Text

2017 Palisades Ile Proposed Written Exam
	ML22167A013
Person / Time
Site:	Palisades
Issue date:	01/23/2017
From:	Randy Baker; NRC/RGN-III/DRS/OLB
To:	; Entergy Nuclear Operations
	Baker R
Shared Package
ML17037C881	List: ML22167A007; ML22167A009; ML22167A011; ML22167A013; ML22167A014; ML22167A016; ML22167A017; ML22167A020;
References
	Download: ML22167A013 (493)
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ES-401 Question 1 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1___ _____

Group # __1___ _____

K/A # CE/E02.EK3.1 Importance Rating __3.2__ _____

K/A statement: Knowledge of the reasons for the following responses as they apply to the (Reactor Trip Recovery): Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.

Proposed Question:

The Plant has been at 100% power for the last 30 days. During the performance of EOP-2.0, "Reactor Trip Recovery," which one of the following describes the expected response of reactor power?

After the initial rapid power reduction, reactor power will stabilize at . . .

A. 10-4 % and then slowly lower over a period of hours.

B. The subcritical multiplication level and then slowly lower.

C. The subcritical multiplication level and then remain at that level.

D. 10-4 % and then rise slowly over a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period as Xenon burns out.

Proposed Answer: B Explanation (Optional):

A. Incorrect, applicant selects incorrect power level (reactor is still critical at this power level)

B. Correct, EOP-2.0 provides this guidance and trend C. Incorrect, applicant selects correct power level but incorrect trend D. Incorrect, applicant selects incorrect power level and misinterprets the significance of Xenon for these conditions Technical Reference(s): EOP-2.0 Basis___________________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: None .

Learning Objective: _________________________ (As available)

Question Source: Bank # __X____

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2003____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

ES-401 Question 2 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1___ _____

Group # __1___ _____

K/A # 008.AK2.02________

Importance Rating __2.7___ _____

K/A statement: Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Sensors and detectors Proposed Question:

The crew is performing AOP-23, Primary Coolant Leak, to attempt to isolate a PCS leak.

Given the following plant conditions:

Containment radiation levels are rising

Pressurizer level is rising

Charging flow has lowered

Letdown flow has risen

Quench Tank level is 72% and stable

Steam Generator levels are at 66% and stable Which of the following is the most likely location of the leakage?

A. Charging Line inside of Containment B. Letdown Line inside Containment C. Pressurizer Power Operated Relief Valve D. Pressurizer Vapor Space Proposed Answer: D Explanation (Optional):

Due to the increasing Containment Radiation levels and the increasing Pressurizer (PZR) level, choice D is correct. If a PZR Safety Valve were leaking, Quench Tank (QT) level would be increasing rather than stable and containment radiation would be stable unless the rupture disk on the QT blows. If the Charging or Letdown line were leaking, PZR level would not be increasing. However, the decreasing Primary System pressure (swell of the PZR) will cause Charging flow to decrease and Letdown flow to increase.

A. Incorrect, a charging line leak inside containment would not cause PZR level to rise.

B. Incorrect, a letdown line leak inside containment would not cause PZR level to rise.

C. Incorrect, a leak from a PZR PORV would cause QT level to rise, not remain stable.

D. Correct, PZR vapor space is leaking into the containment atmosphere, causing containment radiation to rise and PZR level to rise.

Technical Reference(s): PL-PCS, Primary Coolant System Lesson Plan_________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ___None__________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2005________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

Question modified from Palisades 2005 NRC Exam (Question 2). Edited stem and rearranged distractors. Replaced one distractor with new.

PL-PLCS Pressurizer Level Control System Rev. 5 A. Impact on the following systems due to loss or malfunction of Pressurizer Level Control including loss of inputs:

1. CVCS

a. Charging Pumps will start or stop and/or P-55A speed changes depending on nature of malfunction or loss.

b. Letdown Orifices open or close depending on nature of malfunction or loss.

2. PCS

a. A loss or malfunction of PLCS affecting CVCS will cause PCS inventory to be raised or reduced accordingly.

3. Pressurizer Pressure Control

a. Level channel failure can affect Pressurizer Heaters on lo-lo level cutout at 36%

on either Hot Cal channels.

b. Selector switch can remove affected channel from service to restore heaters.

4. Instrument Line Losses

a. A failure of the wet leg (High side of DP cell) would result in indicated level to rise.

1) Minimum Charging and Maximum Letdown.

2) This is also a Vapor space LOCA

b. A failure of the Low side results in a PCS leak! Enter Abnormal Operating Procedures. Depending on the size of the failure, this could result in an indicated PZR level of 0%. Maximum Charging and NO Letdown.

1) Zero percent (0%) indicated PZR level trips all PZR heaters and closes Letdown Orifice Stop Valve (CV-2003)

ES-401 Question 3 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # 00009.EA1.13 Importance Rating __4.4__ _____

K/A Statement: Ability to operate and monitor the following as they apply to a small break LOCA: ESFAS Proposed Question:

Given the following conditions:

The Plant has just tripped from 100% power due to a small break LOCA.

Preferred AC Bus EY-20 was lost during the reactor trip.

Containment pressure is 3.4 psig.

Pressurizer pressure is 1540 psig.

With no operator action, which one of the following describes the response of both the Right and Left Train ESS equipment?

A. NEITHER the Right nor Left Train ESS equipment will be running.

B. BOTH the Right and Left Train ESS equipment are running.

C. ONLY the Right Train ESS equipment is running.

D. ONLY the Left Train ESS equipment is running.

Proposed Answer: D Explanation (Optional):

Preferred AC Bus EY-20 controls the Right Channel ESS equipment. With no control power to actuate the equipment through, the ESS equipment on that Right Channel will not actuate when necessary. Vital AC Bus EY-30 controls the Left Channel ESS equipment. The applicant needs to understand which Vital AC Bus is critical for each Channel and must also understand that a SIAS will occur on low PZR pressure (<1605 psia on 2/4 PZR pressure channels) or on high containment pressure (>3.7 psi to 4.4 psi on 2/4 containment pressure channels).

A. Incorrect, applicant does not believe a valid SIAS will occur under the conditions.

B. Incorrect, applicant believes a loss of EY-20 does not affect Right or Left Channel actuations.

C. Incorrect, applicant believes a loss of EY-20 will only impact the Left Channel.

D. Correct, see explanation Technical Reference(s): DBD-2.05 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ___None__________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam _________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 4 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # 000011.EA2.05 Importance Rating __3.3__ _____

K/A Statement: Ability to determine or interpret the following as they apply to a Large Break LOCA:

Significance of charging pump operation.

Proposed Question:

Given the following:

The Plant tripped from full power due to a large break LOCA.

The Crew has implemented EOP-4.0, LOCA Recovery.

PCS temperature is 540°F and stable

PCS pressure is 1200 PSIA and lowering

The Crew has determined that PCS cooldown is required

2400V Bus 1D is de-energized

P-55C, Charging Pump, is in service providing 40 GPM charging flow Based on the given conditions, PCS cooldown (1) commence (2) .

A. (1) may (2) with no restrictions B. (1) may (2) but must be stopped at 490°F to verify shutdown margin C. (1) cannot (2) until additional charging pumps are started D. (1) cannot (2) until adequate shutdown margin has been verified Proposed Answer: A Explanation (Optional):

EOP-4.0 basis document states it is allowable to commence PCS cooldown as long as emergency boration is in progress. Loss of bus 1D results in loss of LC-12 therefore P-55A and P-55B have no power, in addition this also results in loss of MCC-2 and there is no boric acid pump feed available but boric acid gravity feed is in service (activated on SIAS at PCS pressure of 1605 psia ). Charging pump P-55C is in-service providing 40 GPM charging line flow (this is greater than the minimum of 33 GPM required for emergency boration).

A. Correct, see explanation.

B. Incorrect, part 1 is correct. Part 2 is incorrect as the applicant may confuse this as being correct, this response actually comes from a requirement in EOP-3 for SBO to verify the Reactor will remain shutdown at 50 degree intervals C. Incorrect, the applicant incorrectly believes that additional charging pumps must be started to support emergency boration when in fact P-55C is providing 40 GPM charging flow (procedure requires minimum of 33 GPM)

D. Incorrect, the applicant incorrectly believes that SDM must first be verified prior to commencing cooldown Technical Reference(s): EOP-4.0, EOP-9.0 RA (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __10_

55.43 _____

Comments:

Question from Palisades 2014 Audit Exam.

ES-401 Question 5 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # 000015/17.AA2.08 Importance Rating __3.4__ _____

K/A Statement: Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): When to secure RCPs on high bearing temperature.

Proposed Question:

The Plant is at 100% power when inadvertently, CV-0940, CCW Return From Containment, closes and cannot be opened.

The crew is performing AOP-29, Primary Coolant Pump Abnormal Conditions, in conjunction with AOP-36, Loss of Component Cooling.

Which of the following bearing temperatures would require a manual reactor trip and trip of the affected PCP?

A. Upper Guide bearing temperature 170oF B. Lower Guide bearing temperature 182oF C. Upper Thrust bearing temperature 180oF D. Down Thrust bearing temperature is 171oF Proposed Answer: B Explanation (Optional):

Based on the spurious Containment Isolation Signal, CCW flow to the PCPs is isolated.

Therefore, per AOP-29, the reactor trip criteria are:

Loss of CCW AND o Any PCP bearing temperature exceeding:

Upper Guide bearing - 175oF Lower Guide bearing - 175oF Upper Thrust bearing - 185oF Down Thrust bearing -175oF o Any PCP Lower Seal temperature exceeding 185oF o Any PCP Vapor Seal temperature exceeding185oF A. Incorrect, below the limit of 175oF B. Correct, above the limit of 175oF C. Incorrect, below the limit of 185oF D. Incorrect, below the limit of 175oF

Technical Reference(s): AOP-29, AOP-31_________________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: __None___________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __10_

55.43 _____

Comments:

ES-401 Question 6 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # 000022.G2.4.47 Importance Rating __3.4__ _____

K/A Statement: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Proposed Question:

Given the following conditions:

The Plant is at 100% power.

An instrument failure caused letdown to isolate and letdown cannot be quickly restored.

Charging was manually secured per SOP-2A, Chemical Volume and Control System.

PCS TAVE temperature is stable.

With no further operator action, what is the expected effect of the above conditions?

A. Pressurizer level lowers, Volume Control Tank level rises.

B. Pressurizer level is constant, Volume Control Tank level lowers.

C. Pressurizer level lowers, Volume Control Tank level is constant.

D. Pressurizer level is constant, Volume Control Tank level rises.

Proposed Answer: A Explanation (Optional):

A. Correct, Primary Coolant Pumps controlled bleed-off (CBO) is still rejecting back to the VCT, causing level to rise in the VCT. VCT level will have to be maintained manually by cycling the VCT drain valve. Meanwhile, due to a coolant outflow to CVCS via CBO, PZR level will drop due to a lack of makeup (charging was manually secured).

B. Incorrect, PZR level will drop due to CBO back to CVCS (which is isolated). VCT level will rise due to the input from the PCS via CBO.

C. Incorrect, VCT level will rise due to CBO back to CVCS (which is isolated).

D. Incorrect, PZR level will drop due to CBO back to CVCS (which is isolated).

Technical Reference(s): Primary Coolant System Lesson Plan_________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ___None__________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge ______

Comprehension or Analysis ___X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 7 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # 000025.AK1.01________

Importance Rating __3.9__ _____

K/A statement: Knowledge of the operational implications of the following concepts as they apply to Loss of Residual Heat Removal System: Loss of RHRS during all modes of operation Proposed Question:

Given the following conditions:

The Plant has tripped due to a large break LOCA.

The Crew has transitioned to EOP-4.0, Loss of Coolant Accident Recovery.

The 1C Bus is ENERGIZED.

The 1D Bus FAULTED 10 minutes ago and cannot be restored.

SIRW Tank Level is 2%.

Based on the current plant conditions, which of the following equipment actuations will NOT occur?

A. P-67B LPSI Pump, TRIPS B. CV-3071, P-66A Subcooling Valve, OPENS C. CV-3002, Containment Spray Isolation Valve, THROTTLES D. CV-0826, CCW Heat Exchanger SW Outlet Valve, FULL OPEN Proposed Answer: B Explanation (Optional):

The Recirculation Actuation Signal (RAS) occurs at 2% SIRWT level. As a result, various actions automatically occur (see reference material).

A. Incorrect, LPSI Pumps receive a trip signal generated by the RAS. P-67B will trip. P-67A tripped on the 1D Bus fault prior to the RAS actuation.

B. Correct, CV-3071 requires P-66A to be running in order to open. With the pump tripped, (loss of bus 1D) the valve will remain closed and not open on the RAS.

C. Incorrect, CV-3002 repositions from fully open to throttled to increase the NPSH for the HPSI and CSS Pumps.

D. Incorrect, the CCW Hx SW Outlet valves fully open to allow maximum SWS flow and thus maximum cooling of the CCW flow.

Technical Reference(s): DBD-2.01 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ___None______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam _________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

l ES-401 Question 8 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1___ _____

Group # __1___ _____

K/A # 000026.AK3.02 Importance Rating __3.6__ _____

K/A Statement: Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water: The automatic actions (alignments) within the CCWS resulting from actuation of the ESFAS.

Proposed Question:

Plant conditions are as follows:

A Plant startup is in progress with the Plant in Mode 2.

P-52A, Component Cooling Water (CCW) Pump, is running.

P-52B and P-52C, CCW Pumps, are in standby.

A Loss of Offsite Power (LOOP) has just occurred.

Both Diesel Generator 1-1 and 1-2 are running loaded.

CCW pump P-52A tripped and would NOT restart following the LOOP.

ONE minute later, a Safety Injection actuation occurred due to a small break LOCA.

With no operator action, which ONE of the following describes the configuration of the CCW system at this time?

A. ONLY CCW pump P-52B supplying all CCW loads due to the failure of CCW pump P-52A.

B. ONLY CCW pump P-52C supplying all CCW loads due to the failure of CCW pump P-52A.

C. TWO CCW pumps supplying all CCW loads due to SIAS start of CCW pump P-52B.

D. TWO CCW pumps supplying all CCW loads due to CCW low pressure start of CCW pump P-52B.

Proposed Answer: A Explanation (Optional):

On a SIAS accompanied by a LOOP, P-52C will start only due to a CCW low pressure signal to prevent unnecessary diesel loading. In this case, P-52A tripped upon the NSD sequencer start.

P-52B will start on the NSD sequencer and subsequently again (receive a second start signal) on the DBA sequencer. The DBA sequencer will actuate due to the SIAS that occurred as a result of the LOCA. Since only P-52B will be running at the time of the SIAS, a low pressure signal would not be generated, as the CCW pumps are designed for 100% capacity post-DBA.

A low pressure auto-start condition would be expected upon receipt of a RAS (both CCW HXs inlet valves receive an open signal), which is not present in this case. The applicant must understand the automatic operation of the CCW pumps on a LOOP and a LOCA, as well as the transient operating parameters of the system.

A. Correct, see explanation B. Incorrect, see explanation.

C. Incorrect, see explanation.

D. Incorrect, see explanation.

Technical Reference(s): DBD-1.01, FSAR Chapter 9 Section 9.3.3 Rev 24 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: __None___________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ___________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __8__

55.43 _____

Comments:

ES-401 Question 9 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # 000029.EK1.03____

Importance Rating __3.6__ _____

K/A statement: Knowledge of the operational implications of the following concepts as they apply to the ATWS: Effects of boron on reactivity Proposed Question:

Given the following conditions:

The Plant is at 43% power while shutting down for a refueling outage.

Primary Coolant Pump P-50C spuriously tripped.

The reactor did not automatically trip.

Attempts to trip the reactor per EOP-1.0, Standard Post-Trip Actions, were unsuccessful.

A 40 gpm emergency boration was initiated.

Which statement below correctly describes the expected effect on Moderator Temperature Coefficient (MTC) and Axial Shape Index (ASI) due to the emergency boration?

MTC will be __(1)__ and ASI will be __(2)__.

A. (1) Less negative (2) More negative (less positive)

B. (1) More negative (2) More negative (less positive)

C. (1) Less negative (2) More positive (less negative)

D. (1) More negative (2) More positive (less negative)

Proposed Answer: A Explanation (Optional):

A. Correct, as negative reactivity is added to the PCS and temperature decreases, MTC will increase (become less negative) and ASI will decrease (become more negative). Due to the boration, MTC will add more positive reactivity to the top half of the core than the bottom half of the core and ASI will shift negative.

B. Incorrect, part 1 is incorrect. The applicant believes the value of MTC will become more negative as boron concentration rises.

C. Incorrect, part 2 is incorrect. The applicant believes that ASI will become more positive (less negative) as boron concentration rises.

D. Incorrect, both parts are incorrect.

Technical Reference(s): EM-04-17, General Physics course 192004 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ___None__________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ___________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __1__

55.43 _____

Comments:

ES-401 Question 10 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # 000038.G2.2.44 Importance Rating __4.2_ _____

K/A Statement: 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.

Proposed Question:

Given the following conditions:

The Plant was tripped due to a Steam Generator Tube Rupture.

PCS pressure is 895 psia.

PCS subcooling is 55°F.

Primary Coolant Pumps P-50B and P-50C are running.

A S/G pressure is 790 psia.

A S/G has been isolated per EOP Supplement 12.

A S/G level is 73% NR and rising slowly.

B S/G level is 48% NR and stable.

Which ONE of the following is the preferred method to control level in the isolated steam generator and minimize the spread of contamination?

A. Steam the A S/G to atmosphere via the Atmospheric Dump Valves.

B. Steam the A S/G to the condenser via the Turbine Bypass Valve.

C. Restore S/G blowdown to the condenser from A S/G.

D. Lower PCS pressure below A S/G pressure and allow backflow to the PCS.

Proposed Answer: D Explanation (Optional):

A. Incorrect, this will lower S/G pressure to further below PCS pressure which will increase S/G level and increase the spread of contamination.

B. Incorrect, this will lower SG pressure to further below PCS pressure which will increase S/G level. Steaming to the condenser would minimize the chance of release to the environment, but still spread the contamination to the secondary.

C. Incorrect, re-establishing S/G blowdown will lower S/G level, but will spread contamination to the secondary.

D. Correct, this will lower PCS pressure and reduce SG level by moving water into the PCS. Contamination will be limited by putting the contaminated water back in the PCS.

With PCPs running, the potential for boron dilution is reduced.

Technical Reference(s): EOP-5.0 Bases (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 _10__

55.43 _____

Comments:

ES-401 Question 11 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # CE/E05.EA1.03 Importance Rating __3.4__ _____

K/A Statement: Ability to operate and/or monitor the following as they apply to the (Excess Steam Demand): Desired operating results during abnormal and emergency situations Proposed Question:

The following temperatures exist after an Excess Steam Demand Event in which the B S/G has blown dry:

Loop 1A Tcold: 432oF

Loop 1A Thot: 464oF

Loop 2B Tcold: 424oF

Loop 2B Thot: 456oF The CRS has directed the NCO to stabilize PCS pressure and temperature. Which one of the temperature indications given above should be used to manually control the Atmospheric Dump Valve in order to stabilize temperature with the least amount of PCS heatup?

A. Loop 1A Tcold B. Loop 1A Thot C. Loop 2B Tcold D. Loop 2B Thot Proposed Answer: A Explanation (Optional):

A. Correct, the unaffected S/G Tcold would be used to minimize the heatup.

B. Incorrect, corresponds to the Tcold of the unaffected S/G. Plausible, as the applicant could not understand the least amount of PCS heatup requirement.

C. Incorrect, corresponds to the Thot of the affected S/G. Plausible, as the applicant could chose to use the affected S/G. This could be true in a dual event scenario.

D. Incorrect, corresponds to the Thot of the unaffected S/G. Plausible, as the applicant could chose to use the affected S/G. This could be true in a dual event scenario.

Technical Reference(s): EOP-6.0, Steam Tables (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam St Lucie 2008 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments: Yes, Steam Tables are normally provided as a reference.

ES-401 Question 12 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # CE/E06.EK2.2 Importance Rating __3.5__ _____

K/A Statement: Knowledge of the interrelations between the (Loss of Feedwater) and the following:

Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Proposed Question:

A loss of feedwater accident has occurred at the end of core cycle. When the NCO attempted to trip the Reactor per EOP-1.0, Standard Post-Trip Actions, all Control Rods remained fully withdrawn.

Which of the following describes the initial response of the following plant parameters in response to this event, with no operator action?

Reactor Power Pressurizer Pressure S/G Pressure A. rising rising rising B. rising lowering lowering C. lowering rising rising D. lowering lowering lowering Proposed Answer: C Explanation (Optional):

In an ATWS scenario with a loss of all feedwater, the reduction of secondary system heat removal capability will cause PCS temperature to rise, along with an increase in PCS pressure.

Reactor power will decrease. On the secondary, a loss of feedwater to the S/Gs will cause S/G pressure to increase A. Incorrect, reactor power will decrease due to the addition of negative reactivity via rising PCS temperature.

B. Incorrect, see choice A. Pressurizer pressure and S/G pressure will rise (see explanation).

C. Correct, see explanation.

D. Incorrect, pressurizer pressure and S/G pressure will rise (see explanation).

Technical Reference(s): FSAR Chapter 14 Section 14.13 Rev 24 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ___None__________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 13 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1___ _____

Group # __1___ _____

K/A # 000056.AK1.03 Importance Rating __3.1__ _____

K/A statement: Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite Power: Definition of subcooling: use of steam tables to determine it.

Proposed Question:

Given the following conditions:

The reactor tripped from 100% full power conditions.

The Plant experienced a Loss of Offsite Power and has entered EOP-8.0, Loss of Offsite Power / Forced Circulation Recovery.

Which of the following sets of indications would verify natural circulation flow in at least one Primary Coolant System loop in accordance with EOP-8.0?

PZR Pressure Average CET Loop Thot(s) (oF) Loop Tcold(s) (oF)

A. 1500 psig, stable 571 F, stable o

566 F, lowering o

530oF, stable B. 1550 psig, stable 587oF, stable 576oF, lowering 540oF, lowering C. 1600 psig, stable 578oF, lowering 561oF, stable 530oF, lowering D. 1650 psig, stable 582oF, stable 575oF, lowering 525oF, stable Proposed Answer: A Explanation (Optional):

15 minutes following the SBO, Natural Circulation is developing. Thot and Tcold separate, but the delta T between should be no more than 50oF per Natural Circulation criteria.

The Natural Circulation criteria, per EOP-8.0, are:

1) Core delta T less than 50oF (Average CET minus Tc)

2) Loop Thot(s) and Loop Tcold(s) stable or lowering

3) Average CET at least 25oF subcooled

4) Difference between Loop Thot and Average CET is less than or equal to 15oF A. Correct, all criteria are met (Tsat = 597oF)

B. Incorrect, criteria 3 not met, subcooling = 15oF (Tsat = 602oF)

C. Incorrect, criteria 4 not met, CET - Thot = 17oF (Tsat = 606oF)

D. Incorrect, criteria 1 not met, loop delta T = 57oF (Tsat = 610oF)

Technical Reference(s): EOP-8.0, steam tables (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

ES-401 Question 14 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # 000057. G2.4.03 Importance Rating __3.7_ _____

K/A Statement: Ability to identify post-accident instrumentation.

Proposed Question:

A loss of Preferred AC Bus EY-10 has caused the loss of multiple indications. Which of the following indications will remain available for post-accident use with no operator action?

A. LIA-0102A, PZR Level (Cold Calibrated).

B. LTIR-0101A, Reactor Vessel Level/Qualified Core Exit Thermocouples.

C. SMM-0114, Subcooling Margin Monitor.

D. RIA-2321, Refueling Containment Area Monitor Proposed Answer: C Explanation (Optional):

A. Incorrect, LIA-0102A will fail low on a loss of EY-10. (AOP-12, Att. 1 pg 2 of 5)

B. Incorrect, LTIR-0101A will lose power with a loss of EY-10. (AOP-12, Att. 1 pg 2 of 5)

C. Correct, SMM-0114 is powered by Preferred AC Bus EY-30 and would remain unaffected by a loss of EY-10.

D. Incorrect, RIA-2321 will lose power with a loss of EY-10. (AOP-12, Att. 1 pg 4 of 5)

Technical Reference(s): AOP-12, LCO 3.3.7 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New __X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

Pal 2012 audit exam question used for reference, but question is New.

ES-401 Question 15 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # 000058.AK3.01 Importance Rating __3.4__ _____

K/A Statement: Knowledge of the reasons for the following responses as they apply to the Loss of DC Power: Use of DC control power by D/Gs.

Proposed Question:

Diesel Generator (DG) 1-2 is running partially loaded during a surveillance test run when DC breaker 72-401 (DG 1-2 Field Flashing) on D-21A trips.

Which one of the following describes the effect, if any, on DG 1-2 and why?

A. DG 1-2 output breaker, 152-213, trips on overcurrent to protect the generator from the loss of voltage condition.

B. DG 1-2 output breaker, 152-213, trips on a loss of excitation due to a loss of generator load, to protect the engine from an overspeed condition.

C. No effect on DG 1-2. Field current supplied by the exciter is controlled by the generator voltage regulator automatically after engine startup.

D. No effect on DG 1-2. Field current is not required after the generator develops sufficient voltage upon startup.

Proposed Answer: C Explanation (Optional):

While the DG output breaker will trip on a loss of excitation (as well as an overcurrent condition),

a loss of excitation or overcurrent will not occur due to a loss of DC field flashing power when the DG has reached rated voltage as it has in this case. The DC bus supplies the current for the field flashing as the DG starts, and upon reaching 70% of its rated voltage, the DG exciter will supply the field current. The voltage regulator will maintain this such that there would be no overall impact to the DG.

A. Incorrect, the DG will remain running with the output breaker closed. See explanation.

B. Incorrect, the DG will remain running with the output breaker closed. See explanation.

C. Correct, there is no effect to the DG since the DG is running at rated speed and voltage.

D. Incorrect, field current is required, but the field current is supplied by the DGs exciter.

Technical Reference(s): DBD-5.06, E-8 Sheet 2 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam __________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __8__

55.43 _____

Comments:

Modified from 2007 NRC Exam. Stem of question changed to use D/G 1-2 only partially loaded during the surveillance run. Rearranged and reworded distractors, and changed one distractor (choice B).

EDra PL-EDG Emergency Diesel Generators Revision 11 Page 528 of 53

2) Automatic field flashing is provided from the 125V DC system at 18 Amperes for rapid voltage buildup upon starting.

a) It is automatically removed when the regulator commences operation (when generator voltage is approximately 70% of nominal).

P&ID E-8 Sheet 2 ES-401 Question 16 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # 000062.AA1.06 Importance Rating __2.9__ _____

K/A Statement: Ability to operate and/or monitor the following as they apply to the Loss of Nuclear Service Water (SWS): Control of flow rates to components cooled by SWS Proposed Question:

Given the following conditions:

The Plant was manually tripped from 80% power due to a small break LOCA.

A valid Safety Injection Actuation Signal (SIAS) was received.

A Loss of Offsite Power (LOOP) occurred

Diesel Generator (DG) 1-1 started and loaded normally

DG 1-2 started and tripped on overspeed Service Water cooling demands can be met with __(1)__ Service Water pump(s) if only DG 1-1 is operating, provided manual operator action is taken to isolate service water to __(2)__.

A. (1) ONE (2) non-critical service water header B. (1) ONE (2) containment C. (1) TWO (2) non-critical service water header D. (1) TWO (2) containment Proposed Answer: B Explanation (Optional):

A. Incorrect, while only P-7B will be operating due to the failure of DG 1-2 to energize 1D Bus, the non-critical service water header is isolated automatically on the SIAS. Manual operator action is required to isolate service water cooling flow from containment.

Containment air coolers VHX-1, VHX-2, VHX-3 fans are powered from Bus 1D (as Service Water pump P-7A and P-7C are), water not required for non-operating fans is necessary for other equipment. VHX-4 fan is powered from Bus 1C, however, the service water supply valve to VHX-4 is closed upon the SIAS.

B. Correct, see choice A explanation.

C. Incorrect, see choice A explanation.

D. Incorrect, see choice A explanation Technical Reference(s): EOP-9.0, DBD-1.02 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

Procedural guidance to support required actions are found in AOP-35.

ES-401 Question 17 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1___ _____

Group # __1___ _____

K/A # 000065.AA2.07 Importance Rating __2.8__ _____

K/A statement: Ability to determine and interpret the following as they apply to the Loss of Instrument Air:

Whether backup nitrogen supply is controlling valve position.

Proposed Question:

Given the following conditions:

The Plant is at 100% power.

A transient occurred that resulted in Instrument Air header pressure indicating 62 psig and slowly lowering.

The crew is implementing AOP-37, Loss of Instrument Air.

Which of the following valves will remain unaffected as a result of the transient?

A. CV-2099, PCP Controlled Bleedoff Containment Isolation B. CV-0847, SW Supply to Containment C. CV-1212, Service Air Header Isolation D. CV-0909, CCW Outlet from Letdown Heat Exchanger Proposed Answer: B Explanation (Optional):

A. Incorrect, CV-2099 fails closed on a loss of instrument air pressure.

B. Correct, CV-0847 uses nitrogen backup from nitrogen backup station 1A and >80psig nitrogen will passively allow operation of the valve for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the loss of compressor operability.

C. Incorrect, CV-1212 fails closed on a loss of instrument air pressure to isolate Service Air from Instrument Air.

D. Incorrect, CV-0909 will fail open on a loss of instrument air pressure to ensure letdown is maintained adequately cooled.

Technical Reference(s): AOP-37, DBD-1.05 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ___None__________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New __X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __8__

55.43 _____

Comments:

ES-401 Question 18 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __1__ _____

K/A # 000077.AK2.07 Importance Rating __3.6__ _____

K/A statement: Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: Turbine / generator control Proposed Question:

Given the following conditions:

A grid disturbance occurred due to severe weather in the area.

The Plant is at 825 MWe.

Generator reactive load is 225 MVAR OUT.

Generator Hydrogen pressure is 75 psig.

The following annunciators are LIT.

o EK-0303, VOLTAGE REGULATOR LIMITER OPERATION.

o EK-0310, GENERATOR VOLTAGE REG TRIP.

To restore Generator parameters, the Control Room Supervisor (or CRS) should direct the NCO-T to __(1)__ reactive load by using the __(2)__.

A. (1) RAISE (2) AC Adjuster B. (1) RAISE (2) DC Adjuster C. (1) LOWER (2) AC Adjuster D. (1) LOWER (2) DC Adjuster Proposed Answer: D Explanation (Optional):

The Generator has three protective functions, the Maximum Excitation Limiter (MXL), the On-line Field Forcing (FF) relay, and the Over Excitation Protection (OXP) relay. The MXL will require the Voltage Regulator to attempt to limit field current to 273 amps (as evident by alarm EK-0303) while in auto (i.e. AC Regulate Mode using the AC Adjuster). However, during the transient, the Voltage Regulator tripped (alarm EK-0310), preventing further control in auto. At this point, any control must be made in manual (i.e. DC Regulate Mode using the DC Adjuster).

The unit is in a stable condition, however, based on the information provided, the unit is

operating on the verge of overexcitation. To restore reactive load and maintain a safe operating condition within the bounds of the Generator Capability Curve, the NCO-T must LOWER reactive load using the DC Adjuster. The first part of this question requires the applicant to interpret the generator capability curve and understand that reactive load must be lowered, rather than raised. The second part of the question requires the applicant to realize that, based on the alarms in, the voltage regulator has tripped and action must be taken using the DC adjuster to control reactive load.

A. Incorrect, see explanation B. Incorrect, see explanation C. Incorrect, see explanation D. Correct Technical Reference(s): SOP-8, ARP-2___________________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: SOP-8 Attachment 4 (Generator Capability Curve Only)

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New __X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 19 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __2__ _____

K/A # 000003.AK3.08 Importance Rating __3.1__ _____

K/A Statement: Knowledge of the reasons for the following responses as they apply to the Dropped Control Rod: Criteria for inoperable control rods.

Proposed Question:

In which of the following conditions would a Control Rod be required to be declared inoperable?

(Assume initial plant conditions at full power.)

A. Rod 39 indicates 7 from the other rods within its group.

B. Rod 16 drops to 126" withdrawn.

C. Rod 6 seal leak off high temperature alarm is locked in.

D. CRD matrix is lost due to a failure of Instrument AC Bus EY-01.

Proposed Answer: B Explanation (Optional):

A. Incorrect, the control rod operability misalignment limit is greater than 8.

B. Correct, Rod 16 is a shutdown group B rod. Shutdown group rods are approximately 131 withdrawn at normal full power conditions. Shutdown rods must be > 128 withdrawn to be considered operable.

C. Incorrect, high seal leakoff temperature could be indicative of exceeding the PCS leakage Tech Spec for identified leakage. However, this condition does not cause the control rod to be considered inoperable per LCO 3.1.4 or 3.1.5 D. Incorrect, with a loss of Instrument Bus EY-01, the CRD matrix and CRD LED displays on EC-02 are lost. For the primary rod position indication system to be operable, the digital position readout or the PPC display must provide valid rod position indication, or for regulating and part-length rods, the cam operated red matrix light gives positive indication of rod position. The PPC will remain unaffected as a result of this transient.

Technical Reference(s): Tech Spec 3.1.4 bases, SOP-6, DBD-2.06, ARP-5, PL-CRD Control Rod Drive System Lesson Plan (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2005 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

Modified one distractor, and replaced one distractor with new.

ES-401 Question 20 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __2__ _____

K/A # 000024.AA1.17 Importance Rating __3.9__ _____

K/A Statement: Ability to operate and/or monitor the following as they apply to Emergency Boration:

Emergency borate control valve and indicators Proposed Question:

Given the following conditions:

The Reactor was tripped from 100% power.

Three (3) Control Rods remain fully withdrawn.

P-66B HPSI Pump is running.

P-56A Boric Acid Pump is running.

P-55A Charging Pump is running.

EOP-1.0, Standard Post-Trip Actions, is in progress.

The crew is initiating Emergency Manual Boration per SOP-2A, Chemical Volume and Control System.

Which of the following Emergency Boration flowpaths should be selected if VCT outlet valve (MO-2087) is open and will NOT close from the Main Control Board?

A. Open MO-2169 and MO-2170, Gravity Feed Valves.

B. Open MO-2160, SIRWT to Charging Pump Suction.

C. Open MO-2140, Pumped Feed Valve.

D. Open MO-3072, CVCS to HPSI Train 2.

Proposed Answer: C Explanation (Optional):

A. Incorrect, with the gravity feed valves MO-2169/2170 open, the VCT outlet (MO-2087) must be closed to ensure adequate boration capability is maintained. If the lineup cannot be satisfied, another lineup must be pursued.

B. Incorrect, while borating from the SIRWT is an acceptable emergency boration flowpath, it is not the correct flowpath to use with the given conditions. Per SOP-2A, the pumped feed or gravity feed lineup should be used for performing the emergency boration.

C. Correct, at least one borated flowpath must be lined up. Opening MO-2140 to allow pumped feed to the suction of the charging pumps satisfies that lineup.

D. Incorrect, incorrect procedure adherence. While SOP-2A does allow for alternate PCS injection using HPSI, it is primarily intended to accommodate isolation of the normal

charging path to the PCS when it is anticipated to be of short duration and actual makeup to the PCS will not be required.

Technical Reference(s): SOP-2A (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2003__________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __8__

55.43 _____

Comments:

Modified stem only. All distractors remain unchanged.

ES-401 Question 21 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __2__ _____

K/A # 000028.G2.4.31 Importance Rating __4.2_ _____

K/A Statement: Knowledge of annunciator alarms, indications, or response procedures.

Proposed Question:

Given the following conditions:

The Plant is at 100% power.

Pressurizer level control is selected to Channel B.

Pressurizer Heater Control is selected to Channel A & B.

A transient occurred which resulted in:

Pressurizer level is 51% and lowering.

Annunciator EK-0761 PRESSURIZER LEVEL HI-LO alarming.

Which of the following actions would NOT automatically occur as a result of this transient?

A. P-55A Charging Pump operates at maximum speed.

B. All Pressurizer Heaters de-energize.

C. P-55B and P-55C Charging Pumps start.

D. CV-2004 and CV-2005, Letdown Orifice Stop Valves, close.

Proposed Answer: B Explanation (Optional):

A. Incorrect, P-55A will increase speed to maximize charging flow.

B. Correct, pressurizer heaters only de-energize on a low-low level (36%).

C. Incorrect, both P-55B and P-55C and constant 40gpm pumps when operating and are cycled on/off as necessary to maintain level. Both pumps will start on a low pressurizer level.

D. Incorrect, letdown orifice isolation valves will close in an attempt to maintain maximum charging flow (i.e. maximize the input, minimize the output).

Technical Reference(s): ARP-4, PL-PLCS Pressurizer Level Control System Lesson Plan

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

PL-PLCS Pressurizer Level Control System Rev. 5 A. Alarms (ARP-4)

1. EK-0761, "PRESSURIZER LEVEL HI-LO"

a. Sensor:

1) LIC-0101A or LIC-0101B, whichever is selected by HS 1/LRC-0101

b. Setpoint:

1) 5.75% level deviation from setpoint.

c. Alarm impact on PZR Level Control:

1) AUTO: IF high alarm, THEN:

a) CV-2004 and CV-2005 Orifice Stop Valves open.

b) Charging Pumps P-55B and P-55C stop.

c) Charging Pump P-55A operates at minimum speed (33 gpm).

d) Pressurizer Backup Heaters energize (in "AUTO" position only).

2) AUTO: IF low alarm, THEN:

a) CV-2004 and CV-2005 Orifice Stop Valves close.

b) Charging Pumps P-55B and P-55C start.

c) Charging Pump P-55A operates at maximum speed (53 gpm).

3) Indications to validate alarm:

a) LIC-0101A or LIC-0101B which ever is selected by HS 1/LRC-0101.

b) Check Charging and Letdown response correct; IF not, THEN shift selected level controllers LIC 0101A or LIC 0101B.

2. EK-0763, "PRESSURIZER LEVEL CH 'A' LO-LO"

a. Sensor:

1) LIC-0101AL (Hot Cal)

b. Setpoint:

1) 36%

c. Alarm impact on PZR Level Control:

1) AUTO: IF "A" or "A & B" PZR heater control is selected, THEN all PZR heaters are tripped. Letdown Orifice Stop Valve CV-2003 closes.

d. Indications to validate alarm:

1) All available Hot Cal level instruments in the control room

3. EK-0764, "PRESSURIZER LEVEL CH 'B' LO-LO"

a. Sensor:

1) LIC-0101BL (Hot Cal)

b. Setpoint:

1) 36%

c. Alarm impact on PZR Level Control:

1) AUTO: IF "B" or "A" & "B" PZR heater control is selected, THEN all PZR heaters are tripped. Letdown Orifice Stop Valve CV-2003 closes.

d. Indications to validate alarm:

1) All available Hot Cal level instruments in the control room

ES-401 Question 22 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __2__ _____

K/A # 000032.AK3.02 Importance Rating __3.7__ _____

K/A Statement: Knowledge of the reasons for the following responses as they apply to the Loss of Source Range Nuclear Instrumentation: Guidance contained in EOP for loss of source-range nuclear instrumentation.

Proposed Question:

Given the following conditions:

The Plant was at 89% power when a transient occurred which required a manual reactor trip.

Three (3) control rods did NOT insert into the core.

The crew has just entered EOP-01, Standard Post-Trip Actions.

If both Source Range Nuclear Instruments have become inoperable, what is the effect, if any, on the Reactor Operator's ability to check the status of the Reactivity Control safety function?

A. No effect, since Reactivity Control is satisfied due to Xenon building in for the next approximately 10-12 hours.

B. Reactivity Control must be satisfied by manually driving down ONE of the stuck control rods.

C. Will need to check Reactor power at less than 1E-4 % and constant or lowering on Wide Range Excore indication.

D. Will need to check Reactor power at less than 2% using delta T power indication.

Proposed Answer: C Explanation (Optional):

A. Incorrect, although Xenon will behave this way, this is not an approved methodology for satisfying the Reactivity Control safety function.

B. Incorrect, as, per the stem conditions, this would still result in having more than one control rod stuck fully withdrawn.

C. Correct, per EOP-1.0 Event Diagnostic Flowchart Reactivity Control Requirements.

D. Incorrect, again, this is not an approved methodology for satisfying the Reactivity Control safety function.

Technical Reference(s): EOP-1.0 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ___None__________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2003 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 _10__

55.43 _____

Comments:

Modified question from Palisades 2003 NRC Exam. Modified stem such that Source Range instrumentation is lost rather than Wide Range and changed correct answer to new answer based upon second conditional statement for verification of safety function.

NOTE: 3 Stuck control Rods does not necessarily drive the CRS to transition to EOP 9.0. In EOP 9.0, RC-2 would be the applicable Success Path.

ES-401 Question 23 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __2__ _____

K/A # 000051.AA2.02 Importance Rating __3.9__ _____

K/A Statement: Ability to determine and interpret the following as they apply to the Loss of Condenser Vacuum: Conditions requiring reactor and/or turbine trip.

Proposed Question:

Given the following conditions:

A degrading condenser vacuum condition has caused the Crew to enter AOP-6, Loss of Condenser Vacuum.

Condenser vacuum has degraded to 21.7Hg and is slowly lowering.

Main Generator output is approximately 465 MWe.

Based on these conditions, the crew should A. Trip the reactor and enter EOP-1.0, Standard Post-Trip Actions.

B. Trip the turbine and continue efforts to correct the problem in AOP-6, Loss of Condenser Vacuum.

C. Commence a Power reduction, per GOP-8, POWER REDUCTION AND PLANT SHUTDOWN TO MODE 2 OR MODE 3 > 525 oF, to stabilize condenser vacuum.

D. Commence a Rapid Power Reduction, per AOP-7, RAPID POWER REDUCTION, to stabilize condenser vacuum.

Proposed Answer: A Explanation (Optional):

A. Correct, per AOP-6. The reactor should be tripped if condenser vacuum lowers to less than 22 Hg.

B. Incorrect, while the turbine does need to be tripped under these conditions, the reactor must also be tripped since reactor power is greater than 15%. If reactor power was less than 15%, this would be true.

C. Incorrect, under these conditions, a reactor trip is required. If condenser vacuum were less than 24Hg, this would be a suitable action.

D. Incorrect, under these conditions, a reactor trip is required. If condenser vacuum were less than 24Hg, this would be a suitable action.

Technical Reference(s): AOP-6 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New __X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __10_

55.43 _____

Comments:

Pal 2010 question 25 was used for reference; question as written is new.

ES-401 Question 24 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __2__ _____

K/A # 000067.AK1.01 Importance Rating __2.9 __ _____

K/A Statement: Knowledge of the operational implications of the following concepts as they apply to Plant Fire on Site: Fire classifications, by type.

Proposed Question:

Nuclear Plant Operators have just completed racking in breaker 52-1111, Main Exhaust Fan V-6B. Upon remote closure of the breaker, an arc flash occurred and a sustainable fire followed.

What is the class of fire and what type of extinguisher shall be utilized, per fire protection procedures?

A. Class B fire; use water OR Yellow Foray dry chemical fire extinguisher.

B. Class B fire; use any type of dry chemical OR CO2 fire extinguisher.

C. Class C fire; use ONLY a Purple K type dry chemical fire extinguisher.

D. Class C fire; use any type of dry chemical OR CO2 fire extinguisher.

Proposed Answer: D Explanation (Optional):

Class A fires are ordinary combustible fires. Water extinguishers are used on Class A fires.

Class B fires are flammable and combustible liquid and gas. ABC dry chemical extinguishers are used on Class A, Class B, or Class C fires and are filled with Monoammonium Phosphate, a yellow-colored dry chemical (yellow foray). BC dry chemical extinguishers are used on Class B or Class C fires and are filled with Purple K Potassium Bicarbonate, a purple-colored dry chemical.

Class C fires are electrical fires. Carbon Dioxide (CO2) extinguishers are used on Class B or Class C fires. This is an electrical fire that should be extinguished using CO2 and/or dry chemical extinguishers, per FPIP-4 Fire Protection Systems and Fire Protection Equipment.

A. Incorrect, applicant does not understand the class of fire or the type of extinguisher to use. See explanation.

B. Incorrect, applicant understands the type of extinguisher to use, but not the class of fire.

See explanation.

C. Incorrect, applicant understand the class of fire, but not type of extinguisher to use. See

explanation.

D. Correct, any type of dry chemical (ABC or BC) or CO2 extinguisher is adequate. See explanation.

Technical Reference(s): FPIP-4 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: __ OB 44825(1101)___________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

Question modified from Palisades 2003 Audit Exam. Modified stem to change question from Class A fire to a Class C fire. Two distractors were changed, including changing the answer.

ES-401 Question 25 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __2__ _____

K/A # 069.AK2.03 Importance Rating __2.8__ _____

K/A statement: Knowledge of the interrelations between the Loss of Containment Integrity and the following: Personnel access hatch and emergency access hatch Proposed Question:

Given the following conditions:

The Plant is in Mode 5 on Shutdown Cooling (SDC) preparing to heat up the Primary Coolant System (PCS).

Both Personnel airlock doors are open.

A loss of SDC occurs and PCS temperature rises to 214oF.

Which one of the following statements describes the current status of containment integrity?

Containment integrity:

A. Is not required in the current Mode.

B. Is not required until all OPERABLE PCS Tcold instruments read greater than 220 oF.

C. Has been lost. At least one OPERABLE airlock door must be closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

D. Has been lost. Both OPERABLE airlock doors must be maintained closed at all times.

Proposed Answer: C Explanation (Optional):

A. Incorrect, containment integrity is required for current plant conditions. LCO 3.6.2 is applicable in Modes 1-4. The applicant must understand that due to PCS temperature rising > 200oF, the plant is now in Mode 4, where LCO 3.6.2 is applicable.

B. Incorrect, containment integrity must be restored within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This is incorrect as the applicant chose to maintain the doors open and place the plant into a Mode where LCO 3.6.2 does not apply (Mode 5).

C. Correct, at least one airlock door must be closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months IAW LCO 3.6.2.

D. Incorrect, only one airlock door is required to be closed IAW LCO 3.6.2.B Technical Reference(s): LCO 3.6.2.A, AOP-32_____________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ___None__________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam _______

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _10___

55.43 _____

Comments:

Based upon the initial conditions given in the stem, both airlock doors are still OPERABLE, but the interlock mechanism is not, since both door are open simultaneously. This means that TS 3.6.2.B must be entered and required actions taken upon entering Mode 4.

ES-401 Question 26 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1___ _____

Group # __2___ _____

K/A # CE/A11.AK1.2 Importance Rating __3.0__ _____

K/A Statement: Knowledge of the operational implications of the following concepts as they apply to the (RCS Overcooling): Normal, abnormal and emergency operating procedures associated with (RCS Overcooling).

Proposed Question:

Given the following:

A Main Steam Line Break has occurred upstream of the B Steam Generator MSIV.

Main Steam Line Isolation has automatically actuated.

Which one of the following is of concern if a steaming path from the unaffected Steam Generator is not established immediately following dryout of the affected Steam Generator?

A. Void formation in the Reactor Vessel upper head region.

B. Rapid rise in Average QCET temperatures causing a loss of natural circulation.

C. Rapid repressurization of the PCS and subsequent pressurized thermal shock.

D. Rapid rise in Tcold of the unaffected loop which would result in a loss of natural circulation.

Proposed Answer: C Explanation (Optional):

A. Incorrect, void formation is an undesirable condition, but the concern does not apply here.

B. Incorrect, loss of natural circulation is a concern, but does not apply for given conditions.

C. Correct, prevents an uncontrolled heatup and repressurization due to loss of decay heat removal.

D. Incorrect, loss of natural circulation is a concern, but does not apply for given conditions.

Technical Reference(s): EOP-6.0 Basis___________________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ___None__________

Learning Objective: _________________________ (As available)

Question Source: Bank # __X____

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2003___________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __10_

55.43 _____

Comments:

Rearranged distractors C & D based upon length of distractors. C is now correct answer.

ES-401 Question 27 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1___ _____

Group # __2___ _____

K/A # CE/E09.EA2.1 Importance Rating __3.2__ _____

K/A statement: Ability to determine and interpret the following as they apply to the (Functional Recovery):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Proposed Question:

The CRS has implemented EOP-9.0, Functional Recovery, due to a Steam Generator (S/G)

Tube Rupture on the A S/G and a stuck open Main Steam Safety Valve (MSSV) on the B S/G.

Which of the following describes the required operator actions?

A. Isolate the A S/G. Use the B S/G for heat removal.

B. Isolate the B S/G. Use the A S/G for heat removal.

C. Isolate both S/Gs. Use Once Through Cooling for heat removal.

D. Isolate the B S/G MSSV by gagging it closed, then use the B S/G for heat removal.

Proposed Answer: B Explanation (Optional):

In a dual event scenario, the applicant must determine which S/G is considered the most affected S/G and isolate that S/G. With a dual event scenario, the applicant must assess the current plant conditions and make that choice. With a stuck open MSSV, the S/G will inevitably blow dry and there is little the operator can do to mitigate that. Isolating that S/G will not isolate the steam path past the MSSV. However, once the S/G is blown dry, the operator must use the ruptured S/G to cooldown in order to maintain heat sink availability. In this case, the operator can control the amount of radioactivity being released via the ruptured S/G and minimizing off site dose should be the primary concern in this case.

A. Incorrect, the B S/G is considered the most affected S/G and the ruptured S/G (A) must be used for cooldown.

B. Correct, the B S/G is considered the most affected S/G and the ruptured S/G (A) must be used for cooldown.

C. Incorrect, isolating both S/Gs and utilizing Once Through Cooling for heat removal is a last resort method, even during a dual event scenario.

D. Incorrect, incorrect procedure compliance.

Technical Reference(s): EOP-9 HR-2 and Bases_____________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ___None__________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam _St Lucie 2000_________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __10_

55.43 _____

Comments:

Modified question stem and changed distractor D. Used different S/Gs to change answer.

ES-401 Question 28 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 003.K1.01 Importance Rating __2.6__ _____

K/A statement: Knowledge of the physical connections and/or cause-effect relationships between the RCPS and the following systems: RCP lube oil Proposed Question:

The Plant is on Shutdown Cooling. You have been directed to start Primary Coolant Pump P-50B during plant startup. After starting the AC Oil Lift Pump P-80B, you note that the WHITE PUMP START OIL PERMISSIVE light just above the P-50B handswitch does NOT illuminate.

Which of the following is the alternate method of satisfying the required oil permissive interlock?

A. Start P-50B, Primary Coolant Pump, without delay.

B. Start the DC Oil Lift Pump.

C. Notify Maintenance to prime the Oil Lift Pumps.

D. Wait two minutes and attempt to start P-50B, Primary Coolant Pump.

Proposed Answer: B Explanation (Optional):

A. Incorrect, the breaker will not close without the Pump Start Oil Permissive satisfied B. Correct.

C. Incorrect, this is done if the Pump Start Oil Permissive is not met with both AC and DC Oil Lift Pumps running.

D. Incorrect, the two minute duration for starting the PCP is met after the Lift Oil Pump(s) have run with the Pump Start Oil Permissive satisfied.

Technical Reference(s): SOP-1A________________________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2001___________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 29 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 004.K2.05 Importance Rating __2.7__ _____

K/A Statement: Knowledge of bus power supplies to the following: MOVs Proposed Question:

Given the following conditions:

The Plant tripped from 100% power.

Emergency boration requirements are met.

Bus 19 is faulted and cannot be re-energized.

An emergency boration can be performed from the Control Room using which of the following valves?

A. MO-2170, Boric Acid Tank T-53B Gravity Feed Isolation.

B. MO-2169, Boric Acid Tank T-53A Gravity Feed Isolation.

C. MO-2140, Boric Acid Pump Feed Isolation Valve.

D. MO-2087, VCT Outlet Isolation Valve.

Proposed Answer: C Explanation (Optional):

A. Incorrect, MO-2170 power supply is MCC-1 (fed from Bus 19)

B. Incorrect, MO-2169 power supply is MCC-1 (fed from Bus 19)

C. Correct, MO-2140 power supply is MCC-2 (fed from Bus 20). BA Pump P-56A remains energized to allow for pumped feed boration capability.

D. Incorrect, MO-2087 power supply is MCC-1 (fed from Bus 19)

Technical Reference(s): SOP-2A, PL-CVCS Chemical and Volume Control System Lesson Plan, E-1 Sheet 1, E-4 Sheet 1 & 2 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 30 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2___ _____

Group # __1___ _____

K/A # 004.A1.09 Importance Rating __3.6__ _____

K/A Statement: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: RCS Pressure and Temperature Proposed Question:

Given the following conditions:

The Plant is solid in Mode 5.

One train of Shutdown Cooling (SDC) is in service.

PIC-0202, Intermediate Pressure Letdown Controller, is in MANUAL for control of Intermediate Letdown Regulating Valve, CV-2012.

If CV-3025, SDC Hx Outlet, is throttled CLOSED, what is the effect on PCS temperature AND how would CV-2012 be operated to maintain letdown pressure on setpoint?

PCS temperature will:

A. RISE requiring CV-2012 to be throttled OPEN.

B. LOWER requiring CV-2012 to be throttled CLOSED.

C. RISE requiring CV-2012 to be throttled CLOSED.

D. LOWER requiring CV-2012 to be throttled OPEN.

Proposed Answer: A Explanation (Optional):

If CV-3025 was throttled closed, PCS temperature would rise. With PCS temperature rising, PCS pressure will rise. Letdown regulating valve CV-2012 will auto open to lower pressure back to setpoint. PIC-0202 output raises causing CV-2012 to OPEN to lower letdown pressure.

A. Correct. See explanation.

B. Incorrect. See explanation. CV-3025 operates to control PCS temperature. This would be the response if the valve were throttled closed C. Incorrect. See explanation.

D. Incorrect. See explanation. CV-3025 operates to control PCS temperature. This would be the response if the valve were throttled closed Technical Reference(s): SOP-1B, PL-CVCS CVCS Lesson Plan Rev 7

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ___________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

CVCS Lesson Plan PL-CVCS Revision 7 Page 75 of 76

e. CVs-2012, 2122 (Intermediate Letdown Regulating valves)

1) 2 valves in parallel

2) Maintains letdown pressure above saturation (and less than RV-2006 setpoint) pressure between the letdown heat exchanger and pressure reduction orifices. Prevents flashing (boiling) of letdown fluid prior to being cooled by the letdown heat exchanger.

3) Reduces letdown pressure to design limits of demins

4) Controls PCS pressure during solid plant operations. Varies letdown flow for fixed manual charging rate. Controls PCS pressure by controlling the PCS volume. Selected Intermediate Pressure Letdown Control Valve will control PCS pressure by adjusting letdown flow. This pressure can be set or adjusted via the valve controller on C-02. A change in temperature will cause either a contraction of the water molecules in the PCS (lower temperature) or expansion of the water molecules in the PCS (higher temperature) and Letdown flow will need to be adjusted accordingly.

5) 2 controllers a) PIC-0202 (Intermediate Pressure Letdown Controller) on C-02 Can control in either:

(1) Automatic (a) Auto controls via input from PT-0202 (Low Pressure Letdown Pressure Transmitter) and also receives an anticipatory signal from the Letdown Stop Valves to prevent/-reduce pressure transients and reduce cycling of RV-2006.

(b) PIC-0202 "remembers" what its output should be for 1, 2 and 3 orifices open. When an orifice either opens or closes, a constant is added to the controller output, which will cause a step change in the controller output. Once the constant has been added, the controller feedback will take over and control at its setpoint.

(c) Auto pushbutton will be lit (d) Red pointer shows Letdown pressure (e) Blue pointer shows Letdown pressure setpoint (f) Meter shows output to CV (g) Lever has no function in Auto (h) Fail light lit, call I&C (i) Alarm light lit, call I&C

(2) Manual (a) Manual pushbutton will be lit, (b) Red pointer shows Letdown pressure (c) Blue pointer shows Letdown pressure setpoint, has no function in Manual (d) Meter shows output to CV (e) Lever will open and close CV (f) Fail light lit, call I&C (g) Alarm light lit, call I&C b) HIC-2122 (Intermediate Pressure Letdown Controller) on C-12 Manual control only (1) Red pointer shows Letdown pressure (2) Blue pointer shows signal to CV, corresponds to position (3) Knob opens and closes CV c) HS-0202 (on C-12)

(1) Determines which valve is controlled by which controller.

(2) 2 positions (a) CV-2122 Manual / CV-2012 Auto CV-2122 is controlled by HIC-2122 and CV-2012 is controlled by PIC-0202 (b) CV-2122 Auto / CV-2012 Manual CV-2122 is controlled by PIC-0202 and CV-2012 is controlled by HIC-2122

ES-401 Question 31 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 005.K3.07________

Importance Rating __3.2__ _____

K/A Statement: Knowledge of the effect that a loss or malfunction of the RHRS will have on the following:

Refueling operations Proposed Question:

Given the following conditions:

The Plant is shut down for a refueling outage

Core reload is in progress

LPSI Pump P-67B is in service providing Shutdown Cooling

Qualified CETs indicate 105oF

Reactor cavity level is 647 feet

The following alarms have just alarmed:

o EK-1101, Containment Instr Air Lo Press o EK-1102, Instrument Air Lo Press o EK-1103, Service Air Lo Press

Instrument Air header pressure is lowering at the rate of 15 psig per minute Complete the following statements:

After 5 MINUTES, the Primary Coolant System cooldown rate will __(1)__ and core reload must be suspended __(2)__.

A. (1) Rise (2) immediately B. (1) Rise (2) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months C. (1) Lower (2) immediately D. (1) Lower (2) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Proposed Answer: C Explanation (Optional):

A loss of Instrument Air (~25 psig after the 5 minute duration) will result in the SDC Hx bypass valve to fail open (CV-3006) and the SDC Hx outlet valve (CV-3025) to fail closed. This will result in the PCS cooldown rate to lower (i.e. PCS temperature will rise as a result of the loss of SDC). Per AOP-30 Step 9, fuel movements are to be stopped upon a loss of shutdown cooling, with no allowable duration.

A. Incorrect, the applicant does not understand the relationship between a loss of instrument air and shutdown cooling, specifically the impact on the PCS cooling RATE.

While PCS temperature will RISE, the cooldown rate will LOWER.

B. Incorrect, see choice A. Additionally, the applicant could not understand the procedural requirements for a loss of shutdown cooling.

C. Correct, see explanation.

D. Incorrect, the applicant does not understand the procedural requirements for a loss of shutdown cooling.

Technical Reference(s): DBD-2.01, AOP-37 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ___None______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam __________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _7__

55.43 _____

Comments:

ES-401 Question 32 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 005.A2.03 Importance Rating __2.9_ _____

K/A Statement: Ability to (a) predict the impacts of a RHR pump/motor malfunction, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations.

Proposed Question:

The following conditions exist.

The Plant is in Mode 5.

Shutdown cooling is in-service using LPSI Pump P-67A.

Primary Coolant Pumps P-50B and P-50D are in-service.

All PCS and SDC temperatures are slowly lowering.

The control room receives annunciator EK-1162, LPSI Pump Low Discharge Pressure.

Following the alarms receipt, the control room operators observe the following indications:

PCS Pressure slowly rising.

PCS Temperatures slowly rising.

Pressurizer Level Off-Scale High.

SDC temperatures are stable.

Red indicating light for P-67A is LIT.

SDC Hx valves CV-3006 and CV-3025 position indication remain unchanged.

SDC Flow is zero.

Which of the following is a possible explanation of the indications provided and what would be the appropriate course of action per the applicable procedure?

A. Problems with the SDC Hx valves CV-3006 and/or CV-3025; dispatch an NLO to investigate.

B. Problems with CCW Cooling flow to the SDC Hx; dispatch an NLO to investigate.

C. Problems with LPSI Pump P-67A; trip LPSI Pump P-67A.

D. Problems with LPSI Pump P-67A; start LPSI Pump P-67B.

Proposed Answer: C Explanation (Optional):

A. Incorrect, CV-3006 and CV-3025 positions indications remain unaffected and SDC flow is zero. Both SDC Hx control valves would have to close to lose flow indication.

B. Incorrect, problems with CCW flow to the SDC Hx would not explain a loss of SDC flow and stable SDC temperatures.

C. Correct, per AOP-30 reactor and equipment trip criteria, if shutdown cooling flow is less than 170 gpm with an operating LPSI pump, that pump shall be tripped.

D. Incorrect, incorrect procedure adherence. LPSI Pump P-67A must be tripped and the low shutdown cooling flow conditions resolved prior to starting P-67B.

Technical Reference(s): P&ID 204 sheet A, ARP-7, AOP-30 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

Question modified from Palisades 2006 Audit Exam. Modified question stem, changed correct answer, replaced one distractor.

ES-401 Question 33 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 006.K4.11 Importance Rating __3.9__ _____

K/A Statement - Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following:

Reset of SIS Proposed Question:

Given the following conditions:

The Plant is being cooled down and depressurized in preparation for a refueling outage.

Safety Injection Signal (SIS) has been BLOCKED.

A failure of the Pressurizer pressure controller causes PCS pressure to rise from 1550 psia to the following:

o A Channel - 1700 psia o B Channel - 1685 psia o C Channel - 1695 psia o D Channel - 1705 psia Based on the above conditions, the Safety Injection Signal is:

A. No longer blocked since 3/4 pressure channels have increased above the reset setpoint.

Safety Injection WILL actuate when pressure is lowered to <1605 psia.

B. No longer blocked since 3/4 pressure channels have increased above the reset setpoint.

Safety Injection WILL actuate when pressure is lowered to <1690 psia.

C. Still blocked since not all of the pressure channels have increased above the reset setpoint. Safety Injection WILL NOT actuate when pressure is lowered.

D. Still blocked since the block switches have not been placed to RESET. Safety Injection WILL NOT actuate when pressure is lowered.

Proposed Answer: A Explanation (Optional):

A. Correct, when 3 of 4 pressure channels increase to >1690 psia, the SIS rearms itself (unblocks). When 2/4 pressure channels lower to <1605 psia, SIS will re-actuate.

B. Incorrect, applicant is correct that the SIS is no longer blocked but misapplies the setpoint of initiation.

C. Incorrect, applicant misapplies the reset logic, in that all 4 do not have to rise above 1690.

D. Incorrect, applicant believes that the SIS block must be reset with the switch but it automatically resets on pressure.

Technical Reference(s): E-17 Sheet 3, PL-SIS Safety Injection System Lesson Plan (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2009 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

PL-SIS Safety Injection System Revision 6 Page 13 of 74 A. Blocking SI Actuation

1. What conditions are necessary to block SIS?

A: 3 of 4 pressures (PIA-0102ALL, 0102BLL, 0102CLL, 0102DLL) less than 1690 psia. Then PB 3-1 and PB 3-2, SIS Block Switches are each turned and momentarily held to physically block each train of SIAS.

ES-401 Question 34 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 006.A1.13 Importance Rating __3.5__ _____

K/A Statement: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including: Accumulator pressure (level, boron concentration).

Proposed Question:

Given the following conditions:

The Plant is at 100% power.

Safety Injection Tank (SIT) T-82B has a high level alarm in.

SOP-3 Attachment 3 is being performed to lower level in T-82B.

Chemistry has provided an SIT T-82B boron concentration, at the beginning of the shift, of 2520 ppm

SIT T-82B level is 190. (~38% narrow range)

SIT T-82B pressure is 210 psig.

With the given conditions, what is the current operability condition of SIT T-82B and why?

A. Operable, all SIT parameters are within LCO 3.5.1 limits.

B. Inoperable, the boron concentration is not within LCO 3.5.1 limits.

C. Inoperable, the level is not within LCO 3.5.1 limits.

D. Inoperable, the pressure is not within LCO 3.5.1 limits.

Proposed Answer: B Explanation (Optional):

A. Incorrect, boron concentration is out of spec high.

B. Correct, boron concentration must be 1720-2500 ppm.

C. Incorrect, level is within the volumetric requirement of 1040-1176 ft3 (174-200).

D. Incorrect, pressure is >200 psig, as required per LCO 3.5.1.

Technical Reference(s): LCO 3.5.1 and Bases, ARP-8, SOP-3 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

Is there a learning objective which states the ROs must know these TS surveillance limits?

ES-401 Question 35 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2___ _____

Group # __1___ _____

K/A # 007.K5.02 Importance Rating __3.1__ _____

K/A Statement: Knowledge of the operational implications of the following concepts as they apply to PRTS: Method of forming a steam bubble in the PZR.

Proposed Question:

Given the following conditions:

The Plant has just completed drawing a bubble from a solid plant condition.

Quench Tank, T-73 pressure was noted to be 1 psig.

Pressurizer (PZR) is being maintained at 225 psig by cycling backup PZR heaters.

PZR temperature is 397°F.

If a PZR PORV is venting fluid to the Quench Tank, which of the following states what the expected PORV tailpipe temperature would be AND the Technical Specification LCO leakage limit for this type of leakage?

A. ~320ºF; 1 gpm B. ~320ºF; 10 gpm C. ~387ºF; 1 gpm D. ~387ºF; 10 gpm Proposed Answer: A Explanation (Optional):

By TS definition, Identified leakage is leakage, such that from pump seals or valve packing (except Primary Coolant Pump seal water leakoff), that is captured and conducted to collection systems or a sump or collecting tank. The limit for Identified leakage is 10 gpm (see TS definitions). Quench Tank leakage is not Identified Leakage as the Plant has no accepted method to quantify it per the Leakrate Program.

A. Correct, Find the 240 psia (225 psig) constant pressure line and where it intersects with the saturation line. Draw a straight line (throttling process) from the 240 psig mark on the saturation line to the 16 psia (Quench Tank is at 1 psig) pressure line. The intersecting temperature line is approximately 320ºF. Part 2 correct, 1 gpm is the limit for unidentified leakage.

B. Incorrect, Part 1 is correct. See explanation above for Part 2.

C. Incorrect, applicant uses the saturated temperature (387ºF) for 225 psig. 1 gpm is the limit for unidentified leakage.

D. Incorrect, see explanation in choice C. Part 2 correct.

Technical Reference(s): Steam Tables, Tech Specs Section 1.1 (definitions)______

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: Steam Tables______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ___________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

ES-401 Question 36 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2___ _____

Group # __1___ _____

K/A # 008.K3.01 Importance Rating __3.4__ _____

K/A Statement: Knowledge of the effect that a loss or malfunction of the CCWS will have on the following:

Loads cooled by CCWS Proposed Question:

The Plant is at 100% power with the following conditions:

Component Cooling Water (CCW) Pump P-52A out of service for maintenance.

CCW Pump P-52B is running

CCW Pump P-52C is in standby If the handswitch for CV-0944A, CCW to SFPHXs & RW Evaps, was inadvertently placed in the BYPASS position, which one of the following describes an expected consequence?

A. CV-0944A fails open; CCW Pump P-52C auto-starts on low discharge header pressure.

B. CV-0944 fails closed; the Radwaste Evaporators would lose CCW cooling.

C. Upon receipt of a SIAS, CCW flow to required components would be diverted.

D. Upon receipt of a SIAS, the SFP HXs would lose CCW cooling.

Proposed Answer: C Explanation (Optional):

A. Incorrect, CV-0944A will remain open with the switch in bypass and will not close, as expected, upon receipt of a SIS B. Incorrect, CV-0944 will remain open with the switch in bypass. This allows continued cooling to the Radwaste Evaporators.

C. Correct, upon the SIAS, CV-0944A would remain open, diverting required CCW cooling flow from DBA loads.

D. Incorrect, CV-0944A would remain open and SFP HX cooling would continue.

Technical Reference(s): DBD-1.01______________________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam ___________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

Palisades 2010 Audit Exam. Modified stem, modified (one new) and rearranged distractors.

ES-401 Question 37 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 010.K6.03 Importance Rating __3.2__ _____

K/A Statement: Knowledge of the effect of a loss or malfunction of the following will have on the PZR:

PZR sprays and heaters.

Proposed Question:

Given the following conditions:

The Plant is operating at 92% power during a power ascension.

Pressurizer Pressure Control is selected to Channel B.

Pressurizer Level Control is selected to Channel B.

All Backup heaters are ON.

All Proportional heaters are energized.

Preferred AC Bus EY-20 is lost.

Assuming NO Operator action, which ONE of the following states the response of the Proportional and Backup heaters?

Proportional Heaters Backup Heaters Spray Valves A. De-energized De-energized Open B. De-energized De-energized Closed C. MINIMUM output Energized Open D. MINIMUM output Energized Closed Proposed Answer: B Explanation (Optional):

On a loss of AC Bus EY-20, with the Pressurizer (PZR) pressure control selector switch in the Channel B position, all pressurizer heaters are de-energized. Spray valves will also close. A loss of EY-20 will also impact the pressurizer level controller, LIC-0101BL (fail LOW on loss of power), resulting in a loss of ALL heaters. The effected Pressurizer pressure indicating controller (PIC) will fail to 0 output, in this case there will be NO PZR heaters energized (due to LIC loss of power) AND PZR Spray valves will be closed (due to PIC loss of power).

A. Incorrect, the applicant understands that pressurizer heaters will de-energize, but believes the spray valves will remain unaffected.

B. Correct, see explanation.

C. Incorrect, applicant believes the PZR pressure and level control channels are selected such that a loss of preferred AC bus EY-20 will not impact them.

D. Incorrect, the applicant understands the impact of the loss of power to the PIC (spray valves close), but does not understand the heaters de-energize due to a loss of power to the LIC.

Technical Reference(s): PL-PPCS Pressurizer Pressure Control Lesson Plan, AOP-13 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

PL-PPCS Pressurizer Pressure Control Rev. 6

2. Loss of any Preferred AC Bus

a. Y10/Y20 provides power to Pressurizer Pressure Control circuitry, Channel A/B.

Effected Pressurizer PIC will fail to 0 output.

EO 10a REF ARP-4 windows 63/64

1) Heater Control Selector switch 1/LIC-101 is normally in the A&B position, therefore a loss of Y-10/20 will also impact LIC-0101AL/ LIC-0101BL (fail LOW on loss of power), resulting in a loss of ALL heaters. The Effected Pressurizer PIC will fail to 0 output, in this case there will be NO PZR heaters energized (due to LIC loss of power) AND PZR Spray valves will be closed (due to PIC loss of power)

2) Effect on the PCS -PCS pressure could rise due to loss of power impacting the PLCS (PZR Level Control System) resulting in maximum charging and minmum letdown. LPZR would rise compressing the bubble in the PZR.

3) Without operator action pressure could rise and reach the high PCS pressure Reactor trip setpoint.

b. Y30/Y40 provides power to Channel A/B LTOP circuitry. Effected LTOP circuitry is lost therefore losing overpressure protection from the effected channel.

As long as the other LTOP channel is operable the PCS will be protected from overpressure with one PORV still able to operate.

ES-401 Question 38 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __1__ _____

Group # __2__ _____

K/A # 012.K5.01 Importance Rating __3.3__ _____

K/A Statement: Knowledge of the operational implications of the following concepts as they apply to the RPS: DNB Proposed Question:

Which Reactor Protection System protective function provides Departure from Nucleate Boiling protection?

A. Low Primary Coolant System Flow Trip B. Low Steam Generator Level Trip C. Variable High Power Trip D. High Pressurizer Pressure Trip Proposed Answer: A Explanation (Optional):

A. Correct, the low PCS flow trip provides DNB protection during events which suddenly reduce PCS flow rate during power operation.

B. Incorrect, the low S/G level trips provide protection against PCS overcooling (excessive steam demand event) and PCS overpressurization (loss of feedwater event)

C. Incorrect, the variable high power trip provides protection against positive reactivity excursions.

D. Incorrect, the high pressurizer pressure trip provides protection against PCS overpressure at operating temperature Technical Reference(s): LCO 3.3.1 Bases_________________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis ______

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

Is there a learning objective that supports this as required knowledge for ROs?

ES-401 Question 39 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 013.G2.4.50 Importance Rating __4.2__ _____

K/A Statement: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

Proposed Question:

Given the following conditions:

The reactor has just tripped from full power

A S/G level is 20% and lowering

B S/G level is 22% and lowering

The following annunciator has just LIT:

o EK-16 ARRAY A Annunciator 3-4, P-8A Tripped Under these conditions, the Auxiliary Feedwater Actuation Signal (AFAS) will first start the

__(1)__ pump to deliver a minimum of 100 gpm to __(2)__ S/G(s) to prevent further system actuation(s).

A. (1) P-8B, Steam Driven Auxiliary Feedwater Pump (2) at least ONE B. (1) P-8B, Steam Driven Auxiliary Feedwater Pump (2) BOTH C. (1) P-8C, Auxiliary Feedwater Pump (2) at least ONE D. (1) P-8C, Auxiliary Feedwater Pump (2) BOTH Proposed Answer: C Explanation (Optional):

EK-16 ARRAY A Annunciator 3-4 will alarm if the P-8A breaker opens after receiving a valid auto start signal. The AFAS (Aux Feedwater Actuation Signal) will auto-start the P-8A pump first (in this case, the pump is tripped and will not start). Therefore, if P-8A were to start and establishes flow (100 gpm minimum) to one S/G before Pump P-8C timer times out, the starting of Pump P-8C is blocked. Low flow (less than 100 gpm) to both S/Gs from Pump P-8A will not trip Pump P-8A, but will remove the block for starting of Pump P-8C and then Pump P-8B if low flow exists in Pump P-8C. The pumps continue to operate if low flow is detected but the pumps will not be tripped. If Pump P-8A fails to operate for any reason, the AFAS system will initiate a

start signal for P-8C and then P-8B, should Pump P-8C trip. Therefore, since P-8A is tripped with no flow being provided to either S/G, pump P-8C will start and establish flow to at least one S/G.

A. Incorrect, Pump P-8C will start prior to Pump P-8B.

B. Incorrect, Pump P-8C will start prior to Pump P-8B. Flow > 100 gpm to at least ONE S/G will prevent a low flow condition.

C. Correct.

D. Incorrect, Flow > 100 gpm to at least ONE S/G will prevent a low flow condition.

Technical Reference(s): ARP-36,DBD-1.03________________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 40 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 013.A3.02 Importance Rating __4.1_ _____

K/A Statement: Ability to monitor automatic operation of the ESFAS including: Operation of actuated equipment.

Proposed Question:

Given the following conditions:

The Plant is at 100% power.

Train B Control Room HVAC is in operation.

A spurious Containment High Radiation (CHR) signal was just received.

No further operator actions have been taken.

Which of the following equipment actuations and/or indications is expected given the CHR signal?

A. ES Room Sump Pump West P-73B red light is LIT.

B. Control Room Condensing Unit VC-11 red light is LIT.

C. Control Room Air Filter Unit Fan V-26B red light is LIT.

D. PCP Controlled Bleedoff Valve CV-2083 red light is LIT.

Proposed Answer: C Explanation (Optional):

A. Incorrect, the ESS Room sump pumps auto-start feature is disabled in order to prevent highly radioactive waste from ESS Rooms being transferred to the Dirty Waste Drain Tank during an accident. The pump will be off (green light lit).

B. Incorrect, the control room condensing unit trips (green/off light will be lit) on the CHR signal and are manually restarted.

C. Correct, with the B Train Control Room HVAC in normal operation, the V-26B fan is lined up in AUTO. With the control switch in auto, the fan will start on a CHR signal (red/run light will be lit).

D. Incorrect, PCP CBO valve CV-2083 will isolate (green/closed light will be lit).

Technical Reference(s): PL-CTMT Containment Building Lesson Plan, FSAR Section 9.8.10.b Rev 31, E-17 Sheet 7, AOP-31 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New __X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

PL-CTMT, Containment Building Revision 6 Page 43 of 74

1) Equipment Actuation due to CHR Right & Left Channel a) All automatic containment isolation valves close except:

1. CCW Containment Supply and Returns CV-0910, CV-0911, CV-0940

2. Main Steam Isolations CV-0501, CV-0510

3. Main Feedwater Isolations CV-0701, CV-0703, CV-0734, CV-0735

3) Actuation of CHR also causes the AUTO start feature of the ES Sump Pumps to be disabled.

a) The auto feature is disabled in order to prevent highly radioactive waste from ESS Rooms being transferred to the Dirty Waste Drain Tank and possibly beyond during an accident.

b) Any system leakage will be retained in the sump.

1. Pumps can still be operated in manual.

c) Right Channel - ESS Pumps P-73A, P-72A

1. Left Channel - ESS Pumps P-73B, P-72B d) Right or Left Channel also results in tripping Air Room Purge Fan V-46.

e) CHR also places the Control Room HVAC in EMERGENCY MODE.

1. Right Channel - B Train

2. Left Channel - A Train f) CHR Reset Pushbuttons

1. Left channel RESET pushbutton ONLY resets Left Channel.

2. Right channel RESET pushbutton ONLY resets Right Channel.

3. When CHR is RESET, Containment Isolation Valves will NOT re-open. Valve Hand Switches must be placed to CLOSE, they may then be opened.

ES-401 Question 41 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 022.G2.1.27 Importance Rating __3.9_ _____

K/A Statement: Knowledge of system purpose and/or function.

Proposed Question:

Per the Design Bases Documents DBD-2.03, Containment Spray System, and DBD-2.08, Containment Air Coolers, which of the following design functions are NOT shared by both the Containment Air Cooling system and the Containment Spray system?

A. Act as a barrier to limit radioactive releases from containment.

B. Provide post-accident cooling capability to limit containment pressure to within containment structure design value of 55 psig.

C. Provide post-accident cooling capability to achieve within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a containment pressure which is 50% or less the design pressure.

D. Remove energy from the atmosphere within the Containment Building during normal operation to keep the air temperature below 140oF.

Proposed Answer: D Explanation (Optional):

The CAC and CS systems share multiple design functions: to act as a barrier to limit radiological releases, to limit containment pressure below the design value during an accident, and to provide sufficient cooling to lower containment pressure to 50% the design value within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-accident. The CS system does not have a normal operating condition function, as the CAC system does. The CAC system is designed to remove heat from containment to maintain containment temperature during normal operations to less than 140oF.

A. Incorrect, see explanation.

B. Incorrect, see explanation.

C. Incorrect, see explanation.

D. Correct, see explanation.

Technical Reference(s): DBD-2.03, DBD-2.08 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New __X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 42 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 026.K2.01 Importance Rating __3.4_ _____

K/A Statement: Knowledge of bus power supplies to the following: Containment spray pumps.

Proposed Question:

Complete the following statements:

Power to Containment Spray pump P-54C is normally supplied from __(1)__ and the pump will start on the DBA sequencer, if necessary, __(2)__ after the Diesel Generator output breaker closes.

A. (1) Bus 1C (2) 2 seconds B. (1) Bus 1C (2) 19 seconds C. (1) Bus 1D (2) 2 seconds D. (1) Bus 1D (2) 19 seconds Proposed Answer: B Explanation (Optional):

P-54A is energized from Bus 1D (DG 1-2 on a LOOP). P-54B and P-54C are energized from Bus 1C (DG 1-1 on a LOOP). P-54A and P-54B start 2 seconds after DG output breaker closure. P-54C will not start until 19 seconds after breaker closure. A nominal 15 second auto-start time delay was added to P-54C control circuit to mitigate potential DG overload concerns.

A. Incorrect, see explanation.

B. Correct, see explanation.

C. Incorrect, see explanation.

D. Incorrect, see explanation.

Technical Reference(s): DBD-2.03, DBD-5.05 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New __X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 43 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 039.K1.02 Importance Rating __3.3__ _____

K/A Statement: Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems: Atmospheric relief dump valves.

Proposed Question:

Given the following conditions:

Reactor power is 80%

Tave is 554oF Which one of the following statements describes the expected response of the Atmospheric Steam Dump Valves (ADVs) immediately following a reactor trip from the above initial conditions?

A. ADVs initially modulate open, then modulate closed and are full closed when TAVE is 535°F.

B. ADVs initially modulate open, then modulate closed and are full closed when TAVE is 540°F.

C. ADVs initially quick open, then modulate closed and are full closed when TAVE is 535°F.

D. ADVs initially quick open, then modulate closed and are full closed when TAVE is 540°F.

Proposed Answer: A Explanation (Optional):

The ADVs and TBV modulate on temperature and pressure. The temperature input is from Tave-Tref calculators (TYT-0100 and TYT-0200) providing inputs to the steam dump controller, HIC-0780A. The ADVs and TBV will modulate from full open at 25oF error (Tave minus no load Tave) to full closed at 3oF error. In this scenario, the error signal seen is 554oF-532oF (no load Tave) =

22oF error. At an error of > 25oF, the ADVs and TBV will quick open, upon actuation of the SDCR (steam dump control relay). The 3oF error is used on a decreasing Tave signal. The TBV will cycle with the ADVs, as discussed, but will also cycle to maintain steam header pressure at 900 psia (saturation pressure at no load Tave of 532oF), with a +/- 5 psia band. Therefore, the TBV will be full closed at 895 psia and full open at 905 psia. The TBV does not receive a quick open signal from the TBV pressure controller. PIC-0511.

A. Correct, see explanation.

B. Incorrect, the ADVs and TBV modulate open on increasing Tave at an 8oF error signal

and are full open at 25oF error. However, in this case, a decreasing Tave signal is observed, from 554oF to no load Tave (532oF)

C. Incorrect, the ADVs and TBV receive modulate open signals. See explanation.

D. Incorrect, the ADVs and TBV modulate open on increasing Tave at an 8oF error signal and are full open at 25oF error. However, in this case, a decreasing Tave signal is observed, from 554oF to no load Tave (532oF). See explanation.

Technical Reference(s): DBD-1.09, PL-MSS Main Steam System (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

PL-MSS Main Steam System Revision 7 Page 59 of 74

3. Quick Open Mode

a. When a turbine trip causes the 386 AST relay to energize, a quick open signal is generated.

b. If TAVE is 556.9°F, the steam dump control relay (SDCR) is energized and closes contacts to align the quick open air supply solenoids to the ADV valve actuators and the TBV to open the valves fully.

c. The ADVs and TBV will stay full open until TAVE is less than 556.9°F.

d. When TAVE lowers to less than 556.9°F the SDCR will de-energize and remove the quick open function.

e. The modulating mode will then control the ADVs.

4. Modulate Mode

a. When a turbine trip causes the 386X1 AST relay to energize, a contact is closed to arm Steam Dump Controller HIC-0780A.

b. Steam Dump Controller HIC-0780A will modulate the ADVs and TBV based on a TAVE -

TREF Error Signal.

PL-MSS Main Steam System Revision 7 Page 59 of 74

c. The error signal is developed by comparing actual TAVE to 532°F (Reference No-load Value).

d. The control system will modulate the ADVs and TBV from full open when TAVE is at 556.9°F (25°F error) to full closed at 535°F (3°F error).

e. For increasing TAVE , the control system will modulate the ADVs and TBV from full closed when TAVE is at 540°F (8°F error) to full open at 556.9°F (25°F error).

5. PIC-0511 controls CV-0511 to maintain the steam pressure setpoint.

a. Normally 900 psia (TAVE at 532°F)

b. At 5 psi greater than the setpoint (905 psia, 532.6°F), CV-0511 will be full open.

c. CV-0511 will be fully closed at 5 psi less than the setpoint (895 psia, 531.3°F) on PIC-0511.

d. The TBV scale is 800 psia to 1000 psia.

Since S/G pressure is approximately 770 psia at full power, I&C has set the out of range alarm function (Yellow alarm light in solid) below expected values for full power operation to prevent the yellow light from being illuminated all the time at full power conditions.

e. The TBV pressure control function DOES NOT require a turbine trip (e.g. does not require 386AST relay actuation).

6. PM-0511 auctioneers the signals from PIC-0511 and HIC-0780A, taking the larger of the two signals.

a. Therefore, in addition to the pressure control input, CV-0511 receives a modulate open signal from steam dump controller HIC-0780A through PM-0511.

Example: If the Atmospheric Dump Valves are being operated in manual using HIC-0780A, a signal will also be sent to PM-0711. If this signal is greater than the signal from PT-0510, the TBV will open.

b. This is the same as the signal received by the ADVs from the TAVE Computer. When the TBV opens as a result of input from HIC-0780A, the output meter on PIC-0511 will show zero output. Note also that the TBV can open from HIC-0780A when PIC-0511 is in the manual mode.

c. The 386X1 AST turbine trip relay must be energized to receive this signal.

d. Modulate signal is removed when Tave is lowered to 535°F (+3°F) and will be provided when Tave is greater than 540°F (+8°F). The TBV should already be full open due to the pressure signal if Tave is at 540°F (962.8 psia).

e. CV-0511 also receives the same 'quick opening' signal as the ADVs.

PL-MSS Main Steam System Revision 7 Page 59 of 74

1) Tave at 556.9°F and turbine trip via the 386 AST turbine trip relay

2) Opens SV-0589B and closes SV-0589C to align the quick open air supply and close the modulate air supply.

f. CV-0511 is interlocked to prevent opening and will close if there is less than 5 inches of vacuum in the main condenser.

1) SV-0509A is closed to stop the air supply and SV-0509B is opened to vent the air off CV-0511 actuator

ES-401 Question 44 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 059.A4.12 Importance Rating __3.4__ _____

K/A Statement: Ability to manually operate and monitor in the control room: Initiation of automatic feedwater isolation.

Proposed Question:

Given the following conditions:

An Excess Steam Demand Event occurred, resulting in a reactor trip.

The turbine had to be manually tripped.

Containment pressure is 3.8 psig and rising.

A S/G pressure is 495 psia and lowering.

B S/G pressure is 595 psia and rising.

Assuming NO operator actions, what is the status of each S/Gs Feed Reg Valve and Feed Reg Bypass valve?

A S/G B S/G A. open open B. open closed C. closed open D. closed closed Proposed Answer: C Explanation (Optional):

Feedwater is isolated from either a CHP signal or low S/G pressure. A CHP signal will close FRV and FRV bypass valves for both S/Gs. The setpoint for containment isolation on containment high pressure is 4 psig, so CHP will not cause the FW isolation in this case.

Feedwater will isolate to the A S/G as pressure is less than 512 psia. Feedwater will be unaffected to the B S/G as pressure is > 512 psia.

A. Incorrect, A S/G FW will isolate on low S/G pressure. B S/G FW will remain unaffected.

B. Incorrect, A S/G FW will isolate on low S/G pressure. B S/G FW will remain unaffected.

C. Correct, see explanation.

D. Incorrect, B S/G FW will remain unaffected.

Technical Reference(s): FSAR 7.5.1.3, ARP-8, ARP-21, PL-MSS Main Steam System Lesson Plan (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 45 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 061.K6.01 Importance Rating __2.5_ _____

K/A Statement: Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: Controllers and positioners.

Proposed Question:

Given the following conditions:

The Control Room has been evacuated due to a fire

C-150, Auxiliary Hot Shutdown Panel, has been placed in service As a result of placing C-150 in service, AFW Pump P-8B:

A. Will NOT automatically trip on low suction pressure.

B. Will NOT be available as a source of feedwater.

C. Automatic speed control is disabled.

D. Overspeed trip protection is disabled.

Proposed Answer: A Explanation (Optional):

A. Correct, pump P-8B will not trip on low suction pressure. SV-0522C (C-150) and SV-0522H/G (ATWS) bypass low suction pressure trip, allowing SV-5022B (steam supply valve) to be cycled from C-150.

B. Incorrect, the applicant believes that P-8B cannot be controlled from C-150.

C. Incorrect, the applicant believes that electrical power is required for speed control of P-8B.

D. Incorrect, the applicant believes that electrical power is required for overspeed protection.

Technical Reference(s): PL-AFW Auxiliary Feedwater System Lesson Plan, E-17 Sheet 21A (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2009 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

PL-AFW, Auxiliary Feedwater System Revision 08 Page 70 of 74

16. Local Controls and Indications

g. Panel C-150, Auxiliary Hot Shutdown Panel

1) Enabling instrumentation on Auxiliary Hot Shutdown Panel EC-150/EC-150A disables the corresponding instrumentation in the Control Room and on Redundant Safety Injection Panel EC-33.

2) P-8B - can open and close CV-0522B via HS-0522C a) P-8B will NOT trip on low suction pressure when CV-0522B is opened from C-150A

3) CV-0749 - Can operate HIC-0749C, P-8A/B AFW flow to 'A' S/G if HS-0102A is placed in the "C-150" position

4) CV-0727 - Can operate HIC-0727C, P-8AB AFW flow to 'B' S/G if HS-0102B is placed in the "C-150" position

5) FI-0727B - Flow to 'B' S/G from P-8A/B

6) FI-0749B - Flow to 'A' S/G from P-8A/B 17.a. Control Room Controls

7) Low suction pressure prevents SV-0522B from operating a) SV-0522C (C-150) and SV-0522H/G (ATWS) bypass low suction pressure trip.

ES-401 Question 41 (Old) Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 022.G2.1.27 Importance Rating __3.9_ _____

K/A Statement: Knowledge of system purpose and/or function.

Proposed Question:

Per the Design Bases Documents DBD-2.03, Containment Spray System, and DBD-2.08, Containment Air Coolers, which of the following design functions are shared by both the Containment Air Cooling system and the Containment Spray system?

1. Remove energy from the atmosphere within the Containment Building during normal operation to keep the air temperature below 140oF.

2. Act as a barrier to limit radioactive releases from containment.

3. Provide post-accident cooling capability to limit containment pressure to within containment structure design value of 55 psig.

4. Provide post-accident cooling capability to achieve within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a containment pressure which is 50% or less the design pressure.

E. 1, 2, and 3 F. 1, 2, and 4 G. 1, 3, and 4 H. 2, 3, and 4 Proposed Answer: D Explanation (Optional):

The CAC and CS systems share multiple design functions: to act as a barrier to limit radiological releases, to limit containment pressure below the design value during an accident, and to provide sufficient cooling to lower containment pressure to 50% the design value within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-accident. The CS system does not have a normal operating condition function, as the CAC system does. The CAC system is designed to remove heat from containment to maintain containment temperature during normal operations to less than 140oF.

E. Incorrect, see explanation.

F. Incorrect, see explanation.

G. Incorrect, see explanation.

H. Correct, see explanation.

Technical Reference(s): DBD-2.03, DBD-2.08

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None?________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New __X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 46 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 061.A3.03 Importance Rating __3.9_ _____

K/A Statement: Ability to monitor automatic operation of the AFW, including: AFW S/G level control on automatic start.

Proposed Question:

Given the following sequence of events:

The reactor trips from full power due to a spurious turbine trip

S/G levels lower to 24%

AFW Pump P-8A does not auto start

AFW Pump P-8C auto starts and supplies Auxiliary Feedwater (AFW) to both S/Gs

After one hour, S/G levels are restored to 60-70%

AFAS is NOT reset The NCO then places AFW Pump P-8C in MANUAL and stops AFW Pump P-8C.

Which of the following best describes the AFW system response?

A. AFW Pump P-8B will start approximately 112.5 seconds later.

B. No AFW pumps will automatically start.

C. AFW Pump P-8C will start after 30.5 seconds.

D. AFW Pump P-8B will automatically start immediately.

Proposed Answer: D Explanation (Optional):

A. Incorrect, plausible if the operator believes the timer starts when the running pump is secured.

B. Incorrect, plausible if operator fails to recognize P-8B is still in auto with the timer timed out and a standing AFAS signal.

C. Incorrect, plausible if the operator believes that the pump will start based solely on the timer and the AFAS signal.

D. Correct, P-8B is in automatic, AFAS is present, and the timer has timed out.

Technical Reference(s): PL-AFW Auxiliary Feedwater System Lesson Plan (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2001 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

PL-AFW, Auxiliary Feedwater System Revision 08 Page 3 of 74 13.a.8)

Each Steam Generator supply line for each AFW train contains a flow control valve. These valves will automatically open and control flow at a pre-set value (165 gpm) when the associated AFW pump(s) start.

17. Control Room Controls

c. Operation of FIC-0727/0749

1) Control flow from P-8A/B through 4 lines.

2) FIC and HIC powered from EY-10. Auto swaps to EY-30 if EY-10 lost.

3) Controller response when P-8A and P-8B OFF.

a) If FIC placed to AUTO, BLUE pen drops to 0%. Operator can then adjust setpoint as required.

4) Controller response when P-8A or P-8B started.

a) If FIC is in AUTO, valve opens to BLUE setpoint flow rate.

b) If FIC is in CASCADE, valve opens to obtain 165 gpm. Adequate flow for removing decay heat.

5) Controller response when P-8A or P-8B running a) If FIC is transferred from CASCADE to AUTO, setpoint remains at 165 gpm.

6) FIC will transfer to AUTO and track the HIC if:

a) FIC was originally in MANUAL b) HIC is then swapped from AUTO to MANUAL

7) FIC-0727/0749 Operability Requirements. Considered operable if:

a) In CASCADE, or b) In AUTO AND no MFW in service AND S/G levels and PCS temps maintained within bands required for decay heat removal.

c) NOT operable if FIC or HIC in MANUAL.

b. Operation of FIC-0737A/0736A

1) This controller can control either the 1 1/2 bypass valves (CV-0736/CV-0737) or the main 4 valves (CV-0736A/CV-0737A).

2) HIC (on panel C-33) and FIC powered from EY-20. No auto swap to alternate source. Loss of EY-20 also TRIPS P-8C since the pressure switches lose power for the low suction pressure trip.

3) P-8C is preferred for plant startups and shutdowns. WHY?

a) PF Mode (Program Function) controls the small bypass valves for finer control of AFW flow.

b) If in AUTO or PF Mode, and an AFAS occurs, the controller auto swaps to CASCADE, and the bypass valves close.

4) P-8C must be operating before FIC can be swapped to AUTO. If you try to select AUTO with P-8C OFF, the FIC swaps to CASCADE!

17.7)

Q: Why are FIC-0736A/0737A controllers operable in AUTO always but FIC-0727/0749 controllers have restrictions?

A: FIC-0736A/0737A controllers will automatically swap to CASCADE upon an AFAS and FIC-0727/0749 controllers will not.

ATTACHMENT 3: Worksheet and Answer Key WORKSHEET AND ANSWER KEY

8. Given the following conditions:

All AFW HICs are in the Auto position

All FICs are in Auto Mode

AFW Pump P-8A and P-8C are operating A. What controller mode would FIC-0736A and FIC-0737A transfer to following an AFAS?

Cascade B. What Controller Mode would FIC-0727 and FIC-0749 transfer to following an AFAS?

They would remain in Auto Mode at the current set point

ES-401 Question 47 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 062.K4.02 Importance Rating __2.5_ _____

K/A Statement: Knowledge of the AC distribution system design feature(s) and/or interlock(s) which provide for the following: Circuit breaker automatic trips.

Proposed Question:

Given the following conditions:

The Plant is at 35% power.

GOP-5, Power Escalation in Mode 1, is in progress.

The Crew is preparing to start the second Feedwater Pump.

The following annunciators have just alarmed:

o EK-0334, Switchyard Critical Trouble o EK-5004, Bkr 29R8 Fail to Trip Operated Given the conditions noted above, what is the expected status of the Switchyard Rear R Bus and what is the expected status of the Plant?

A. De-energized; Plant remains online.

B. De-energized; Plant is tripped.

C. Energized, Plant remains online.

D. Energized, Plant is tripped.

Proposed Answer: A Explanation (Optional):

The rear bus is de-energized due to the 486BF relay tripping 29H9, actuation of a transfer trip on the Vergennes Line, and actuation of both the 486-P/R (Bus Protection Lockout Primary) and 486-B/R (Bus Protection Lockout Backup) relays tripping 25R8, 27R8, and 31R8. The Plant will remain online as the 4160VAC busses are on Station Power transformers and not the Start-Up transformers (the swap is performed at 20% power, per GOP-5).

A. Correct, see explanation.

B. Incorrect, the applicant incorrectly believes that the Start-Up transformers are supplying power.

C. Incorrect, the applicant does not understand the impacts of the lockout relay actuation.

D. Incorrect, the applicant does not understand the impacts of the lockout relay actuation and incorrectly believes that the Start-Up transformers are supplying power.

Technical Reference(s): ARP-13, DBD-6.02 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam ___________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

Question modified from 2010 Palisades NRC Exam. Modified question to utilize switchyard breaker failure to trip in lieu of a Start-up transformer sudden pressure relay.

Significantly modified question stem. Reworded and reorganized distractors.

ES-401 Question 48 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 062.A4.07 Importance Rating __3.1_ _____

K/A Statement: Ability to manually operate and/or monitor in the control room: synchronizing and paralleling of different AC supplies.

Proposed Question:

An operator is attempting to parallel an available offsite source with a loaded Diesel Generator per SOP-22, Emergency Diesel Generators. The operator notices the synchroscope rotating in the fast direction at 10 seconds per revolution.

What should the operator do in order to perform the synchronization?

A. Raise DG speed until synchroscope turns slowly in the slow direction.

B. Lower DG speed until synchroscope turns slowly in the fast direction.

C. Raise DG speed until synchroscope turns slowly in the fast direction.

D. Lower DG speed until synchroscope turns slowly in the slow direction.

Proposed Answer: D Explanation (Optional):

The synchroscope must be rotating slowly in the slow direction, per SOP-22, in order to synch a loaded DG to an offsite source. Since the synch scope is rotating fast in the fast direction (as evidenced by 10 second per revolution on the synchroscope), the DG speed must be lowered.

SOP-22 requires 45 seconds (or more) per revolution to ensure adequate operation of the protective synch relays. The applicant needs to interpret the 10 seconds per revolution and apply that to being fast in the fast direction.

A. Incorrect, see explanation. The applicant misunderstands the operation of adjusting DG speed in relation to the synchroscope. If speed is raised, the synchroscope will rotate faster in the fast direction.

B. Incorrect, see explanation. The applicant is confusing the synchronization of an unloaded DG onto an offsite source, rather than a loaded DG onto an offsite source.

C. Incorrect, both parts are incorrect, see explanation.

D. Correct, see explanation.

Technical Reference(s): SOP-22 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 49 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2___ _____

Group # __1___ _____

K/A # 063.K1.03 Importance Rating __2.9__ _____

K/A Statement: Knowledge of the physical connections and/or cause effect relationships between the DC electrical system and the following systems: Battery charger and battery Proposed Question:

The Plant was at 100% power with ED-16, Station Battery Charger No 2, in service supplying the No 2 DC Bus.

FIVE minutes ago, the Plant experienced:

A Loss of Offsite Power

The AC supply breaker (52-225) for ED-16, Station Battery Charger No 2, tripped and cannot be reclosed.

Given the plant conditions noted above, DC Bus No 1 is currently powered from __(1)__ and DC Bus No 2 is currently powered from __(2)__.

A. (1) ED-15, Station Battery Charger No 1 and Station Battery # 1.

(2) ED-18, Station Battery Charger No 4 and Station Battery # 2.

B. (1) ONLY Station Battery # 1.

(2) ONLY Station Battery # 2.

C. (1) ED-17, Station Battery Charger No 3 and Station Battery # 1.

(2) ED-18, Station Battery Charger No 4 and Station Battery # 2.

D. (1) ED-15, Station Battery Charger No 1 and Station Battery # 1.

(2) ONLY Station Battery # 2.

Proposed Answer: D Explanation (Optional):

A. Incorrect, see explanation for choice D B. Incorrect, see explanation for choice D C. Incorrect, see explanation for choice D D. Correct, the DC bus will remain energized due to power being supplied from Station Battery No 2. Battery Chargers ED-15 and ED-17 are capable of powering DC Bus No 1, with only one charger normally lined up to the bus. Battery chargers ED-16 and ED-18 are capable of powering DC Bus No 2, with only one charger normally lined up to the bus. DG 1-1 supplies chargers ED-15 and ED-18 while DG 1-2 supplies chargers ED-17 and ED-16. Power to ED-18, Battery charger No 4, will need to be restored from a

backup power source (DG 1-1).

Technical Reference(s): SOP-30, DBD-4.02_______________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 50 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 064.K2.02 Importance Rating __2.8__ _____

K/A Statement: Knowledge of bus power supplies of the following: Fuel oil pumps Proposed Question:

Fuel Oil Transfer Pump P-18B is normally energized from __(1)__ and can NOT be energized by __(2)__ in an emergency situation.

A. (1) MCC-1 (2) DG 1-1 B. (1) MCC-1 (2) DG 1-2 C. (1) MCC-8 (2) DG 1-1 D. (1) MCC-8 (2) DG 1-2 Proposed Answer: B Explanation (Optional):

A. Incorrect, MCC-1 normally supplies power to P-18B, while MCC-8 supplies power to P-18A. P-18B can only be powered from offsite power or from DG 1-1. DG 1-2 can supply power to P-18A, but not P-18B. DG 1-1 can supply power to P-18B if necessary.

B. Correct, see choice A.

C. Incorrect, see choice A.

D. Incorrect, see choice A.

Technical Reference(s): TS Bases 3.8.3, PL-EDG Emergency Diesel Generator Lesson Plan, E-4 Sheet 1&2 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam __________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

Lesson Content Instructor notes

5) The system automatically transfers fuel oil to a day tank via the P-18A pump (pump P-18B is operated manually only).

a) A low level switch opens the solenoid isolation valve to the day tank and starts the P-18A pump.

b) A high level switch on the day tank will close the solenoid isolation valve and stop pump P-18A when the tank is full.

EO 6 List the power supplies for the following Emergency Diesel Generator system components:

Fuel Oil Transfer Pumps c) Pump P-18A is powered by 480V MCC-8.

d) Pump P-18B is powered by 480V MCC-1.

VA-39

6) Fuel oil is gravity-fed through a level actuated solenoid valve from the day tank to its respective belly tank.

7) The engine driven Fuel Oil Booster Pump (P-209A, P-209B) supplies fuel from the belly tank to the fuel injection pumps

8) The system normal operating pressure is between 40 psig and 60 psig.

a) Measured in the engine to bedplate day tank fuel return line.

PL-EDG Emergency Diesel Generators Revision 11 Page 23 of 47

ES-401 Question 51 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 064.A3.01 Importance Rating __4.1__ _____

K/A Statement: Ability to monitor automatic operation of the ED/G system, including: Automatic start of compressor and ED/G Proposed Question:

The Diesel Generator Starting Air Compressor(s) auto-start at what pressure?

A. 245 psig B. 235 psig C. 220 psig D. 215 psig Proposed Answer: B Explanation (Optional):

A. Incorrect, 245 psig is the normal Air Start Tank pressure, maintained by the Starting Air compressors.

B. Correct, at 235 psig, D/G starting air compressors automatically start on low Air Start Tank pressure C. Incorrect, at 220 psig, the D/G Trouble alarm comes in due to low Air Start Tank pressure.

D. Incorrect, at 215 psig, the D/G becomes inoperable due to insufficient Starting Air pressure.

Technical Reference(s): DBD-5.01 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New __X___

Question History: Last NRC Exam __________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 52 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 073.K5.02 Importance Rating __2.5_ _____

K/A Statement: Knowledge of the operational implications as they apply to concepts as they apply to the PRM system: Radiation intensity changes with source distance.

Proposed Question:

A hot spot has been identified six (6) feet upstream of RIA-1049, Radwaste Discharge Monitor, at Position X. Assume that during a release, the hot spot moved five (5) feet downstream towards the radiation monitor to Position Y. The change in radiation intensity seen at the radiation monitor detector, RIA-1049, with the hot spot at Position Y would be how many times greater than that at Position X?

RIA-1049 Position X Position Y Detector 5

6 A. 5 B. 6 C. 25 D. 36 Proposed Answer: D Explanation (Optional):

A. Incorrect, the applicant applies a linear relationship between intensity and distance and does not account for the 1 between the detector and the final Position Y.

B. Incorrect, the applicant applies a linear relationship between intensity and distance.

C. Incorrect, the applicant misapplies the inverse square law and applies 5 as the final distance, rather than the correct 1. This would be due to misinterpreting the figure.

D. Correct, due to the inverse square law, the count rate decreases by 1/4 when the distance is doubled (Ix/Iy)=(Xx/Xy)2, where I is intensity in cps and X is distance from the Technical Reference(s): GFES Lesson Plan N-RO-01-L-044-I, Radiation Protection (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2012 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _12__

55.43 _____

Comments:

Modified from Palisades 2012 NRC Exam. Modified stem to change scenario, the radiation monitor, and the particle distances from the monitor. The answer and one distractor were changed.

Palisades 2012 #50 Ref DOE-HDBK-1130-2008 Module 4 Palisades 2010 #51 Ref DOE-HDBK-1130-2008

ES-401 Question 53 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 076.A2.02 Importance Rating __2.7_ _____

K/A Statement: Ability to (a) predict the impacts of the following malfunctions or operations of the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Service water header pressure.

Proposed Question:

Given the following Plant conditions:

The Plant is at 100% power.

Today is January 12th.

Service Water (SW) Pumps P-7A and P-7C are running.

SW Pump P-7B is in Standby.

Dilution Water Pumps P-40A and P-40B are running.

Frazil ice begins to develop on Traveling Screens F-4B and F-4C and it is noted that SW Header Pressures and SW Pump Bay Levels are lowering.

Which one of the following describes __(1)__ The impact of these conditions, and

__(2)__ The correct action to take?

A. (1) SW Pump P-7B will start at 45 psig discharge pressure for P-7A or P-7C.

(2) Trip reactor if SW Pump Bay level lowers to 572.

B. (1) SW Pump P-7B will start at 40 psig discharge pressure for P-7A or P-7C.

(2) Trip reactor if SW Pump Bay level lowers to 574.

C. (1) SW Pump P-7B will start at 40 psig discharge pressure for P-7A or P-7C.

(2) Trip reactor if SW Pump Bay level lowers to 572.

D. (1) SW Pump P-7B will start at 45 psig discharge pressure for P-7A or P-7C.

(2) Trip reactor if SW Pump Bay level lowers to 574.

Proposed Answer: C Explanation (Optional):

At 40 psig SW header pressure on the running SW pump(s), the standby SW pump will auto-start. At 45 psig, a Non-Critical Service Water Low Pressure alarm will annunciate, alerting the operators of the degrading conditions. Per AOP-35, the reactor is required to be tripped at 572 SW Bay Level. SW Bay Level of 574 is an entry criteria into AOP-35 and will also annunciate a SW Pump Bay Low Level alarm.

A. Incorrect, part 1 is incorrect. Part 2 is correct. See explanation.

B. Incorrect, part 1 is correct, part 2 is incorrect. See explanation.

C. Correct, see explanation.

D. Incorrect, part 1 and part 2 are both incorrect. See explanation.

Technical Reference(s): ARP-7, AOP-35 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _______None______

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

Question used from Palisades 2009 Audit Exam.

ES-401 Question 54 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 078.K4.02 Importance Rating __3.2_ _____

K/A Statement: Knowledge of IAS design feature(s) and/or interlocks which provide for the following:

Cross-over to other air systems.

Proposed Question:

The Service Air system will isolate from the Instrument Air system when Instrument Air header pressure lowers to what setpoint?

A. 92 psig B. 85 psig C. 80 psig D. 60 psig Proposed Answer: B Explanation (Optional):

A. Incorrect, this is the pressure when the standby IA compressor auto-starts.

B. Correct, the SA system isolates from the IA system by the automatic closure of CV-1212 Service Air Header Isol.

C. Incorrect, this is the Service Air low pressure alarm setpoint.

D. Incorrect, this is the pressure at which certain valves (specifically Aux Feedwater) will be supplied with Nitrogen backup.

Technical Reference(s): AOP-37, DBD-1.05 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New __X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 55 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __1__ _____

K/A # 103.A2.03 Importance Rating __3.5_ _____

K/A Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Phase A and B isolation.

Proposed Question:

Given the following conditions:

The Plant tripped in response to a Loss of Coolant Accident.

EOP-1.0, Standard Post-Trip Actions, is in progress.

Pressurizer pressure is 1565 psia and lowering.

Pressurizer level is off-scale low.

Containment pressure is 4.1 psig and slowly rising.

Containment Area Rad Monitors indicate:

o RIA-1805 - 8 R/hr o RIA-1806 - 9 R/hr o RIA-1807 - 8 R/hr o RIA-1808 - 9 R/hr

Alarm EK-1342, Safety Injection Initiated, has annunciated.

Alarm EK-1126, CIS Initiated, has NOT annunciated.

Which one of the following actions is required to be performed based on the above conditions?

A. Ensure operating all Containment Air Cooler A and B fans.

B. Push left and right Injection Initiate pushbuttons.

C. Close both Steam Generator Main Steam Isolation Valves.

D. Ensure open all Containment Air Cooler inlet and outlet valves.

Proposed Answer: C Explanation (Optional):

A. Incorrect, this is the action to take if containment temperature is > 125oF when a SIAS is not present. Only the Containment Air Cooler A fans are started when containment pressure > 4.0 psig.

B. Incorrect, EOP-1.0 requires validating the SIAS initiated alarm is in OR, if the alarm is not in, pushing the left and right pushbuttons to actuate SI when PZR pressure is less than 1605 psia. In this case, a valid SIAS has occurred, as evidenced by the SI Initiated

alarm that is locked in and manually actuating SIAS would not result in any additional action(s).

C. Correct, this action is required when containment pressure setpoint for containment isolation is exceeded (4.0 psig).

D. Incorrect, there is an associated action with containment air coolers when containment pressure is > 0.85 psig, but the action is to ensure open all outlet valves only not inlet valves; air cooler #4 inlet valve will automatically close on a safety injection signal.

Technical Reference(s): EOP-1.0, ARP-7, ARP-8 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2012 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _10__

55.43 _____

Comments:

Question modified from Palisades 2012 Exam. Modified question stem to provide individual rad monitor readings, used a high containment pressure rather than high containment radiation, and required applicants to determine whether or not a valid Safety Injection occurred.

Replaced two distractors and changed overall answer.

PL-CTMT, Containment Building Revision 6 Page 42 of 74

7. Containment Isolation A. The specific plant parameters that will cause an automatic containment isolation are:

1) Containment High Pressure, initiates 3.7 psig 4.3 psig

2) Containment High Radiation of 10 R/hr on RIA-1805, 1806, 1807, 1808

3) Refueling Monitors RIA-2316 and RIA 2317 a) Refueling Radiation Monitors are placed in service by placing the Cutout Switches for RIA-2316 & 2317 in the "IN" position using the key switches on panel C-11 rear.

B. EOP Supplement 6 Checklist contains all Containment Isolation Valves for CHR and CHP.

C. Containment High Pressure Logic

1) 8 pressure sensing bellows, each of which operates two independently adjustable micro-switches

2) The switches are set as follows:

a) Pressure Switches 1801, 1802, 1803, 1804

1. 1 contact opens @ ~ 3.4 psig (94.1 inches)

2. 1 contact opens @ ~ 4 psig (110.7 inches) b) Pressure Switches 1801A, 1802A, 1803A, 1804A

1. Both contacts close > 4 psig c) To get a CHP Left Channel Signal you must have a 2/4 logic using Pressure Switches 1801, 1802A, 1803, 1804A.

d) To get a CHP Right Channel Signal you must have 2/4 from Pressure Switches 1801A, 1802, 1803A, 1804.

e) Use E-17 logic print, Sheet 6, to show the result of various pressure switch combinations.

3) Equipment Actuation due to CHP a) CHP Actuation Tables given for LEFT and RIGHT Channel b) Some equipment actuated from both left and right channels.

c) Relays which actuate equipment are dependent upon power

1. Y-40 for right channel

2. Y-10 for left channel d) SIAS actuation places the Containment Spay Pumps in "STANDBY".

1. If CHP is Reset, but NOT SIAS, then if another CHP occurs due to high Containment pressure, the Containment Spray Pumps will NOT automatically start.

2. Pushing Right Channel CHP Reset only resets the Right Channel, it DOES NOT RESET the SIAS channel that the CHP actuated.

3. Pushing Left Channel CHP Reset only resets the Left Channel, it DOES NOT RESET the SIAS channel that the CHP actuated.

D. Containment High Radiation Logic

1) Use E-17 Logic Print, Sheet 7, to show the results of various radiation monitor combinations.

a) Pushing either High Rad Initiate P/Bs (CHR-L CS or CHR-R CS) will result in BOTH Left and Right CHR Actuation.

b) Any combination of the RIA-1805, 1806, 1807 or 1808 in a 2 out of 4 coincidence will result in a Left and Right Channel Actuation

ES-401 Question 56 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __2__ _____

K/A # 001.K1.03 Importance Rating __3.4_ _____

K/A Statement: Knowledge of the physical connections and/or cause effect relationships between the CRDS and the following systems: CRDM.

Proposed Question:

Which of the following describes how control rods function on a reactor trip signal?

A. The magnetic clutches and CRDM motors all de-energize, allowing the control rods to fall into the core.

B. The magnetic clutches and CRDM motors all energize, allowing the control rods to be driven into the core.

C. The CRDM motors are energized, driving the control rods into the core. If the rod does not move, the magnetic clutches are de-energized, allowing the rods to fall into the core.

D. The magnetic clutches de-energize, allowing the control rods to fall into the core. The CRDM motors energize, so that if a rod does not drop, the rod is driven into the core.

Proposed Answer: D Explanation (Optional):

A. Incorrect, the CRDM motors are energized and the motor will rotate in the downward direction, forcing the rods into the core if the rods do not drop due to the de-energizing of the magnetic clutches.

B. Incorrect, de-energizing the magnetic clutch drops the control rod into the reactor by allowing the interconnecting pinion gears and drive shaft to rotate freely as the rod free falls.

C. Incorrect, if the rod does not drop due to de-energizing the magnetic clutches, then the CRDM motors and brakes energize, rotating downward and forcing the rods into the core.

D. Correct, de-energizing the magnetic clutch drops the control rod into the reactor by allowing the interconnecting pinion gears and drive shaft to rotate freely as the rod free falls. If the rod does not drop, the CRDM motors and brakes energize, rotating downward and forcing the rods into the core.

Technical Reference(s): PL-CRD Control Rod Drive System Lesson Plan (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __6__

55.43 _____

Comments:

Lesson Content Instructor Notes OBJ 3: Describe the design feature and interlock that provides for automatic rundown of control rods (1-

41) after a Reactor trip.

7. When a reactor trip signal is present, the CRDM motors and the brakes will energize and the motor will rotate in the downward direction, forcing the control rod into the core. This would be used in case the rod does not drop for some reason; the motor can push the rod down by driving through an anti-reverse clutch. The motor cannot drive the rod up under these conditions since the anti-reverse clutch will transmit torque only in the drive down direction. This would be necessary in the situation if gravity were not sufficient such as if the control rod were to get stuck due to excessive friction or possibly a LOCA in the reactor head region with large flow velocities in the upward direction.

OBJ 2: Describe the operational design of the Control Rod Drive Magnetic Clutch.

CRD-VA-18, 19, 20 - Magnetic Clutch

9. Magnetic Clutch a) Magnetic clutch, when energized, connects the motor to the vertical drive shaft to allow raising and lowering of:

1. Shutdown rods (Groups A and B)

2. Regulating rods (Groups 1, 2, 3 and 4) b) When de-energized, the drive shaft is disconnected from the motor and brake, allowing the shutdown and regulating rods to fall into the core by gravity.

c) Part-length control rods DO NOT have magnetic clutches, therefore they do NOT trip.

d) The magnetic clutch has an anti-reverse feature. The anti-reverse feature functions when the clutch is de-energized. Its function is to ensure a control rod is forced down into the core by the motor if necessary, in the event of a reactor trip. It also prevents upward forces from forcing the rod out of the core, during this time also.

1. The anti-reverse feature of the clutch has "pawl" like mechanisms, which allow rotation only in the downward direction once the clutch is disengaged.

2. Additionally, once the control rod has reached the bottom after a reactor trip, a limit switch (which we will learn more about later) de-energizes the motor and the brake, resulting in the brake becoming engaged. Then the anti-reverse feature along with the brake prevents any outward motion of the control rod.

PL-CRD Revision 8 Page 52 of 74

ES-401 Question 57 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __2__ _____

K/A # 014.K3.02 Importance Rating __2.5_ _____

K/A Statement: Knowledge of the effect that a loss or malfunction of the RPIS will have on the following:

Plant computer.

Proposed Question:

Control rod position will NOT be available from the Palisades Plant Computer (PPC) if, at a minimum, which of the following is/are lost:

Note: Primary Indication Position (PIP), Secondary Position Indication (SPI).

A. ONLY the PIP is lost.

B. ONLY the SPI is lost.

C. BOTH the PIP and SPI are lost.

D. ONLY the RODMON program is lost.

Proposed Answer: C Explanation (Optional):

A. Incorrect, both PIP and SPI can provide rod position to the PPC. While losing the PIP would result in the loss of a large amount of control rod monitoring capabilities, the SPI node would still be able to provide adequate indication of Control Rod position.

B. Incorrect, both PIP and SPI can provide rod position to the PPC. By losing the SPI only, the following indications are lost: 1) Secondary Control Rod Position Indication, 2) redundant PPDIL and PDIL alarms on the PPC, 3) redundant 4 and 8 deviation alarms on the PPC.

C. Correct, both PIP and SPI would need to be lost to lose PPC indication for rod position.

D. Incorrect, both PIP and SPI can provide rod position to the PPC. RODMON uses inputs from PIP both SPI. "RODMON" will use the synchro rod positions (since they are more accurate) and the Loop 2 T (from PIP) as long as they are valid. If they go invalid, RODMON will swap over to using the reed switch rod positions (SPI inputs) and/or Loop 1 T (SPI input).

Technical Reference(s): PL-CRD Control Rod Drive System Lesson Plan, SOP-6 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __6__

55.43 _____

Comments:

Lesson Content Instructor Notes PIP (PRIMARY INDICATION POSITION)

A. Palisades Plant Computer (PPC) Interface (PIP). The PIP software does calculations related to the Control Rods on the basis of analog and digital inputs and, as a result of these calculations, generates digital outputs and provides analog and digital data to the PPC host system.

1. Analog Inputs:

45 synchro positions

Loop 2 delta temp

2. Digital Inputs:

45 rotary switch positions

Rod drop timing start

3. Analog Outputs:

45 rod position (in inches) to host

Loop 2 delta temperature "core power"

Rod drop timing data

4. Digital Outputs:

Rod Control Relays

Annunciators

Digital Rod displays CRD-VA PIP Processor - Target Rod B. The PIP will put -199.9 in place of a rod position when it detects a bad synchrotransmitter to digital converter.

1. Signal from synchro not good

2. Switch contacts not made up or more than one made up C. Group Target Rod Processing

1. The PIP software determines the 7 group target rod selections given the input from the Rod Control Selector Switches.

2. The target rod positions are provided internally to the PPC for use in Manual Sequencing (MS) and in determining the Power Dependent Insertion Limit (PDIL), and Pre-Power Dependent Insertion Limit (PPDIL).

3. If the target rod position is determined to be "bad",

PPC processing is suspended for the programs using the bad group target rod.

a) No contacts or more than one contact closed on switch CRD-VA PIP Processor - Upper Rod Stop D. Upper Rod Stop Processing

1. The PIP software determines the state of the 45 sets of upper rod stop (URS) contacts.

2. This processing is suspended for any rod with an invalid position (bad synchrotransmitter).

3. The URS contacts stay "as is" if either the PIP loses power or if the synchrotransmitter fails.

4. The URS contacts are used in the following:

a) Core Matrix light logic b) Manual Sequential (MS) RAISE logic c) Manual Group (MG) regulating and power shaping rod RAISE logic CRD-VA PIP Processor - Lower Rod Stop

5. Lower Rod Stop Processing a) The PIP software determines the state of the 25 lower rod stop (LRS) contacts for the rods in the regulating and power shaping rods.

b) This processing is suspended for any rod with an invalid position (bad synchrotransmitter).

c) The LRS contacts are used in the following:

1) Manual Sequential (MS) LOWER logic

2) Manual Group (MG) LOWER logic CRD-VA PIP Processor - Upper/Lower Sequential Permissive

6. Upper Sequential Permissive Processing

a) The PIP software determines the state of the 3 upper sequential permissive (USP) contacts based on the position of the group target rods for regulating groups 1, 2, and 3.

b) The USP contacts are used in Manual Sequential (MS) RAISE logic.

7. Lower Sequential Permissive Processing a) The PIP software determines the state of the 3 lower sequential permissive (LSP) contacts based on the position of the group target rods for regulating groups 2, 3, and 4.

b) The LSP contacts are used Manual Sequential (MS) LOWER logic.

CRD-VA PIP Processor - 4 and 8 inch Deviation

8. Four Inch Deviation Processing a) The PIP software determines the state of the 7 four inch group deviation contacts b) If the highest rod in a group is more than 4 inches from the lowest rod in the group, the four-inch deviation contact for that group is opened.

c) If the deviation is less than 4 inches minus the deadband, the contact is closed.

d) The contacts are used to initiate the annunciator EK-0911 "Rod Position 4 inches Deviation" and to illuminate the appropriate group deviation light on C-02.

9. Eight Inch Deviation Processing a) The PIP software determines the state of the 7 eight inch group deviation contacts b) If the highest rod in a group is more than 8 inches from the lowest rod in the group, the eight-inch deviation contact for that group is opened.

c) If the deviation is less than 8 inches minus the deadband, the contact is closed.

d) The contacts are used to initiate the annunciator EK-0912 "Rod Position 8 inches Deviation" and to illuminate the appropriate group deviation light on C-02.

CRD-VA PIP Processor - Out of Sequence

10. Out of Sequence Processing a) The PIP software determines the state of the out of sequence (OOS) contact based on the positions of the group target rods for the four regulating groups. This processing is intended to limit the ways in which the operators can move the control rods of the regulating groups.

b) If any of the regulating group target rods have bad synchrotransmitters, no OOS processing is completed between adjacent rod groups.

c) If an OOS condition is detected, annunciator EK-0916 "Control Rods Out-Of-Sequence" is received.

CRD-VA PIP Processor - PDIL

11. Power Dependent Insertion Limit (PDIL) Processing a) The PIP software calculates power dependent insertion limits (PDIL) for each regulating group and determines the state of four PDIL contacts based on the position of the group target rod.

b) If the core power signal is invalid, the PDIL alarm for regulating group 1 is alarmed (opened) to annunciate this fact.

c) Core power is generated using the TDY_0200 analog input.

d) If a PDIL condition is detected, annunciator EK-0924 (Group 1), EK-0930 (Group 2), EK-0936 (Group 3), or EK-0942 (Group 4) will be received.

CRD-VA PIP Processor - PPDIL

12. Pre-Power Dependent Insertion Limit (PPDIL)

Processing a) The PIP software calculates the four pre-power dependent insertion limits (PPDIL) and determines the state of the four PPDIL contacts based on the position of the group target rod for the corresponding four regulating groups.

b) The calculations for PPDIL utilize the fact that PDIL has already been calculated. Since the PPDIL curves are offset by a fixed amount from the PDIL curves using the same core power estimate, the PPDIL can be quickly calculated by adding an offset to the PDIL. In addition, the PPDIL is clamped at a maximum value.

c) If a PPDIL condition is detected, annunciator EK-0923 (Group 1), EK-0929 (Group 2), EK-0935 (Group 3), or EK-0941 (Group 4) will be received.

CRD-VA PIP Processor - Watchdog Timer/Rod Drop Timing

13. Watchdog Timer Processing a) The PIP software must also maintain the watchdog timer, which is implemented using a programmable delay timer card. If the software for any reason fails to update this watchdog timer or power is lost, the watchdog will cause EK-0918, PIP Trouble, alarm to annunciate.

14. Rod Drop Timing a) This mode allows the operators to acquire a drop time profile of a rod under test. During this mode, all normal PIP activities are suspended (except for watchdog timer) and the PIP system is used exclusively for acquiring the profile of the rod drop. Even normal rod position indications will not change during the test.

D. Secondary Position Indication

1. The reed switches measure control rod position by use of control rod actuated magnetic reed switches.

They transmit rod height signals to the secondary position indication and rod matrix light display.

CRD-VA SPI

2. An assembly containing a number of series resistors to form a voltage divider network with reed switches (approximately 2 inches apart) connected at each junction is attached to the control rod extension housing. A voltage is applied to the network; output voltage depends on which reed switches are closed in the voltage divider. A magnet on top of the control rod extension will close the reed switches as the control rod moves. Overlap between adjacent reed switches is provided. The output is a voltage directly proportional to control rod position.

3. Output voltage is provided to the Secondary Position Indication (SPI) Node, which interprets the output voltage and provides position indication to the PPC.

1. EK-0971, SPI TROUBLE a) The PPC SPI Node resets a Watchdog Timer every couple of seconds. A failure of the SPI Node will cause the Watchdog Timer to not be reset, which will bring in the alarm.

b) If the SPI Node is lost the following Control Rod Drive System functions are affected:

1) The Secondary Control Rod Position Indication is lost.

2) The redundant channel of the Control Rod Out of Sequence alarm, which alarms on the PPC, is lost.

3) The redundant PPDIL and PDIL alarms, which alarm on the PPC, are lost.

4) The redundant 4 inch and 8 inch deviation alarms, which alarm on the PPC, are lost.

5) The PPC status page and Control Rod page can be checked to help validate this alarm.

10. EK-0918, PIP TROUIBLE CRD-VA-114 - EK-0918, PIP Trouble a) If the PIP is not functioning and does not reset its Watchdog Timer, this alarm will come in.

b) If the PIP is not functioning, all the Control Rod functions performed by the PIP will be lost. See the System Interrelationships section for more detail on the impact of a loss of the PIP.

c) Check the PPC status page as well as the Control Rod page on the PPC to validate this alarm.

A. On the PPC a program called RODMON is performing all the same calculations as the PIP; however, it does NOT interface with rod control, annunciators or the LEDs.

1. "RODMON" will use the synchro rod positions (since they are more accurate) and the Loop 2 T (from PIP) as long as they are valid. If they go invalid, RODMON will swap over to using the reed switch rod positions (SPI inputs) and/or Loop 1 T (SPI input).

2. RODMON will determine and display the following and will provide an audible PPC alarm for those conditions marked with an *:

a) 45 validated rod positions b) 7 validated Group Target Rod (GTR) c) Upper Rod Stop (URS) d) Lower Rod Stop (LRS).

e) Shutdown Rod Insertion (SRI)*

f) Upper Sequential Permissive (USP) g) Lower Sequential Permissive (LSP) h) Four Inch Deviation i) Eight Inch Deviation*

j) Out of Sequence (OOS)*

k) Power Dependent Insertion Limit*

l) Pre-power Dependent Insertion Limit (PPDIL)

PL-CRD Revision 8 Page 61 of 74

ES-401 Question 58 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __2__ _____

K/A # 016.K5.01 Importance Rating __2.7_ _____

K/A Statement: Knowledge of the operational implication of the following concepts as they apply to NNIS:

Separation of control and protection circuits.

Proposed Question:

Given the following conditions:

The Plant is at 100% power.

RPS Channel B for A Steam Generator Low Level is BYPASSED due to a failure of LI-0751B, Steam Generator E-50A Low Level Indicator.

Refer to the below graphic of A Steam Generator level instrumentation.

Which one of the following additional instrument failures will result in a Reactor trip? (Assume no operator action.)

A. LI-0751A, Steam Generator E-50A Low Level Indicator, fails LOW.

B. LI-0751A, Steam Generator E-50A Low Level Indicator, fails HIGH.

C. LIA-0702, Steam Generator E-50A Level Alarm Indicator, fails LOW.

D. LIA-0702, Steam Generator E-50A Level Alarm Indicator, fails HIGH.

Proposed Answer: D Explanation (Optional):

A. Incorrect, there are now 2 channels that feed the RPS that exceed a setpoint, however, a trip will not be processed because RPS channel B is bypassed.

B. Incorrect, there are now 2 channels that feed the RPS that exceed a setpoint, however, RPS channel B is bypassed and there is no RPS trip for high S/G level.

C. Incorrect, there are now 2 channels that feed the RPS that exceed a setpoint, however, RPS channel B is bypassed and LIA-0702 does not provide an input to RPS.

D. Correct, if LIA-0702 fails high, the feed regulating valve associated with A S/G would close which would cause an actual low level condition (a high level override from LIA-0702 closes CV-0701 A FRV). An RPS trip would be generated from the remaining 3 channels that are not bypassed.

Technical Reference(s): ARP-5, FSAR 7.2, page 7.2-2, 7.2-9 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2007 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

ES-401 Question 59 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __2__ _____

K/A # 029.A1.02 Importance Rating __3.4_ _____

K/A Statement: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Containment Purge System controls including: Radiation levels.

Proposed Question:

Given the following conditions:

Core offload has just begun during a refueling outage.

The Air Space Purge Fan, V-46 is running with the Containment Purge Supply Valves (CV-1813 and CV-1814) open, per SOP-24, Ventilation and Air Conditioning System.

Containment background radiation is 15 mR/hr.

The following readings were just noted on the Containment Refueling Radiation Monitors:

RIA-2316: 98 mR/hr

RIA-2317: 76 mR/hr With the given conditions, what is the status of the Containment Purge system and why?

A. In-progress; the setpoint of 80 mR/hr has not been reached on 2/2 Containment Refueling Monitors.

B. In-progress; the setpoint of 80 mR/hr above background radiation has not been reached on 2/2 Containment Refueling Monitors.

C. Isolated; the setpoint of 80 mR/hr above background radiation has been reached on 1/2 Containment Refueling Monitors.

D. Isolated; the setpoint of 80 mR/hr has been reached on 1/2 Containment Refueling Monitors.

Proposed Answer: C Explanation (Optional):

The applicant must both the setpoint and coincidence for a Containment High Radiation (CHR) condition during refueling operations. A CHR during refueling operations occurs when 1 of 2 Containment Refueling Monitors exceeds 80 mR/hr above the background radiation level. In this case, only one rad monitor would need to exceed 90 mR/hr to actuate a CHR, which RIA-2316 has.

A. Incorrect, applicant assumes both an incorrect setpoint and an incorrect coincidence for actuation.

B. Incorrect, applicant assumes an incorrect coincidence for actuation.

C. Correct, see explanation above.

D. Incorrect, applicant assumes an incorrect setpoint for actuation.

Technical Reference(s): PL-RMS Radiation Monitoring System Lesson Plan (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

ENTERGY NUCLEAR LESSON PLAN Containment Refueling Monitors RIA-2316/2317 The monitors are energized and calibrated prior to starting core alterations; they are de-energized at the conclusion of Mode 6 activities.

Each channel has a RIA in the Control Room; the RIAs are identical to the typical digital ratemeters The Containment Refueling Monitors provide inputs to the Containment Isolation logic during refueling operations (containment isolation logic is described later in Section Error! Reference source not found., Error! Reference source not found..)

RE-2316 and RE-2317 are removed and stored after Mode 6 operations.

Each monitor has a key switch located on the back of Panel C-11.

VA-1 RIA-2316 Key Switch RF-1 The key switches, RF-1 and RF-2, require Keys 54 and 55 for channels RIA-2316 and 2317, respectively.

When the key switch is in the IN position, the associated monitor inputs to a 1/2 high alarm logic to initiate containment isolation.

The keys are removable only in OUT position.

The high alarm setpoint for RIA-2316 and RIA-2317 is 80 mR/hr above background.

Containment High Radiation (CHR)

A CHR signal is generated by the following area radiation monitors:

2/4 Containment Radiation Monitors RIA-1805/1806/1807/1808 High (Trip 2)

Setpoint 10 R/hr 1/4 channels High (Trip 2) actuates Control Room alarm EK-1363, CONTAINMENT HI RADIATION.

1/2 Containment Refueling Monitors RIA-2316/2317 High with associated keyswitch in IN position.

Setpoint 80 mR/hr above background

RIA RIA RIA RIA RF-1 RIA RIA RF-2 1805 1806 1807 1808 IN 2316 2317 IN 2/4 1/4 EK 1363 1/2 CHR Figure 1. CHR Logic The CHR signal initiates the following:

Actuates Containment Isolation Containment Isolation actuated from CHR does not close CCW supply to containment valves (these are closed on Containment High Pressure signal).

Trips Air Room Purge Fan V-46.

Actuates Control Room HVAC Emergency System (Trains A and B)

Disables automatic operation of Safeguards Rooms Sump Pumps (P-72A, 72B, 73A, 73B)

The pumps can be operated in manual mode.

Actuates Control Room alarm EK-1363, CONTAINMENT HI RADIATION only when initiated by RIA-1805/1806/1807/1808.

Actuates Control Room alarm EK-1126, CIS INITIATED.

c. CHR Signals must be manually reset at Panel C-13.

PL-RMS Revision 4 Page 70 of 74

ES-401 Question 60 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __2__ _____

K/A # 035.K6.01 Importance Rating __3.2_ _____

K/A Statement: Knowledge of the effect of a loss or malfunction on the following will have on the S/Gs:

MSIVs.

Proposed Question:

Given the following conditions:

The Plant is at 24% power.

The Main Generator is synchronized to the grid.

A single Main Steam Isolation Valve closes on a spurious signal.

Assuming the reactor does NOT trip, which ONE of the following correctly describes the initial response of S/G Pressure and Level in the affected loop?

S/G Pressure S/G Level A. Rises Rises B. Lowers Rises C. Lowers Lowers D. Rises Lowers Proposed Answer: D Explanation (Optional):

A. Incorrect, part 1 is correct, part 2 is incorrect. The applicant believes SG level will rise due to the loss of steam flow while maintaining feedwater flow. This is an incorrect initial response which does not take into account the SG shrink/swell effect.

B. Incorrect, this is an incorrect initial response which does not take into account the SG shrink/swell effect. The applicant believes heat is still removed from the SG or SG pressure spikes causing an ASD to open to lower pressure. This is not the initial response. Also, the applicant believes SG level will rise do to the loss of steam flow while maintaining feedwater flow.

C. Incorrect, part 1 is incorrect, part 2 is correct. The applicant believes heat is still remove from the SG or SG pressure spikes causing an ASD to open to lower pressure. This is NOT the initial response.

D. Correct, SG pressure initially rises due to heat no longer being removed from the SG.

With this pressure increase, the SG level will lower (shrink) due to the saturation

pressure rising.

Technical Reference(s): EOP-2.0 Basis (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Turkey Point 2011 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

ES-401 Question 61 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __2__ _____

K/A # 041.A4.06 Importance Rating __2.9_ _____

K/A Statement: Ability to manually operate and/or monitor in the control room: Atmospheric relief valve controllers.

Proposed Question:

Given the following conditions:

The Plant is at 3% power.

Main Turbine trip testing was completed and all Main Turbine protective relays have NOT been reset.

PCS temperature is 532oF.

HIC-0780A, Steam Dump Control, is in AUTO.

PIC-0511, Steam Bypass Pressure Controller, is in MANUAL with 0% demand.

If reactivity is added to the core to cause PCS temperature to rise, at what PCS temperature will steam flow stop the temperature rise?

A. 532oF B. 535oF C. 540oF D. 545oF Proposed Answer: C Explanation (Optional):

A. Incorrect, applicant incorrectly assumed the TBV will use the steam pressure signal to maintain its setpoint of 900 psia (saturation temperature of 532oF). This signal is overriden by the controller being in MANUAL.

B. Incorrect, applicant incorrectly assumed the ADV would open at an incorrect temperature of 535oF. At 535oF, on a decreasing temperature, the modulate signal is removed and open ADVs will close, to allow for the TBV to control at its 900 psia main steam header pressure setpoint.

C. Correct, the ADVs and TBV will modulate open at 540oF to control temperature. The ADVs will open at 540oF, the deadband value for the valves to open on increasing temperature from HIC-0780A. This 8oF deadband for the ADVs and TBV to open is to allow the TBV to open and maintain main steam header pressure at 900 psia, however the TBV will not open with PIC-0511 in MANUAL at 0%. The output from HIC-0780A to

the TBV and the ADVs remains enabled by not resetting the 386/AST relay (turbine trip signal still present).

D. Incorrect, the applicant incorrectly assumed the ADVs and/or the TBV will not open and that temperature will increase until pressure causes the first bank of safeties to open (approximately 1000 psia; saturation temperature of 545oF).

Technical Reference(s): PL-MSS Main Steam System (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 1999 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

Modified question stem to provide information about main turbine testing and turbine trip relays have not been reset. Changed correct answer and changed one distractor.

Lesson Content Instructor Notes

1. HIC-0780A, Steam Dump Controller inputs

a. TAVE from the TAVE/TREF Controller TAVE, - TREF Calculators TYT-0100 and TYT-0200 provide the TAVE signal input to the steam dump controller.

b. Both modes (Modulate and Quick Open) of operation require a turbine trip signal to cause the ADVs to open.

1) 386 AST relay arms the quick open function.

2) 386X1 AST relay arms the modulate function.

c. Upon a Turbine trip and receipt of the 386 AST relay actuation, contacts close in the ADV circuitry. When these contacts close, the controller then uses TAVE to open/close the ADVs.

2. Quick Open Mode

a. When a turbine trip causes the 386 AST relay to energize, a quick open signal is generated.

b. If TAVE is 556.9°F, the steam dump control relay (SDCR) is energized and closes contacts to align the quick open air supply solenoids to the ADV valve actuators and the TBV to open the valves fully.

c. The ADVs and TBV will stay full open until TAVE is less than 556.9°F.

d. When TAVE lowers to less than 556.9°F the SDCR will de-energize and remove the quick open function.

e. The modulating mode will then control the ADVs.

3. Modulate Mode

a. When a turbine trip causes the 386X1 AST relay to energize, a contact is closed to arm Steam Dump Controller HIC-0780A.

b. Steam Dump Controller HIC-0780A will modulate the ADVs and TBV based on a TAVE -

TREF Error Signal.

c. The error signal is developed by comparing actual TAVE to 532°F (Reference No-load Value).

d. The control system will modulate the ADVs and TBV from full open when TAVE is at 556.9°F (25°F error) to full closed at 535°F (3°F error).

e. For increasing TAVE , the control system will modulate the ADVs and TBV from full closed when TAVE is at 540°F (8°F error) to full open at 556.9°F (25°F error).

f. The 3°F error offset on decreasing TAVE and the 8°F error offset on increasing Tave allows Turbine Bypass Valve CV-0511 and Controller PIC-0511 to control TAVE.

B. Turbine Bypass Valve CV-0511

1. The 6 inch, air operated, automatically actuated turbine bypass valve (TBV) has a capacity of approximately 4.5% of rated steam flow (508,000 lbm/hr).

2. The turbine bypass provides reactor decay heat removal following reactor shutdown and provides PCS cooldown capability.

3. The turbine bypass valve may be manually operated from the Control Room.

4. The TBV has a red (open) and a green (closed) position indicating lamps on C-01.

a. As the TBV is opened, the GREEN closed light stays illuminated until PIC-0511, TBV Controller, output signal shows 25% open. At 25% output signal from the controller until the TBV is full open, both the RED and GREEN light stay illuminated. When the TBV is full open the GREEN light goes out.

b. This is different from the ADVs as the ADV valve position lights are both off when the ADVs are in an intermediate valve position.

5. The TBV is normally lined up to control in automatic from either the Steam Dump controller signal or main steam header pressure signal, whichever is greater.

6. CV-0511 is controlled by PIC-0511, which receives its signal from PT-0510.

a. PT-0510 is located on the steam line to the STOP valves.

C. PIC-0511 Turbine Bypass Valve Controller Operation

1. Controller is a typical Yokogawa controller and similar to the ADV Controller with respect to controller features.

2. Auto - Controls steam header pressure at selected set point (900 psia).

Set point is Operator selected. The controller will maintain +/-5 psia of chosen set point.

3. Manual - Operator controls signal output to the valve using the slide bar.

Valve opens based upon output signal.

4. Input is steam header pressure from PT-0510.

5. Alarm indication (Same as ADV controller)

a. Yellow light in solid indicates a process input/output failure or other internal problems.

b. Yellow light flashing indicates a low voltage condition in the controller battery. This should not affect operation as long as power is available to the controller. Controller program will be lost if the controller loses power with a flashing yellow light.

c. Red light indicates a controller computer failure. Controller should fail to last good value but may not be held for long. Manual control may be available.

6. PIC-0511 controls CV-0511 to maintain the steam pressure setpoint.

a. Normally 900 psia (TAVE at 532°F)

b. At 5 psi greater than the setpoint (905 psia, 532.6°F), CV-0511 will be full open.

c. CV-0511 will be fully closed at 5 psi less than the setpoint (895 psia, 531.3°F) on PIC-0511.

d. The TBV scale is 800 psia to 1000 psia.

Since S/G pressure is approximately 770 psia at full power, I&C has set the out of range alarm function (Yellow alarm light in solid) below expected values for full power operation to prevent the yellow light from being illuminated all the time at full power conditions.

e. The TBV pressure control function DOES NOT require a turbine trip (e.g. does not require 386AST relay actuation).

7. PM-0511 auctioneers the signals from PIC-0511 and HIC-0780A, taking the larger of the two signals.

a. Therefore, in addition to the pressure control input, CV-0511 receives a modulate open signal from steam dump controller HIC-0780A through PM-0511.

Example: If the Atmospheric Dump Valves are being operated in manual using HIC-0780A, a signal will also be sent to PM-0711. If this signal is greater than the signal from PT-0510, the TBV will open.

b. This is the same as the signal received by the ADVs from the TAVE Computer. When the TBV opens as a result of input from HIC-0780A, the output meter on PIC-0511 will show zero output. Note also that the TBV can open from HIC-0780A when PIC-0511 is in the manual mode.

c. The 386X1 AST turbine trip relay must be energized to receive this signal.

d. Modulate signal is removed when TAVE is lowered to 535°F (+3°F) and will be provided when TAVE is greater than 540°F (+8°F). The TBV should already be full open due to the pressure signal if TAVE is at 540°F (962.8 psia).

e. CV-0511 also receives the same 'quick opening' signal as the ADVs.

1) TAVE at 556.9°F and turbine trip via the 386 AST turbine trip relay

2) Opens SV-0589B and closes SV-0589C to align the quick open air supply and close the modulate air supply.

PL-MSS Main Steam System Revision 7 Page 5 of 71

ES-401 Question 62 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __2__ _____

K/A # 056.G2.1.23 Importance Rating __3.9_ _____

K/A Statement: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Proposed Question:

Given the following conditions:

The Plant is at 35% power during a power escalation.

P-10A, Heater Drain Pump (the first heater drain pump) was just started.

BOTH Condensate Pumps are operating.

ONE Main Feedwater Pump is operating.

One of the operating Condensate Pumps trips.

Which of the following describes the impact on the Condensate System Recirculation Valve, CV-0730, and what must the operator do?

CV-0730 will throttle in the . . .

A. OPEN direction and direct more flow to feedwater trains. Monitor Heater Drain Pump for normal operation.

B. CLOSED direction and direct more flow to feedwater trains. Monitor Heater Drain Pump for normal operation.

C. OPEN direction and direct more flow to the Main Condenser Hotwell. Align alternate Gland Seal Exhauster to maintain vacuum.

D. CLOSED direction and direct more flow to the Main Condenser Hotwell. Align alternate cooling to Air Ejector Condenser to maintain vacuum.

Proposed Answer: B Explanation (Optional):

CV-0730 modulates on a flow signal from FC-0730 to maintain a minimum flow of 6800 gpm (1600 gpm through the gland seal condenser and 5200 gpm through the air ejector condenser).

At low power (i.e. <25%, the valve is full open to provide a flow path for the Condensate pumps). At approximately 35-40% power, the valve should be full closed to ensure adequate cooling flow and adequate NPSH for the Feedwater Pumps. At 30% power, in this case, the valve would be partially open. If a condensate pump were to trip, the valve would close in order to maintain its minimum flow requirement and support FW pump NPSH.

A. Incorrect, CV-0730 would throttle closed. Throttling open would not allow more flow to the FW pumps, instead directing more flow to the hotwell.

B. Correct, see explanation.

C. Incorrect, CV-0730 would throttle closed.

D. Incorrect, while CV-0730 throttling closed is correct, doing so would not result in more flow to the hotwell and would allow more flow to the FW pumps.

Technical Reference(s): PL-CDFW Main Condenser, Condensate and Feedwater (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2003 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _10__

55.43 _____

Comments:

PL-CDFW - Main Condenser, Condensate and Feedwater Revision 08 Page 8 of 71

f. Condensate Recirc Control Valve CV-0730

1) 12" valve modulates on a flow signal from FC-0730 to maintain minimum flow of 6800 gpm, 1600 gpm through the gland seal condenser and 5200 gpm through the air ejector condenser.

2) Also provides a flow path for the Condensate Pumps during plant start up and shutdown.

3) Red and green position indicating lights on C-01

4) Fails closed on loss of air.

5) On a down power, CV-0730 should start opening at approximately 25%

power.

On a power escalation, CV-0730 should be full closed at 35 to 40% power.Have class review Gary Katt memo from 1998, Attachment 1 of this lesson plan.

ATTACHMENT 1: GKatt Memo - Condensate Recirculation Valve CV-0730 Author: Gary Katt at CPC-PA1 Date: 2/3/98 11:50

Subject:

Condensate Recirculation Valve CV-0730 CV-0730 is designed to maintain ~6800 gpm through the condensate system to provide adequate flow path for the condensate pumps and adequate cooling for the gland steam and air ejector condensers. There is some concern whether CV-0730 will open (due to a previous problem during a forced outage), and what to do if it won't open.

1. FC-0730 was completely overhauled during the last forced outage, and was working properly.

2. The valve should start to open at ~25% power. Monitoring the valve as power is lowered from 25% will enable Operations to determine if the controller is operating properly. If problems develop, this should give I&C time to fix the controller prior to causing problems with the secondary side.

3. The minimum required flow for each condensate pump is approximately 1400 gpm. The pumps should not be operated if there is no flow path available.

4. CV-0730 can be manually failed open by closing a small metal flapper in FC-0730. This flapper is located inside the panel below the flow indicator. Moving this small flapper tight against the adjacent nozzle will result in a full open air signal being sent to CV-0730 causing the valve to open. This flapper would have to be secured in place to keep CV-0730 in the full open position.

5. Alternate flow paths are available if the valve is stuck closed and all other means to fix it have failed:

As long as a feedwater pump is still operating there is no concern for CV-0730 failing to open as the feedwater pump recirculation valve will be open maintaining an adequate flow path for the condensate pumps. If the feed pumps are tripped then the feed pump recirculation valves could be failed open to maintain the desired flow path.

Caution:

Another flow path if the feed pumps are tripped would be to recirculate the condensate back to the condenser through the E-6A/B recirculation line.

Ensure Main Feed Pumps Aux Oil Pump is running prior to establishing flow through an idle feed pump.

Refer to SOP-11, "To Recirculate Condensate/Feedwater System Using P-2A or P-2B.

ES-401 Question 63 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __2__ _____

K/A # 071.K4.04 Importance Rating __2.7_ _____

K/A Statement: Knowledge of design feature(s) and/or interlock(s) which provide for the following:

Isolation of waste gas release tanks.

Proposed Question:

A waste gas release is in progress from Waste Gas Decay Tank T-68A. Waste Gas Radiation Monitor, RIA-1113, spikes above the HIGH alarm setpoint. As a result, the release is automatically terminated by which combination of the following actions:

1. CV-1113, Waste Gas Surge Tank to Stack, closes

2. CV-1119A, T-68A Discharge Control Valve, closes

3. CV-1123, Waste Gas Decay Tank to Stack, closes

4. The operating Main Exhaust fan, V-6A or V-6B, trips A. 1 and 2 B. 2 and 3 C. 1 and 3 D. 2 and 4 Proposed Answer: C Explanation (Optional):

On a high radiation condition sensed by RIA-1113, valves CV-1113 and CV-1123 will close, isolating the waste gas decay tank and waste gas surge tank from the vent stack. CV-1119A will not automatically close on the high radiation condition, it must be manually closed. A Main Exhaust Fan is required to be operating during a release, and if a fan were to trip during the release, the release would have to be manually secured. The tripping of V-6A or V-6B will not isolate the release.

A. Incorrect, see explanation.

B. Incorrect, see explanation.

C. Correct, see explanation.

D. Incorrect, see explanation.

Technical Reference(s): PL-RMS Radiation Monitoring System Lesson Plan, M-211 Sheets 2 and 3 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ENTERGY NUCLEAR LESSON PLAN RIA-1113, Waste Gas Radiation Monitor VA-66 RIA-1113 Purpose Purpose Digital channel RIA-1113 monitors radioactive waste gas discharges from the waste gas decay tanks to the stack.

It is designed to detect higher-than-expected radiation levels in the waste gas release and terminate the release upon detection of such.

VA-67 RIA-1113 Flow Path Flow Path RE-1113 is a full flow in-line detector that monitors the waste gases that are released from the waste gas decay tanks.

Effluent from the detector outlet goes to the stack.

OBJ 12 Describe the consequences of operating the Radiation Monitoring System with Waste Gas Monitor RE-1113 improperly isolated in accordance with SOP-38.

Waste Gas Surge Tank Relief Valve RV-1111 discharges through RE-1113 to the stack. If RE-1113 is improperly isolated, then RV-1111 will also be isolated and thereby disabled. See Figure 1 below.

When isolating RE-1113, bypass valve MV-WG118 must be opened before closing either RE-1113 inlet isolation valve MV-WG117 or RE-1113 outlet isolation valve MV-WG119.

Failure to perform the alignment as specified will isolate the RV-1111 release path and disable the WGST overpressure protection.

Generally, RE-1113 would be isolated only for maintenance needs.

EK 1371 L

RIA H 1113 EK C-11 1364 HS PURGE TRIP CLOSED VALVES 2317 PI CV-1113 & 1123 CHECK 2317 SOURCE WGDT RV-1111 WG-625 WG-117 WG-119 (L.O.) (L.O.)

WGST STACK PI SV 2317A 2317 CA-332 WG-523 RE PURGE 1113 AS CA-226 FCV 2317 WG-118 (L.C.)

Figure 1. RIA-1113 Flow Path Alarms and Setpoints RIA-1113 HIGH alarm activates Control Room annunciator EK-1364, GASEOUS WASTE MONITORING HI RADIATION.

Addressed in The HIGH alarm setpoint is variable and is calculated specifically for System each waste gas release. Operations Section.

RIA-1113 low signal output actuates Control Room annunciator EK-1371, RADIATION MONITOR SYSTEM CKT FAILURE.

VA-68 RIA-1113 Auto Actions Automatic Actuations On a HIGH alarm, RIA-1113 automatically closes waste gas discharge valves CV-1113 and CV-1123, thereby terminating the waste gas release.

ES-401 Question 64 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __2__ _____

K/A # 072.A2.02 Importance Rating __2.8_ _____

K/A Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the ARM system- and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure.

Proposed Question:

The Plant is at 100% power. Containment Radiation Monitor, RIA-1805, has recently experienced multiple high spikes and has been removed from service by I&C Technicians, who have removed the fuse for the monitor.

An operator accidently tests Containment Radiation Monitor RIA-1806, per SOP-39, Area Radiation Monitoring System, rather than RIA-1805 which has been removed from service.

Containment Radiation Monitors RIA-1807 and RIA-1808 remain unaffected.

Given the conditions above, does a Containment Isolation Signal result and what is the correct action?

A. Containment Isolation occurs; Enter AOP-31, Spurious Containment Isolation.

B. Containment Isolation occurs; Enter EOP-1.0, Standard Post-Trip Actions.

C. Containment Isolation does NOT occur; attempt to reset RIA-1806 per SOP-39, Area Radiation Monitoring System.

D. Containment Isolation does NOT occur; Initiate Work Request to repair.

Proposed Answer: A Explanation (Optional):

A. Correct, a containment isolation occurs on Containment High Radiation (CHR), as the coincidence for a CHR is 1/3 with RIA-1805 removed from service (fuse pulled). Testing the rad monitor RIA-1806 places the monitor in tripped condition (Trip 2) and initiates internal self-check. The self-check applies a voltage to the channel circuitry that corresponds to 103 R/hr, which trips the channel (exceeding the 10 R/hr trip setpoint for a CHR).

B. Incorrect, the applicant incorrectly believes a reactor trip is required. A reactor trip is only required per AOP-31 if the containment isolation is caused by a high containment pressure condition.

C. Incorrect, the applicant incorrectly believes a containment isolation does not occur and

does not believe that removing a monitor from service will allow a 1/3 coincidence to actuate a CHR. This action would be correct if a monitor were to not reset during a test.

D. Incorrect, the applicant incorrectly believes a containment isolation does not occur and does not believe that removing a monitor from service will allow a 1/3 coincidence to actuate a CHR. This action is a correct action per ARP-8 (#63) if a single monitor were to fail.

Technical Reference(s): SOP-39, PL-RMS Radiation Monitoring System, AOP-31, ARP-8 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ENTERGY NUCLEAR LESSON PLAN Containment Radiation Monitors RIA-1805/1806/1807/1808.

The Containment Radiation Monitors are analog channels.

VA-1 RIA-1805/1806/1807/1808 Each channel has a RIA in the Control Room; the four RIAs are identical and include the following controls and indications:

Analog meter Three-position function selector switch Trip 1, Trip 2, and Operate lamps/pushbuttons Analog meter - provides measured radiation field indication Vertical, single-scale meter Meter spans six decades, from 10-2 to 104 R/hr Three-position function selector switch CHECK - Places monitor in tripped condition (Trip 2) and initiates internal self-check.

The self-check applies a voltage to the channel circuitry that corresponds to 103 R/hr, which trips the channel.

The function selector switch spring returns to OPERATE from this position.

OPERATE - Places monitor in service.

TRIP ADJ - Causes the channel output to go full-scale high.

This switch position should NOT be used by Operations personnel Pushbutton lamps:

Operate (Green)

When illuminated, indicates that the monitor is in service.

If not illuminated, then the monitor is not in service, or may have failed low.

Trip 1 (Alert - Yellow)

Lamp illuminates when measured radiation meets or exceeds the alert alarm setpoint.

Alert alarm locks in and yellow lamp remains illuminated until alarm is manually reset by pressing the Trip 1 button Alert setpoint is approximately 1 R/hr.

Trip 2 (High - Red)

Lamp illuminates when measured radiation meets or exceeds the high alarm setpoint.

High alarm locks in and red lamp remains illuminated until alarm is manually reset by pressing the Trip 2 button Actuates Control Room alarm EK-1363, CONTAINMENT HI RADIATION.

High alarm setpoint is approximately 10 R/hr.

The Trip 2 bistable provides an input to the Containment Isolation logic (described later in Section Error! Reference source not found., Error! Reference source not found..)

PL-RMS Revision 4 Page 17 of 71

ES-401 Question 65 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __2__ _____

Group # __2__ _____

K/A # 086.A3.02 Importance Rating __2.9_ _____

K/A Statement: Ability to monitor automatic operation of the Fire Protection System including: Actuation of the FPS.

Proposed Question:

Given the following conditions:

The Plant is at 100% power.

Multiple fire alarms were just received.

Fire Protection header pressure lowered to 88 psig prior to stabilizing at 105 psig.

With NO operator action, what is the status of the Fire Protection pumps?

Fire Jockey Electric Fire Diesel Fire Diesel Fire Pump P-13 Pump P-9A Pump P-9B Pump P-41 A. ON OFF OFF OFF B. OFF ON OFF OFF C. OFF ON ON OFF D. ON ON ON ON Proposed Answer: B Explanation (Optional):

The Fire Jockey Pump (P-13) is rated for 50 gpm at 115 psi; it is designed to handle small leaks and maintain fire header pressure. The electric Fire Pump (P-9A) auto-starts when header pressure drops to 98 psig. The two diesel Fire Pumps (P-9B and P-41) auto-start on low header pressure of 83 psig and 68 psig, respectively. Other than the Jockey Pump, all fire pumps need to be manually secured upon a low pressure auto-start. In this scenario, header pressure is not maintained with the Jockey Pump and the Electric Fire Pump will start upon reaching the auto-start setpoint of 98 psig. The Fire Header Lo Pressure alarm comes in at 95 psig, indicating that the Electric Fire Pump should have auto-started prior to reaching the alarm. The Diesel Fire Pumps P-9B and P-41 will not start since the low header pressure setpoints are not reached.

As pressure recovers above the auto-start setpoints, the Electric Fire Pump remains running until manually secured.

A. Incorrect, the Electric Fire Pump will start. The applicant must understand the auto-start

setpoints of each FP pump and if the pumps must be manually secured or secure automatically.

B. Correct, see explanation.

C. Incorrect, the Diesel Fire Pump P-9B will not start.

D. Incorrect, the Diesel Fire Pumps P-9B and P-41 will not start.

Technical Reference(s): DBD-1.10, AOP-40 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __7__

55.43 _____

Comments:

ES-401 Question 66 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __3__ _____

Group # _____ _____

K/A # G2.1.25 Importance Rating __3.9_ _____

K/A Statement: Ability to interpret reference materials, such as graphs, curves, tables, etc.

Proposed Question:

Given the following conditions:

The Plant has experienced a small break LOCA and has implemented EOP-4.0, Loss of Coolant Accident Recovery.

Pressurizer Level indicates 60% on LIC-0101B.

Pressurizer pressure is 1500 psia.

Containment temperature is 205oF.

What is the actual Pressurizer level?

A. 50%

B. 56%

C. 82%

D. 86%

Proposed Answer: A Explanation (Optional):

A. Correct, at 205oF, the applicant must use page 1 of Supplement 9 to determine the corrected pressurizer indicated level, which is 54% (60% indicated minus 6% error). At 54% pressurizer corrected indicated level and 1500 psia, the actual pressurizer level is 50%.

B. Incorrect, the applicant used the correct hot calibrated (Supplement 9), but did not account for error due to containment temperature.

C. Incorrect, the applicant used the cold calibrated (Supplement 10).

D. Incorrect, the applicant used the cold calibrated (Supplement 10) and did not account for error due to containment temperature.

Technical Reference(s): EOP-4.0, EOP Supplement 9, EOP Supplement 10 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: EOP Supplement 9, 10 Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __10_

55.43 _____

Comments:

ES-401 Question 67 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __3__ _____

Group # _____ _____

K/A # G2.1.30 Importance Rating __4.4_ _____

K/A Statement: Ability to and operate components, including local controls.

Proposed Question:

Component Cooling Water (CCW) has been lost to Containment for greater than 10 minutes.

Per AOP-36, Loss of Component Cooling, why is CCW flow manually re-initiated and where is the preferred restoration performed from?

Assume access to all plant areas is possible, all plant conditions are stable, and CCW flow restoration is desired.

A. Manual flow is re-initiated to prevent thermal shock and possible equipment damage.

This is performed from inside Containment 590' level, using PCP & CRDM return isolation valves.

B. Manual flow is re-initiated to prevent a possible low system pressure auto start on a standby CCW pump. This is performed from inside Containment 590' level, using PCP &

CRDM return isolation valves.

C. Manual flow is re-initiated to prevent a possible low system pressure auto start on a standby CCW pump. This is performed from inside the CCW Pump Room, 590' level, using the CCW Return from Containment isolation (MV-CC713).

D. Manual flow is re-initiated to prevent thermal shock and possible equipment damage.

This is performed from inside the CCW Pump Room, 590' level, using the CCW Return from Containment isolation (MV-CC713).

Proposed Answer: A Explanation (Optional):

A. Correct, flow is manually re-initiated to allow a controlled restoration of flow to components in Containment, to prevent thermal shock and potential equipment damage when cooling water flow is restored. Per AOP-36, with access to containment, flow is to be restored using the PCP and CRDM return isolation valves.

B. Incorrect, see Choice A, flow is slowly restored to minimize the potential for thermal shock to the system and equipment. Doing this will provide the operators better system pressure control.

C. Incorrect, see Choice A, flow is slowly restored to minimize the potential for thermal shock to the system and equipment. Doing this will provide the operators better system

pressure control. The applicant believes containment access is not available or not prudent and restoration should be performed from the CCW Pump room.

D. Incorrect, see Choice A. The applicant believes containment access is not available or not prudent and restoration should be performed from the CCW Pump room.

Technical Reference(s): AOP-36 and AOP-36 bases (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __12_

55.43 _____

Comments:

Question modified from Palisades 2005 NRC Exam. Modified all distractors to better comply with procedural validity. Stem changed to ask for the preferred restoration method.

ES-401 Question 68 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __3__ _____

Group # _____ _____

K/A # G2.1.44 Importance Rating __3.9__ _____

K/A Statement: Knowledge of RO duties in the control room during fuel handling such as alarms from fuel handling area, communication with fuel storage facility, systems operated from the control room in support of fueling operations, and supported instrumentation.

Proposed Question:

The Plant is in the middle of core offload during a refueling outage. In accordance with GOP-11, Refueling Operations and Fuel Handling, and LCO 3.7.12, Fuel Handling Area Ventilation System, core alterations, movement of irradiated fuel, or cask movements are required to be suspended if which of the following Fuel Handling Area Ventilation System alignments is true:

A. Only V-8/B, Fuel Handling Exhaust Fan, is operating.

B. V-7, Fuel Handling Area Supply Fan, is operating.

C. Both V-70A/B, Fuel Handling Area Exhaust Fans, OFF.

D. V-69, Fuel Handling Area Supply Fan, OFF.

Proposed Answer: B Explanation (Optional):

A. Incorrect, per GOP-11, no more than one, V-8A or V-8B, can be operating to comply with LCO 3.7.12.

B. Correct, per GOP-11, V-7 must be OFF.

C. Incorrect, per GOP-11, V-70A/B must be OFF.

D. Incorrect, per GOP-11, V-69 must be OFF.

Technical Reference(s): GOP-11 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 _10__

55.43 _____

Comments:

ES-401 Question 69 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __3__ _____

Group # _____ _____

K/A # G2.2.12 Importance Rating __3.0_ _____

K/A Statement: Knowledge of surveillance procedures.

Proposed Question:

Diesel Generator (DG) 1-1 is running for a monthly surveillance test per MO-7A-1, Emergency Diesel Generator 1-1. While raising load on DG 1-1 to gather two-hour load limit data, the NCO incorrectly stabilized load at 2790 kW.

For DG 1-1 to be below the two-hour load limit, which one of the following is the least amount of load that must be reduced?

A. 50 kW B. 100 kW C. 150 kW D. 300 kW Proposed Answer: A Explanation (Optional):

A. Correct, the two-hour load limit is 2750 kW. Reducing DG load by 50 kW would place DG loading at 2740 kW, under the maximum two-hour load limit.

B. Incorrect, the applicant is applying the maximum value of the DBA load band of 2700 kW.

C. Incorrect, the applicant is applying the minimum value of the DBA load band of 2650 kW.

D. Incorrect, the applicant is applying the maximum continuous load limit of 2500 kW.

Technical Reference(s): MO-7A-1, SOP-22, PL-EDG Emergency Diesel Generators Lesson Plan (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2010 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __10_

55.43 _____

Comments:

Question modified from Palisades 2010 NRC Exam. Modified stem to ask for 2-hr load limit rather than continuous load limit. Changed distractors to accommodate the change in the stem.

ES-401 Question 70 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __3__ _____

Group # _____ _____

K/A # G2.2.35 Importance Rating __3.6_ _____

K/A Statement: Ability to determine Technical Specification Mode of Operation.

Proposed Question:

Given the following plant conditions:

The current time is 0400.

Reactor power is 0%.

All control rods are fully inserted.

PCS average temperature is 320°F.

PCS cooldown rate is 37°F/hr.

Assuming PCS cooldown rate remains stable, what MODE, as defined by Technical Specifications, will the Plant be in at 0800?

A. MODE 3 B. MODE 4 C. MODE 5 D. MODE 6 Proposed Answer: C Explanation (Optional):

A. Incorrect, the Plant is in Mode 3 as of 0400, but will not be in Mode 3 at 0800 given the conditions.

B. Incorrect, the Plant will pass through Mode 4 during the cooldown between 0400 and 0800, however, the Plant will exit Mode 4 prior to 0800.

C. Correct, after 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of a 37oF/hr cooldown rate, PCS temperature will be 172oF. This would place the Plant in Mode 5.

D. Incorrect, no information was provided for head tensioning. Mode 6 requires one or more head bolts less than fully tensioned.

Technical Reference(s): Technical Specifications, Section 1.1 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __10_

55.43 _____

Comments:

ES-401 Question 71 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __3__ _____

Group # _____ _____

K/A # G2.3.4 Importance Rating _3.2__ _____

K/A Statement: Knowledge of radiation exposure limits under normal or emergency conditions.

Proposed Question:

Given the following conditions:

The Plant automatically tripped and Safety Injection actuated due to a Large Break LOCA that occurred inside containment.

LPSI Pump P-67B suction piping was damaged from a water hammer event.

All attempts to isolate the leak from the Control Room have been unsuccessful.

An NPO was attempting to isolate the leak locally when he slipped and was injured. The NPO remains in the area and needs assistance.

Another NPO, stating that he fully understands the potential health risks, has volunteered to find the injured NPO and bring them to a low dose area.

What is the maximum allowed Total Dose (TEDE) exposure the Emergency Plant Manager can authorize the NPO to receive while performing this task?

A. 5 REM.

B. 10 REM.

C. 25 REM.

D. No upper limit for TEDE exposure.

Proposed Answer: D Explanation (Optional):

A. Incorrect, this limit is the 10CFR20 annual limit for the whole body.

B. Incorrect, this limit applies to the protection of property.

C. Incorrect, this limit applies to life-saving or protection of large populations, not on voluntary basis.

D. Correct, no upper TEDE limit applies to life-saving or protection of large populations only on a voluntary basis to persons who are fully aware of the risks involved.

Technical Reference(s): Site Emergency Plan SEP

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 _12__

55.43 _____

Comments:

ES-401 Question 72 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __3__ _____

Group # _____ _____

K/A # G2.3.14 Importance Rating __3.4_ _____

K/A Statement: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Proposed Question:

A welding contractor arrived on site today (9/24/16) and will be performing weld overlay work on the Reactor Head during the upcoming outage three weeks from now. The contractor workers radiation exposure history in the last year is given as follows:

200 mR whole body from a medical procedure three weeks ago.

300 mR TEDE while at Monticello Nuclear Plant from 12/3/15 to 12/6/15

500 mR TEDE while at Perry Nuclear Plant from 3/6/16 to 3/13/16

800 mR TEDE while at Palisades Nuclear Plant from 5/1/16 to 5/14/16

The contract worker has provided completed NRC Form 5s for each quarter for the past 2 years.

Assuming no dose extensions have been authorized for the worker beyond the Annual Entergy Administrative Dose Guideline (ADG), which one of the following values is the maximum amount of whole body radiation the worker can receive at Palisades during the upcoming refueling outage and not exceed the Annual ADG?

A. 200 mR.

B. 400 mR.

C. 700 mR.

D. 1200 mR.

Proposed Answer: C Explanation (Optional):

Per EN-RP-201-004, the ADG is a company-imposed occupational dose guideline used for the purposes of maintaining doses below the regulatory dose limits established for 10CFR Part 20.

As this is occupational dose only, the 200 mR from the medical procedure does not count towards the ADG. Additionally, the 300 mR from working at Monticello does not count as it was during the last calendar year. Only the current calendar year accumulated dose counts towards the ADG (500 mR + 800 mR = 1300 mR). Therefore, 700 mR of margin exists to reach the ADG limit.

A. Incorrect, the applicant believes that the medical exposure counts as well as the prior

year exposure from Monticello.

B. Incorrect, the applicant understands that the medical exposure does not count, but incorrectly counts the dose obtained from working at Monticello, which occurred in the prior year.

C. Correct, the occupational dose accumulated in within the calendar year counts. (2000-(500+800) = 700 mR margin D. Incorrect, the applicant believes the ADG is specific to dose accumulated within the Entergy fleet.

Technical Reference(s): EN-RP-201 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _12__

55.43 _____

Comments:

ES-401 Question 73 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __3__ _____

Group # _____ _____

K/A # G2.4.2 Importance Rating __3.9_ _____

K/A Statement: Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

Proposed Question:

Which of the following conditions would result in raising the pressure setpoint portion of the Thermal Margin/Low Pressure reactor trip?

A. Operating with Group 4 rods inserted 4 into the core instead of full out.

B. Operating with power at 95% instead of 100%.

C. Operating with Tave 2oF above program.

D. Operating with pressurizer pressure 15 psia below normal.

Proposed Answer: C Explanation (Optional):

A. Incorrect, the applicant incorrectly believes that this will cause ASI to be more positive, but actually lowers temperature in the top of the core, thus allowing a greater margin to DNB.

B. Incorrect, the applicant incorrectly believes that this will result in a lower inlet temperature to the core, but at a lower power less heat is added so the likelihood of DNB is lowered.

C. Correct, Operating at an elevated temperature places the plant closer to DNB conditions.

This is actually determined by Tcold temperatures which will be higher with a higher Tave and the same power level.

D. Incorrect, the applicant incorrectly believes that a lower pressure affects the trip setpoint, but actually affects the margin to trip.

Technical Reference(s): LCO 3.3.1, PL-RPS Reactor Protection System Lesson Plan (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 1999 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __5__

55.43 _____

Comments:

Lesson Content Instructor Notes

a. Protection Against Single Failures (1) Field Instrument Failure OBJ 28 Predict how the following conditions will impact operation of the Reactor Protective System:

high failure of monitored parameter input signal (a) Failure in the high direction

i. If the trip unit is designed to trip on a high value of the measured plant parameter, then a field instrument failure in the high direction will cause the trip unit to trip.

This will not cause a reactor trip because the logic requires two channels to reach the set point, and only one channel is in the trip condition.

This will not prevent a reactor trip because three operable redundant instruments remain in service, and any two of the operable channels going into trip will cause a reactor trip.

The trip logic is changed from 2/4 to 1/3, unless the failed instruments trip unit is bypassed.

o If the failed instruments trip unit is bypassed, then the trip logic is changed to 2/3.

This will not cause a reactor trip because the trip unit has not actuated a trip signal in any of the three associated logic matrices.

This will not prevent a reactor trip because three operable redundant instruments remain in service, and any two of the operable channels going into trip will cause a reactor trip.

o Low failure of a pressurizer pressure instrument will prevent the pressurizer high pressure trip unit from actuating on the affected channel, but will actuate the Thermal Margin/Low Pressure trip unit.

ii. If the trip unit is designed to trip on a low value of the plant parameter, then a field instrument failure in the high direction will prevent the trip unit from tripping.

This will not cause a reactor trip because the trip unit has not actuated a trip signal in any of the three associated logic matrices.

This will not prevent a reactor trip because three operable redundant instruments remain in service, and any two of the operable channels going into trip will cause a reactor trip.

Lesson Content Instructor Notes

High failure of a pressurizer pressure instrument will prevent the Thermal Margin/Low Pressure trip unit from actuating on the affected channel, but will actuate the pressurizer high pressure trip unit.

The trip logic is changed from 2/4 to 2/3, unless the failed instruments trip unit is placed in the trip condition.

o If the failed instruments trip unit is placed in the trip condition, then the trip logic is changed to 1/3.

OBJ 28 Predict how the following conditions will impact operation of the Reactor Protective System:

low failure of monitored parameter input signal (b) Failure in the low direction

i. If the trip unit is designed to trip on a high value of the measured plant parameter, then a field instrument failure in the low direction will prevent the trip unit from tripping.

This will not cause a reactor trip because the trip unit has not actuated a trip signal in any of the three associated logic matrices.

This will not prevent a reactor trip because three operable redundant instruments remain in service, and any two of the operable channels going into trip will cause a reactor trip.

o Low failure of a pressurizer pressure instrument will prevent the pressurizer high pressure trip unit from actuating on the affected channel, but will actuate the Thermal Margin/Low Pressure trip unit.

PL-RPS Revision 5 SLIDE 1 Basis - Thermal Margin / Low Pressure (TM/LP) Trip (a) Thermal Margin / Low Pressure (TM/LP) Trip

i. The TM/LP trip is provided to prevent reactor operation when the DNBR is insufficient. The TM/LP trip protects against slow reactivity or temperature increases, and against pressure decreases.

ii. The trip set points are derived from the core thermal limits through application of appropriate allowances for measurement uncertainties and processing errors. The allowances specifically account for instrument drift in both power and inlet temperatures, calorimetric power measurement, inlet temperature measurement, and primary system pressure measurement.

iii. Other uncertainties including allowances for assembly power tilt, fuel pellet manufacturing tolerances, core flow measurement uncertainty and core bypass flow, inlet temperature measurement time delays, and ASI measurement, are included in the development of the TM/LP trip set point used in the accident analysis.

ES-401 Question 74 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __3__ _____

Group # _____ _____

K/A # G2.4.6 Importance Rating __3.7_ _____

K/A Statement: Knowledge of EOP mitigation strategies.

Proposed Question:

The Plant has entered EOP-7.0, Loss of All Feedwater Recovery. Both Steam Generator levels are at 10% narrow range and slowly decreasing. Other than attempting to re-establish feedwater, which of the following is performed to mitigate the event?

A. Cooldown the PCS Tave to less than 525oF.

B. Maintain PCS Tave less than 540oF.

C. Minimize PCS subcooling.

D. Secure ONLY one PCP in each loop.

Proposed Answer: B Explanation (Optional):

A. Incorrect, a cooldown to 525oF-532oF is performed to allow feeding the S/Gs with AFW, however, since adequate shutdown margin may not be established at this point, 525oF is the bounding low temperature.

B. Correct, the S/G steaming strategy to maintain Tave less than 540oF (the maximum post-trip temperature) is intended to maintain existing PCS temperature and prevent uncontrolled heatup; ensuring PCS heat removal is maintained throughout the event.

C. Incorrect, subcooling should be maximized in order to minimize potential for voiding and to provide sufficient margin for reestablishing HPSI flow if the minimum value cannot be maintained.

D. Incorrect, all PCPs are secured to minimize the heat input to the PCS.

Technical Reference(s): EOP-7.0 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 __10__

55.43 _____

Comments:

ES-401 Question 75 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # __3__ _____

Group # _____ _____

K/A # G2.4.11 Importance Rating __3.4_ _____

K/A Statement: Knowledge of abnormal condition procedures.

Proposed Question:

Which of the following combination of conditions meets the reactor trip criteria per AOP-35, Loss of Service Water?

1. Annunciator EK-1124, Traveling Screen Hi DP in alarm.

2. Service Water Bay level is 574

3. Annunciator EK-1165, Non Critical Serv Water Lo Press in alarm

4. EK-0259, Exciter Cooler Hi Temp in alarm.

A. 1 and 2 B. 1 and 3 C. 2 and 3 D. 3 and 4 Proposed Answer: D Explanation (Optional):

A. Incorrect, traveling screen high P alarms are not reactor trip criteria. If the alarm is in and bay level is dropping, the Dilution Water Pump(s) must be secured and reactor power reduced. Service Water Bay level has reactor trip criteria, however, the level is 572, which is not met in this case.

B. Incorrect, traveling screen high P alarms are not reactor trip criteria. If the alarm is in and bay level is dropping, the Dilution Water Pump(s) must be secured and reactor power reduced.

C. Incorrect, Service Water Bay level has reactor trip criteria, however, the level is 572, which is not met in this case.

D. Correct, Non-Critical Service Water pressure low combined with high exciter temperatures meets reactor trip criteria per AOP-35.

Technical Reference(s): AOP-35 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 __10__

55.43 _____

Comments:

ES-401 Question 76 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __1__

Group # _____ __2__

K/A # 009.EA2.34 Importance Rating _____ _4.2__

K/A Statement: Ability to determine or interpret the following as they apply to a small break LOCA:

Conditions for throttling or stopping HPI.

Proposed Question:

A small break Loss of Coolant Accident (LOCA) has occurred. The crew is at Step 23 of EOP-4.0, Loss of Coolant Accident Recovery, SI Pump throttling.

The following plant conditions are observed:

Containment Pressure is 3.4 psig and slowly rising.

Containment temperature is 130oF and slow rising.

PCS Pressure is 1400 psia and rising.

Average CET temperature is 543oF and lowering.

PCS subcooling exceeds the minimum required and is rising.

Actual S/G levels are 64% and stable for both S/Gs.

Actual Pressurizer level is 26% and slowly rising.

RVLMS channels indicate 4 Red lights and 4 Green lights.

ONLY HPSI Pump P-66A is running.

For the given conditions:

The required action in response to SI pump throttling criteria is to __(1)__. The basis of the action to throttle SI pump flow is to __(2)__?

A. (1) RAISE HPSI flow AND START HPSI Pumps as necessary (2) Reduce potential for PCS overpressurization B. (1) THROTTLE HPSI flow OR STOP one HPSI Pump at a time (2) Reduce potential for PCS overpressurization C. (1) RAISE HPSI flow AND START HPSI Pumps as necessary (2) Reduce potential for PCS overcooling D. (1) THROTTLE HPSI flow OR STOP one HPSI Pump at a time (2) Reduce potential for PCS overcooling Proposed Answer: A

Explanation (Optional):

The purpose of throttling SI pump flows is to reduce the potential for PCS overpressurization and have better control of PZR pressure and level via normal methods. The applicant could misunderstand the basis for this action by believing the influx of cold, borated water in excess of the minimum required could cause an overcooling concern.

SI Pump throttling criteria is met if ALL of the following are met:

1) Average of QCET is at least 25oF subcooled or greater than the minimum subcooling curve on EOP Supplement 1 (degraded containment conditions)

2) Corrected PZR level is greater than 20% (40% for degraded containment) and controlled.

3) At least one S/G is available for PCS heat removal with corrected level being maintained or being restored to between 60% and 70%.

4) Operable RVLMS channels indicate greater than 102 inches above the bottom of fuel alignment plate (6218)

A. Correct, SI pump throttling criteria is not met. Containment is degraded in this case, as containment pressure is greater than 3 psig. With degraded containment, pressurizer level must be greater than 40% and it is not. Saturation temperature at 1400 psia is 587oF, therefore, subcooling is 34oF. S/G level and RVLMS indicated level are both satisfactory. Perform RNO action.

B. Incorrect, SI pump throttling criteria is not met, see choice A for explanation.

C. Incorrect, SI pump throttling criteria is not met, see choice A for explanation. The purpose of SI pump throttling is to reduce the possibility for PCS overpressurization.

D. Incorrect, while SI Pump throttling criteria is not met, the purpose of SI pump throttling is to reduce the possibility for PCS overpressurization.

Technical Reference(s): EOP-4.0, EOP-4.0 Basis___________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ___________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This meets SRO criteria 10CFR55.43 (b)5 since the applicant must assess plant conditions and determine the required procedural action in response to those conditions.

The applicant must also know the basis behind the procedural action.

ES-401 Question 77 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __1__

Group # _____ __1__

K/A # 000038.G2.1.20 Importance Rating _____ __4.6_

K/A Statement: Ability to interpret and execute procedure steps.

Proposed Question:

Given the following:

EOP-5.0, Steam Generator Tube Rupture Recovery, has been implemented due to a Steam Generator Tube Rupture in progress on the A S/G.

A S/G has been isolated per EOP-5.0.

A S/G pressure is 840 psia.

B S/G pressure is 560 psia.

Pressurizer pressure is 930 psia.

Pressurizer level is 42%.

PCS temperature is 508oF.

For these conditions, which of the following actions, regarding Pressurizer pressure, will the CRS direct in accordance with EOP-5.0?

A. Raise Pressurizer pressure to re-establish adequate subcooling margin B. Raise Pressurizer pressure to prevent backflow dilution of the PCS C. Lower Pressurizer pressure to ensure Main Steam Safety Valves remain closed D. Lower Pressurizer pressure to minimize leakage into the A S/G from the PCS Proposed Answer: D Explanation (Optional):

Pressurizer pressure shall be maintained per the following criteria of EOP-5.0 Step 17 (continuously applicable):

Less than 940 psia

Within the limits of EOP Supplement 1

Preferably within 50 psid of the isolated S/G pressure A. Incorrect, subcooling margin is adequate at approximately 33oF. EOP Supplement 1 requires a minimum of 25oF subcooling.

B. Incorrect, raising Pressurizer pressure would prevent backflow to the PCS; however, per the EOP basis document, the amount of dilution would not jeopardize shutdown margin C. Incorrect, Pressurizer pressure is maintained less than 940 psia in order to prevent the

MSSVs from opening.

D. Correct, the Pressurizer pressure and ruptured S/G pressure should be within 50 psid of each other. Maintaining PCS pressure approximately equal to or less than the affected S/G pressure allows for minimum from the PCS into the S/G.

Technical Reference(s): EOP-5.0, EOP-5.0 Basis, EOP Supplement 1 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This meets SRO-only criteria 10CFR55.43 (b)5 as the applicant must assess plant conditions and then direct procedural action to mitigate the event/transient as well as understanding the basis of why the action is being taken.

ES-401 Question 78 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __1__

Group # _____ __1__

K/A # CE/E06.EA2.1 Importance Rating _____ _3.9__

K/A Statement: Ability to determine and interpret the following as they apply to the (Loss of Feedwater):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Proposed Question:

Given the following:

The Plant has entered EOP-7.0,Loss of All Feedwater Recovery.

The Plant has experienced a loss of all instrument air.

P-8A and P-8B, Auxiliary Feedwater Pumps, are NOT available.

Which one of the following procedures will the CRS use to feed the Steam Generators using P-8C, Auxiliary Feedwater Pump, for the above conditions?

A. EOP Supplement 19, "Alternate Auxiliary Feedwater Methods."

B. AOP-37, "Loss of Instrument Air."

C. SOP-12, "Feedwater System."

D. EOP Supplement 31, "Supply AFW Pumps from Alternate Sources."

Proposed Answer: A Explanation (Optional):

A. Correct, EOP Supplement 19 contains directions for manually initiating AFW flow by manually operating AFW control valves without air.

B. Incorrect, the applicant may believe there is guidance in AOP-37 for manually operating the AFW control valves without instrument air.

C. Incorrect, the applicant may believe there is guidance in SOP-12 for manually operating the AFW control valves without instrument air.

D. Incorrect, the applicant may believe there is guidance in EOP supplement 31 for manually operating the AFW control valves, however, this procedure is used to establish a water source if T-2 is not available.

Technical Reference(s): EOP-7.0, EOP Supplement 19 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2010 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This question meets the criteria for an SRO-only question because the applicant must assess the facility conditions given in the stem and use those conditions to select the appropriate procedure to mitigate the consequences a loss of all feed water using AFW flow control valves in manual.

ES-401 Question 79 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __1__

Group # _____ __1__

K/A # 055.EA2.02 Importance Rating _____ __4.6__

K/A Statement: Ability to determine or interpret the following as they apply to a Station Blackout: RCS core cooling through natural circulation cooling to S/G cooling Proposed Question:

At time 1330, a Station Blackout occurred.

The crew entered EOP-3.0, Station Blackout Recovery, with the following conditions:

PCS pressure is 1905 psig and slowly lowering.

P-8B AFW Pump tripped on overspeed and cannot be reset.

Both S/G Levels are lowering.

Loop Tcold temperatures are 495oF and lowering rapidly.

Loop Thot temperatures are 535oF and rising slowly.

Qualified CET indicate 542oF and rising slowly.

Complete the following statements:

To ensure that the PCS is cooled by Natural Circulation, safety related 2400 VAC power must be restored from an Emergency Diesel Generator by time __(1)__ in accordance with the FSAR, Station Blackout Analysis.

To maintain Natural Circulation conditions for the temperatures given above, the CRS should direct the NCO to throttle the ADVs __(2)__.

A. (1) 1430 (2) OPEN B. (1) 1430 (2) CLOSED C. (1) 1730 (2) OPEN D. (1) 1730 (2) CLOSED Proposed Answer: D Explanation (Optional):

15 minutes following the SBO, Natural Circulation is developing. Thot and Tcold separate, but the delta T between should be no more than 50oF per Natural Circulation criteria. For the given conditions, the delta T is at 40oF with Thot slowly rising. Since Tcold is lowering at an accelerated rate, the ADVs need to be throttled closed to reduce the cooldown rate in order to maintain loop delta T < 50oF.

A. Incorrect, see explanation and selection B B. Incorrect, Palisades is a DC coping unit and the FSAR SBO analysis assumes a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months duration until AC power has to be restored. The applicant could chose the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time requirement as the station batteries are rated at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months without performing the load shed, however, this is not in accordance with EOP-3.0. By performing the DC load shed, within the 30 minute time requirement per the FSAR, the 4160VAC safety bus power restoration requirement is expanded to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

C. Incorrect, see explanation.

D. Correct, see explanation. FSAR SBO safety analysis takes credit for the operator action of establishing 4160VAC safety bus power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Technical Reference(s): EOP-TCA 20, EOP-3.0____________________________

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam ____________

(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge ____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This meets SRO criteria 10CFR55.43 (b)5, for the assessment and procedural selection to comply with the SBO safety analysis in the FSAR which has time requirements as part of the facility license to ensure that PCS Core Cooling is established and maintained through Natural Circulation.

ES-401 Question 80 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __1__

Group # _____ __1__

K/A # 000056.G2.2.22 Importance Rating _____ _4.7__

K/A Statement: Knowledge of limiting conditions for operations and safety limits.

Proposed Question:

Given the following:

The Plant is in MODE 6 with core offload in progress.

All 2400VAC busses are being supplied by Safeguards Transformer 1-1.

Switchyard Rear "R" bus has been de-energized for maintenance.

A Loss of Offsite Power occurs due to a Safeguards Transformer 1-1 fault and both Diesel Generators (DG) energize their respective 2400VAC busses.

For these conditions, which one of the following describes the Technical Specification 3.8.2 AC Sources - Shutdown implications for this event and why?

A. LCO 3.8.2 is met because Station Power Transformer 1-2 is a qualified offsite source in Mode 6.

B. LCO 3.8.2 is met because both DGs are operable and supplying the 2400 VAC safety-related busses.

C. LCO 3.8.2 must be entered because no required offsite source is available to supply the 2400 VAC safety-related busses.

D. LCO 3.8.2 must be entered because only one of two qualified offsite sources required to be operable in MODE 6 is available.

Proposed Answer: C Explanation (Optional):

A. Incorrect, although Station Power Transformer 1-2 is considered a qualified offsite source in Mode 5-6 or defueled, it is not considered available unless backfeed is actually aligned through the Main Transformer No 1 to be able to power 2400 VAC busses. One operable offsite circuit ensures that all required loads may be powered from offsite power. Any of the three offsite supplies, Safeguards Transformer 1-1, Station Power Transformer 1-2, or Startup Transformer 1-2 is acceptable as a qualified circuit.

B. Incorrect, LCO 3.8.2 requires at least one qualified offsite source and at least one DG operable.

C. Correct, there must be a minimum of one of three qualified offsite sources operable to supply the 2400 VAC busses.

D. Incorrect, the applicant believes that two offsite power sources are required to be operable in Mode 6, which is true for Modes 1-4.

Technical Reference(s): LCO 3.8.1, LCO 3.8.2 & Bases (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2008 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __2__

Comments:

This question meets the criteria for an SRO-only question as the applicant must understand and apply Tech Spec conditions and actions in accordance with rules of application requirements and understand the bases for the spec.

Question modified from Palisades 2008 NRC Exam. Modified stem to change scenario to a Loss of Offsite Power with information that both D/Gs started to supply their respective busses. Additionally, information that Station Power Transformer 1-2 was capable of being used via Main Transformer backfeed was removed.

ES-401 Question 81 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __1__

Group # _____ __1__

K/A # 000058.G2.2.37 Importance Rating _____ _4.6__

K/A Statement: Ability to determine operability and/or availability of safety related equipment.

Proposed Question:

Electrical Maintenance has just completed the Monthly Battery Check surveillances ME-12A and ME-12B on both Station Batteries ED-01 and ED-02.

Given the following battery parameters:

Station Battery ED-01 Station Battery ED-02 Electrolyte Level 1/2 inch below max 1/2 inch below max Electrolyte Temp. 68oF 76oF Float Voltage 2.11V 2.12V Specific Gravity 1.205 1.192 Note - Battery cell parameters provided are representative of both the pilot cell and each connected cell.

Based on the given conditions, which of the following is the maximum time allowed before the Plant must be in Mode 3, based on the applicable Technical Specifications?

A. 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months B. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C. 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months D. 31 hours3.587963e-4 days 0.00861 hours 5.125661e-5 weeks 1.17955e-5 months Proposed Answer: A Explanation (Optional):

Station Battery ED-01 has low electrolyte temperature, which provides direct entry into LCO 3.8.6 Condition C, requiring ED-01 to be declared inoperable. With the ED-01 inoperable, LCO 3.8.4 Condition B must be applied, requiring restoration of the battery within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Station Battery ED-02 has low specific gravity, lower than the Category C limits of LCO 3.8.6, requiring entry into LCO 3.8.6 Condition C. Since the battery cell parameters are given (assumed that the pilot cell and each connected cell values are the same), Condition C must be entered since it is known that Category C limits for specific gravity are not met for ED-02. One is not allowed the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to verify this in Condition A as it is already known and given in the

question stem. This (LCO 3.8.6 Condition C) requires the battery to be immediately declared inoperable. With ED-02 inoperable, LCO 3.8.4 Condition B must be applied, however, now both batteries are inoperable and LCO 3.0.3 must be applied, requiring the plant to be in Mode 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months ; the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of shutdown preparation time does not extent the maximum allowable time to reach Mode 3.

A. Correct, see explanation.

B. Incorrect, the applicant is directly entering LCO 3.0.3, but not appropriately applying the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months shutdown preparation time allowance. The applicant could also be incorrectly using the LCO 3.8.6 Condition A 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to verify battery ED-02 is not within Category C limits and then incorrectly applying LCO 3.8.4 Condition C, which would not be applied as both batteries are inoperable.

C. Incorrect, the applicant is incorrectly applying LCO 3.8.6 Condition A requirements to verify battery cell parameters within Category C limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The battery cell parameters (in this case, a low specific gravity not within Category C limits) is known.

However, in this case, the applicant also does not apply LCO 3.0.3 for 2 inoperable batteries and incorrectly enters LCO 3.8.4 Condition C, to be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months + 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ).

D. Incorrect, the applicant is incorrectly applying LCO 3.8.6 Condition A requirements to verify battery cell parameters within Category C limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The battery cell parameters (in this case, a low specific gravity not within Category C limits) is known.

The applicant is correctly applying LCO 3.0.3 time requirements (7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months to be in Mode 3), but incorrectly applying LCO 3.8.6 requirements.

Technical Reference(s): LCO 3.8.6, LCO 3.8.4 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: TS 3.8.4, TS 3.8.6 Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This question meets the criteria for an SRO-only question as the applicant must apply TS Required Actions in accordance with rules of application requirement, of which are not specifically immediate or within one hour actions requirements.

ES-401 Question 82 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __1__

Group # _____ __2__

K/A # 000033.AA2.10 Importance Rating _____ _3.8__

K/A Statement: Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Tech-Spec limits if both intermediate range channels have failed.

Proposed Question:

Given the following conditions:

A Reactor start-up is in progress in Mode 3

All shutdown rods and part-length rods are withdrawn

Group 1 regulating rods remain fully inserted

Source Range NI-1 indicates 500 cps

Source Range NI-2 indicates 500 cps

Wide Range NI-3 indicates 3x10-7%

Wide Range NI-4 indicates 2x10-7%

What actions are required per Technical Specifications?

A. Enter Tech Spec 3.0.3 due to BOTH Wide Range NI channels inoperable.

B. Immediately stop all positive reactivity additions and restore BOTH Wide Range NI channels to OPERABLE status prior to MODE 2.

C. Immediately stop all positive reactivity additions and restore ONE Wide Range NI channel to OPERABLE status prior to MODE 2.

D. Remain in Mode 3 with power less than 1E-4% and High Startup Rate Trips bypassed.

Proposed Answer: B Explanation (Optional):

The neutron flux monitoring channels consist of two combined source range/wide range channels, designated NI-1/3 and NI-2/4. The wide range portions, (NI-3 and NI-4) provide neutron flux power indication from < 1E-7% RTP to > 100% RTP. The source range portions, designated NI-1 and NI-2, provide source range indication over the range of 0.1 to 1E+5 cps.

A. Incorrect, in accordance with LCO 3.3.9, two neutron flux channels (NI-1/3 and NI-2/4) are required to be operable. LCO 3.3.9 has required action A.1 if no neutron flux monitoring channel is operable (the action is the same as one channel inoperable). In this case, NI-3 and NI-4 are not operable as they are indicating significantly lowered than expected for an approach to critical. Therefore, LCO 3.0.3 is not applicable due to the

NI-3 and NI-4 failures.

B. Correct, both Wide Range NIs are required in Mode 3.

C. Incorrect, both Wide Range NIs are required in Mode 3. The applicant could believe only one Wide Range NI is required to be operable in Mode 3.

D. Incorrect, the High Startup Rate Trip uses the Wide Range NIs and is automatically bypassed when power is <1x10-4 % and is not applicable in this Mode. The applicant could believe that this trip is required in Mode 3.

Technical Reference(s): Tech Spec 3.3.1 and Bases, Tech Spec 3.3.9 and Bases, GOP-3 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This question meets the criteria for an SRO-only question as the applicant must apply TS Required Actions in accordance with rules of application requirement, which are not specifically immediate or within one hour actions requirements.

Lesson Content Instructor Notes

1. Excore - Source and Wide Range Nuclear Instruments

a. Source and Wide Ranges cover approximately 12 decades.

1) SR covers 6 decades from 0.1 to 105 cps (approximately 10-10 to 10-4 % power).

2) WR covers 10 1/3 decades from 10-8 to 200 % power.

3) When SR is ~ 3 cps, WR should read ~ 1x10-7 %.

PL-NI Revision 5 Page 43 of 72 GOP-3 Section 5.2

ES-401 Question 83 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __1__

Group # _____ __2__

K/A # 000061.G2.2.25 Importance Rating _____ __4.2_

K/A Statement: Knowledge of the bases in Technical Specification limiting conditions for operations and safety limits.

Proposed Question:

Given the following conditions:

The Plant is in Mode 4 coming out of a refueling outage.

The Spent Fuel Pool is full of irradiated fuel.

The Mode Change Checklist is complete with one discrepancy.

o A channel check on Spent Fuel Pool Area Monitor, RIA-2313, was not performed during SHO-1, Operators Shift Items Modes 1, 2, 3 and 4.

The Plant Manager wants to continue the Plant heatup and transition into Mode 3. Can the transition to Mode 3 continue; why or why not?

A. Yes, the Plant can transition to Mode 3, provided no fuel is being moved within the SFP.

B. Yes, the Plant can transition to Mode 3, provided a risk assessment is performed.

C. Yes, the Plant can transition to Mode 3, as the monitor is not in the Mode of applicability.

D. No, the Plant must remain in Mode 4 until the monitor is restored.

Proposed Answer: A Explanation (Optional):

A. Correct, ORM 3.7.16 requires both fuel pool area radiation monitors to be operable with fuel in the fuel pool area. Specification 3.0.4c is applicable to this specification, per ORM Table 3.17.6. As such, entry into a different mode with an inoperability provided the associated actions to be entered (restore in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in this case) do not provide for continued operation for an unlimited period of time and a risk assessment has not been performed.

B. Incorrect, the plant can transition to Mode 3, as discussed in explanation A, however, the allowance to change modes is not due to performing a risk assessment. The applicant misunderstands Specification 3.0.4c in this case.

C. Incorrect, while the plant can transition to Mode 3, ORM 3.17.6 is always applicable with fuel in the fuel pool area. The applicant misunderstands Table 3.17.6.

D. Incorrect, specification 3.0.4c is applicable for the fuel pool area radiation monitors. The applicant misunderstands the application of Specification 3.0.4c.

Technical Reference(s): ORM 3.17.6, ORM Specification 3.0.4 Bases (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ORM 3.17.6 Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __2__

Comments:

This question meets SRO-only criteria as the applicant must understand and apply conditions in the Operational Requirements Manual and must also apply generic LCO requirements.

ES-401 Question 84 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __1__

Group # _____ __2__

K/A # 000076.G2.2.38 Importance Rating _____ _4.5__

K/A Statement: Knowledge of conditions and limitations in the facility license.

Proposed Question:

Given the following conditions:

The Plant is at 100% power.

At 0800 on 9/26/16, Surveillance Requirement 3.4.16.2 was complete.

The initial dose equivalent I-131 sample result was 2 µCi/gm.

Assuming Dose Equivalent I-131 concentration were to continue to rise by 2 µCi/gm upon each subsequent 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months sample, and is not restored, what is the latest time the Plant is required to be in Mode 3 with Tave < 500oF, per LCO 3.4.16?

A. 1400 on 9/28/16 B. 1500 on 9/28/16 C. 1800 on 9/29/16 D. 1900 on 9/29/16 Proposed Answer: A Explanation (Optional):

A. Correct, using a linear relationship of 2 µCi/gm rise upon each sample (every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per LCO 3.4.16 Required Action A.1, the Dose Equivalent I-131 (DE I-131) would reach 26 µCi/gm after 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . LCO 3.4.16 Condition B is required to be entered if the Required Action and associated Completion Time of Condition A not met OR DE 1-131 >

40 µCi/gm OR Gross specific activity of the primary coolant not within limit. In this case, the required action and associated completion time of Condition A is not met. Adding 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to the Condition B entrance time, the latest time the plant would be required to be in Mode 3 with Tave < 500oF is 1400 on 9/28/16.

B. Incorrect, the applicant entered LCO 3.0.3 upon reaching the 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of DE 1-131 exceeding 1.0 µCi/gm. LCO 3.0.3 allows an extra hour to prepare for plant shutdown, whereas LCO 3.4.16 Condition B does not.

C. Incorrect, the applicant waited to enter LCO 3.4.16 Condition B until DE 1-131 reached 40 µCi/gm (40 µCi /gm reached at 1200 on 9/29/16), rather than entering Condition B after 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months per Required Action A.2.

D. Incorrect, the applicant entered LCO 3.0.3 upon reaching 40 µCi/gm DE 1-131 rather

than entering Condition B.

Technical Reference(s): LCO 3.4.16 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: TS LCO 3.4.16 Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __1__

Comments:

This question meets the criteria for an SRO-only question as the applicant must apply TS Required Actions in accordance with rules of application requirements as well as the application of generic LCO requirements.

Need to include the next TS page which shows the TS SR 3.4.16.2 requirements - < or = 1.0 µCi/gm.

ES-401 Question 85 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __1__

Group # _____ __2__

K/A # CE/A11.AA2.02 Importance Rating _____ _3.4__

K/A Statement: Ability to determine and interpret the following as they apply to the (RCS Overcooling):

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Proposed Question:

Given the following conditions:

The Plant is in Mode 3 at Normal Operating Temperature and Pressure.

A Main Steam Line break occurs on the A Main Steam Line outside Containment, upstream of the MSIV.

After 15 minutes, A Steam Generator blows dry.

Average of the Qualified Core Exit Thermocouples (QCET) lowers to 442°F

Tcold lowers to 425°F.

With Primary Coolant Pumps in operation ___(1)___ should be used to monitor PCS cooldown limits. The Primary Coolant System cooldown limits ___(2)___ been exceeded.

___(1)___ ___(2)___

A. QCET have B. QCET have NOT C. Tcold have D. Tcold have NOT Proposed Answer: C Explanation (Optional):

A. Incorrect, part 1 is incorrect. Tcold should be used to monitor cooldown and not QCET.

QCET is used when determining subcooling for PCP operation. The cooldown calculation is correct.

B. Incorrect, both parts are incorrect. Tcold should be used to monitor cooldown with PCPs running. Cooldown rate has been exceeded.

C. Correct, EOP-6 requires the operator to determine the cooldown rate during an ESDE per EOP Supplement 2. EOP Supplement 2 uses Tcold, as this would be the temperature

of the PCS water entering the reactor vessel. Per EOP-6.0, the operators are bounded by TS 3.4.3 values of 100oF/hr maximum cooldown. During this event, Tcold dropped greater than 100°F over 5 minutes (no-load Tcold is 532oF, but could be as high as 540oF due to ADV control) and the cooldown rate limit of 100°F/hr was exceeded.

D. Incorrect, the use of Tcold is correct, however, the cooldown has been exceeded. The applicant was using the average QCET value for the cooldown rate.

Technical Reference(s): EOP-6, TS 3.4.3 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This meets SRO criteria 10CFR55.43 (b)5 since the safety analysis in the FSAR and TSs have requirements as part of the facility license to ensure that PCS system cooldown rate limitations are maintained.

ES-401 Question 86 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __2__

Group # _____ __1__

K/A # 010.G2.2.42 Importance Rating _____ _4.6_

K/A Statement: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Proposed Question:

Given the following conditions:

The Plant is at 100% power

Pressurizer Pressure indication, PIA-0102BLL, indicates 1500 psig

The following alarms are in:

o EK-0756, Pressurizer Safety Inj Signal B Lo-Lo Press o EK-0601C, TM/LO Pressure Channel Trip o EK-0605C, TM/LO Pressure Channel Pre-Trip Based on the given conditions, the CRS will enter which of the following Tech Specifications?

A. TS 3.3.1, Reactor Protection System (RPS) Instrumentation and TS 3.3.3, Engineered Safety Features (ESF) Instrumentation B. TS 3.3.1, Reactor Protection System (RPS) Instrumentation and TS 3.4.12, Low Temperature Overpressure Protection (LTOP) System C. TS 3.3.3, Engineered Safety Features (ESF) Instrumentation and TS 3.3.7, Post Accident Monitoring (PAM) Instrumentation D. TS 3.3.3, Engineered Safety Features (ESF) Instrumentation and TS 3.4.12, Low Temperature Overpressure Protection (LTOP) System Proposed Answer: A Explanation (Optional):

A. Correct, the applicant should identify that a Pressurizer Pressure transmitter (PT-0102B) has failed low and provides inputs into RPS and SIS, given the alarms provided. The applicant should understand the TS implication and that TS 3.3.1 (RPS) and TS 3.3.3 (SIS) are applicable.

B. Incorrect, TS 3.3.7 is not applicable to this pressure transmitter. The applicant is confusing this pressure transmitter for one that provides input to WR pressure indication and LTOP (PT-0105A/B).

C. Incorrect, TS 3.3.8 is not applicable to this pressure transmitter. The applicant is confusing this pressure transmitter for one that provides input to the LTOP and SDC

Interlocks (PT-0104A/B).

D. Incorrect, Technical Reference(s): TS 3.3.1, TS 3.3.3, TS 3.3.1 Bases (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __3__

Comments:

This questions meets SRO-only criteria as the applicant must use Tech Spec Bases knowledge to apply given entry-level conditions to Tech Spec applicability.

ES-401 Question 87 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __2__

Group # _____ __1__

K/A # 012.A2.03 Importance Rating _____ _3.7_

K/A Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the RPS:

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Incorrect channel bypassing.

Proposed Question:

Given the following conditions:

The Plant is at 100% power

Source/Wide Range NI-2/4 is being removed from operation

The NCO incorrectly performed the following on the RPS channels associated with NI-2/4:

o D RPS channel in bypass o C RPS channel in trip What actions should the NCO have taken and what procedure should be used to remove the Source/Wide Range NI from operation?

A. C channel should have been bypassed and D tripped. SOP-36, Reactor Protective System and Anticipated Transient Without Scram (ATWS) System.

B. B channel should have been bypassed and D tripped. SOP-36, Reactor Protective System and Anticipated Transient Without Scram (ATWS) System.

C. C channel should have been bypassed and D tripped. SOP-35, Neutron Monitoring System.

D. B channel should have been bypassed and D tripped. SOP-35, Neutron Monitoring System.

Proposed Answer: D Explanation (Optional):

A. Incorrect, C channel should only be tripped or bypassed when removing NI-1/3 from service. SOP-36 is not used to remove the NI from operation, but is used to bypass the channel.

B. Incorrect, SOP-36 is not used to remove the NI from operation, but is used to bypass the channel.

C. Incorrect, C channel should only be tripped or bypassed when removing NI-1/3 from service. SOP-35 is the correct procedure.

D. Correct, when taking Source/Wide Range NI out of service, the NCO should use SOP-35 Section 7.1.2. This section requires the NCO to place in bypass one affected channel Hi SUR Trip unit AND in trip any other affected channel Hi SUR Trip unit. SOP-36 is used to bypass the channel.

Technical Reference(s): SOP-35 (section 7.1.2), SOP-36 (section 7.4)

(Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None_____

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This question meets SRO-only criteria as the applicant must have detailed knowledge of a System Operating procedure.

ES-401 Question 88 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __2__

Group # _____ __1__

K/A # 059.A2.04 Importance Rating _____ _3.4_

K/A Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the MFW:

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Feeding a dry S/G Proposed Question:

Given the following conditions:

EOP-7.0, Loss of All Feedwater, is in progress.

A Steam Generator level: -75%.

B Steam Generator level: -95%.

Both A and B S/G pressures: 410 psia.

The crew is preparing to use the Condensate Pumps for feeding the Steam Generators.

Feed Regulating Bypass Valves (CV-0734, CV-0735) have been opened to 20%.

Feed pump discharge pressure 500 psia.

Which of the following describes:

1) Whether the resulting amount of feed flow will be acceptable, and

2) Gives the bases for subsequent selection of procedure actions?

A. 1) Feed flow to both S/Gs will be acceptable.

2) Both S/Gs may be fed at this rate until levels are restored to between 60% - 70%.

B. 1) Feed flow to neither S/G will be acceptable.

2) Once-through-cooling must be initiated as the heat sink is lost.

C. 1) Feed flow to A S/G will be acceptable.

2) Feed flow to B S/G must be reduced to avoid the potential for significant S/G tube bundle damage.

D. 1) Feed flow to A S/G will be acceptable.

2) Feed flow to B S/G must be raised to raise level above -84% to avoid the need to initiate once-through-cooling.

Proposed Answer: C Explanation (Optional):

A S/G is considered dry at less than -125% and feedwater flow to a dry S/G could cause significant tube bundle damage. With S/G level less than -84%, feedwater flow should be limited to less than 300 gpm.

EOP Supplement 41 must be utilized to determine the expected FW flow compared to the differential pressure. In this case, with the feed regulating bypass valves 20% open and a 90 psid differential pressure, the expected feed flow is approximately 345 gpm.

A. Incorrect, the applicant believes the feed flow to both S/Gs is adequate, while feed flow to the B S/G must be reduced to less than 300 gpm in order to protect the tube bundles from potential damage. S/G level of 60-70% is normal operational level and feed to that desired level should only be to S/Gs with a level greater than -84%.

B. Incorrect, the applicant believes the once-through-cooling (OTC) criterion are met and the heat sink is lost. OTC must be initiated if any of the following conditions are met: 1)

Both S/Gs levels are below -84% and are not being restored or 2) Tc rises uncontrollably 5oF or greater. See explanation and choice A for explanation on whether feed flow is adequate.

C. Correct, see explanation and choice A.

D. Incorrect, feed flow to A S/G is acceptable but with B S/G less than -84% it is not desirable to raise feed flow, as the limit for this condition is 300 gpm.

Technical Reference(s): EOP-7.0 and basis, EOP Supplement 41 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: EOP Supplement 41 Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This question meets SRO-only criteria as the SRO must use procedures to determine procedurally required actions.

ES-401 Question 89 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __2__

Group # _____ __1__

K/A # 063.G2.1.7 Importance Rating _____ _4.7_

K/A Statement: Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.

Proposed Question:

Given the following Plant conditions:

PCS temperature is 420ºF

LTOP System is armed in LTOP MODE

Charging Pump P-55A is in operation

Letdown is in service Then, the following occurs:

125 VDC Panel ED-11-1 de-energizes due to a fault

EK-0702, RELIEF VALVE 2006 DISCH HI TEMP, annunciates

The NCO attempts to close CV-2003, CV-2004, and CV-2005, Letdown Orifice Stop Valves, in accordance with AOP-17, "Loss of 125V DC Panel(s)," but CV-2003 does NOT close To address these conditions, the CRS will . . .

A. Direct the crew to re-establish Charging and Letdown flow per SOP-2A, "Chemical and Volume Control."

B. Implement AOP-23, "Primary Coolant Leak."

C. Direct the crew to bypass the CVCS purification demineralizers per SOP-2B, "Chemical and Volume Control System - Purification and Chemical Injection."

D. Implement AOP-31, "Spurious Containment Isolation."

Proposed Answer: B Explanation (Optional):

A. Incorrect, re-establishing charging and letdown would not address the problem (a lifted relief valve causing a PCS leak).

B. Correct, a loss of D-11-1 causes CV-2009 (letdown containment isolation) to close; however, letdown is still flowing from PCS into the letdown line upstream of CV-2009.

Pressure upstream of CV-2009 will remain high with a failure of CV-2003 to close,

causing RV-2006 to lift due to the higher letdown pressure.

C. Incorrect, the applicant believes that excessive letdown flow (and higher temperature) is the result of the DC loss.

D. Incorrect, CV-2009 fails closed on a loss of ED-11-1. The reasoning it closed was not due to a spurious containment isolation signal, but rather a loss of DC control power.

AOP-31 will not correct the underlying cause of the failure of CV-2003 to close.

Technical Reference(s): AOP-23, AOP-17, ARP-4 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2010 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This meets SRO-only criteria 10CFR55.43 (b)5 as the applicant must assess plant conditions and then select a procedure to mitigate the event/transient.

Bank question was modified to accommodate current Palisades terminology and procedures.

ES-401 Question 90 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __2__

Group # _____ __1__

K/A # 103.G2.4.41 Importance Rating _____ __4.6_

K/A Statement: Knowledge of the emergency action level thresholds and classifications.

Proposed Question:

Given the following conditions:

A large break LOCA occurred 30 minutes ago.

Bus 1C is faulted and cannot be energized.

P-54A CS Pump tripped on overcurrent.

Containment pressure 60 psia and stable.

Containment Hydrogen concentration 5% and stable.

Containment Radiation is 17,000 R/hr as read on RIA-2321, Containment High Range Radiation Monitor.

Does a Loss or Potential Loss of the Containment Fission Product Barrier currently exist and if so why?

A. Yes, a Potential Loss of Containment condition exists due to conditions described in subcategory C.6 B. Yes, a Potential Loss of Containment condition exists due to conditions described in subcategory B.3 C. Yes, a Potential Loss of Containment condition exists due to conditions described in subcategory B.5 D. No, a Potential Loss of Containment does not currently exist Proposed Answer: C Explanation (Optional):

A. Incorrect, containment radiation is 17,000 R/hr on the containment high range monitors, whereas the potential loss of containment value is > 20,000 R/hr, as defined in subcategory C.6.

B. Incorrect, containment pressure is 60 psia (~45 psig), less than the potential loss of containment value of > 55 psig.

C. Correct, with a loss of Bus 1C, CS Pumps P-54B and P-54C are lost. Since P-54A tripped on overcurrent, one full train of containment cooling (TS 3.6.6 Bases) is not in operation as required by subcategory B.5.

D. Incorrect, See explanation in C above.

Technical Reference(s): SEP Supplement 1 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: SEP Supplement 1 Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2005 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This question meets SRO-only criteria as the applicant must assess plant conditions and then determine the Emergency Action Level threshold.

Modified stem to include an electrical fault and additional (and modified) initial conditions. Modified potential answers and changed correct answer.

ES-401 Question 91 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __2__

Group # _____ __2__

K/A # 015.A2.01 Importance Rating _____ _3.9_

K/A Statement: Ability to (a) predict the impacts on the following malfunctions or operations on the NIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Power supply loss of erratic operation Proposed Question:

Given the following conditions:

The Plant is in MODE 3 at normal operating pressure and temperature.

The PCS is being diluted in preparation for critical approach two shifts from now.

The Wide Range flux level plasma indication for Source/Wide Range NI-1/3 on Panel C-06 begins to act erratically and is determined to be unreliable.

The NI-1/3 analog count rates on Panel C-02 are indicating normal.

NI-1/3 indications on Panel C-02 channel check to NI-2/4 indications.

Given these conditions, which one of the following identifies if LCO 3.3.9, Neutron Flux Monitoring Channels, and LCO 3.3.7, Post Accident Monitoring Instrumentation, are satisfied and identifies the minimum ACTIONS, if any, that are required?

A. LCO 3.3.9 is met, LCO 3.3.7 is met. No ACTIONS are required.

B. LCO 3.3.9 is NOT met, LCO 3.3.7 is met. Immediately stop dilution of the PCS and perform a SDM verification within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

C. LCO 3.3.9 is met, LCO 3.3.7 is NOT met. Restore NI-1/3 plasma indication to OPERABLE status within 30 days.

D. LCO 3.3.9 is met, LCO 3.3.7 is NOT met. Restore NI-1/3 plasma indication to OPERABLE status within 7 days.

Proposed Answer: C Explanation (Optional):

A. Incorrect, the applicant believes that plasma indication is not a PAM instrument.

B. Incorrect, the applicant believes that a channel of Neutron Flux Monitoring is lost.

C. Correct, the plasma display is used to satisfy LCO 3.3.7, per SHO-1. Both channels must be within 1/2 decade of one another to be considered operable.

D. Incorrect, the applicant believes that LCO action 3.3.7.C applies, which is incorrect, as there is only one function affected (the Wide Range). The applicant believes the Wide Range and Source Range are both affected.

Technical Reference(s): LCO 3.3.7, LCO 3.3.1 Bases Table B 3.3.1-1, LCO 3.3.9, LCO 3.3.9 Bases, SHO-1 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: LCO 3.3.7, LCO 3.3.9 Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2009 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __2__

Comments:

This question meets the SRO-only criteria as the applicant must apply knowledge from Tech Spec bases information to the Tech Spec to determine the required actions to take.

ES-401 Question 92 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __2__

Group # _____ __2__

K/A # 034.K4.02 Importance Rating _____ _3.3_

K/A Statement: Knowledge of design feature(s) and/or interlock(s) which provide for the following: Fuel movement.

Proposed Question:

A new fuel bundle is being lowered into the Reactor Core during core reload, when suddenly the Refueling Machine Hoist movement stops. What interlock is responsible for the Refueling Machines response to the conditions?

A. Extreme hoist travel limit interlock B. Cable slack interlock C. Grapple open hoist interlock D. Hoist load bypass interlock Proposed Answer: B Explanation (Optional):

A. Incorrect, the extreme hoist travel limit interlock is an up limit only. This would not be reached while lowering a fuel bundle. Incorrect, the extreme hoist travel interlock is only to stop the hoist on rising motion if the upper limit does not stop the machine. This interlock will shut off the RFM with no indication as to why.

B. Correct, the Refueling Machine hoist is stopped during lowering on hoist underload or cable slack.

C. Incorrect, the hoist cannot be raised passed the UGOZ (Upper Grapple Operating Zone) with an open grapple.

D. Incorrect, the hoist load bypass is used when lowering an empty grabble.

Technical Reference(s): PL-IOTD Refueling Operations Lesson Plan (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 _____

55.43 __6__

Comments:

This question meets SRO-only criteria as fuel movements and operation of the Refueling Machine must be observed and approved by an SRO.

PL-IOTD Revision 11 Page 37 of 73 C.2.f. RFM Hoist operations.

1) Hoist load bypass a) Must use hoist load bypass when lowering empty grapple

2) Electronic load cell used for sensing:

a) Under load b) Over load c) Cable slack

3) Weighing zones:

a) Control Rod or fuel plus grapple b) Control Rod or fuel plus grapple plus hoist c) Exception: RFM computer adjusts for weight of hoist box when hoist begins to pick up fuel assembly and hoist when fuel assembly is in the hoist box.

4) Interlocks a) The computer stops hoist movement on:

(1) Under or overload. Hoist over load/under load setpoints are based upon recommendations by the vendor and include expected maximum and minimum fuel weights to the end of plant life to eliminate the need for an equipment modification and reduce the number of setpoint changes.

(2) There is an extreme hoist up limit that shuts the RFM and SFHM down.

(a) If you are raising the hoist and for some reason it does not stop at up limit, the extreme travel limit shuts the machine down with no other messages to tell you what happened. This has tripped up some Operators in the past.

(b) Commonly encountered fuel handling difficulties tells you how to deal with this.

b) Bridge or trolley energized stops hoist movement c) Up motion stops:

(1) hoist at up limit (green light)

(2) Hoist Overload (computer screen and digital display)

(3) Grapple open leaving UGOZ. (Orange light out for UGOZ and orange grapple open light on) d) Down motion stops:

(1) Hoist under load (computer screen)

(2) Cable slack (orange light)

5) Hoist not at UP limit and in either the core slow zone or the high speed zone.

a) In the core clear or high speed zone, may operate the bridge and trolley with the hoist not at up limit with an empty grapple b) May move B/T by pushing the BTI pushbutton. This is the way minor adjustments are made to the B/T position to get fuel into the applicable location occasionally and is addressed procedurally.

6) Fuel hoist overload and underload stops.

a) Point out that SFHM has only fuel assembly or control rod over/underloads. RFM has hoist box + fuel or CR over/underloads since the RFM computer has to take into account the added weight of the hoist box for over/underloads.

7) Core area hoist speed restrictions. Hoist automatically slows down in transition zones to avoid overload (underload) while accepting (shedding) weight of hoist box.

8) Grapple operate zone interlocks.

9) Cable slack Example: After dropping off a fuel bundle into the Core (opening the grapple) an attempt is made to raise the hoist. The machine shuts off at the upper grapple operate zone and will not rise further.

Determination: Grapple open hoist interlock is in effect. Therefore, closing the grapple will allow continued operation.

ES-401 Question 93 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __2__

Group # _____ __2__

K/A # 071.G2.1.32 Importance Rating _____ _3.8_

K/A Statement: Ability to explain and apply all system limits and precautions.

Proposed Question:

A liquid release is being prepared for release per SOP-17B, Dirty Radioactive Waste System.

Which of the following statements is correct if no Dilution Water pumps are operating at the start of the release?

A. The release cannot occur with no Dilution Water pumps operating.

B. CV-1054, Treated Waste Discharge valve, cannot open without the installation of a Temporary Modification.

C. RIA-1049, Liquid Radwaste Monitor, cannot exceed 20,000 cpm greater than background.

D. Only one Service Water pump shall be operating during the release.

Proposed Answer: B Explanation (Optional):

A. Incorrect, a release can occur, however it must be sampled, calculated, and lined up using independent verifications.

B. Correct, per the Note on Step 7.10.a.7 of SOP-17B, CV-1054 is interlocked with the Dilution Water pump breakers, such that the interlock must be defeated before it can be opened.

C. Incorrect, with no Dilution Water pumps operating, RIA-1049 cannot exceed 10,000 cpm greater than background. The applicant is confusing the maximum rad monitor setpoints with only one Dilution Water pump running (20,000 cpm greater than background).

D. Incorrect, an additional dilution requirement applies during such situation where there shall be at least one additional Service Water Pump either operating or available with the pump in Standby. The applicant is misunderstanding the additional dilution requirements established.

Technical Reference(s): SOP-17B (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This meets SRO-only criteria 10CFR55.43 (b)5 as the applicant must assess plant conditions and apply procedural requirements, including precautions and limitations, from a System Operating Procedure. The applicant must also know the basis behind the procedural action.

ES-401 Question 94 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __3__

Group # _____ _____

K/A # G2.1.4 Importance Rating _____ _3.8_

K/A Statement: Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.

Proposed Question:

Three SROs have completed proficiency watchstanding in order to maintain Active License status per 10CFR55.53(e). The following table depicts the position, date and hours stood for each watch by three SROs since the beginning of the year:

Assume today is May 16, 2016.

SRO Hours SRO Hours SRO Hours

1 Position stood #2 Position stood #3 Position stood 1/7/16 CRS 12 1/7/16 CRS 12 1/9/16 CRS 12 1/19/16 SM 12 1/14/16 CRS 12 2/9/16 CRS 12 2/8/16 SM 12 2/3/16 CRS 12 2/18/16 CRS 12 2/18/16 CRS 12 2/16/16 CRS 12 3/22/16 CRS 12 3/22/16 CRS 12 2/19/16 STA 12 3/28/16 STA 12 3/28/16 CRS 12 2/28/16 STA 8 4/9/16 CRS 12 4/7/16 STA 8 3/8/16 CRS 12 4/16/16 CRS 12 4/22/16 CRS 12 3/19/16 STA 8 5/1/16 CRS 12 5/13/16 SM 12 4/4/16 CRS 12 5/6/16 CRS 12 Which of the following statements is correct regarding the SROs active license status?

A. Only SRO #1 failed to maintain active license status B. Only SRO #2 failed to maintain active license status C. Only SRO #3 failed to maintain active license status D. ALL SROs successfully maintained active license status Proposed Answer: C Explanation (Optional):

A. Incorrect, SRO #1 has 6 full 12-hour watches within the first quarter at either the CRS or SM position.

B. Incorrect, SRO #2 has 5 full 12-hour watches within the first quarter at the CRS position.

C. Correct, SRO #3 only has 4 full 12-hour watches within the first quarter at the CRS

position.

D. Incorrect, SRO #3 does not meet the 10CFR55.53.(e) requirements. See Choice C for explanation.

Technical Reference(s): 10CFR55.53.(e), Palisades Admin 4.00 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __2___

Comments:

This question meets the criteria for an SRO-only question as the applicant must understand the License maintenance requirements specifically for the SRO job function.

ES-401 Question 95 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __3__

Group # _____ _____

K/A # G2.1.37 Importance Rating _____ _4.6_

K/A Statement: Knowledge of procedures, guidelines, or limitations associated with reactivity management.

Proposed Question:

The design bases event for Limiting Condition of Operation 3.1.1, Shutdown Margin, is:

A. Positive reactivity addition resulting from a Rod Ejection event at beginning of core life from 0% power conditions.

B. Positive reactivity addition resulting from a Rod Ejection event at end of core life from 100% power conditions.

C. Excessive cooldown resulting from a Main Steam Line Break at beginning of core life from 100% power conditions.

D. Excessive cooldown resulting from a Main Steam Line Break at end of core life from 0% power conditions Proposed Answer: D Explanation (Optional):

A. Incorrect, PDILs are based on Rod Ejection event.

B. Incorrect, PDILs are based on Rod Ejection event.

C. Incorrect, MTC less negative during beginning of life conditions. At Hot Full Power, there is less mass in the S/G for boil-off, as evident by a lower S/G pressure.

D. Correct, from Hot Zero Power, there is more mass in the SG as evident by a higher S/G pressure. End of cycle conditions have the most negative MTC. Therefore, this reactivity transient is the most severe Technical Reference(s): LCO 3.1.1 bases (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam St Lucie 2015 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 _____

55.43 __6___

Comments:

This question meets SRO-only criteria as the applicant is required to understand the Tech Spec bases as it applies to plant conditions.

ES-401 Question 96 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __3__

Group # _____ _____

K/A # G2.2.11 Importance Rating _____ _3.3_

K/A Statement: Knowledge of the process for controlling temporary design changes.

Proposed Question:

An emergency temporary modification can be implemented in the event of an imminent threat to the safety or reliability of the plant due to an unforeseen plant event:

When implementing an emergency temporary modification, which of the following are required?

1. The Shift Manager, with the concurrence of the Engineering Director, or designee, may direct the installation or removal of a Temporary Modification to the plant without approved controlling documentation, as long as the Temporary Modification does not adversely affect nuclear safety.

2. As soon as conditions permit, the Operations Manager and the Systems &

Components Manager, or their designee, shall be verbally notified of the modification and a Condition Report shall be initiated by Maintenance.

3. A Temporary Modification or a permanent Engineering Change shall be completed within 7 calendar days after installation.

4. The Responsible Engineer should coordinate with other Departments (i.e., the Systems & Components Engineer, Operations, Maintenance, Training, Planner and Installer) to ensure they are cognizant of the change and have provided appropriate input.

5. If the Temp Mod is a Comp action, a separate CR will be written by the Shift Manager to track the comp measure.

A. 1, 2 and 3 only B. 2, 3 and 4 only C. 2, 4 and 5 only D. 1, 3 and 5 only Proposed Answer: D Explanation (Optional):

EN-DC-136, Temporary Modifications, contains this requirement in section 5.3

5.3 EMERGENCY TEMPORARY MODIFICATION IMPLEMENTATION

[1] In the event of an imminent threat to the safety or reliability of the plant due to an unforeseen plant event:

(a) The Shift Manager, with the concurrence of the Engineering Director, or designee, may direct the installation or removal of a Temporary Modification to the plant on an emergency basis without approved controlling documentation, as long as the Temporary Modification does not adversely affect nuclear safety.

(b) As soon as conditions permit, the Operations Manager and the Systems & Components Manager or their designee shall be verbally notified of the emergency modification and a Condition Report shall be initiated by Engineering. The CR issued shall be used to track the installation of the Emergency Temporary Modification. Following installation, removal of the Emergency Temporary Modification shall follow the applicable steps of this procedure.

(c) IF the Temporary Modification is also a compensatory measure (operational), THEN the Shift Manager will ensure that a Condition Report is issued to track the compensatory measure. This is a separate CR from step 5.3.(1)(b).

(d) A Temporary Modification or a permanent Engineering Change shall be completed within 7 calendar days after installation.

A. Incorrect, #2 is wrong due to the CR is to be initiated by Engineering not Maintenance, and #5 is also required.

B. Incorrect, #2 is wrong due to the CR is to be initiated by Engineering not Maintenance, and #4 is incorrect C. Incorrect, #2 is wrong due to the CR is to be initiated by Engineering not Maintenance and #4 is Incorrect D. Correct, see explanation Technical Reference(s): EN-DC-136 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Grand Gulf 2015 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 _____

55.43 __3__

Comments:

This question meets SRO-only criteria as the SRO must have knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

ES-401 Question 97 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __3__

Group # _____ _____

K/A # G2.2.39 Importance Rating _____ _4.5_

K/A Statement: Knowledge of less than or equal to one hour technical specification action statements for systems.

Proposed Question:

Given the following conditions:

The Plant is in MODE 5 during a forced outage.

PCS temperature is at 180°F on Shutdown Cooling (SDC).

SDC is in-service using P-67A LPSI Pump at a flowrate of 3000 gpm.

Primary Coolant Pumps (PCPs) P-50A and P-50D are operating.

Both S/G levels are 65%.

P-8C, Auxiliary Feedwater Pump, is removed from service for maintenance.

Then, the following simultaneously occur:

2400VAC Bus 1C de-energizes due to a fault on the bus

P-50D, Primary Coolant Pump, trips Given these current conditions, which one of the following identifies whether the LCO for LCO 3.4.7, PCS Loops - MODE 5, Loops Filled, is currently satisfied and identifies the minimum action, if any, required to satisfy the LCO without reliance on any action statement?

A. LCO 3.4.7 is met; no action statements are required to be entered.

B. LCO 3.4.7 is NOT met; starting P-50B PCP will allow all action statements to be exited.

C. LCO 3.4.7 is NOT met; restoring P-8C and making it available for operation will allow all action statements to be exited.

D. LCO 3.4.7 is NOT met; starting P-50C PCP will allow all action statements to be exited.

Proposed Answer: C Explanation (Optional):

LCO 3.4.7 requires one SDC train operable and in operation with > 2810 gpm flow through the reactor core and either:

A. One additional SDC train shall be operable, or B. The secondary side water level of each S/G shall be > -84%.

A. Incorrect, the applicant does not recall the requirement to have AFW available in order to take credit for an operable heat sink.

B. Incorrect, the applicant does not recall the requirement to have AFW available in order to take credit for an operable heat sink and believes that restoring 2 PCPs will allow action statement to be exited.

C. Correct, additional requirements are necessary for a S/G to replace a standby SDC train, per requirement B in the explanation. A S/G can act as a heat sink via natural circulation if the S/G has the minimum water level specified in SR 3.4.7.2, the S/G is operable, the S/G has available method of feedwater addition and a controllable path for steam release, and the ability to pressurize and control pressure in the PCS. In this case, AFW Pump P-8C would satisfy that requirement.

D. Incorrect, the applicant does not recall the requirement to have AFW available in order to take credit for an operable heat sink and the belief that restoring 2 PCPs will allow action statement to be exited.

Technical Reference(s): LCO 3.4.7 and bases (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ____None_________

Learning Objective: _________________________ (As available)

Question Source: Bank # ___X___

Modified Bank # _______ (Note changes or attach parent)

New _______

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __2__

Comments:

This exam question meets the criteria for an SRO-only criteria as the applicant must apply specific knowledge from the Tech Spec Bases to given plant conditions and then determine if the LCO is met.

Question used from Palisades 2010 Audit Exam.

ES-401 Question 98 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __3__

Group # _____ _____

K/A # G2.3.6 Importance Rating _____ __3.8_

K/A Statement: Ability to approve release permits.

Proposed Question:

Which of the following personnel have the ability to approve a Containment Purge release permit in accordance with Procedure No CH 6.27, Containment Purge?

A. Shift Engineer and Heath Physics Technician B. Shift Manager and RETS Analyst C. Duty Station Manager and RETS Analyst D. Shift Engineer and Chemistry Technician Proposed Answer: B Explanation (Optional):

A. Incorrect, an HP Technician cannot be designated to approve the release per CH 6.27.

B. Correct, per CH 6.27, only the following individuals (or designees) can approve a Containment Purge release: Shift Manager and RETS Analyst, Chemistry Supervision, or designees.

C. Incorrect, the Duty Station Manager cannot be designated to approve the release per CH 6.27.

D. Incorrect, a Chemistry Technician cannot be designated to approve the release per CH 6.27.

Technical Reference(s): CH 6.27 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: _____None________

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41 _____

55.43 __4__

Comments:

This questions meets SRO-only criteria as the applicant must use procedure knowledge outside to perform a function that is only performed by an SRO licensed individual.

ES-401 Question 99 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __3__

Group # _____ _____

K/A # G2.4.16 Importance Rating _____ _4.4_

K/A Statement: Knowledge of EOP implementation hierarchy and coordination with other support procedures.

Proposed Question:

Given the following:

The Plant is at 100% power.

Emergency Diesel Generator 1-2 out of service.

Then, the following occurs:

The Plant is manually tripped due to indications of a Steam Generator Tube Rupture.

EOP-5.0, Steam Generator Tube Rupture Recovery, is in-progress.

A valid Safety Injection Actuation Signal is received.

Bus 1D de-energizes due to a ground/overcurrent fault.

Pressurizer level is currently 33% and slowly rising.

Which one of the following procedures provides the most expeditious method of restoring available Pressurizer Heater capability for the above plant conditions?

A. AOP-9, Loss of Bus 1D.

B. AOP-10, Loss of Bus 1E.

C. EOP Supplement 29, Restore Busses 1C, 1D, 1E Power From Offsite Source.

D. SOP-30, Station Power.

Proposed Answer: D Explanation (Optional):

A. Incorrect, AOP-9 contains directions for transfer heater power from Bus 1E to Bus 1C but this could take up to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (AOP-9 refers the operator to SOP-30 when restoring pressurizer heaters).

B. Incorrect, AOP-10 Attachment 1 Emergency Feed of PZR Heater Transformer #15 from Bus 1C is directed per EOP-5.0 Step 42 since Bus 1D and 1E are not energized, but this will require an extended amount of time compared to SOP-30. (Bus 1E undergoes a load shed upon a Safety Injection Signal to ensure adequate voltage levels on the ESF buses when the D/Gs are called on to supply power.)

C. Incorrect, EOP Supplement 29 directs starting DG 1-3 but this DG can only supply 1C or 1D bus.

D. Correct, SOP-30 contains a section for restoring Bus 1E following an SIAS and EOP-5.0

directs the use of this procedure, this action does not account for Bus 1D being de-energized. This is the purpose of the RNO of this step.

Technical Reference(s): EOP-5.0, AOP-9, AOP-10, SOP-30, EOP Supplement 29 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # ___X___ (Note changes or attach parent)

New _______

Question History: Last NRC Exam Palisades 2009 (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __5__

Comments:

This meets SRO-only criteria as the applicant must assess plant conditions and apply procedural requirements/steps outside of immediate actions and/or entry conditions.

Question modified from Palisades 2009 NRC Exam. Modified question to accommodate current Palisades terminology. Changed correct answer and one distractor.

ES-401 Question 100 Form ES-401-5 Examination Outline Cross-

Reference:

Level RO SRO Tier # _____ __3__

Group # _____ _____

K/A # G2.4.29 Importance Rating _____ _4.4_

K/A Statement: Knowledge of the emergency plan.

Proposed Question:

Complete the following statements regarding the Emergency Plan:

An Assembly is the process of gathering personnel in designated areas __(1)__ following the declaration of an __(2)__ or higher.

A. (1) Outside the protected area (2) Alert B. (1) Inside the protected area (2) Alert C. (1) Outside the protected area (2) Unusual Event D. (1) Inside the protected area (2) Unusual Event Proposed Answer: A Explanation (Optional):

A. Correct, an Assembly is performed at an Alert or higher and assembles personnel outside of the PA.

B. Incorrect, part 1 is incorrect, Accountability is performed inside the PA.

C. Incorrect, part 2 is incorrect, see Choice A.

D. Incorrect, both parts are incorrect, see Choice A Technical Reference(s): EI-12.1 (Attach if not previously provided, _______________________________________________

including version/revision number) _______________________________________________

Proposed references to be provided to applicants during examination: ______None_______

Learning Objective: _________________________ (As available)

Question Source: Bank # _______

Modified Bank # _______ (Note changes or attach parent)

New ___X___

Question History: Last NRC Exam (Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content: 55.41 _____

55.43 __1__

Comments:

This question meets SRO-only criteria as it is an SRO-only duty to direct the performance of an Assembly when required.

ML22167A013

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