ABN-RCC, Loss of RCC Alarm Response Procedure 4.602.A6.6-6 SOP-CN-CONT-VENT, Containment Vent, Deinert, Purge, and Attached w/ Revision # See Ventilating Comments / Reference PPM 1.3.1, Operating Policies, Programs and Practices OI-09, Operations Standards and Expectation SWP-PRO-01, Procedure and Work Instruction Use and Adherence Proposed references to be provided during examination: None Learning Objective: LO5706 - Explain the interlocks associated with the following components or system conditions, including setpoints: a. RCC pump auto start, b. RCC Pump trips.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.6 Page 3 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-RCC, Loss of RCC Revision: Major 006 Minor 003 Page 4 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

ABN-RCC - Bases Revision: Major 006 Minor 003 Page 5 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Alarm Response Procedure 4.602.A6.6-6 Revision: 032 Page 6 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-CN-CONT-VENT Revision: 025 Page 7 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

OI-09 Revision: 064 Comments /

Reference:

PPM 1.3.1 Revision: 120 mr 005 Page 8 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SWP-PRO-01 Revision: 030 Page 9 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 2 Group 2 K/A 219000.A2.13 Level of Difficulty: 4 Importance Rating 3.7 RHR/LPCI: Torus/Pool Cooling Mode: Ability to (a) predict the impacts of the following on the RHR/LPCI:

TORUS/SUPPRESSION POOL COOLING MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High suppression pool temperature.

Question # 91 CGS is operating in Mode 4.

Reactor coolant temperature is 165°F.

RPV time to boil is 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

RRC-P-1A is in operation.

RHR Loop B is in shutdown cooling (SDC).

RHR Loops A and C are in standby.

Annunciator 601.A11.1-3, DRYWELL/SUPPRESSION POOL TEMPERATURE HIGH, is in alarm due to high suppression pool temperature.

Suppression pool temperatures:

CMS-TR-5, point 220, Average Supp Pool Temp (Upper Level): 88.7 AVE °F

CMS-TR-5, point A02, Average Supp Pool Temp (Lower Level): 82.2 AVE °F

CMS-TR-6, point 220, Average Supp Pool Temp (Upper Level): 87.2 AVE °F

CMS-TR-6, point A02, Average Supp Pool Temp (Lower Level): 81.5 AVE °F What actions should be taken?

The CRS should direct the crew to place A. RHR Loop A in suppression pool cooling per SOP-RHR-SPC, Suppression Pool Cooling/Spray/Discharge/Mixing, and maintain RHR Loop B in SDC.

B. RHR Loop B in suppression pool cooling per SOP-RHR-SPC, Suppression Pool Cooling/Spray/Discharge/Mixing, and place RHR Loop A in SDC.

C. RHR Loop C in suppression pool mixing per SOP-RHR-SPC, Suppression Pool Cooling/Spray/Discharge/Mixing, and maintain RHR Loop B in SDC.

D. LPCS system in suppression pool mixing per SOP-LPCS-SP, LPCS Suppression Pool Mixing, and maintain RHR Loop B in SDC.

Answer: A Page 1 of 6

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 K/A Match:

This question requires the candidate to demonstrate an understanding of methods available to lower suppression pool temperature, including the impact that these methods have on RHR system operability, and select the correct method for given plant conditions.

Page 2 of 6

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is an A2 statement and Page 3 of 6

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. Suppression pool bulk temperature is > 85°F. Therefore, suppression pool mixing will not lower bulk suppression pool temperature below the alarm setpoint.

B. Incorrect. Plausible since RHR Loop B may be used to lower suppression pool temperature.

However, RHR Loop A should not be placed in SDC with RRC-P-1A running.

C. Incorrect. Plausible since placing RHR Loop C in suppression pool mixing could reduce suppression pool temperature by reducing stratification. However, suppression pool bulk temperature is > 85°F and suppression pool mixing should not be used.

D. Incorrect. Plausible since placing LPCS in suppression pool mixing is allowed for the given plant conditions. However, suppression pool bulk temperature is > 85°F and suppression pool mixing should not be used.

Technical Reference(s) Attached w/ Revision # See SOP-RHR-SPC, Suppression Pool Cooling/Spray/Discharge/Mixing Comments / Reference SOP-RHR-SDC, RHR Shutdown Cooling Proposed references to be provided during examination: None Learning Objective: 11808 Describe the effect that a loss or malfunction of the Suppression Pool Cooling mode of the Residual Heat Removal System will have on suppression pool temperature control.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 4 of 6

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-RHR-SPC Revision: Major 008 Minor 003 Page 5 of 6

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-RHR-SDC Revision: Major 025 Minor 002 Page 6 of 6

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/6/2017 Tier 2 Group 2 K/A 234000 A3.02 Level of Difficulty: 3 Importance Rating 3.7 Ability to monitor automatic operation of the FUEL HANDLING EQUIPMENT including: Interlock operation Question # 92 CGS is in Mode 5.

Core refueling is in progress.

The refueling bridge crane main hoist grapple has been lowered to the grapple position for a fuel assembly in the core.

The Grapple Engage/Release switch has been toggled to the ENGAGE position.

The GRAPPLE ENGAGED light on the Right-Hand Controller is illuminated.

Prior to raising the fuel assembly, what action is required to verify that the grapple is properly engaged per SOP-REFUEL-OPS, Refueling Bridge Operation?

Verify the A. HOIST LOADED light on the Left-Hand Controller is illuminated.

B. back and forth rotation of mast is consistent with a seated fuel assembly.

C. SLACK CABLE light on the Left-Hand Controller is extinguished.

D. grapple is at the correct position by checking the tape markers on the main hoist cable.

Answer: B K/A Match:

The question requires the candidates to demonstrate knowledge of the procedure required to verify that the MAIN HOIST FUEL LOADED interlock will be satisfied after grappling a fuel bundle.

Page 1 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a Tier 2, Group 2 selection related to fuel handling facilities and procedures and Page 2 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since the HOIST LOADED light should be illuminated while removing the fuel from the core (step 5.2.14). However, the light will not illuminate when verifying that grapple is engaged while the grapple is resting on the fuel cell.

B. Correct. In addition to verifying that the GRAPPLE ENGAGED light is illuminated OR that the grapple is fully engaged by visual inspection, operators are required to verify back and forth rotation of mast limited to that allowed by seated fuel assembly.

C. Incorrect. Plausible since verifying the SLACK CABLE light is required to be verified prior to engaging the grapple (step 5.2.9). However, it is not an indication used to verify proper grapple engagement prior to lifting the fuel assembly.

D. Incorrect. Plausible since tape markers may be used on the Frame Mounted Hoist and the Monorail Mounted Hoist. However, it is not acceptable to use tape on the cable of the Main Hoist.

Technical Reference(s)

SD000207, Fuel Handling System Description Attached w/ Revision # See SOP-REFUEL-OPS, Refueling Bridge Operation Comments / Reference Proposed references to be provided during examination: None Learning Objective: 8968 - Discuss the indications that should be verified to ensure a bundle is properly grappled.

Question Source: Bank #

Modified Bank # LO01485 (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.7 Page 3 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SD000207, Fuel Handling System Revision: Major 013 Minor 002 Description Page 4 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-REFUEL-OPS, step 5.2.14 Revision: 009 Page 5 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-REFUEL-OPS, step 5.2.11 Revision: 009 Page 6 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-REFUEL-OPS, step 5.2.9 Revision: 009 Page 7 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SOP-REFUEL-OPS, Precautions Revision: 009 Page 8 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Bank Question LO01485, Revision: Original Procedurally, which of the following provides adequate demonstration that a fuel assembly is properly engaged?

A. Grapple elevation reading of 552" B. Slack Cable light is illuminated C. The grapple Engage Switch is in the 'ENGAGE' position D. Grapple Engaged light is illuminated E. Visual observation of the grapple jaws fully engaged F. Back and forth rotation of the mast A. D and F B. C and D C. A, C, and F D. B, D, and E Answer: A Page 9 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 2 Group 2 K/A 256000.2.4.47 Level of Difficulty: 3 Importance Rating 4.2 Reactor Condensate: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Question # 93 CGS is operating in Mode 1. Reactor power is 50%.

Primary chemistry parameters for the past 2 days are as follows:

Condensate Reactor Water Reactor Water Feedwater Demineralizer Inlet Conductivity Sulfate Conductivity (CDI) Conductivity Date Time (µS/cm) (ppb) (µS/cm) (µS/cm) 1/1/2017 1200 0.10 4 .055 0.03 1/1/2017 1800 0.23 4 .056 0.04 1/2/2017 0000 0.33 4 .058 0.08 1/2/2017 0600 0.37 7 .060 0.12 1/2/2017 1200 0.40 9 .055 0.25 1/2/2017 1800 0.52 11 .057 0.65 1/3/2017 0000 0.57 12 .059 0.75 1/3/2017 0600 0.68 14 .062 0.88 1/3/2017 1200 0.70 14 .059 0.92 Using the reference provided, and based on the trend information, what is the next Primary Chemistry Action Level (PCAL) to be entered?

Enter PCAL (1) based on (2) .

A. (1) 2 (2) reactor water conductivity B. (1) 1 (2) feedwater conductivity C. (1) 2 (2) reactor water sulfate D. (1) 3 (2) condensate demineralizer inlet conductivity Answer: D K/A Match:

This question requires the candidate to evaluate conductivity trends and determine required actions.

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ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a G statement and Page 2 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since Primary Chemistry Action Level (PCAL) 2 for reactor water conductivity is > 1.0 µS/cm. However, trend data shows that Condensate Demineralizer Inlet (CDI) conductivity will reach PCAL 2 first.

B. Incorrect. Plausible since feedwater conductivity is close to its PCAL 1 limit of .065 µS/cm.

However, trend data shows that Condensate Demineralizer Inlet (CDI) conductivity will reach PCAL 2 first.

C. Incorrect. Plausible since PCAL 2 for reactor water sulfate is >20 ppb and this parameter will reach PCAL 2 in about 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . However, trend data shows that Condensate Demineralizer Inlet (CDI) conductivity will reach PCAL 2 in less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D. Correct. Trend data shows that Condensate Demineralizer Inlet (CDI) conductivity will reach PCAL 3 in less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , while the other parameters will take at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to reach their next PCAL.

Technical Reference(s) Attached w/ Revision # See SWP-CHE-02, Chemical Process Management and Control Comments / Reference Proposed references to be provided during examination: SWP-CHE-02, tables 6.1.13, 6.1.15 Learning Objective:

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 3 of 9

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SWP-CHE-02 Revision: 026 Page 4 of 9

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ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 6 of 9

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ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 5 Date: 2/9/2017 Tier 3 Group 1 K/A 2.1.9 Level of Difficulty: 2 Importance Rating 4.5 Ability to direct personnel activities inside the control room.

Question # 94 When may control room operators perform actions that deviate from technical specifications or license conditions per 10CFR50.54.(x)?

If no adequate means of protecting public health and safety in an emergency is apparent consistent with tech specs A. , OR if failing to deviate from technical specifications will result in unnecessary emergency equipment damage, with operations manager and NRC approval at a minimum.

B. , OR if failing to deviate from tech specs will result in unnecessary emergency equipment damage, with CRS or SM approval at a minimum.

C. ONLY, with Ops Manager and NRC approval at a minimum.

D. ONLY, with CRS or SM approval at a minimum.

Answer: D K/A Match:

This question requires the candidate to demonstrate when the CRS may direct control room operators during conditions requiring actions per 10 CFR 50.54X.

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ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a G statement and Page 2 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since actions should be taken to prevent equipment damage, such as a reactor scram. However, this is not a reason to take actions outside technical specifications per 10 CFR 50.54X. Additionally, the minimum approval for invoking 10CFR50.54X is CRS/SM.

B. Incorrect. Plausible since the minimum approval for invoking 10CFR50.54X is CRS/SM , and actions should be taken to prevent equipment damage, such as a reactor scram. However, 10CFR50.54X is invoked to protect the health and safety of the public, not to prevent equipment damage.

C. Incorrect. Plausible since 10CFR50.54X is invoked to protect the health and safety of the public only. However, the minimum approval for invoking 10CFR50.54X is CRS/SM.

D. Correct. Though considered the exception, reasonable action that departs from Technical Specifications and licensing conditions are permitted per 10 CFR 50.54(x) provided an emergency exists and such action is immediately needed to protect the health and safety of the public when no adequate or equivalent means of protection consistent with Technical Specifications or License Conditions are apparent. Although other individuals/entities should be contacted prior to invoking 10CFR50.54X, such as the NRC, operations manager, etc., the minimum approval for invoking 10CFR50.54X is CRS/SM.

Technical Reference(s) Attached w/ Revision # See PPM 1.3.1, Operating Policies, Programs and Practices Comments / Reference Proposed references to be provided during examination: None Learning Objective: 6073 - State when deviations from Technical Specifications and Licensing conditions are permitted per 10CFR50.54X. [PPM 1.3.1]

Question Source: Bank # LX00897 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 43.1 Page 3 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 1.3.1, section 4.3 Revision: Major 120 Minor 005 Page 4 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 1.3.1, section 4.8.3 Revision: Major 120 Minor 005 Page 5 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 3 Group 1 K/A 2.1.23 Level of Difficulty: 3 Importance Rating 4.4 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Question # 95 CGS is operating in Mode 3.

A normal plant shutdown is in progress per PPM 3.2.1, Normal Plant Shutdown.

RPV pressure is 355 psig.

DEH is lowering RPV pressure 5 psig/min.

The crew notes evidence of level notching on RFW-LI-606A, Narrow Range Reactor Water Level.

The other 3 RPV narrow range level instruments are tracking normally.

How should the CRS control the plant cooldown?

A. Reduce the RPV depressurization rate per PPM 3.2.1, section 5.6, RPV Depressurization, to allow the condensing pot to refill the reference leg.

B. Maintain the RPV depressurization rate and lineup for reference leg backfill per PPM 10.27.64, Continuous Backfill System Operation.

C. Stop the RPV depressurization and perform an operability evaluation per PPM 3.2.1, attachment 7.1, RPV Depressurization with Evidence of Notching.

D. Stop the RPV depressurization and declare the affected water level trip functions inoperable per TS 3.3.1.1, Reactor Protection System Instrumentation.

Answer: C K/A Match:

This questions requires the candidate to understand how to proceed with an integrated plant procedure (normal plant shutdown) with an abnormal condition (notching) that requires the use of an attachment.

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ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a G statement tied to 10CFR.55.43 and Page 2 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since slowing the RPV depressurization rate would assist in refilling the reference leg. However, attachment 7.1, step 7.3.1 requires operators to stop RPV depressurization and evaluate instrument operability.

B. Incorrect. Plausible since lining up continuous backfill is an approved method to recover the affected instrument. However, attachment 7.1, step 7.3.1 requires operators to stop RPV depressurization and evaluate instrument operability.

C. Correct. With evidence of level instrument notching, attachment 7.1, step 7.3.1 requires operators to stop RPV depressurization and evaluate instrument operability.

D. Incorrect. Plausible since attachment 7.1, step 7.3.1 requires operators to stop RPV depressurization and evaluate instrument operability with evidence of level notching. However, TS 3.3.1.1 does not apply in the stated mode of operation.

Technical Reference(s) Attached w/ Revision # See PPM 3.2.1, Normal Plant Shutdown Comments / Reference CGS Technical Specifications, LCO 3.3.1.1, Reactor Protection System Instrumentation.

Proposed references to be provided during examination: None Learning Objective: 9875 Given appropriate conditions, indications and copies of plant procedures, determine the operator actions necessary to mitigate a plant transient from an analysis of plant conditions. [Plant Shutdown]

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 3 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 3.2.1, attachment 7.1, RPV Revision: Major 083 Minor 001 Depressurization with Evidence of Notching Page 4 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 5 of 7

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ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

TS LCO 3.3.1.1, Applicability Revision: 21 Page 7 of 7

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 3 Group K/A 2.2.11 Level of Difficulty: 4 Importance Rating 3.3 Knowledge of the process for controlling temporary design changes.

Question # 96 A Temporary Modification (TM) has been approved for installation.

The TM includes jumper installation in the control room.

Maintenance personnel are in the control room, ready to install the jumper.

Which of the following list the requirements for TM installation in accordance with PPM 1.3.9, Temporary Modifications?

Prior to installing the jumper, the CRS should verify that (1) verification will be used during jumper installation. Top tier drawings are updated (2) jumper installation.

A. (1) independent (2) before B. (1) simultaneous (2) before C. (1) independent (2) after D. (1) simultaneous (2) after Answer: D K/A Match:

The question requires the candidate to demonstrate knowledge of SRO responsibilities when installing a temporary modification.

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ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a G statement tied to 10CFR.55.43 and Page 2 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since independent verification is a method authorized for use by STANDARD-01, Worker Error Prevention Tools. This method could be used if significant radiation exposure was likely during TM installation. Additionally, updating drawings prior to TM installation will ensure that the drawings are correct prior to equipment configuration changes. However, per PPM 1.3.9, Temporary Modifications, simultaneous verification is required since incorrectly installing the jumper could lead to immediate and possibly irreversible harm to the plant or personnel. Since the question stem states that the installation is in the control room, significant radiation exposure is not likely.

PPM 1.3.9 directs that drawings are updated after TM installation.

B. Incorrect. Plausible since simultaneous verification is the correct method specified by PPM 1.3.9.

However, per PPM 1.3.9, drawings are update after the TM is installed. See distractor A discussion.

C. Incorrect. Plausible since drawings are updated after the TM is installed. However, simultaneous verification is required for jumper installation. See distractor A discussion.

D. Correct. Per PPM 1.3.9, step 4.2.1.d, the CRS/SM should verify that provisions have been made to use simultaneous verification. Step 4.2.7 directs the responsible engineer to process drawings changes after the TM installation is complete (WO in the FINISHED condition). See distractor A discussion.

Technical Reference(s)

A PPM 1.3.9, Temporary Modifications Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: 11251 - Knowledge of the process for making field changes Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 43.3 Page 3 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 1.3.9, Temporary Modifications Revision: Major 53, Minor N/A Page 4 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 5 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 3 Group K/A 2.2.20 Level of Difficulty: Importance Rating 3.8 Knowledge of the process for managing troubleshooting activities.

Question # 97 When should the Outage Control Center (OCC) be activated to assist in troubleshooting an equipment fault in accordance with SWP-MAI-03, Emergent Issue Management?

The OCC should be activated A. when the fault causes entry into a LCO action statement of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or less.

B. after any reactor power reduction of 5%.

C. on an emergency classification of Alert or higher.

D. when the Engineering VP determines that it is warranted.

Answer: A K/A Match:

The question requires the candidate to demonstrate knowledge of criteria for activating additional resources for troubleshooting activities.

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ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

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ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Correct. In accordance with SWP-MAI-03, the OCC is activated when a LCO action statement of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is entered.

B. Incorrect. Plausible since the OCC is activated for significant unplanned power derates. However, SWP-MAI-03 defines significant derates as > 10% power.

C. Incorrect. Plausible since the OCC may be activated on an emergency classification level of Unusual Event. It is not activated on a higher classification since the Emergency Response Organization (ERO) will be activated instead. If activated, the ERO will assume the responsibilities of the OCC until the ERO is deactivated.

D. Incorrect. Plausible since the OCC may be activated during engineering events and the engineering VP participates in management discussions leading to OCC activation. However, activation of the OCC is determined by the shift manager, work week manager, or duty manager, and approved by the plant general manager.

Technical Reference(s)

SWP-MAI-03, Emergent Issue Management Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: LO6188 - Identify concurrences needed for troubleshooting activities that are high risk to plant operations.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 3 of 6

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

SWP-MAI-03, emergent Issue Revision: 013 Management Page 4 of 6

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ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 3 Group 3 K/A 2.3.11 Level of Difficulty: 2 Importance Rating 4.3 Ability to control radiation release.

Question # 98 PPM 5.4.1 Radioactivity Release Control requires Emergency Depressurization if the exclusion area boundary release rate approaches or exceeds the General Emergency limit.

Which of the following describes the specific reason for this requirement per PPM 5.0.10, Flowchart Training Manual?

Emergency Depressurization A. allows the containment vent path (MSIVs) to open and vent the primary system to the condenser.

B. allows low pressure systems to inject into the reactor, limiting the release to the environment.

C. reduces the flow of primary systems that are unisolated and discharging outside of containment.

D. slows the rate of fuel damage in the reactor core and heat addition to the primary system.

Answer: C K/A Match:

The question requires that the candidate understand when an emergency depressurization is used to mitigate an off site radiation release based on emergency action levels.

SRO Only:

K/A is a G statement tied to 10CFR.55.43 and Page 1 of 4 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 2 of 4 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since venting the primary to the condenser would be desirable to reduce radiation release. However, emergency depressurization is not a prerequisite to reopening MSIVs.

B. Incorrect. Plausible since this is the result of performing an emergency depressurization (ED),

especially during a low RPV level condition. However, the question is asking for the reason to perform an ED during a radioactive release that threatens to reach the General Emergency EAL level. At this point, if a primary system is discharging outside the primary and secondary containment, an ED is performed independent of other parameters, such as RPV level and pressure, primary and secondary containment parameters, etc. Although allowing low pressure injection sources to inject into the RPV may, under specific circumstances, eventually reduce offsite dose, with the conditions specified in the stem, an ED is performed to immediately reduce RPV pressure to minimize the flow of primary systems outside containment, thereby immediately reducing the rate of release to the general population.

C. Correct. RPV depressurization places the primary system in the lowest possible energy state and reduces the driving head and flow of primary systems that are unisolated and discharging outside of containment. This immediately reduces the rate of release to the general population.

D. Incorrect. Plausible since emergency depressurization may slow the rate of fuel damage under specific condtions. However, this will not necessarily immediately reduce off site radiation release.

See explanation for distractor B.

Technical Reference(s)

PPM 5.0.10, Flowchart Training Manual Attached w/ Revision # See PPM 5.4.1, Secondary Containment Control Comments / Reference Proposed references to be provided during examination: None Learning Objective: 8481 - Identify the statement that describes the two reasons for emergency depressurizing the RPV if radioactivity release rates approach or exceed the General Emergency level and a primary system is discharging reactor coolant outside primary and secondary containment. (PPM 5.4.1)

Question Source: Bank # LO00181 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 43.4 Page 3 of 4 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 5.0.10, Flowchart Training Manual Revision: Major 21 Minor 001 Page 4 of 4 Written Exam Question Single Question Template

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 1 Date: 2/6/2017 Tier 3 Group K/A 2.4.29 Level of Difficulty: 3 Importance Rating 4.4 Knowledge of the emergency plan.

Question # 99 CGS is operating in Mode 1.

The shift manager declared an Unusual Event at 1525 due to a casualty.

The casualty was upgraded to an Alert at 1535, prior to initiating off site notifications.

Which of the following correctly describes when the initial off site notifications are required to be completed?

Notifications to state and local agencies must be completed by (1) . The NRC must be notified by (2) .

A. (1) 1550 (2) 1635 B. (1) 1540 (2) 1625 C. (1) 1540 (2) 1635 D. (1) 1550 (2) 1625 Answer: B K/A Match:

The question requires SRO only knowledge of the emergency plan; specifically, time requirements for off site reporting of emergency declarations.

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ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a G statement tied to 10CFR.55.43 and Page 2 of 6

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since these are the required times for the second emergency declaration.

However, required off site notifications are timed from the initial emergency declaration.

B. Correct. In accordance with PPM 13.4.1, Emergency Notifications, Initial notification of Washington State and local authorities must be made within 15 minutes following declaration of the emergency event, and Initial notification of the NRC via the Emergency Notification System (ENS) should be made immediately after notification of the appropriate state and local authorities, and must be made not later than one (1) hour after emergency event declaration. The time for the initial notifications starts from the initial emergency declaration.

C. Incorrect. Plausible since the time for local notification is correct. However, time to notify NRC starts from the initial notification not from the second notification.

D. Incorrect. Plausible since time for local notification is from the second emergency declaration and the NRC time is correct. However, required off site notifications are timed from the initial emergency declaration.

Technical Reference(s)

PPM 13.4.1, Emergency Notifications Attached w/ Revision # See Comments / Reference Proposed references to be provided during examination: None Learning Objective: 10222 - Identify the methods used to make emergency response notifications.

Question Source: Bank #

Modified Bank # LR01288 (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 43.5 Page 3 of 6

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

Original Bank Question LR01288 Revision: N/A Columbia was operating at full power when a series of events occurred that caused a reactor scram. Due to low RPV water level the Shift Manager has declared an Alert at 1525.

Which of the following correctly describes when notification to the State and local authorities must be made?

Notifications must be made no later than:

A. 1540 B. 1555 C. 1610 D. 1625 Page 4 of 6

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 13.4.1, Emergency Notifications Revision: Major 43, Minor 02 Page 5 of 6

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 6 of 6

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Examination Outline Cross-reference: Level RO SRO Rev. 2 Date: 2/6/2017 Tier 3 Group 1 K/A 2.4.31 Level of Difficulty: 3 Importance Rating 4.1 Knowledge of annunciator alarms, indications, or response procedures.

Question # 100 CGS is operating in Mode 1.

A control room annunciator is cycling at the setpoint due to setpoint shift:

What action must be taken prior to disabling the annunciator?

A. Notify the Crew Assistant Operations Manager.

B. Place a Caution Tag on the annunciator.

C. Write a work request to adjust the annunciator setpoint.

D. Perform a 50.59/72.48 evaluation to validate FSAR compliance.

Answer: D K/A Match:

This question requires the candidate to demonstrate knowledge of requirements to disable a valid annunciator.

Page 1 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 SRO Only:

K/A is a G statement and Page 2 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Explanation:

A. Incorrect. Plausible since the Crew Assistant Operations Manager should be informed if an annunciator remains disabled for greater than one shift. However, this is not required prior to disabling the annunciator.

B. Incorrect. Plausible since a caution clearance should be intiated if an annunciator remains disabled for greater than one shift. However, this is not required prior to disabling the annunciator.

C. Incorrect. Plausible since a work request should be written if an annunciator is disabled for greater than one shift due to inoperable source inputs. However, this is not required prior to disabling the annunciator due to setpoint shift.

D. Correct. Prior to disabling an annunciator that is cycling at its setpoint (not due to active maintenance), a 50.59/72.48 evaluation should be performed.

Technical Reference(s) Attached w/ Revision # See PPM 1.3.1, Operating Policies, Programs and Practices Comments / Reference Proposed references to be provided during examination: None Learning Objective: 11270 Knowledge of annunciators alarms and indications / and use of the response instructions.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 43.3 Page 3 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Comments /

Reference:

PPM 1.3.1, Operating Policies, Programs Revision: Major 120, Minor and Practices 003 Page 4 of 5

ES-401 CGS NRC 2017 SRO Written Exam Worksheet Form ES-401-5 Page 5 of 5

CGS 2017 NRC Exam Examination Light Reference Some questions contained in this examination use graphical representations of control board equipment and indications.

Use this reference as an aid in determining the status of the indications given.

Round Indicating Lights Square Indicating Lights Indication Light On Light Off Indication Light On Light Off White White Indicating Indicating Light Light Amber Amber Indicating Indicating Light Light Green Blue Indicating Indicating Light Light Red Indicating Light Page 1 of 1

Page 1 Redacted 2017 NRC RO Written Exam Reference Question: RO-19 Number: PFP-RB-522 Use Category: N/A Major Rev: 005 Minor Rev: N/A Title: REACTOR 522 Fire Area: R-1, 18, 21, M-27 Page: 11 of 14 Location: SE QUADRANT Fire Area: R-1/1, 18/2, 21/2 Page 2 of 2 Areas covered by this Pre-Fire Plan:

General Floor Area ................................................................................................ R404 Pipe Space ............................................................................................................ R405 RWCU Pump Room 1B.......................................................................................... R406 RWCU Pump Room 1A.......................................................................................... R407 Access Area above RWCU Pump Rooms .............................................................. R409 MCC Room ............................................................................................................ R410 Post Fire Safe Shutdown Equipment or Associated Cabling:

Fire Area R-1/1, Room R404 SW-42-8BA10C Motor Starter for SW-M-V/187B (NEMA 1) N7/3.9 Fire Area R-21/2, Room R405; RHR-V-42B LPCI Isolation N1/5.7 Fire Area R-18/2, Room R410; E-MC-8B Motor Control Center 8B N/3.8 E-MC-8BA Motor Control Center 8BA+ N/3.8 Special Entry Considerations:

RWCU Pump Rooms are Locked HIGH HIGH Radiation Areas. Contact HP for keys to access.

Fire Detection:

Ionization detectors.

Manual Pull Station Alarms Suppression Systems and Isolation Location:

N/A Electrical Disconnects or Special Shutoffs:

None Exposure or Fire Extension Routes:

Smoke could extend throughout general floor area and up or down through open hatches to elevations. Because of the small amount of combustible materials, fire extension is improbable from this fire area.

Radiological Considerations:

Potential To Be High. Consult HP Tech for current conditions.

Combustibles or Gasses:

Cables, Hydraulic fluid, Cabinets of protective clothing Fire Loading:

LOW Emergency Ventilation:

Building Exhaust System Special Hazards:

None Page 2 of 10

2017 NRC RO Written Exam Reference Question: RO-19 Number: PFP-RB-522 Use Category: N/A Major Rev: 005 Minor Rev: N/A Title: REACTOR 522 Fire Area: R-1, 18, 21, M-27 Page: 14 of 14 Location: FIRE FIGHTING EQUIPMENT LEGEND Fire Area: None Page 1 of 1 Page 3 of 10

2017 NRC RO Written Exam Reference Question: RO-19 Number: ABN-FIRE Use Category: CONTINUOUS Major Rev: 036 Minor Rev: 001 Title: Fire Page: 39 of 67 APPENDIX R FIRE AREA OPERATOR MANUAL ACTIONS Fire Area R-1 Location: Reactor Building General Area all levels (excluding Containment and ECCS Pump Rooms)

Protected Level Inst: MS-LR/PR-623B Required Actions:

USE Division 2 Post Fire Safe Shutdown systems.

CAUTION Fire in the RB 471 east could result in RB door locking, and preventing access.

PERFORM the following: {AR-12926}

a. NOTIFY CAS/SAS of intent to open E-CB-DP/SS/IN4/A/4

b. OBTAIN key 157 from the Control Room key locker.

c. UNLOCK and OPEN grated door C-245 (RW 467 vestibule to the Division 1 battery room)

d. OPEN E-CB-DP/SS/IN4/A/4 CAUTION Loss of keep fill pressure may cause a water hammer during pump start. The following actions should be taken from the Control Room.

START RHR-P-2A within 30 minutes (Control Room). {12.1}, {P-80442}

START RHR-P-2B within 30 minutes (Control Room). {12.1}, {P-80442}

NOTE: Fire in this area may affect RHR-FIS-10B, which could affect operation of RHR-FCV-64B.

VERIFY RHR-FCV-64B opens upon the start of RHR-P-2B (Control Room)

{AR-42457}

Attachment 13.1, Appendix R Fire Area Operator Manual Actions Page 4 of 10

2017 NRC RO Written Exam Reference Question: RO-19 Number: ABN-FIRE Use Category: CONTINUOUS Major Rev: 036 Minor Rev: 001 Title: Fire Page: 40 of 67 IF RHR-FI-603B is erratic or inoperable, THEN USE RHR-P-2B ammeter (RHR-AM-P/2B) to verify RHR flow exists (H13-P601).

CAUTION Radiological monitoring is suggested prior to entry. The use of SCBA may be required.

VERIFY CLOSE RWCU-V-32 within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , Manual Blowdown Isolation (RB 501 SW).

NOTE: Door wedges are in portable lantern boxes.

OPEN the door R408 to MCC 8B/8BA within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months to provide passive cooling to the room (RB 522 SE).

CLOSE RCC-V-50 within 3.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of the start of the fire (RCC Surge Tank Isolation) (RB 572 NE). {12.1}

NOTE: The Shift Managers extension 2441 is PFSS credited and should be used as the primary control room contact number, the CRS extension is to be used as the backup.

MONITOR the Spent Fuel Pool temperature and level locally within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Infrared temperature monitoring device is available in the tool crib. REFER to ABN-FPC-LOSS.

{P-80442}

Attachment 13.1, Appendix R Fire Area Operator Manual Actions Page 5 of 10

2017 NRC RO Written Exam Reference Question: RO-19 Number: ABN-FIRE Use Category: CONTINUOUS Major Rev: 036 Minor Rev: 001 Title: Fire Page: 41 of 67 Fire Area R-4 Location: RHR-P-2B Pump Room, RHR B Pipe Chase from RB 470 to RB 548, RB 548 South Valve Room, RHR-HX-B room RB-548 and 572.

Protected Inst: MS-LR/PR-623A Required Actions:

USE Division 1 Post Fire Safe Shutdown systems.

CLOSE RCC-V-130 from the Control Room (FPC HX RCC Outlet) (H13-P626) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of the start of the fire. This action is required since a fire may cause SW-V-187B to open and cause service water to be cross-tied to RCC. Resulting flooding and a loss of LPCS-P-2 (RHR A/LPCS keep fill pump). {12.1}

Fire Area R-6 Location: RCIC Pump Room, RB 422 and 444 Protected Level Inst: MS-LR/PR-623A Required Actions:

USE Division 1 Post Fire Safe Shutdown systems.

DECLARE RCIC Inoperable and NOTIFY Security Fire Area R-7 Location: RHR-P-2C Pump Room, RB 422 and 444 Protected Inst: MS-LR/PR-623A Required Actions:

USE Division 2 Post Fire Safe Shutdown systems.

CAUTION Loss of keep fill pressure may cause a water hammer during pump start. The following actions should be taken from the Control Room.

START RHR-P-2B within 30 minutes. {12.1},{P-80442}

Attachment 13.1, Appendix R Fire Area Operator Manual Actions Page 6 of 10

2017 NRC RO Written Exam Reference Question: RO-19 Number: ABN-FIRE Use Category: CONTINUOUS Major Rev: 036 Minor Rev: 001 Title: Fire Page: 42 of 67 Fire Area R-8 Location: LPCS Pump Room, RB 422 and 444 Protected Level Inst: MS-LR/PR-623B Required Actions:

USE Division 2 Post Fire Safe Shutdown systems.

CAUTION Loss of keep fill pressure may cause a water hammer during pump start. The following actions should be taken from the Control Room.

START RHR-P-2A within 30 minutes (Control Room). {12.1}, {P-80442}

Fire Area R-18 Location: DIV 2 MCC Room, RB 522 Protected Level Inst: MS-LR/PR-623A Required Actions:

USE Division 1 Post Fire Safe Shutdown systems.

CAUTION Loss of keep fill pressure may cause a water hammer during pump start. The following actions should be taken from the Control Room.

START RHR-P-2B within 30 minutes. {12.1}, {P-80442}

CLOSE RCC-V-130 from the Control Room (FPC HX RCC Outlet) (H13-P626) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of the start of the fire. This action is required since a fire may cause SW-V187B to open and cause service water to be cross-tied to RCC. Resulting flooding and a loss of LPCS-P-2 (RHR A/LPCS keep fill pump). {12.1}

Attachment 13.1, Appendix R Fire Area Operator Manual Actions Page 7 of 10

2017 NRC RO Written Exam Reference Question: RO-19 Number: ABN-FIRE Use Category: CONTINUOUS Major Rev: 036 Minor Rev: 001 Title: Fire Page: 56 of 67 FIRE AREAS PSFF DIV FP VITAL FIRE AREA LOST FIRE LOCATION AREA DG-1 1 DG 441-455 HPCS Generator Room Yes DG-2 1 DG 441-455 Diesel Generator 1 Room Yes DG-3 2 DG 441-455 Diesel Generator 2 Room Yes DG-4 1 DG 441 Diesel Generator 1 Diesel Oil Tank Pump Room Yes DG-5 2 DG 441 Diesel Generator 2 Diesel Oil Tank Pump Room Yes DG-6 N DG 441 HPCS Diesel Oil Tank Pump Room No DG-7 N DG 441 HPCS Diesel Day Tank Room No DG-8 1 DG 441 DG-1 Diesel Oil Day Tank Room Yes DG-9 2 DG 441 DG-2 Diesel Oil Day Tank Room Yes DG-10 N DG 455 Deluge Valve Equipment Room No R-1 1 Reactor Building 471-606 General Equipment Area Yes including 422 NW Stairwell, 441 Train Bay, and NE Vestibule 471-606.

(Does not include instrument rack fire areas)

R-2 U Primary Containment (Inerted during operation) No R-3 N RB 422 HPCS Pump Room No R-4 2 RB 422 RHR-B Pump Room, 470-548 Pipe Chase, Yes 471 SW Valve Room, 492 & 563 Pipe Tunnels, 548 South Valve Room, 548 & 572 Heat Exchanger Equipment Rooms and 572 Hydrogen Recombiner Room R-5 1 RB 422 RHR-A Pump Room, 470-548 Chase, 492 & 563 Yes Pipe Tunnels (west half) and 548 Heat Exchanger Equipment Room R-6 2 RB 422 RCIC Pump Room Yes R-7 1 RB 422 RHR-C Pump Room Yes R-8 1 RB 422 LPCS Pump Room Yes R-9 N RB 422-623 SW Stair A6 No R-10 N RB 422-623 SW Elevator 2 No R-11 N RB 422-623 NE Stair A5 No Attachment 13.2, Fire Areas Page 8 of 10

2017 NRC RO Written Exam Reference Question: RO-19 Number: ABN-FIRE Use Category: CONTINUOUS Major Rev: 036 Minor Rev: 001 Title: Fire Page: 57 of 67 PSFF DIV FP VITAL FIRE AREA LOST FIRE LOCATION AREA R-12 N RB 441-623 NE Elevator 1 No R-15 N RB 422 Lobby Outside of Stair A5 & Elevator 1 No R-18 2 Division 2 MCC Room Yes R-21 2 RB 522 South Valve Room Yes M-9 2 RB 471 Instrument Rack E-IR-H22/P009 Enclosure Yes M-21 2 RB 501 Instrument Rack E-IR-H22/P021 Enclosure Yes M-27 2 RB 522 Instrument Rack E-IR-H22/P027 Enclosure Yes RC-1 N RW 437-507 General Equipment Areas No (excluding below rooms)

RC-1 2 RW 437 Rms C-106 & C-108 Yes (Vital Portion)

RC-2 1 RW 484 Cable Spreading Room Yes RC-3 1 RW 467-525 Cable Chase Yes RC-4 1 RW 467 Division 1 Electrical Equipment Rooms Yes (Battery Charger Room No. 1 and RPS Room No. 1)

RC-5 1 RW 467 Battery Room 1 Yes RC-6 2 RW 467 Battery Room 2 Yes RC-7 2 RW 467 Division 2 Electrical Equipment Rooms Yes (Battery Charger Room No. 2 and RPS Room No. 2)

RC-8 2 RW 467 Switchgear Room 2 Yes RC-9 2 RW 467 Remote Shutdown Room Yes RC-10 U RW 510 Main Control Room Yes RC-11 1 RW 525 Unit A Air Condition Room Division 2 Yes PFSS Feeders RC-12 2 RW 525 Unit B Air Conditioning Room Yes RC-13 2 RW 525 Communications Room, Emergency Chiller Area Yes and HVAC Chase RC-14 1 RW 467 Switchgear Room 1 Yes RC-19 2 RW 467 Vital Island Corridor C-205 Yes RC-20 1 RW 467 Pipe Chase and 487 PASS Area Yes Attachment 13.2, Fire Areas Page 9 of 10

2017 NRC RO Written Exam Reference Question: RO-19 Number: ABN-FIRE Use Category: CONTINUOUS Major Rev: 036 Minor Rev: 001 Title: Fire Page: 58 of 67 PSFF DIV FP VITAL FIRE AREA LOST FIRE LOCATION AREA SW-1 1 Standby Service Water Pump House 1A Yes SW-2 2 Standby Service Water Pump House 1B Yes TG-1 N Turbine Generator Building General Areas 441-501 No (including E-W corridors adjacent RW & RB)

Zone TG-2 N TB 441 East Turbine Oil Storage Room No TG-3 N TB 441-518 East Stair A1 No TG-4 N TB 441-518 East Elevator 3 No Zone TG-5 N TB 441 Auxiliary Boiler Room No TG-6 N TB 441-501 North Stair A3 No Zone TG-7 N TB 441 Hydrogen Seal Oil Room No TG-8 N TB 441-501 NW Stair A4 No Zone TG-9 N TB 471 Turbine Oil Reservoir Room No Zone TG-10 N TB 441 West Transformer Vault No Zone TG-11 N TB 441 East Transformer Vault (by Aux. Boiler) No Zone TG-12 2 441 Corridors between RW & RB, DG & RB (includes Yes Rooms C121, D104 and D113)

NOTE: Plant areas not listed do not contain post-fire safe shutdown equipment or cables and can be considered non-FP vital.

Division Code Legend DIV 1 Fire area contains Div. 1 post-fire safe shutdown equipment or cables which may be lost.

Use Division 2 PFSS systems for a fire in this area.

DIV 2 Fire area Contains Div. 2 post-fire safe shutdown equipment or cables which may be lost. Use Division 1 PFSS systems for a fire in this area.

DIV N Fire area does not contain equipment or cables for either division of post-fire safe shutdown equipment.

DIV U This code indicates that this fire area has been uniquely analyzed for post fire safe shutdown.

END Attachment 13.2, Fire Areas Page 10 of 10

2017 NRC SRO Exam Reference Question: SRO-77 AC Sources - Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating LCO 3.8.1 The following AC electrical power sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electric Power Distribution System; and

b. Three diesel generators (DGs).

APPLICABILITY: MODES 1, 2, and 3.

NOTE--------------------------------------------

Division 3 AC electrical power sources are not required to be OPERABLE when High Pressure Core Spray System is inoperable.

ACTIONS

NOTE----------------------------------------------------------

LCO 3.0.4.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One offsite circuit A.1 Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. OPERABLE offsite circuit.

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND Columbia Generating Station 3.8.1-1 Amendment No. 169,187 225 Page 1 of 5

2017 NRC SRO Exam Reference Question: SRO-77 AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 Declare required feature(s) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from with no offsite power discovery of no offsite available inoperable when power to one division the redundant required concurrent with feature(s) are inoperable. inoperability of redundant required feature(s)

AND A.3 Restore offsite circuit to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

AND 6 days from discovery of failure to meet LCO when not associated with Required Action B.4.2.2 AND 17 days from discovery of failure to meet LCO B. One required DG B.1 Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. OPERABLE offsite circuit(s). AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND B.2 Declare required feature(s), 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from supported by the inoperable discovery of DG, inoperable when the Condition B redundant required concurrent with feature(s) are inoperable. inoperability of redundant required feature(s)

AND Columbia Generating Station 3.8.1-2 Amendment No. 195,197 225 Page 2 of 5

2017 NRC SRO Exam Reference Question: SRO-77 AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3.1 Determine OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DG(s) are not inoperable due to common cause failure.

OR B.3.2 Perform SR 3.8.1.2 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if not OPERABLE DG(s). performed within the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND B.4.1 Restore required DG to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from OPERABLE status. discovery of an inoperable DG AND 6 days from discovery of failure to meet LCO OR B.4.2.1 Establish risk management 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months actions for the alternate AC sources.

AND B.4.2.2 Restore required DG to 14 days OPERABLE status.

AND 17 days from discovery of failure to meet LCO Columbia Generating Station 3.8.1-3 Amendment No. 197,205 225 Page 3 of 5

2017 NRC SRO Exam Reference Question: SRO-77 AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Two offsite circuits C.1 Declare required feature(s) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from inoperable. inoperable when the discovery of redundant required Condition C feature(s) are inoperable. concurrent with inoperability of redundant required feature(s)

AND C.2 Restore one offsite circuit to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

D. One offsite circuit --------------------NOTE-------------------

inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.7, AND "Distribution Systems - Operating,"

when Condition D is entered with no One required DG AC power source to any division.

inoperable. ------------------------------------------------

D.1 Restore offsite circuit to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OPERABLE status.

OR D.2 Restore required DG to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OPERABLE status.

E. Two required DGs E.1 Restore one required DG to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months inoperable. OPERABLE status.

OR 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if Division 3 DG is inoperable Columbia Generating Station 3.8.1-4 Amendment No. 169,197 225 Page 4 of 5

2017 NRC SRO Exam Reference Question: SRO-77 AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and F.1 ---------------NOTE--------------

associated Completion LCO 3.0.4.a is not Time of Condition A, B, applicable when entering C, D, or E not met. MODE 3.

Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months G. Three or more required G.1 Enter LCO 3.0.3. Immediately AC sources inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated 7 days power availability for each offsite circuit.

SR 3.8.1.2 ------------------------------NOTES-----------------------------

1. All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.

2. A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR as recommended by the manufacturer. When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.

Verify each required DG starts from standby 31 days conditions and achieves steady state:

a. Voltage 3910 V and 4400 V and frequency 58.8 Hz and 61.2 Hz for DG-1 and DG-2; and

b. Voltage 3910 V and 4400 V and frequency 58.8 Hz and 61.2 Hz for DG-3.

Columbia Generating Station 3.8.1-5 Amendment No. 169,181,225,236 Page 5 of 5

2017 NRC SRO Written Exam Reference Question: SRO-79 Number: ABN-CR-EVAC Use Category: CONTINUOUS Major Rev: 035 Minor Rev: N/A Title: Control Room Evacuation and Remote Cooldown Page: 4 of 63 ABN-CR-EVAC FLOW CHART When CR evacuation is Immediate Actions required and 1. Mode Switch to Shutdown immediate actions 2. MSIV Arm and Depress are complete, 3. Security performs Attachment 7.1 Then 4. Announce: Scram CR Evacuation Perform Attachments 7.2-7.10 Control Room Subsequent Actions

Initiate RCIC by Arm & Depress

Open CB-1/7 Control Room

If CRDs failed to insert, Initiate ARI conditions permit Yes

Place MSIV control switches to closed subsequent

Ensure MS-V-16 or MS-V-19 are closed actions

Ensure SW-P-1A is started

Close RWCU-FCV-33 using RWCU-RMC-606

Ensure the Main Turbine is tripped No SM/CRS/IA/STA CRO1 CRO2 CRO3 DGSS EOSS Proceed to SD Panel Isolate SM-8 Open CB-B/8 Stop DG-2 Trip Pump Bkrs Remote S/D Activation Attachment 7.3 Attachment 7.4 Attachment 7.5 Attachment 7.7 if Room Attachment 7.2 SM-1,2,3 engergized Reset 8/DG2 Isolate SM-8 Ensure door Align RHR & Attachment 7.3 Attachment 7.4 C239 is open SW for Start Isolate Loads &

between RSD Attachment 7.2 Extended and SM-7 AND Monitoring Attachment 7.10 Classify as an Alert and Start DG-2 send IA/STA Attachment 7.5 for notification Monitor DG-2 Yes DG-2 If evacuating Attachment 7.6 Starts due to fire, Then enter ABN-FIRE No Yes Close CB-B/8 Attachment 7.20 When SD Panel is Activated and SM-8 is Isolated and Energized, Then Start SW-P-1B Within 10 Required to be Ready to Emergency Depressurize Minutes of DG2 Attachments Completed By Start Attachment 7.8

7.2 SD Panel Activation CRO1

7.3 Isolate SM-8 & Reset 8/DG2 CRO2 Start RHR-P-2B

7.4 Open CB-B8, Isolate SM-8, and RPS Breakers CRO3 to Control RPV

SM-8 Energized from 7.5 DG-2 DGSS Level as or 7.20 CB-B/8 DGSS/EOSS necessary

7.7 Trip Pump Breakers EOSS Attachment 7.9 If Div 1 125v Battery is LE 108 VDC Emergency Depressurize or RPV level is LE -147, THEN Required See A-2 Page 1 of 2

2017 NRC SRO Written Exam Reference Question: SRO-79 Number: ABN-CR-EVAC Use Category: CONTINUOUS Major Rev: 035 Minor Rev: N/A Title: Control Room Evacuation and Remote Cooldown Page: 5 of 63 When the Control Room is habitable THEN Return to the Control Room per Attachment 7.21 A-2 IF Div 1 125 Battery LE 108 VDC OR RPV Level LE-147" THEN EMERGENCY DEPRESSURIZATION RPV Level RPV Pressure WW Temperature WW Level Start RCIC Control RPV Pressure If not required for RPV per per RHR-P-2B RHR-P-2A injection, place RHR-P-2B in Attachment 7.11 Attachment 7.12 Suppression Pool Cooling per Attachment 7.15 Restore Restore When plant Is PV Level RPV Level conditions have RHR-P-2A Control Suppression

-50

+13" to to +54 54" stabilized Then discharge Pool Level per With RCIC with RCIC pressure GT 75 psig?

Attachment 7.16 Yes No Conduct a cooldown per If Attachment 7.22 Place SW-P-1A Perform RPV Level in operation per SOP-RHR-FILL LT-147" Then Attachment 7.13 When RPV Emergency Depressurization pressure LT 48 psig Then Reinstall Control power fuses for RHR-P-2A Shutdown Cooling Place RHR-P-2A in operation per Attachment 7.14 If not required for RPV injection, then place RHR-P-2A/B in Suppression Pool Cooling per Attachment 7.15 Before WW Temp Reaches 150 °F Shutdown Cooling Emergency Depressurization Shutdown Cooling

1. Open six (6) SRVs

2. When RPV pressure LT 470 psig, then open RHR-V-42B for injection If RHR-V-8 is available if required then Initiate Shutdown

3. When RHR-P-2B flow is GT 2,000 Cooling on RHR-P-2B per gpm then close RHR-FCV-64B Attachment 7.18

4. Maintain RPV Level +13" to 54.5" OR or Initiate Alternate Shutdown Initiate Alternate Cooling per Attachment 7.19 Shutdown Cooling per

5. Initiate Shutdown Cooling Attachment 7.19.

Page 2 of 2

2017 NRC SRO Written Exam Reference Question: SRO-80 ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.4.b is not applicable to HPCS.

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.1 Restore low pressure 7 days(1) injection/spray ECCS injection/spray subsystem inoperable. subsystem to OPERABLE status.

B High Pressure Core B.1 Verify by administrative Immediately Spray (HPCS) System means RCIC System is inoperable. OPERABLE when RCIC System is required to be OPERABLE.

AND B.2 Restore HPCS System to 14 days OPERABLE status.

(1)

The Completion Time that one train of RHR (RHR-B) can be inoperable as specified by Required Action A.1 may be extended beyond the 7 day completion time up to 7 days to support restoration of RHR-B from the modification activity. Upon successful restoration of RHR-B, this footnote is no longer applicable and will expire at 05:00 PST on February 9, 2015.

Columbia Generating Station 3.5.1-1 Amendment No. 187 225 230 Page 1 of 3

2017 NRC SRO Written Exam Reference Question: SRO-80 ECCS - Operating 3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Two ECCS injection C.1 Restore ECCS 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months subsystems inoperable. injection/spray subsystem to OPERABLE status.

OR One ECCS injection and one ECCS spray subsystem inoperable.

D. Required Action and D.1 --------------NOTE--------------

associated Completion LCO 3.0.4.a is not Time of Condition A, B, applicable when entering or C not met. MODE 3.

Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months E. One required ADS valve E.1 Restore ADS valve to 14 days inoperable. OPERABLE status.

F. One required ADS valve F.1 Restore ADS valve to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

AND OR One low pressure ECCS F.2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months injection/spray ECCS injection/spray subsystem inoperable. subsystem to OPERABLE status.

Columbia Generating Station 3.5.1-2 Amendment No. 149,169,225,236 Page 2 of 3

2017 NRC SRO Written Exam Reference Question: SRO-80 ECCS - Operating 3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and G.1 --------------NOTE--------------

associated Completion LCO 3.0.4.a is not Time of Condition E or F applicable when entering not met. MODE 3.

OR Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Two or more required ADS valves inoperable.

H. HPCS and Low H.1 Enter LCO 3.0.3. Immediately Pressure Core Spray (LPCS) Systems inoperable.

OR Three or more ECCS injection/spray subsystems inoperable.

OR HPCS System and one or more required ADS valves inoperable.

OR Two or more ECCS injection/spray subsystems and one or more required ADS valves inoperable.

Columbia Generating Station 3.5.1-3 Amendment No. 149,169,225,236 Page 3 of 3

2017 NRC SRO Exam Reference Question: SRO-88 RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.1-1 after implementation of Power Range Neutron Monitor (PRNM) upgrade.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months channels inoperable.

OR

NOTE---------------

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

A.2 Place associated trip 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months system in trip.

NOTE-------------- B.1 Place channel in one trip 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Not applicable for system in trip.

Functions 2.a, 2.b, 2.c, 2.d, or 2.f. OR

B.2 Place one trip system in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months B. One or more Functions trip.

with one or more required channels inoperable in both trip systems.

Columbia Generating Station 3.3.1.1-9 Amendment No. 226 Page 1 of 4

2017 NRC SRO Exam Reference Question: SRO-88 RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions C.1 Restore RPS trip capability. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with RPS trip capability not maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, Table 3.3.1.1-1 for the or C not met. channel.

E. As required by Required E.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action D.1 and POWER to < 30% RTP.

referenced in Table 3.3.1.1-1.

F. As required by Required F.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully insert Immediately Action D.1 and all insertable control rods in referenced in core cells containing one or Table 3.3.1.1-1. more fuel assemblies.

Columbia Generating Station 3.3.1.1-10 Amendment No. 169 225 226 Page 2 of 4

2017 NRC SRO Exam Reference Question: SRO-88 RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I. As required by Required I.1 Initiate alternate method to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and detect and suppress referenced in thermal hydraulic instability Table 3.3.1.1-1. oscillations.

AND

NOTE-------------

LCO 3.0.4 is not applicable.

I.2 Restore required channels 120 days to OPERABLE J. Required Action and J.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER to less than the Time of Condition I not value specified in the met. COLR.

SURVEILLANCE REQUIREMENTS

NOTES----------------------------------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Columbia Generating Station 3.3.1.1-11 Amendment No. 169 225 226 Page 3 of 4

2017 NRC SRO Exam Reference Question: SRO-88 RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 Table 3.3.1.1-1 (page 1 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. Intermediate Range Monitors

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 122/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.10 SR 3.3.1.1.14 (a) 5 3 H SR 3.3.1.1.1 122/125 SR 3.3.1.1.4 divisions of full SR 3.3.1.1.10 scale SR 3.3.1.1.14

b. Inop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 (a) 5 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14

2. Average Power Range Monitors (b)

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 20% RTP (Setdown) SR 3.3.1.1.6 SR 3.3.1.1.7 (d),(e)

SR 3.3.1.1.10 SR 3.3.1.1.16 (b)

b. Simulated Thermal 1 3 F SR 3.3.1.1.1 0.63W + 64.0% RTP (c)

Power - High SR 3.3.1.1.2 and 114.9% RTP SR 3.3.1.1.7 (d),(e)

SR 3.3.1.1.10 SR 3.3.1.1.16 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Each APRM/OPRM channel provides inputs to both trip systems.

(c) 0.63W + 60.8% RTP and 114.9% RTP when reset for single loop operation per LCO 3.4.1, Recirculation Loops Operating.

(d) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and as-left tolerances are specified in the Licensee Controlled Specifications.

Columbia Generating Station 3.3.1.1-15 Amendment No. 169 225 226 Page 4 of 4

2017 NRC SRO Written Exam Reference Question: SRO-93 Number: SWP-CHE-02 Use Category: INFORMATION Major Rev: 026 Minor Rev: N/A Title: Chemical Process Management and Control Page: 22 of 62 6.1.13 Table 6.1.13, Reactor Water Control Parameters - Power Operation, >10% Power Frequency of Additional Category Chemistry Regime Value Measurement Guidance (Needed)

Conductivity All 0.15 µS/cm Continuous (with zinc injection)

All 0.08 µS/cm Continuous 1 Good Practice (without zinc injection)

All (corrected with zinc 0.08 µS/cm Quarterly injection)

Primary Chemistry All >0.30 µS/cm Continuous 2, 3 Action Level 1 Primary Chemistry All >1.0 µS/cm Continuous Action Level 2 4

Primary Chemistry All >5 µS/cm Continuous Action Level 3 Chloride Good Practice All 1 ppb Daily 5 Primary Chemistry All 5 ppb 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Action Level 1 Primary Chemistry All >20 ppb 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 4, 6 Action Level 2 Primary Chemistry All >100 ppb 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 4, 6 Action Level 3 Sulfate Good Practice All 2 ppb Daily 5 Primary Chemistry All >5 ppb 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Action Level 1 Primary Chemistry All >20 ppb 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 4, 6 Action Level 2 Primary Chemistry All >100 ppb 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 4, 6 Action Level 3 Page 1 of 6

2017 NRC SRO Written Exam Reference Question: SRO-93 Number: SWP-CHE-02 Use Category: INFORMATION Major Rev: 026 Minor Rev: N/A Title: Chemical Process Management and Control Page: 23 of 62 Table 6.1.13 Additional Guidance: Reactor Water Control Parameters, Power Operation,

>10% Power:

1. Good Practice Conductivity Values The Good Practice continuous reactor water conductivity value of 0.15 µS/cm for plants with zinc injection is achievable with Good Practice levels of chloride and sulfate and up to approximately 20 ppb soluble zinc; the Good Practice continuous conductivity value of 0.08 µS/cm is for plants without zinc injection. The Good Practice quarterly conductivity value of 0.08 µS/cm is the value after correction for contributions from soluble iron and soluble zinc; monitoring is performed quarterly based on grab samples. Guidance for performing conductivity balances and corrections is provided in BWRVIP-190 Volume 2 Appendix C.

2. Conductivity An ion-conductivity balance should be performed to estimate the concentration of unidentified and potentially corrosive ions. Conductivity may be corrected for contributions from soluble iron and zinc to confirm Action Level entry and for crack growth rate calculations. Guidance for performing conductivity balances and corrections is provided in BWRVIP-190 Volume 2 Appendix C.

3. Action Level 1 Conductivity Value during OLNC Injection During noble metal injection for OLNC, the conductivity may exceed the Action Level 1 value due to increases in sodium from the platinum compound, Na2Pt(OH)6 (see BWRVIP-190 Volume 2 Chapter 1). Short duration conductivity increases from this known source of sodium can be predicted in advance and are not a concern for accelerated IGSCC or fuel cladding corrosion.

Reactor water sampling and analysis results during the OLNC application should be reviewed to confirm that the increased conductivity is not due to aggressive anions or ionic impurities from other sources. If confirmed, no additional corrective actions or Action Level 1 responses are required and the time in excess of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that reactor water conductivity exceeds 0.30 µS/cm does not have to be considered when calculating Hydrogen Availability (see BWRVIP-190 Volume 2 Appendix E (Mitigation Performance Indicator) for the definition of Hydrogen Availability).

4. Action Level 2 and Action Level 3 A plant-specific analysis should be available to determine whether plant shutdown or continued Power Operation is the most prudent approach with regard to IGSCC and fuel damage (see BWRVIP-190 Volume 2 Appendix B). When the Action Level 3 value is exceeded, and such an analysis is not available and cannot be conducted within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following the excursion, orderly shutdown of the plant is the most prudent approach.

5. Chloride and Sulfate Sampling and Analysis Recognizing that chloride and sulfate have associated near-term operational actions for off-normal conditions, Daily sampling frequency is Needed and should not be relaxed. For additional information see BWRVIP-190 Volume 2 Chapter 7.

6. Action Level 2 and Action Level 3 Values for OLNC+HWC Evaluation of the effects of a chemistry transient that exceeds Action Level 2 or Action Level 3 should be based on welds or components of concern as defined by the utility and on the mitigation status as provided in BWRVIP- 62 Revision 1 (Chapter 4 and Table 4-1). Welds and components of concern and their mitigation status may be defined in the plant-specific Strategic Water Chemistry Plan; see BWRVIP-190 Volume 2 Appendix A (Guidance on Developing a BWR Strategic Water Chemistry Plan).

Page 2 of 6

2017 NRC SRO Written Exam Reference Question: SRO-93 Number: SWP-CHE-02 Use Category: INFORMATION Major Rev: 026 Minor Rev: N/A Title: Chemical Process Management and Control Page: 28 of 62 6.1.15 Table 6.1.15, Reactor Feedwater/Condensate Control Parameters - Power Operation, >10% Power Frequency of Chemistry Additional Category Value Measurement Regime Guidance (Needed)

Feedwater Conductivity Good Practice All 0.060 µS/cm Continuous Primary Chemistry All >0.065 µS/cm Continuous 1 Action Level 1 Feedwater Dissolved Hydrogen 0.21 ppm or 10 SCFM Hydrogen Needed OLNC+HWC Continuous 2, 3 Flow Rate Feedwater Dissolved Oxygen Good Practice All 40 - 100 ppb Continuous Primary Chemistry <30 ppb 4,5 All Continuous Action Level 1 >200 ppb CDI Conductivity Primary Chemistry All >0.10 µS/cm Continuous Action Level 1 Primary Chemistry All >1.0 µS/cm Continuous 6 Action Level 3 Condensate Dissolved Oxygen Good Practice All 40 - 100 ppb Continuous Primary Chemistry <30 ppb 6,7 All Continuous Action Level 1 >200 ppb Feedwater Total Iron Good Practice All 1.0 ppb Integrated 8,9 (Cycle average) 3.0 ppb Fe with FW Cu 0.05 ppb and FW Zn 0.5 ppb 2.0 ppb Fe with FW Cu 0.05 ppb and FW Zn >0.5 ppb Needed (Quarterly All Integrated 9, 10, 11 average) 2.0 ppb Fe with FW Cu >0.05ppb and FW Cu+Zn >0.5 ppb or Cycle-Specific Values established by the Cross Discipline Review Team Primary Chemistry All >5.0 ppb Integrated 96 hr 9, 10 Action Level 1 Page 3 of 6

2017 NRC SRO Written Exam Reference Question: SRO-93 Number: SWP-CHE-02 Use Category: INFORMATION Major Rev: 026 Minor Rev: N/A Title: Chemical Process Management and Control Page: 29 of 62 Frequency of Chemistry Additional Category Value Measurement Regime Guidance (Needed)

Feedwater Total Zinc 0.5 ppb quarterly average 0.4 ppb cycle average Needed (Quarterly or All Integrated 9, 11 or cycle average) Cycle-Specific Values established by the Cross Discipline Review Team Feedwater Total Copper Good Practice All 0.05 ppb Integrated 9, 12 (Cycle average)

Needed (Quarterly All 0.10 ppb Integrated 9, 12, 13 average)

Primary Chemistry All >0.2 ppb Integrated 96 hr 9, 13 Action Level 1 Table 6.1.15 Additional Guidance: Reactor Feedwater/Condensate Control Parameters - Power Operation, >10% Power:

1. Feedwater Conductivity Plant-specific Action Levels should be adopted if elevated conductivity results from the feedwater iron injection program or increased soluble iron due to the catalytic effects of noble metal deposits in the sample line.

2. Feedwater Hydrogen When the feedwater hydrogen concentration (or calculated concentration based on the feedwater hydrogen injection rate and feedwater flow rate) decreases below the plant-specific Needed value established for IGSCC mitigation, hydrogen injection should be restored to the required rate as soon as possible to minimize the impact on HWC Availability.

3. OLNC+HWC Plant-Specific Feedwater Hydrogen With NMCA/OLNC+HWC, the plant-specific hydrogen injection rate is determined by:

- Performing an HWC benchmark test to determine the hydrogen injection rate that provides an H2:O2 molar ratio of 2, as described in BWRVIP-245.

- Running BWRVIA at the plant conditions at the time of the HWC benchmark test and determining a scaling factor relating BWRVIA predictions to the plant-specific measured hydrogen injection rate as described in BWRVIP-245.

- Applying the BWRVIA model to determine the hydrogen injection rate required for a molar ratio 3 at the upper downcomer location at beginning of cycle (BOC),middle of cycle (MOC) and end of cycle full power (EOC) conditions.

Page 4 of 6

2017 NRC SRO Written Exam Reference Question: SRO-93 Number: SWP-CHE-02 Use Category: INFORMATION Major Rev: 026 Minor Rev: N/A Title: Chemical Process Management and Control Page: 30 of 62

4. Feedwater Dissolved Oxygen If noble metal contamination or other causes of oxygen depletion in the sample line leads to observations of lower than expected oxygen concentrations in the feedwater or when feedwater dissolved oxygen monitoring is unavailable, condensate oxygen may be used to assure sufficient dissolved oxygen to control FAC. If oxygen is injected into the condensate, the alternate sample point should be located downstream of the oxygen injection point

5. Feedwater and Condensate Dissolved Oxygen The technical bases for the low dissolved oxygen Action Level and Good Practice values for FAC control are given in BWRVIP-190 Volume 2 Chapter 3. A feedwater dissolved oxygen Good Practice value <40 ppb but 30 ppb may be adopted based on supporting operating experience. A plant-specific upper limit may have to be adopted to ensure consistency with ECP or hydrogen:oxidants molar ratio requirements under OLNC+HWC. The BWRVIA radiolysis model should be run if dissolved oxygen in the feedwater exceeds the plant-specific upper limit to determine whether an increase in the hydrogen injection rate is required. In addition, the licensing basis for plant life extension should be reviewed to determine if a commitment to limit the maximum feedwater oxygen has been made in the analysis of the fatigue life of feedwater piping (see BWRVIP-190 Volume 2 Chapter 2).

6. Condensate (CDI) Conductivity An evaluation is required to establish a plant-specific limit, which depends on the condensate polishing system design and cooling water chemistry.

7. Condensate Dissolved Oxygen The sample may be obtained from either the condensate pump discharge or condensate polisher outlet.

8. Feedwater Total Iron, Feedwater Total Zinc and Feedwater Total Copper (Action Level 1, Quarterly and Cycle Averages)

Power and time weighting may be used over the integrated sample collection period. Details are given in BWRVIP-190 Volume 2 Chapter 7. Corrective actions for exceeding Action Level 1 for feedwater iron and feedwater zinc apply after 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months .

9. Feedwater Total Iron (Effects of Noble Metal in Sample Lines)

If noble metals deposit in the Feedwater sample lines, the measured FW Fe concentration may be elevated due to catalytic effects in the sample line (see BWRVIP-190 Volume 2 Chapter 7). In this case, an alternate Feedwater sample point (further upstream) or the condensate polishing system effluent sample point can be used. Plants with forward-pumped drains must also take into account the Fe contribution from these drains to the final Feedwater.

Page 5 of 6

2017 NRC SRO Written Exam Reference Question: SRO-93 Number: SWP-CHE-02 Use Category: INFORMATION Major Rev: 026 Minor Rev: N/A Title: Chemical Process Management and Control Page: 31 of 62

10. Feedwater Total Iron (Quarterly Average)

The following values apply, based on quarterly averages for feedwater iron, copper and zinc, unless cycle-specific values based on an approved fuel risk assessment or change management assessment are established:

- FW Fe 3.0 ppb with FW Cu 0.05 ppb and FW Zn 0.5 ppb

- FW Fe 2.0 ppb with FW Cu 0.05 ppb and FW Zn >0.5 ppb

- FW Fe 2.0 ppb with FW Cu >0.05 ppb and FW Cu+Zn >0.5 ppb If any of the cycle-specific values, or the above values if cycle-specific values are not developed, are exceeded, then notify Fuels that criteria for performing a fuel risk assessment in accordance with the Fuel Reliability Guidelines: BWR Fuel Cladding Crud and Corrosion, Revision 1 [2-8] and change management assessment in accordance with Fuel Reliability Guidelines: Fuel Surveillance and Inspection Revision 2 may have been met.

11. Feedwater Total Zinc (Quarterly Average and Cycle Average)

The following values apply unless cycle-specific values based on an approved fuel risk assessment or change management assessments are established:

- FW Zn 0.5 ppb quarterly average

- FW Zn 0.4 ppb cycle average If any of the cycle-specific values, or the above values if cycle-specific values are not developed, are exceeded, then notify Fuels that criteria for performing a fuel risk assessment in accordance with the Fuel Reliability Guidelines: BWR Fuel Cladding Crud and Corrosion, Revision 1 [2-8] and change management assessment in accordance with Fuel Reliability Guidelines: Fuel Surveillance and Inspection Revision 2 may have been met.

12. Feedwater Total Copper (Sampling Frequency and Good Practice Value)

Frequency of measurement may be reduced for plants without copper-alloy condenser tubes and for which long-term trending has shown copper levels 0.05 ppb.

13. Feedwater Total Copper (Quarterly Average)

The following value applies unless a cycle-specific value based on an approved fuel risk assessment or change management assessment is established:

- FW Cu 0.1 ppb quarterly average If the cycle-specific value is exceeded, then notify Fuels that the criterion for performing a fuel risk assessment in accordance with Fuel Reliability Guidelines: BWR Fuel Cladding Crud and Corrosion, Revision 1 and change management assessment in accordance with Fuel Reliability Guidelines: Fuel Surveillance and Inspection Revision 2 may have been met.

Page 6 of 6

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