Text

2020 Palisades Ile Proposed Written Exam with Answer Key - Redacted
	ML20267A359
Person / Time
Site:	Palisades
Issue date:	07/06/2020
From:	; Entergy Nuclear Operations
To:	; NRC/RGN-III/DRS/OLB
	Bergeon B
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ML19213A184	List: ML20267A310; ML20267A321; ML20267A332; ML20267A334; ML20267A343; ML20267A344; ML20267A349; ML20267A352; ML20267A356; ML20267A359; PNP 2020-003, 2020 Palisades Ile Proposed Exam Submittal Letter;
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EXAMINATION ANSWER KEY Daft N 1 1 ID: 852830 Points: 1.00 After a trip from 100% power, EOP-1.0, Standard Post Trip Actions, is complete.

The crew has entered EOP-8.0, Loss of Offsite Power / Forced Circulation Recovery.

The Safety Function Status Checks (SFSCs) are being performed and the following are the current conditions:

NO PCPs are running

PZR Level is 50% with Charging and Letdown in service

PZR Pressure is 1620 psia

ALL RVLMS lights are green

AFW is maintaining BOTH S/Gs at 65% level

Tcolds are 540°F

Thots are 560°F

Tave indicates 550°F

Average of Qualified CETs (KCETA) is 584°F

ASDVs are throttled and lowering PCS temperatures Given these conditions, complete the following statements:

The Safety Function Status Check for:

PCS Heat Removal ___(1)___ satisfied.

Core Heat Removal ___(2)___ satisfied.

(1) (2)

A. is is B. is is NOT C. is NOT is D. is NOT is NOT Answer: B Answer Explanation Answer: B B. is, is NOT - is correct.

Explanation:

Per EOP-8.0 Attachment 1, page 7 and 8 of 14:

To Satisfy PCS Heat Removal, At least one S/G with level between 60 and 70% with FW available and Tave or Tc is stable or lowering are all met.

PLP - 2018 NRC EXAM Page: 1 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 To Satisfy Core Heat Removal, PCS Loop Delta T is less than 50°F with natural circulation (met), AND PCS Subcooling is greater than or equal to 25°F based on the average of Qualified CETs (NOT met).

Subcooling calculated by:

Tsat for 1620 psia is 606.17°F Ave QCET is 584°F Difference is 22°F subcooling.

Distractors:

A. Part one is correct. Part two is plausible for the student that doesn't know the Core Heat Removal requirements but is incorrect because they are NOT met as explained.

C. Parts one and two are plausible for the student that doesn't know either requirement but are incorrect because neither is met as explained.

D. Part one is plausible for the student that doesn't know the PCS Heat removal requirement but is incorrect as explained. Part two is correct.

References:

EOP-8.0, Loss of Offsite Power / Forced Circulation Recovery, Revision 19, Attachment 1, pages 7 and 8, Steps 5 and 6.

EOP-1.0 SPTA, Section 5.0 Steps 6 and 7 (pages 14 and 15 of 23)

Note: Meeting the provisions of ONE of the Conditions will satisfy the Safety Function:

5. Core Heat Removal - Condition 1:

a. PCS Loop delta T is less than 50°F with natural circulation.

b. PCS subcooling is greater than or equal to 25°F based on the average of qualified CETs.

OR Condition 2:

a. PCS loop delta T is less than 10°F with forced circulation.

b. PCS subcooling is greater than or equal to 50°F based on the average of qualified CETs.

6. PCS Heat Removal

a. At least one S/G has (1) level being maintained between 60% and 70% with feedwater available OR (2) Level greater than -84% and restoring to between 60% and 70%

b. PCS Tave or Tc is stable or lowering KA:

007 Reactor Trip / stabilization / Recovery 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

(CFR: 41.7 / 43.5 / 45.12)

IMPORTANCE RO 4.0 SRO 4.6 HIGH COG - Comprehension of what parameters and values are assessed is necessary to determine the status of a safety function from the data given in the stem.

Test item meets KA. To answer the question, requires knowledge of what parameters and logic are assessed to determine status of the PCS Heat Removal and Core Heat Removal safety functions after a reactor trip.

Objective:

PLP - 2018 NRC EXAM Page: 2 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Given post reactor trip conditions, analyze a given parameter trend and determine if the trend is responding as expected in accordance with EOP-1.0 Basis.

Question 1 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 852830 User-Defined ID: 007G2.4.21 Cross Reference Number: TBAB_E01.01 After a trip from 100% power, EOP-1.0, Standard Post Trip Topic:

Actions, is complete. The crew has enter Num Field 1:

Num Field 2:

Text Field:

Comments: different version of 847344 New PLP - 2018 NRC EXAM Page: 3 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 4 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 2 ID: 847350 Points: 1.00 (Reference Provided)

PCS Heatup is in progress in accordance with SOP-1C, "Primary Coolant System - Heatup".

A bubble has been formed in the Pressurizer.

At time 1000:00,

PCS is at 400°F

Pressurizer Pressure is at 450 psia

3 PCPs are operating within the required parameters of the procedures

The LTOP Switches are in 'ENABLE'

BOTH PZR PORV Block valves are OPEN At time 1005:00,

PT-0104B, Pressurizer Pressure input to LTOP Controller PY-0105B fails to 1500 psia Complete the following:

At 1005:00, the number of OPEN Pressurizer PORVs is ___(1)___.

At 1006:00, if TS-0115, PCS Temperature input to LTOP Controller PY-0105A fails to 450°F, then

___(2)___ PZR PORV can provide automatic programmed LTOP protection throughout the heatup.

(1) (2)

A. ONLY ONE ONLY ONE B. ONLY ONE NEITHER C. NONE NEITHER D. NONE ONLY ONE Answer: C Answer Explanation Answer: C C. NONE, NEITHER; is correct.

Provide EOP Supplement 1 Rev 6 Pressure and Temperature Limit curves page 1 (upper portion)

Explanation:

The VLTOP Setpoint at 400°F is 1600 psia per EOP Supplement 1 (TS limit of 2100 psia per TS page 3.4.12-1). The failure of this train's pressure input to 1500 psia is below the automatic opening setpoint for PORV 1043. Thus, part one is correct, NO PZR PORV is open due to the instrumentation problem, and no vapor space LOCA is occurring. As pressure rises throughout the heatup, this PORV can NOT respond.

PLP - 2018 NRC EXAM Page: 5 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 When the TS-0115 (which is on the other train) fails to 450°F, pressure would have to rise to 2200 psia before the PORV would open. That would be okay once temperature got above 430°F, but it would not provide the proper automatic over-pressure protection from 400 to 430°F. Thus part two is correct.

Logic / coincidence:

1 temperature switch and 1 pressure transmitter for each PORV.

Reference:

EOP Supplement 1 Rev 6 Pressure and Temperature Limit curves page 1 (upper portion)

M-201 sht 2.

Distractors:

A. Part one and Part two are plausible for the examinee that can't figure out the setpoint from the provided reference, and/ or doesn't know the logic. Both parts wrong.

B. Part one is plausible for the examinee that can't figure out the setpoint from the provided reference, or doesn't know the logic. Part one is wrong, part two is correct.

D. Part one is correct as explained. Part two is plausible for the examinee that can't figure out the setpoint from the provided reference, or doesn't know the logic. Part two is wrong.

KA:

008 Vapor space LOCA AA2. Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident:

(CFR: 43.5 / 45.13)

AAA2.06 PORV logic control under low-pressure conditions . . . . . . . . . . . . . . . . . . . . 3.3 3.6 Test item meets KA. Question poses a low pressure situation where the logic and coincidence inputs to the PZR PORVs in LTOP mode must be known to answer it correctly (see justifications for the distractors-failure to know the logic results in getting it wrong.) An inadvertent opening of a PORV would be a Vapor Space LOCA.

HIGH COG - comprehension of the logic / train inputs to each PZR PORV, and knowledge of how the required protection changes at 430°F is required.

Objective:

From memory, describe the design features and interlocks that provide the following Pressurizer Pressure Control system functions:

a. Automatic protective de-energization of Pressurizer heaters (010 K4.02)

b. PCS pressure control (manual and auto)

c. PCS overpressure protection (010 K4.03) in accordance with DBD-2.11 and FSAR 4.3.7.

PLP - 2018 NRC EXAM Page: 6 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 2 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 847350 User-Defined ID: 008AA2.06 Cross Reference Number: PPCS_CK09.0-17 (Reference Provided) PCS Heatup is in progress in accordance Topic:

with SOP-1C, "Primary Coolant System Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 7 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 8 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 3 ID: 847669 Points: 1.00 Given the following:

At time 1000:00

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

2400V Bus 1C is de-energized (faulted)

BOTH Main Feedwater Pumps have tripped Given these conditions, Auxiliary Feedwater flow ___(1)___ automatically initiate.

At time 1005:00

Highest S/G pressure is 900 psia

The actions of EOP-1.0, Standard Post Trip Actions, have been completed

The CRS has given the direction to commence a PCS cooldown and depressurization When AFW flow is established, AFW flow will be from ___(2)___ .

(1) (2)

A. will P-8A B. will P-8C C. will NOT P-8A D. will NOT P-8C Answer: B Answer Explanation Answer: B B. will, P-8C; is correct.

Explanation /

References:

With no MFW pumps at the time of the plant trip from 100% power, the SG levels will lower to the AFAS setpoint (30%) in a short time due to decay heat boiling away the SG inventory through the TBV / ADVs.

The AFW pumps have a specific start sequence:

If there is power, P-8A starts first. If there is no flow from P-8A to BOTH S/Gs, then after 30.5 seconds, P-8C starts. If there is no flow from P-8C, after 112.5 seconds, P-8B starts.

P-8C is a converted HPSI pump. It will deliver the minimum required flow for decay heat removal at less than or equal to 900 psia with all 4 PCPs in operation, or at 1000 psia or less with all PCPs secured.

Because of the faulted 2400V bus 1C, AFW pump P-8A will not start on the AFAS. However, P-8C will start because it has power from 2400V bus 1D, and the AFAS exists, and P-8A has produced no flow.

Dwg E-1 shows the power supplies to the AFW pumps.

Dwg E-17 sht 21 show the logic to start P-8A and P-8C PLP - 2018 NRC EXAM Page: 9 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Distractors:

At Palisades, the ESS buses are 'C' and 'D'. In addition to 'A' pumps, there are 'B' and 'C' pumps that are on the 'C' Bus; and there are 'A', 'B' and 'C' pumps on the 'D' Bus.

A. will, P-8A is plausible because Part one is correct as explained above, and Part two is plausible for those who don't know the power source for the AFW pumps.

C. will NOT, P-8A is plausible but incorrect. Part one is plausible for those who believe the AFAS won't occur, either due to level not lowering enough, or the power source failure. Part two is plausible as explained in A. Part two does not rule out manual actuation.

D. will NOT, P-8C is plausible but incorrect. Part one is plausible as explained in C. Part two is correct as explained above. Part two does NOT rule out manual actuation.

KA:

009 SB LOCA EA1 Ability to operate and monitor the following as they apply to a small break LOCA:

(CFR 41.7 / 45.5 / 45.6)

EA1.11 AFW/MFW . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 4.1 Test item meets KA as the question poses a small break LOCA and requires the examinee to determine if the AFW pumps have to be manually started (Operate) or will auto start (monitor) and then also determine which AFW pump is providing flow (monitor).

HIGH COG - There are multiple mental steps involved with the loss of power to a bus, and the development of an AFAS signal, and how it affects the AFW pumps, and is pressure low enough to allow AFW flow.

Objective: From memory, for automatic actions associated with the Auxiliary Feedwater System:

- AFAS Actuation (including C-187 sensors) (K/A 061 K4.0.2)

- Pump Auto Start Sequence (K/A 061 K4.0.6)

- Pump Trip (K/A 061 K4.0.7)

a. state the parameter and value (setpoint) at which each automatic action occurs

b. explain the purpose of each automatic actuation in accordance with E-17 Sheets 21, 21A, 22. AFW_CK10.0 PLP - 2018 NRC EXAM Page: 10 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 3 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 847669 User-Defined ID: 009EA1.11 Cross Reference Number: AFW_CK10.0 Given the following: At time 1000:00 The Plant has been Topic:

tripped from 100% power due to a small br Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 11 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 12 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 4 ID: 848570 Points: 1.00 The Plant is at 100% power, with ALL systems normally aligned.

A seismic event occurs, and the operators note that PCS pressure rapidly lowers to containment pressure.

While performing EOP-1.0, Standard Post Trip Actions,

The operators will stop ___(1)___ PCPs.

The reason for the above action is to ___(2)___.

(1) (2)

A. ONLY two prevent unwanted additional loss of mass from PCS B. ONLY two provide protection against cavitation C. ALL four provide protection against cavitation D. ALL four prevent unwanted additional loss of mass from PCS Answer: C Answer Explanation Answer: C C. ALL four, provide protection against cavitation; is correct.

Explanation:

The situation is a LARGE break LOCA has occurred, as implied by the rapid drop in PCS pressure to containment pressure. SIAS occurs, and the SITs inject. There is nothing the operators can do to prevent the loss of any PCS mass on so large a break. Thus, stopping PCPs for that reason is N/A. Per the reference (EOP-1.0 Basis Document), the reason for tripping the PCPs is to protect the PCPs from operating with insufficient NPSH, which is described by EOP Supplement 1 curves. Per the EOP Supplement 1 Basis, the minimum pressure for PCP operation curve provides protection against cavitation for the PCPs.

Distractors:

Those distractors with 'ONLY two' are plausible for the examinee that applies the 'trip two/leave two' strategy applicable to smaller break LOCAs. This is incorrect because this is a large break.

Those distractors with 'prevent unwanted additional loss of mass' are plausible for the examinee that applies the 'trip two/ leave two' strategy applicable to smaller break LOCAs.

Tripping ALL four PCPs to prevent unwanted additional loss of PCS mass is plausible, because the loss of pumping pressure head would reduce the additional loss of mass, at the expense of forced circulation, but that is NOT why all four PCPs are tripped for LARGE break LOCAs as stated in the explanation.

(Can't prevent the total / complete blowdown of the PCS mass to the containment)

Reference:

EOP-1.0 SPTA Basis, and EOP-1.0 Step 5 RNO 5.3, and 5.4.

PLP - 2018 NRC EXAM Page: 13 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 KA: 011 Large Break LOCA EK3 Knowledge of the reasons for the following responses as the apply to the Large Break LOCA:

(CFR 41.5 / 41.10 / 45.6 / 45.13)

EK3.14 RCP tripping requirement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 4.2 Test item meets KA. Question asks how many PCPs are tripped upon diagnosing a LBLOCA and asks for the reason all are tripped.

HIGH COG - must diagnose a LBLOCA vice a small break LOCA from the conditions given in the stem; PCP tripping requirements are different.

Objective: Given conditions involving a reactor trip, determine any required EOP 1.0 right-hand contingency action(s) in accordance with EOP-1.0.

Question 4 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 848570 User-Defined ID: 011EK3.14 Cross Reference Number: TBAB_E01.03 The Plant is at 100% power, with ALL systems normally Topic:

aligned. A seismic event occurs, and the ope Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 14 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 15 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 5 ID: 847690 Points: 1.00 Note to Examinee: There is a color image provided as a handout for this question The plant is in MODE 3 at normal operating temperature and pressure

PCS Flow on Panel C-12 is shown below:

Then, EK-0902, PRI COOLANT PUMP P-50B TRIP annunciates.

Given these conditions, complete the following statements:

PCS Flow indication on ___(1)___ Flow Indicator(s) lower(s).

The automatic start of a Lift Oil Pump is indicated by the illumination of a(n) ___(2)___ light on Panel C-02.

(1) (2)

A. ONLY one Amber B. ONLY one Red C. ALL four Amber D. ALL four Red Answer: D Answer Explanation Answer: D.

D. ALL four, Red is correct. Each PCP has 25% input into each channel of flow indication on Panel C-12, therefore flow on each indicator will lower by 25%. Thus, part one is correct.

When properly aligned for auto start (as indicated by an illuminated amber 'AUTO' light), upon a trip of a PCP, the AC Lift Oil pump starts first. If after 5 seconds, oil pressure is < 1700 psi, then the DC lift oil pump starts. When either lift oil pump is running after an auto start, both the amber 'auto' light and the red running light are lit.

DISTRACTOR ANALYSIS A. Part one is plausible for the examinee that doesn't understand how the flow instruments are connected and interrelated. Part one is incorrect, as the student believes only Loop 1 FIs will lower. Part two is plausible for the examinee that thinks the 'AUTO' light illuminating occurs when the oil lift pump auto starts. Part two is incorrect as explained in D.

B. Part one is plausible but incorrect as described in A. Part two is correct as explained in D.

PLP - 2018 NRC EXAM Page: 16 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 C. Part one is correct as explained in D. Part two is plausible but incorrect as explained in A.

KA:

015 RCP Malfunctions AK2. Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following:

(CFR 41.7 / 45.7)

AK2.10 RCP indicators and controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.8* 2.8 Test item meets KA. Answer requires knowledge of what happens to the PCS Flow indications and the control lights for lift oil pumps when a PCP inadvertently trips causing a loss of flow.

Reference:

PL-PCP rev 6, page 24-25. and 27. and power point slide 15, 24, 28 PL-PCS rev 6 pg 29-30.

HIGH COG - prediction of response of flow indication, and control lights for a trip of one PCP.

Objective:

From memory, for the automatic actions associated with the PCPs

a. state the parameter and value (setpoint) at which each automatic action occurs

b. explain the purpose of each automatic actuation

c. explain the setpoints for automatic actuations associated with the Oil Lift Pump and Backstop Oil Pumps in accordance with E-184, E-185 and E-186. (003 K4.03, 003 K4.05) PCP_CK10.0 PLP - 2018 NRC EXAM Page: 17 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 5 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847690 User-Defined ID: 015AK2.10 Cross Reference Number: PCP_CK10.0 Note to Examinee: There is a color image provided as a Topic:

handout for this question The plant is in Num Field 1:

Num Field 2:

Text Field:

Comments: Modified from PAL-LOI-6706 - 2013 NRC, to be a two part question:

Given the following initial indications for Primary Coolant Flow on Panel C-12:

Which one of the following would validate alarm EK-0902, PRI COOLANT PUMP P-50B TRIP? Assume all four PCPs are initially in service.

A. FI-0102A FI-0102B FI-0102C FI-0102D lowers lowers lowers lowers B. FI-0102A FI-0102B FI-0102C FI-0102D lowers lowers unchanged unchanged C. FI-0102A FI-0102B FI-0102C FI-0102D lowers lowers rises rises D. FI-0102A FI-0102B FI-0102C FI-0102D unchanged lowers unchanged unchanged Answer: A PLP - 2018 NRC EXAM Page: 18 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 19 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 20 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 6 ID: 853530 Points: 1.00 The plant is at 100% power, with all systems in automatic control

Single Charging is aligned in accordance with SOP-2A, Chemical and Volume Control System P-55A, Charging Pump, trips, and NO other pump starts.

Given these conditions and with NO operator action, complete the following statements:

Letdown isolates ___(1)___ EK-0763, Pressurizer Level CH A LO-LO, or EK-0764, Pressurizer Level CH B LO-LO annunciates.

In 15 minutes, Pressurizer Level ___(2)___.

(1) (2)

A. before stabilizes B. before continues to lower C. after stabilizes D. after continues to lower Answer: B Answer Explanation Answer: B.

B. before, continues to lower; is correct Reference/ Explanation:

The loss of charging flow, which is cooling the letdown flow through the Regen Hx, causes the temperature of the PCS being letdown to rise above the high temperature auto closure setpoint of CV-2001, Letdown Stop Valve in a very short time (< 1 minute). (

Reference:

ARP-4, EK-0701, Regen HT EX Tube Outlet Hi Temp.

The ARP for the setpoint of the LO-LO Level alarms states the setpoint as 36%.

SOP-1A Attachment 10 PZR Level Program.

PL-PLCS rev 5 page 16:

Pressurizer level change of 1% requires 66 gallons of makeup water per AOP-23 (Attachment 1 page 2 of 4.)

PL-PLCS rev 5 page 16:

'...and closing the #1 Letdown Orifice Stop on a Pressurizer level <36%.'

At 100% power, PZR level is 57%.

Letdown isolates at 36%.

PZR Level must lower by 21% to auto isolate letdown by closing CV-2003.

There are 66 gallons of water per 1% PZR level.

66 x 21 = 1386 gallons must be removed from the PZR and not replaced.

One letdown orifice removes 40 gpm PLP - 2018 NRC EXAM Page: 21 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PCP Bleed off is an additional 4 gpm loss.

Total loss per minute is 44 gallons.

1386 / 44 = 31.5 minutes (31 minutes 30 seconds).

This is a lot longer than it takes to reach the Hi Temp closure of CV-2001. Thus, part one is correct.

When Letdown Isolates, PCP Bleedoff continues at a total of 4 gpm. This drains mass from the PCS and PZR Level continues to lower.

Distractors:

A. before, stabilizes is plausible but incorrect. Part one is correct as explained in the answer. Part two is plausible for the examinee that forgets about PCP bleedoff.

C. After, stabilizes is plausible but incorrect. Part one is plausible for the examinee that overlooks the high temp closure and recalls the isolation at 36%. Part one is incorrect as explained in the answer. Part two is plausible but incorrect as explained in A.

D. After, continues to lower is plausible but incorrect. Part one is plausible but incorrect as explained in C. Part two is correct as explained in the answer.

KA:

APE 022 Loss of Reactor Coolant Makeup AK1. Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: l (CFR 41.8 / 41.10 / 45.3)

AK1.03 Relationship between charging flow and PZR level . . . . . . . . . . . . . . . . . 3.0 3.4 Test item meets KA as answer requires knowledge of the operational implications of losing charging to the PCS (loss of makeup), which results in the isolation of letdown due to high temperature first, then a secondary reason, and the impact on PZR level after letdown isolates.

HIGH COG - the knowledge of the operational implication of the relationship between charging flow and PZR Level is tested by comprehending the impact of a loss of charging on the cooling of the letdown flow, and what happens to PZR level due to PCP Bleedoff flows, and PZR level and letdown isolation setpoints.

Objective:

From memory, and given a loss or malfunction of the Pressurizer Level Control System, including loss of inputs, describe the effects on the following:

a. CVCS (011 K3.01)

b. PCS (011 K3.02)

c. Pressurizer Pressure Control System (011 K3.03)

d. Pressurizer Level Control System in accordance with DBD-1.04 and FSAR Chapter 4.

PLP - 2018 NRC EXAM Page: 22 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 6 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 5 Difficulty: 3.00 System ID: 853530 User-Defined ID: 022AK1.03 Cross Reference Number: PLCS_CK11.0 The plant is at 100% power, with all systems in automatic Topic:

control Single Charging is aligned in ac Num Field 1:

Num Field 2:

Text Field:

Comments: Modified from 843639 - LOI20 Audit question number 44, which asked a similar question from a power level of 50% - which has a different initial PZR level and results in a different answer.

The plant is at 50% power, with all systems in automatic control

CV-2003, Letdown Orifice Isolation valve, is OPEN in AUTO

P-55A, Charging Pump 'A', is running in AUTO A loss of the operating charging pump occurs, and NO other pump starts.

With NO operator action, how long until letdown isolates?

A. less than 20 minutes B. 20-21 minutes C. 22-23 minutes D. more than 23 minutes Answer: B PLP - 2018 NRC EXAM Page: 23 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 24 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 7 ID: 853534 Points: 1.00 The plant has just been shutdown from a 100 day run at full power.

Component Cooling Water Temperature is 90°F

E-60A, Shutdown Cooling (SDC) Heat Exchanger is unavailable due to the shell side (CCW) being isolated

E-54A/B, CCW Heat Exchanger Differential Pressures are 7 psid Given these conditions, complete the following statements:

The MAXIMUM number of CCW pumps that can be operating is ___(1)___.

If SDC were in service and a loss of EY-01 occurs with NO operator action for 5 minutes, then PCS temperature would ___(2)___.

(1) (2)

A. ONE lower B. ONE rise C. TWO rise D. TWO lower Answer: B Answer Explanation Answer: B.

B. ONE, rise; is correct.

Explanation

References:

Per SOP-3 and SOP-16 (and DBD-2.01 Section 3.3.6), there are general plant requirements that limit the amount of CCW flow through the SDC heat exchangers. The CCW flows through the shell side of the CCW heat exchangers and the SDC heat exchangers. When only one SDC HX is available, the limit on operating CCW pumps is one per SOP-3 section 4.1.4. When only one CCW pump is operating, SOP-16 Section 4.3.2 describes the optimum CC Hx DP at 6-8 psid. Thus, part one is correct.

A loss of EY-01 causes a loss of power to the controllers (HIC-3025A and HIC-3025B) for the SDC Hx common outlet (CV-3025), and the power to the controllers (FIC-0306 and HIC-0306) for the common SDC HX Bypass valve (CV-3006). The HX outlet fails closed, and the Bypass fails open. NO PCS flow is being cooled, and PCS temperature rises. Thus, part two is correct.

Distractors are all plausible for the examinee that doesn't know the impact of the failure on the operation of the controllers of the valves that allow flow through and / or around the SDC Heat exchangers and the limits on CCW with only one SDC HX available. The distractors with 'TWO' in part one are plausible for the examinee that doesn't realize the impact on the flowpath through the SDC heat Exchangers. The distractors with 'lower' in part two are plausible for the examinee that gets the fail positions of the valves that lost power to their controllers reversed.

PLP - 2018 NRC EXAM Page: 25 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 KA: 025 Loss of RHRS AK2. Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following:

(CFR 41.7 / 45.7)

AK2.01 RHR heat exchangers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.9 2.9 Test item meets KA. At Palisades the SDC system is the RHR system. Question poses a loss of the RHR system / SDC HX (via an unavailable SDC HX, then also a loss of power), and requires knowledge of the interrelation between PCS, CCW, and SDC to determine the impact on the PCS.

HIGH COG - need to predict / draw conclusion based on knowledge of the fail positions of system valves, the effect on flows through and around the heat exchangers, and how the PCS temperature responds.

Objective: From memory, describe the effects a loss or malfunction of the Component Cooling Water system has on Shutdown Cooling operations in accordance with the FSAR. (005 A2.01)

Question 7 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 853534 User-Defined ID: 025AK2.01A Cross Reference Number: SDC_CK08.0 The plant has just been shutdown from a 100 day run at full Topic:

power. Component Cooling Water Temper Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 26 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 27 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 8 ID: 847954 Points: 1.00 The plant is in MODE 1,

A Loss of Component Cooling Water has occurred

The appropriate AOP has been entered The following PCP Temperatures are ALL noted to be rising at a rate of 1°F per minute:

P-50A Upper Guide Bearing 170°F P-50B Upper Thrust Bearing 171°F P-50C Lower Seal 172°F P-50D Vapor Seal 173°F Which PCP temperature will require an immediate reactor trip FIRST?

A. P-50A B. P-50B C. P-50C D. P-50D Answer: A Answer Explanation Answer: A.

A. P-50A, is CORRECT.

Explanation:

The temperature limit for tripping P-50A on high Upper Guide Bearing is 175°F - 5 minutes away.

The other parameters require at least 10 minutes to reach their setpoints requiring a reactor trip.

Distractors are all plausible for examinees that don't know the operating limits and limiting temperatures associated with PCPs during a loss of CCW per AOP-29, PCP Abnormal Conditions.

The temperature limit for tripping P-50B on Upper Thrust bearing is 185°F, which occurs in 14 minutes.

The temperature limit for tripping P-50C on Lower Seal is 185°F, which is 13 minutes away.

A trip of the Reactor and PCP is required when during a loss of CCW, any vapor seal temperature exceeds 185°F. That will occur in 12 minutes.

Reference:

AOP-29 Reactor and Equipment Trip Criteria, page 5 of 29 KA:

026 Loss of CCW 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

(CFR: 41.10 / 43.5 / 45.12)

IMPORTANCE RO 4.2 SRO 4.2 PLP - 2018 NRC EXAM Page: 28 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Test item meets KA. The question requires an assessment of temperature trends in a timely manner during a loss of CCW, that will lead to a trip of the reactor and PCPs. The control room reference material is the trip criteria that operators have committed to memory.

HIGH COG - recall trip setpoints for each monitored value, meaningful and operationally oriented calculation of time to reach trip setpoint, compare times to determine quickest.

Objective:

Given Abnormal Operating plant conditions, determine if an immediate, manual Reactor Trip is required without error.

Question 8 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847954 User-Defined ID: 026G2.4.47 Cross Reference Number: IOTF_CK05.0 The plant is in MODE 1, A Loss of Component Cooling Water Topic:

has occurred The appropriate AOP has bee Num Field 1:

Num Field 2:

Text Field:

Comments: Modified from 727692 (PAL-LOI-498) which required memory of the limit for any bearing temperature, and did not involve trending any values of seal or bearing temperatures. The answers of these two questions are different.

Which one of the following temperatures would require a reactor trip during a loss of CCW?

A. Highest PCP Lower Seal temperature is 183ºF.

B. Highest Controlled Bleedoff temperature is 176ºF.

C. Highest PCP Bearing temperature is 186ºF.

D. One CRDM Leakoff temperature is 202ºF all others are <

200ºF Answer: C PLP - 2018 NRC EXAM Page: 29 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 30 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 9 ID: 848571 Points: 1.00 The Plant is at 100% power.

Which of the following would cause the Primary Coolant System to get closer to saturated conditions?

A. In service Pressurizer Pressure Controller output fails low B. Pressurizer Spray valve CV-1057 loses Instrument Air C. In service Pressurizer Pressure Controller output fails high D. Steam Generator ADV inadvertently opens Answer: C Answer Explanation Answer: C C. In service Pressurizer Pressure Controller output fails high; is correct Explanation:

When the actual pressure rises, the output of this controller goes up, which will open the spray valves and lower pressure back to the setpoint. A failure of the output in the high direction causes the sprays to open, lowering pressure. Because the output is failed, there is no way for actual pressure which is lowering, to override the output signal and the sprays stay open. Even though the PZR is saturated and some flashing to maintain PCS pressure would occur, the sprays would continue to remain open, lowering pressure. As pressure drops for the same Thot, conditions get closer to Saturation in the PCS.

Distractors:

Inservice Pressurizer Pressure Controller Output fails low is plausible for the examinee that doesn't know how the controllers work (gets the output control signals and responses reversed). It is incorrect because a low output would turn on heaters and close sprays, both of which raise pressure taking it further from Saturation conditions.

Spray valve loses Instrument Air is plausible for the examinee that believes the valve would fail open. (it fails closed). If heaters were on, or an in surge occurs, pressure would rise, not fall.

Steam Generator ADV opens is plausible for the examinee that believes the cooling effect from removing more heat from the PCS would be large enough to reduce PCS pressure. It is incorrect because the sprays would close, and any drop in PCS pressure would just cause more flashing in the PZR to maintain pressure.

Reference:

Pressurizer Pressure Control System Lesson Plan pages 11, 18, 31 KA:

027 Pressurizer Pressure Control System Malfunction AA2. Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions:

(CFR: 43.5 / 45.13)

AA2.02 Normal values for RCS pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8 3.9 PLP - 2018 NRC EXAM Page: 31 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Test item meets KA. Question poses a situation at 100% power without stating actual PCS pressure.

Examinee must recall what 'closer to saturation conditions' means, based on what actual RCS pressure is at 100% power. Then determine how it is affected by a PPCS malfunction.

HIGH COG - comprehension of how controller failures impact PCS conditions.

Objective: From memory, and given a loss or malfunction of the Pressurizer Pressure Control system, describe the effects on the following:

a. Primary Coolant System

b. Reactor Protective System

c. Safety Injection System without error. (010 K3.01, 010 K3.02, 010 K3.03)

Question 9 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 2.00 System ID: 848571 User-Defined ID: 027AA2.02 Cross Reference Number: PPCS_CK11.0-11 The Plant is at 100% power. Which of the following would Topic:

cause the Primary Coolant System to ge Num Field 1:

Num Field 2:

Text Field:

Comments: PAL-LOR-4422 Bank - not used on 2014, 2017 NRC license exams PLP - 2018 NRC EXAM Page: 32 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 33 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 10 ID: 852014 Points: 1.00 The plant is at power, when a transient occurs causing PCS Pressure to reach 2385 psia.

The Reactor has NOT yet tripped.

Given these conditions, complete the following statements:

Depressing the Manual Reactor Trip Pushbutton on Panel ___(1)___ is supposed to trip the reactor trip breakers (42-1/RPS and 42-2/RPS) similarly to the actions of the ATWS Trip circuitry.

If ONLY 42-1/RPS opens, the reactor ___(2)___ trip.

(1) (2)

A. C-02 does B. C-02 does NOT C. C-06 does D. C-06 does NOT Answer: D Answer Explanation Answer: D.

D. C-06, does NOT; is correct.

Explanation:

There are two manual reactor trip pushbuttons (PB). (One on Panel C-02, and one on Panel C-06).

The PB on C-02 causes a reactor trip by de-energizing all four M relays, which then de-energize the four clutch power supplies. ALL four clutch power supplies must be de-energized to cause the rods to fall by gravity.

The PB on C-06 causes a reactor trip by de-energizing the undervoltage coils on reactor trip breakers 42-1/RPS and 42-2/RPS, causing them to open, interrupting the voltage from the Preferred AC Buses Y-30 and Y-40 to ALL four clutch power supplies. Rods fall by gravity.

In series with 42-1/RPS and 42-2/RPS are the lines from the ATWS trip circuitry. Reaching the ATWS setpoint of 2375 psia, opens contacts which de-energize the undervoltage coils on reactor trip breakers 42-1/RPS and 42-2/RPS, causing them to open, interrupting the voltage from the Preferred AC Buses Y-30 and Y-40 to ALL four clutch power supplies. Rods fall by gravity. This is similar to but independent from Manual Reactor Trip Pushbutton on C-06. Thus, part one is correct.

The opening of ONLY 42-1/RPS, results in a half trip. Only two of the Clutch Power Supplies would get de-energized. Power Supplies 3 and 4 immediately keep powering the clutches, thus NO free fall of the rods by gravity, and NO reactor trip. Part two is correct.

Distractors:

All are plausible for the examinee not familiar with the ATWS circuitry, and the arrangement of the Reactor Trip Breakers and Clutch Power Supply redundancy.

Distractors with C-02 are plausible because one of the two reactor trip pushbuttons is located there. It is incorrect because its action is NOT similar to action of the one on C-06.

PLP - 2018 NRC EXAM Page: 34 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Distractors with 'does' (in part two) are plausible for the examinees that believe the reactor trip breakers are in series, vice parallel. It is incorrect because both reactor trip breakers must be opened to cause a reactor trip.

Reference:

PL-RPS, rev 6, pages 36-39; 101 KA: 029 ATWS EA1 Ability to operate and monitor the following as they apply to an ATWS:

(CFR 41.7 / 45.5 / 45.6)

EA1.12 M/G set power supply and reactor trip breakers . . . . . . . . . . . . . . . . . . . . . . . 4.1 4.0 Test item meets KA. Palisades does NOT have MG sets specifically for powering the reactor trip breaker arrangement.

The question poses a situation where a reactor trip setpoint (hi pressure) has been exceeded and the reactor has not automatically tripped. This is an ATWS situation.

Question requires knowledge of reactor trip breaker circuitry, operation / how it works, and comparison to Manual Trip Circuits. Answer also requires knowledge circuit breaker arrangement for keeping rods withdrawn and impact of the loss of one reactor trip breaker. This is the ability to operate and or monitor the reactor trip breakers.

MEMORY level Objective: From memory, explain how the two manual reactor trip pushbuttons in the Reactor Protective System differ from:

- each other

- the automatic trip function in accordance with DBD-2.05. (012 K4.04, 012 A4.01)

Question 10 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 852014 User-Defined ID: 029EA1.12 Cross Reference Number: RPS_CK27.0 The plant is at power, when a transient occurs causing PCS Topic:

Pressure to reach 2385 psia. The Reacto Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 35 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 36 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 11 ID: 847990 Points: 1.00 A SGTR occurred while the plant was in MODE 1

EOP-5.0, Steam Generator Tube Rupture Recovery, is in progress.

Given these conditions, complete the following statements:

The requirement to maintain PCP NPSH for any running PCP ___(1)___ take precedence over the equalization of primary and secondary pressure.

The reason for lowering PCS pressure to the affected S/G pressure is to ___(2)___.

(1) (2)

A. does raise injection flow from the SIS pumps B. does minimize the potential for overfilling the affected S/G C. does NOT raise injection flow from the SIS pumps D. does NOT minimize the potential for overfilling the affect S/G Answer: B Answer Explanation Answer: B.

B. does, minimize the potential for overfilling the affected S/G; is CORRECT.

Explanation:

The bases document for the SGTR Recovery procedure states the requirement for maintenance of NPSH for the PCPs takes precedence over the strategy of equalizing primary pressure and secondary pressure.

The bases document for the SGTR Recovery procedure also states that maintaining the PCS Pressure within the limits of the Post Accident PT Curves, approximately equal to the isolated S/G pressure and below the lowest MSSV lift setpoint will minimize the loss of primary fluid to the secondary side and the possibility of overfilling the isolated S/G.

Reference:

SGTR Basis Document Rev 11, pages 44, and 45.

Distractors:

Part one that states 'does NOT' is plausible for examinees that believe reducing primary pressure is most important for SGTR recovery (Approaching the same pressures where PCP NPSH is a concern) and confuse this strategy with that of lowering PCS pressure to the pressure of the affected Steam Generator.

These are incorrect as stated in the explanation.

Part two that states 'raise injection flow from SIS pumps' is plausible because flow will rise as pressure is reduced, but incorrect because too much flow when throttling criteria are met, can overfill the S/G, or violate C/D limits on the PZR or PCS.

KA: 038 SGTR PLP - 2018 NRC EXAM Page: 37 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 EK3 Knowledge of the reasons for the following responses as the apply to the SGTR:

(CFR 41.5 / 41.10 / 45.6 / 45.13)

EK3.01 Equalizing pressure on primary and secondary sides of ruptured S/G . . . . . . 4.1 4.3 Test item meets KA. Answer requires knowledge of reason for equalizing pressures of the primary and secondary on a SGTR.

MEMORY Objective:

From memory, discuss the concerns associated with effected SG overfill as it relates to a SGTR without error.

Question 11 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847990 User-Defined ID: 038EK3.01 Cross Reference Number: TBAF_E03.02 A SGTR occurred while the plant was in MODE 1 EOP-5.0, Topic:

Steam Generator Tube Rupture Recovery, is i Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 38 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 39 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 12 ID: 852250 Points: 1.00 (Reference Provided)

Given the following Plant conditions during an Excess Steam Demand Event (ESDE):

Containment temperature is 250°F

'B' steam generator pressure is 300 psia

'B' steam generator level indicated by LI-0752A is 67%

PCS Pressure is 900 psia Which of the following describes the actual level in 'B' steam generator?

A. 53%

B. 56%

C. 58%

D. 65%

Answer: B Answer Explanation Answer: B B. 56%.

Provide EOP Supplement 11 page 1 and 2.

Explanation:

Reference EOP supplement 11:

Subtract 7% (page 1 of 2) from 67% and then use 60% on page 2 of 2 up to the 300 psia line and obtain 56%.

Distractors:

A. 53% is plausible for the examinee that starts with 67%, finds the error of 7% and uses the 100 psia line on page 2.

C. 58% is plausible for the examinee that uses PCS pressure of 500 psia on page 2.

D. 65% is plausible for the examinee that uses 67% on page 1, subtracts the error; then on page 2 uses the graph incorrectly by coming in at 60 on the S/G Level Actual axis, and goes to the 300 psia intercept and looks at the corrected indicated level.

KA: CE E05 Excess Steam Demand EK2. Knowledge of the interrelations between the (Excess Steam Demand) and the following:

(CFR: 41.7 / 45.7)

EK2.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

IMPORTANCE RO 3.3 SRO 3.6 Test item meets KA.

PLP - 2018 NRC EXAM Page: 40 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question tests knowledge of adverse containment conditions having an impact on instrumentation which is needed to make proper decisions.

HIGH COG - requires multiple step use and interpretation of graphs.

Objective:

Given an event involving Emergency Operating Procedures and indicated Steam Generator levels, determine corrected Steam Generator levels in accordance with EOP Supplement 11.

Question 12 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 852250 User-Defined ID: CE E05EK2.1 Cross Reference Number: TBAD_E05.02-1 (Reference Provided) Given the following Plant conditions Topic:

during an Excess Steam Demand Event (ESD Num Field 1:

Num Field 2:

Text Field:

Comments: Modified (changed initial parameters) from Bank not used on 2014 or 2017 NRC exams PAL-LOR-95:

Given the following Plant conditions:

Containment temperature is 250°F

'B' steam generator pressure is 300 psia

'B' steam generator level indicated by LI-0752A is 65%

Which ONE of the following describes the actual level in 'B' steam generator?

A. 43%

B. 54%

C. 60%

D. 67%

Answer: B PLP - 2018 NRC EXAM Page: 41 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 42 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 43 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 13 ID: 847995 Points: 1.00 A Station Blackout occurred 30 minutes ago.

The crew is evaluating indications to verify Natural Circulation is occurring:

Pressurizer Pressure is 1700 psia

Average Qualified CETs is 548°F

Pressurizer level is 36%

Auxiliary Feedwater is keeping each SG level at 40%

Given these conditions, complete the following statements:

If PCS temperatures are slowly lowering, then Natural Circulation is occurring if

the PCS Loop Tcolds are between ___(1)___,

AND

the PCS Loop Thots are between ___(2)___.

(1) Tcolds (2) Thots A. 485°F and 495°F 520°F and 530°F B. 485°F and 495°F 535°F and 545°F C. 500°F and 510°F 520°F and 530°F D. 500°F and 510°F 535°F and 545°F Answer: D Answer Explanation Answer: D B. 500°F and 510°F, 535°F and 545°F is correct.

Reference:

EOP-3.0 rev 19, step 19 page 17/45 Step 24:

If all PCPs are stopped, (and they are because of the SBO) then verify natural circulation in at least one PCS loop by all of the following:

Core DT is less than 50°F (Ave of QCET minus Tc)

Loop Ths and Loop Tcs constant or lowering Ave of QCETs at least 25°F subcooled Difference between Loop Th and Ave of QCET is less than or equal to 15°F.

Distractors are plausible for the examinee that doesn't know the qualitative attributes necessary for natural circulation.

A. plausible but incorrect because difference between Loop Th and Ave of QCET is NOT less than or equal to 15°F, and the Tcs are more than 50°F below the Average of the QCETs, making the Core DT not met.

PLP - 2018 NRC EXAM Page: 44 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 B. plausible but incorrect because difference between Loop Th and Ave of QCET is NOT less than or equal to 15°F. The Ths are okay.

C. Plausible but incorrect because the Ths are too low. The range of Tcs is okay.

D. Correct, all criteria met.

KA:

055 SBO EK1 Knowledge of the operational implications of the following concepts as they apply to the Station Blackout :

(CFR 41.8 / 41.10 / 45.3)

EK1.02 Natural circulation cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 4.4 Test item meets KA as answer requires the ability to determine what the appropriate indications (operational implications) should read for natural circulation to be occurring.

HIGH COG - after recalling the attributes necessary for natural circulation to be occurring, perform a meanningful calcualtion to determine the required temperatures.

Objective:

From memory, determine the impact a Loss of Forced Circulation has on each Safety Function without error.

PLP - 2018 NRC EXAM Page: 45 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 13 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847995 User-Defined ID: 055EK1.02 Cross Reference Number: TBAC_E02.02 A Station Blackout occurred 30 minutes ago. The crew is Topic:

evaluating indications to verify Natural C Num Field 1:

Num Field 2:

Text Field:

Comments: Modified from 843623 (2020 Audit) which has different initial temperatures and pressure, and different bands for distractors and the answer.

A Loss of Forced Circulation has occurred 30 minutes ago.

The crew is evaluating their indications to verify Natural Circulation is occurring.

Pressurizer Pressure is 1820 psia

Average Qualified CETs is 582°F

Pressurizer level is 36%

Auxiliary Feedwater is keeping each SG level at 40%

Given these conditions, complete the following statements:

If the PCS temperatures are slowly lowering, then Natural Circulation is occurring if the PCS Loop...

Tcolds are between ___(1)___,

and the Thots are between ___(2)___.

(1) (2)

A. 532°F and 550°F 557°F and 565°F B. 532°F and 550°F 567°F and 577°F C. 500°F and 530°F 567°F and 577°F D. 500°F and 530°F 557°F and 565°F Answer: B PLP - 2018 NRC EXAM Page: 46 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 47 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 14 ID: 847996 Points: 1.00 A Loss of Offsite Power has occurred.

EOP-8.0 Loss of Offsite Power/ Forced Circulation Recovery is in progress at step 13b, which states:

Corrected PZR level is greater than 20% (40% for degraded Containment) and controlled.

REFER to EOP Supplement 9 (provided)

The following indications are noted:

Containment Ambient Temperature is 250°F

Pressurizer Pressure is 1000 psia

LIC-0101A and 0101B indicate 60% PZR Level What is the actual pressurizer level?

A. greater than 30% but less than 40%

B. greater than 40% but less than 50%

C. greater than 50% but less than 60%

D. greater than 60%

Answer: B Answer Explanation Answer: B B. greater than 40% but less than 50%.

Provide EOP Supplement 9 PZR Level Correction - Hot Cal Explanation:

Step 13 of EOP-8.0, Loss of Offsite Power/ Forced Circulation directs the determination of the corrected PZR level per EOP Supplement 9 for Hot Cal (or 10 for Cold Cal).

Page 1 of Supplement 9 determines the error to subtract from the indications on the hot cal PZR level channels. The error for 250°F Containment Temperature is 9%. Subtracting 9% from the indication of 60% gives 51%.

Using page 2 of EOP Supplement 9, a corrected level of 51% at 1000 psia gives an ACTUAL LEVEL of 43%. Thus B is correct.

Distractors are plausible for misuse of the graphs.

A. is plausible for the examinee that determines the error of 9% and subtracts from 60 to get 51%. then enters page 2 at 51% on the corrected axis, goes to 1000psia line and over to the actual level axis, gets 44%. then subtracts the 9% error (again).

C is plausible for the examinee that only subtracts the error from 60% = 51%.

D is plausible for the examinee that determines a 9% error, but uses page 2 of supp 9 wrong by coming in at 51% on the actual axis, and going across to the corrected level on the 1000 psia line for 61%

KA:

PLP - 2018 NRC EXAM Page: 48 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 056 Loss of Offsite Power 2.1.20 Ability to interpret and execute procedure steps.

(CFR: 41.10 / 43.5 / 45.12)

IMPORTANCE RO 4.6 SRO 4.

Test item meets KA. Answer requires interpreting and executing a step to determine actual pressurizer level in the EOP for a loss of offsite power.

HIGH COG - given parameters, determine PZR Level by read/ interpret graphs .

Objective:

Given plant voiding conditions, determine the potential effect on Pressurizer Level without error.

Question 14 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847996 User-Defined ID: 056G2.1.20 Cross Reference Number: TBAC_E01.01 A Loss of Offsite Power has occurred. EOP-8.0 Loss of Offsite Topic:

Power/ Forced Circulation Recovery Num Field 1:

Num Field 2:

Text Field:

Comments: new, because the initial conditions are different a similar question was on the 2017 NRC exam as question 66:

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%

PLP - 2018 NRC EXAM Page: 49 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 50 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 51 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 15 ID: 847830 Points: 1.00 Given the following conditions with the plant at 65% power during Escalation:

Preferred AC Bus Y-20 de-energizes With NO operator action, then over the next 30 seconds the response of level in the Steam Generators is...

A. A rises, and B lowers B. B rises, and A lowers C. BOTH rise D. BOTH lower Answer: D Answer Explanation Answer: D is correct.

D. BOTH lower.

Explanation /

Reference:

PL-SIS: PL-EPS; AOP-13, Loss of Preferred AC Bus EY-20, Item on Reactor and Equipment Trip Criteria page:

CV-0710, Feed Pump P-1B Recirculation Valve Fails Open.

At 65% power both P-1A, and P-1B MFPs are in service. On a loss of power (EY-20), the recirc valve for P-1B fails open diverting MFP flow from going to the SG to going to the main condenser. Actual level lowers in both SGs.

Distractors:

A. A rises and B lowers is plausible for the examinee that believes the loss of power affects one SG in one way, and the other in an opposite way. This would be possible if the level control system were thought to be Preferred Bus on one SG, and a different Preferred Bus on the other.

B. A lowers and B rises. Like 'A' but reversed.

C. BOTH rise. Plausible for the examinee that believes the loss of power caused the recirc to go from partially open, to closed, resulting in a sudden increase in water flow to the SGs.

Objective:

From memory, and given a loss or malfunction of the 125V DC and AC Power system, describe the effects on the following:

- Major system loads (062 K3.01, 063 K3.02)

- Emergency Diesel Generators (062 K3.02, 063 K3.01)

- DC Electrical system (062 K3.03) without error.

KA:

057 Loss of Vital AC Electrical Instrument Bus AA2. Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:

PLP - 2018 NRC EXAM Page: 52 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 (CFR: 43.5 / 45.13)

AA2.19 The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.0 4.3 Test item meets KA. A loss of Vital bus is proposed in the question. The answer requires comprehension of how that loss of power automatically affects feed flow to the SGs (an automatic action that will occur on the loss of a vital bus.)

HIGH COG - calls for a prediction of response Question 15 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847830 User-Defined ID: 057AA2.19 Cross Reference Number: EPS_CK11.0 Given the following conditions with the plant at 65% power Topic:

during Escalation: Preferred AC Bus Y-Num Field 1: 013A2.04 Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 53 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 54 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 16 ID: 848010 Points: 1.00 The Plant was at 100% power with the following alignments:

ED-17, Station Battery Charger No 3, is in service supplying the No 1 DC Bus

ED-18, Station Battery Charger No 4, is in service supplying the No 2 DC Bus FIVE minutes ago, the Plant experienced a Loss of Offsite Power ALL systems and equipment responded as designed EXCEPT:

2400V bus 1C faulted and remains de-energized

AC Supply Breaker (52-285) for ED-17, Station Battery Charger No 3, tripped and can NOT be reclosed Given these conditions, complete the following statements:

Assuming NO operator actions,

DC Bus No 1 is currently powered from ___(1)___.

DC Bus No 2 is currently powered from ___(2)___.

A. (1) ONLY Station Battery No 1 (2) ONLY Station Battery No 2 B. (1) ED-15, Station Battery Charger No 1, and Battery No 1 (2) ONLY Station Battery No 2 C. (1) ONLY Station Battery No 1 (2) ED-18, Station Battery Charger No 4, and Station Battery No 2 D. (1) ED-15, Station Battery Charger No 1, and Station Battery No 1 (2) ED-18, Station Battery Charger No 4, and Station Battery No 2 Answer: A Answer Explanation Answer: A A. (1) ONLY Station Battery No 1 (2) ONLY Station Battery No 2 A. correct.

Explanation:

With the loss of the feeder breaker to the No 3 Battery Charger, and no automatic pickup by Battery Charger No 1, the only power source for DC bus No 1 is Battery No 1.

The fault of the 2400V Bus 1C, causes MCC-1 to be de-energized. This de-energizes ED-18, Battery Charger No 4. Battery Charger No 2 (the backup power supply to DC Bus No 2) does NOT auto energize the DC bus, leaving the only power for the DC Bus as Battery No 2.

PLP - 2018 NRC EXAM Page: 55 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 The DC buses 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 aligned to the bus. Battery Chargers ED-16 and ED-18 are capable of powering DC Bus No 2, with only one charger normally aligned to the bus.

DG 1-1 supplies chargers ED-15 and ED-18 (except when Bus 1C is faulted), while DG 1-2 supplies chargers ED-16 and ED-17.

Distractor Analysis:

All distractors plausible for examinees that don't know the capabilities of the DC system components, and the impact of the malfunctions imposed.

B. Part one is plausible for the examinee that believes the Battery Charger will be auto re-energized due to the loss of power. It is incorrect because the transfer of the power source is NOT automatic to the charger. Part two is correct.

C. Part one is correct. Part two is plausible for the examinee that believes the Battery charger will be auto re-energized due to the loss of power. It is incorrect because Bus 1C which supplies the charger, is faulted and the EDG won't close in on the faulted bus.

D. Part one is plausible but incorrect as explained in B. Part two is plausible but incorrect as explained in C.

Reference:

SOP-30, DBD-4.02 KA:

058: Loss of DC power AA1. Ability to operate and / or monitor the following as they apply to the Loss of DC Power:

(CFR 41.7 / 45.5 / 45.6)

AA1.01 Cross-tie of the affected dc bus with the alternate supply . . . . . . . . . . . . . . . 3.4* 3.5 Test item meets KA. Answer requires knowledge of how the alternate power supplies operate or are aligned when a DC bus is lost.

HIGH COG - requires a prediction of equipment response to a loss of power with an initially degraded lineup.

Objective:

Given Abnormal Operating plant conditions and control room references, determine the subsequent actions/operator actions to mitigate the event and stabilize the plant in accordance with the applicable Abnormal Operating Procedure(s).

PLP - 2018 NRC EXAM Page: 56 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 16 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848010 User-Defined ID: 058AA1.01 Cross Reference Number: IOTF_CK08.0 The Plant was at 100% power with the following alignments:

Topic:

ED-17, Station Battery Charger No 3, is Num Field 1: 058AA2.01 Num Field 2:

Text Field:

Comments: Modified from NRC 2017 RO question 49 (745633) which posed a different set of initial conditions followed by a different loss of power.

LO-2017N49 (745633)

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 ALL systems responded as designed, EXCEPT:

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.

(2) ED-18, Station Battery Charger No 4.

B. (1) ONLY Station Battery # 1.

(2) ONLY Station Battery # 2.

C. (1) ED-17, Station Battery Charger No 3.

(2) ED-18, Station Battery Charger No 4.

D. (1) ED-15, Station Battery Charger No 1.

(2) ONLY Station Battery # 2.

Answer: D PLP - 2018 NRC EXAM Page: 57 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 58 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 17 ID: 848599 Points: 1.00 While at 100% power, entry conditions for AOP-35, Loss of Service Water, were met.

Below is an excerpt from AOP-35, Loss of Service Water, Section 5.0 IMMEDIATE ACTIONS:

© 1. VERIFY EK-1165, "NON CRITICAL 1.1 IF any of the following annunciators are in SERV WATER LO PRESS" is clear. alarm, THEN TRIP the Reactor, if reset. Refer to EOP-1.0, "Standard Post-Trip Actions."

EK-0259, "EXCITER COOLER HI TEMP" (50C)

PPC Urgent Alarm, "EXC Field Cold Air RTD-31" (48°C)

PPC Urgent Alarm, " EXC Diode Cold Air RTD-32" (48°C) 1.2. IF the following occur, THEN PERFORM Attachment 1, "Break Condenser Vacuum":

Non-Critical Service Water has been isolated or lost

Exciter Cold Air Temperature is greater than 80C or rising as indicated on REC-C11A-01, Generator Temperature Recorder The reason for tripping the reactor per step 1.1 is because ___(1)___.

The reason for breaking condenser vacuum per step 1.2 is to ___(2)___.

A. (1) exciter damage occurs within 10 seconds (2) protect the condenser from overpressure B. (1) exciter damage occurs within 10 seconds (2) minimize the amount of time the exciter is subjected to overheating conditions C. (1) the heat load in containment needs to be reduced to within the capacity of the Containment Air Coolers (2) minimize the amount of time the exciter is subjected to overheating conditions D. (1) the heat load in containment needs to be reduce to within the capacity of the Containment Air Coolers (2) protect the condenser from overpressure Answer: B Answer Explanation Answer: B.

B. is Correct.

(1) exciter damage occurs within 10 seconds (2) minimize the amount of time the exciter is subjected to overheating conditions PLP - 2018 NRC EXAM Page: 59 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Explanation:

Without proper cooling (as indicated by the presence of at least one of the temperature alarms and the low non critical SW header pressure), exciter damage occurs within 10 seconds as stated in the basis document for the AOP, and the CAUTION statement prior to the step but omitted from the stem of the question.

Breaking vacuum because of other extreme conditions, slows the turbine quicker, which slows the exciter quicker, which results in less windage heating.

Distractors:

The distractors with 'reduce heat load in containment' are plausible for examinees that incorrectly conclude that CRITICAL SERVICE WATER header pressure is less than 42 psig. (But entry conditions for this procedure are met even with Critical and Non Critical SW header pressures > 42 psig. They need only be less than 45 psig.) A low Critical SW header pressure of less than 42 psig would apply to step 17 of this AOP, where CV-0824 is closed (which isolates the SW flow to the Containment Air Coolers -

CACs), thereby raising pressure elsewhere in the SW system.

The distractors with 'protect the condenser from overpressure' are plausible for the examinee that thinks a loss of Service Water impacts the Circ Water System which would affect the condenser.

KA: 062 AK3. Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water:

(CFR 41.4, 41.8 / 45.7)

AK3.03 Guidance actions contained in EOP for Loss of nuclear service water . . . . 4.0 4.2 Test item meets KA. Palisades does not have an EOP for loss of service water. Question/ answer requires knowledge of the reason for steps in the Loss of Service Water AOP.

MEMORY level Objective: Given Abnormal Operating plant conditions, determine if an immediate, manual Reactor Trip is required without error.

Question 17 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 848599 User-Defined ID: 062AK3.03 Cross Reference Number: IOTF_CK05.0 While at 100% power, entry conditions for AOP-35, Loss of Topic:

Service Water, were met. Below is an ex Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 60 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 61 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 62 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 18 ID: 848572 Points: 1.00 The Plant is at 825 MWe:

Generator Hydrogen pressure is 75 psig A grid disturbance occurred due to severe weather in the area resulting in:

Generator reactive load is 225 MVAR OUT

The following annunciators are LIT o EK-0303, VOLTAGE REGULATOR LIMITER OPERATION o EK-0310, GENERATOR VOLTAGE REG TRIP Initial attempts to clear the alarms per the ARPs are unsuccessful.

Given these conditions, complete the following statements:

When using the DC Adjuster under these conditions:

the Automatic Regulator Limits ___(1)___ function.

390CS, Voltage Regulator Control Switch should be in the ___(2)___ position.

(1) (2)

A. will AUTO or ON B. will OFF or TEST C. will NOT AUTO or ON D. will NOT OFF or TEST Answer: D Answer Explanation Answer: D D. Correct. The generator has three protective functions, the Maximum Excitation Limiter (MXL), the Online field Forcing Relay (FF), and the over excitation protection (OXP) relay. The MXL will require the Voltage Regulator to attempt to limit field current to 273 amps (as evidenced 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 EK0310) preventing further control in auto. At this point any control must be made in MANUAL (i.e. DC Regulator Mode using the DC Adjuster). The unit is in a stable condition but 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 EK-0303 alarm indicates the generator is on the verge of overexcitation. The initial operator action is to attempt to adjust the Generator Terminal Voltage using the AC Adjuster to clear the alarm. The stem of the question states that doesn't work. The follow-up action is to control generator terminal voltage using the DC adjuster. A note before this step states the automatic regulator limits won't function when using the DC adjuster.

PLP - 2018 NRC EXAM Page: 63 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 The actions require placing the 390CS in OFF or TEST to make adjustments.

Distractors are plausible based on various misunderstandings of how to control reactive load, and the procedurally required position of the control switch to attempt to clear alarms.

The distractors with 'will' are plausible because it is left up to the knowledge of the examinee with regard to DC Adjuster manipulations to determine what positions the controls are in (or placed in) under these conditions.

The distractors with AUTO or ON are plausible for the examinee that believes the operation of the DC adjuster is like the AC Adjuster, where the position of the control switch 390CS is in AUTO or ON.

Reference:

ARP-2, EK-0303, and EK0310.

KA: 077 Generator Voltage and Electric Grid Disturbances AK1. Knowledge of the operational implications of the following concepts as they apply to Generator Voltage and Electric Grid Disturbances:

(CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)

AK1.02 Over-excitation.......................................................................... 3.3 3.4 Test item meets KA. Question poses a situation where overexcitation is likely / occurring due to a grid disturbance. Answer requires knowledge of the operational implications of this overexcitation by requiring knowledge of how the generator protective and control systems work.

MEMORY level Objective:

From memory, describe the design features and interlocks that provide the following Main Generator System functions:

a. Exciter Protective Relays (K/A 045 K1.20)

1) Underexcitation

2) Loss of Field

3) Overexcitation

4) Volts per Hertz

5) Field Ground

6) Field Overcurrent

7) Field Forcing

b. Main Generator Protective Relays (K/A 045 K1.20)

1) Generator Differential

2) Overall Unit Differential

3) Reverse Power

4) Ground Protection

5) Voltage Balance

6) Negative Phase Sequence in accordance with E-2, E-9, E-17 Sh 8, ARP-2 and SOP-8.

Question tests objective and meets KA. answer requires knowledge of how to operate MG controls during a grid disturbance that causes generator parameters to waiver from normal.

PLP - 2018 NRC EXAM Page: 64 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 18 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848572 User-Defined ID: 077AK1.02 Cross Reference Number: MGEN_CK09.0 The Plant is at 825 MWe: Generator Hydrogen pressure is 75 Topic:

psig A grid disturbance occurred due t Num Field 1:

Num Field 2:

Text Field:

PLP - 2018 NRC EXAM Page: 65 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Comments: NEW because this is two parts, but has some similarities to RO Question 18 on the 2014 NRC Exam, which is provided below for comparison:

Given the following with the Plant at full power:

A voltage disturbance occurs on the Grid Main Generator reactive load changes from zero (0) MVARS to 300 MVARs IN Alarm EK-0303, VOLTAGE REGULATOR LIMITER OPERATION, annunciates The Control Room team determines that the Minimum Excitation Limiter has actuated The Main Generator Voltage Regulator is then transferred to Direct Control (DC) in accordance with the associated Alarm Response Procedure Which one of the following describes the impact of using the DC Regulator to maintain Main Generator terminal voltage?

a. The Generator Loss of Field Relay (340) is blocked from actuating.

b. The Volts/HZ Limiter Relay (395) is blocked from actuating.

c. The Main Generator Capability curves are not valid.

d. Automatic Voltage Regulator limits will not function.

There are some similarities to question 18 on the 2017 NRC Exam which is included below:

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 PLP - 2018 NRC EXAM Page: 66 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 D. (1) LOWER (2) DC Adjuster PLP - 2018 NRC EXAM Page: 67 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 68 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 19 ID: 853531 Points: 1.00 The plant is at 70% power, with the controlling group of Regulating Rods at 120 inches withdrawn.

One of the rods in this controlling group drops to the bottom of the core.

The reactor remains critical.

Given these conditions, complete the following statements:

If NO operator action is taken, the average indicated Reactor Power (NI-5, 6, 7, 8) will stabilize at a level ___(1)___ the power level prior to the dropped rod.

In accordance with AOP-5, Control Rod Drop, the operators must ___(2)___ to maintain Tave-Tref mismatch within the prescribed band.

(1) (2)

A. equal to dilute the PCS B. equal to lower turbine load C. lower than lower turbine load D. lower than dilute the PCS Answer: C Answer Explanation Answer: C.

C. lower than, lower turbine load; is correct Explanation:

AOP-5, Control Rod Drop, directs the operator to stop adding positive reactivity, and verify Tave within 3°F of Tref, in the first two steps. The dropping of the rod adds negative reactivity which is offset by the effects of MTC which results in a lower Tave. Because Tave is lower, the water is denser. Since denser water leaks less neutrons, and indicated power is based on leakage, the INDICATED power is lower after the rod drops.

The procedure does NOT allow withdrawing rods, or diluting, and does NOT require a reactor trip. The recovery of Tave is directed by reducing Turbine load.

Reference:

AOP-5, Control Rod Drop.

Distractors:

A. equal to, dilute is plausible but incorrect. Part one is plausible, for examinees that don't understand the change in water density, and the effect on indicated power (vice ACTUAL POWER which would be correct, as there was no change in steam demand). Part two is plausible because diluting would raise Tave but is incorrect because it is not directed by AOP-5.

B. equal to, lower turbine load is plausible but incorrect. Part one is plausible but incorrect as explained in A. Part two is correct as explained in the answer.

PLP - 2018 NRC EXAM Page: 69 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 D. lower than, dilute the PCS is plausible but incorrect. Part one is correct as explained in the answer.

Part two is plausible but incorrect as explained in A.

KA: 003 Dropped Control Rod AA1. Ability to operate and / or monitor the following as they apply to the Dropped Control Rod:

(CFR 41.7 / 45.5 / 45.6)

AA1.05 Reactor power - turbine power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 4.1 Test item meets KA. Question asks what happens to indicated reactor power after a dropped rod and requires knowledge of what to do with turbine power to restore Tave-Tref mismatch from a dropped rod.

HIGH COG - comprehension of how the changing density of the PCS affects the indicated reactor power is necessary to answer the question.

Objective: Given Abnormal Operating plant conditions and control room references, determine the subsequent actions/operator actions to mitigate the event and stabilize the plant in accordance with the applicable Abnormal Operating Procedure(s).

Question 19 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 853531 User-Defined ID: 003AA1.05 Cross Reference Number: IOTF_CK08.0 The plant is at 70% power, with the controlling group of Topic:

Regulating Rods at 120 inches withdrawn.

Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 70 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 71 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 20 ID: 852018 Points: 1.00 A reactor start up is in progress. The following Nuclear Instrument readings are obtained:

Nuclear Instrument Initial Reading 1/M Plot Reading NI-1A, Source Range 3.0 cps 290 cps NI-2A, Source Range 3.3 cps 360 cps NI-3A, Wide Range 1.2 x 10E-7% power 9.5 x 10E-5% power NI-4A, Wide Range 1.3 X 10E-7% power 9.5 x 10E-6% power ALL Nuclear Instruments are responding properly EXCEPT:

A. NI-1A, Source Range B. NI-2A, Source Range C. NI-3A, Wide Range D. NI-4A, Wide Range Answer: C Answer Explanation C. NI-3A, Wide Range is correct.

Per the reference: GOP-3 5.2 NEUTRON FLUX MONITORS 5.2.1 When Source Range Neutron Flux Monitors, NI-1 and NI-2, indicate 3.0 CPS, the Wide Range Neutron Flux Monitors, NI-3 and NI-4, should be responsive and indicate approximately 1x10-7% power. (From the table this is met) 5.2.2 When the Wide Range Neutron Flux Monitors, NI-3 and NI-4, begin to indicate, for each decade of Source Range change, the Wide Range Neutron Flux Monitors should change by approximately one decade. (this is NOT met for NI-3A, for a two decade change in NI-1A/2A, there was NOT a two decade change in NI-3A. 9.5 X 10E-5 is almost 1.0X10E-4 which corresponds to almost a 3 decade change from 1.2 X 10E-7). Thus, C is correct.

5.2.3 When performing 1/M plots, ensure Source/Wide Range Neutron Flux Monitors, NI-1/3 and NI-2/4, responding as expected.

Distractor Analysis:

A. Source Range NI-1A is plausible because the initial and 1/M plot readings are different but close to the readings of Source Range NI-2A. It is incorrect because the test item is looking for the NI that is NOT responding properly.

B. Source Range NI-2A is plausible because the initial and 1/M plot readings are different but close to the readings of Source Range NI-1A. It is incorrect because the test item is looking for the NI that is NOT responding properly.

PLP - 2018 NRC EXAM Page: 72 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 D. Wide Range NI-4A is plausible because the initial and 1/M plot readings are close to Wide Range NI-3A but are within the requirements of the reference.

Reference:

GOP-3 5.2 NEUTRON FLUX MONITORS 5.2.1 When Source Range Neutron Flux Monitors, NI-1 and NI-2, indicate 3.0 CPS, the Wide Range Neutron Flux Monitors, NI-3 and NI-4, should be responsive and indicate approximately 1x10-7% power.

5.2.2 When the Wide Range Neutron Flux Monitors, NI-3 and NI-4, begin to indicate, for each decade of Source Range change, the Wide Range Neutron Flux Monitors should change by approximately one decade.

5.2.3 When performing 1/M plots, ensure Source/Wide Range Neutron Flux Monitors, NI-1/3 and NI-2/4, responding as expected.

KA: 033 Loss of Intermediate Range Nuclear Instrumentation AA2 Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation:

(CFR: 43.5/45.13)

AA2.04 Satisfactory overlap between source-range, intermediate-range and power-range instrumentation. 3.2 / 3.6 Test item meets KA because it requires understanding of the limits on verification of proper functioning of the NI with regard to proper SR to IR overlap.

HIGH COG - Knowledge and comprehension of procedure requirements, then application to a given set of data.

Objective: From memory, describe the design features and interlocks that provide for the following NI System functions: Redundant Nuclear Instrumentation power level indications (015 K4.10) in accordance with FSAR Chapter 7 and DBD-2.05.(NI_CK09.0)

Question 20 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 852018 User-Defined ID: 033AA2.04 Cross Reference Number: NI_CK09.0 From memory, describe the design features and interlocks that Topic:

provide for the following NI System f Num Field 1:

Num Field 2:

Text Field:

Comments: LO-NI OVERLAP Bank - NOT used on 2014 or 2017 NRC exams PLP - 2018 NRC EXAM Page: 73 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 74 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 21 ID: 848052 Points: 1.00 The Plant was at 100% power,

AOP-6, Loss of Condenser Vacuum, has been entered, due to P-39A, Cooling Tower Pump trip

Attachment 1, Cooling Tower Pump Trip actions, are in progress Complete the following statements in accordance with AOP-6, Attachment 1:

MV-AE128, Main Condenser E-10 West Side Air Off Take needs to be ___(1)___ .

MV-AE128 is located ___(2)___.

(1) (2)

A. OPENED in the 'A' MFP Condenser Bay B. OPENED on the Hydrogen Cooler Mezzanine C. CLOSED in the 'A' MFP Condenser Bay D. CLOSED on the Hydrogen Cooler Mezzanine Answer: D Answer Explanation Answer: D D. CLOSED, on the Hydrogen Cooler Mezzanine is correct.

Explanation / Reference of AOP-6, specifies this position and location for a trip of P-39A.

Distractors:

OPENED is plausible, because there are 2 valves on a list of valves to ENSURE OPEN for a Trip of P-39A.

In the 'A' MFP Condenser Bay is plausible because it is one of the first places you check on a loss of condenser vacuum, and it has a lot of valves that penetrate the main condenser and it is the location of the Turbine Bypass Valve isolations.

Reference:

AOP-6 Rev 1, Attachment 1.

KA: 051 Loss of Condenser Vacuum 2.1.30 Ability to locate and operate components, including local controls.

(CFR: 41.7 / 45.7)

IMPORTANCE RO 4.4 SRO 4.0 Test item meets KA. Answer requires knowledge of which controls are operated locally and where they are located for a loss of vacuum.

MEMORY PLP - 2018 NRC EXAM Page: 75 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Objective: Given Abnormal Operating plant conditions and control room references, determine the subsequent actions/operator actions to mitigate the event and stabilize the plant in accordance with the applicable Abnormal Operating Procedure(s).

Question 21 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 848052 User-Defined ID: 051G2.1.30 Cross Reference Number: IOTF_CK08.0 The Plant was at 100% power, AOP-6, Loss of Condenser Topic:

Vacuum, has been entered, due to P-39A, Cool Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 76 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 77 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 22 ID: 848584 Points: 1.00 T-101A, Waste Gas Decay Tank, develops a leak on its manway.

The Radiation Monitor that will indicate this release FIRST is...

A. RIA-1113, Waste Gas B. RIA-2326, Stack Gas Effluent C. RIA-1809, Radwaste Ventilation D. RIA-1810, East Safeguards Equipment Room Answer: C Answer Explanation Answer: C C. RIA-1809, Radwaste Ventilation is correct.

Explanation:

The air flow in Waste Gas Decay Room (where the Waste Gas Decay Tanks are located), is drawn into a HEPA filter (V-54) by the Radwaste Area Exhausters. RIA-1809 monitors this flowpath and an increase in this monitor's reading would be the first indication of this accidental gaseous radwaste release.

Distractors:

A. RIA-1113, Waste Gas is plausible because this is in the normal gaseous radwaste release path, but incorrect because a leak at the gas decay tank manway would not flow through the normal release path.

B. RIA-2326, Stack Gas Effluent is plausible because the Aux Building ventilation does provide some flow up the stack past this monitor, but this is downstream of RIA-1809, and would not be the first indication.

D. RIA-1810, East Engineered Safeguard Equipment Room, is plausible because this flowpath is from this room to the Radwaste Area Exhauster, similar to the flowpath FROM the WASTE GAS Decay Room.

But the RIA-1810 would be sampling the SafeGuards Equipment Room, NOT the Waste Gas Decay Room.

References:

M-218 Sheet 4.

KA: 060 Accidental Gaseous Radwaste Release AK1. Knowledge of the operational implications of the following concepts as they apply to Accidental Gaseous Radwaste Release:

(CFR 41.8 / 41.10 / 45.3)

AK1.01 Types of radiation, their units of intensity and the location of sources of radiation in a nuclear reactor power plant . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5 3.1*

Test item meets KA. Answer requires knowledge of arrangement of rad monitors and their proximity to a source of gaseous activity that is leaking. The operational implication is answering which rad monitor indicates specific accidental release first.

HIGH COG - Assess conditions against system lineup knowledge, predict outcome PLP - 2018 NRC EXAM Page: 78 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Objective:

From memory, explain the purpose of the interfaces (physical connections) between the Waste Gas System and the following plant systems:

a. Chemical Volume Control System (CVCS)

b. Component Cooling Water System (CCW)

c. Nitrogen Gas System

d. Plant HVAC System

e. Instrument Air System

f. Radiation Monitoring System

g. Radioactive Liquid Waste System without error.

Question 22 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 848584 User-Defined ID: 060AK1.01 Cross Reference Number: ISEE_CK06.0 T-101A, Waste Gas Decay Tank, develops a leak on its Topic:

manway. The Radiation Monitor that will indi Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 79 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 80 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 23 ID: 848600 Points: 1.00 The plant was just tripped from 100% power,

Conditions in the Main Control Room degrade to the point where Evacuation is required.

After leaving the Main Control Room, the normal feed breaker to Bus 1C trips on OVERCURRENT.

NO Emergency Diesel Generators automatically start due to UV relay failures.

Given these conditions, complete the following statements:

EDG 1-1 ___(1)___ be started from the local control panel in the Emergency Diesel Generator Room.

Once EDG 1-1 is started, the EDG output breaker ___(2)___ AUTOMATICALLY close.

(1) (2)

A. can will B. can will NOT C. can NOT will D. can NOT will NOT Answer: B Answer Explanation Answer: B B. can, will NOT; is correct.

Explanation:

The tripping of the normal feed breaker creates an instantaneous undervoltage condition on 2400V Bus 1C. This is a start signal for EDG 1-1 even though there is an overcurrent condition on the bus. The UV Start is not prevented. Additiionally, there is a local start switch available in the DG room for locally starting the DG. The bus overcurrent condition does NOT prevent this switch from starting the DG. The automatic closure of the EDG Output Breaker is prevented by the overcurrent lockout condition present on the bus and requires manual reset before the breaker can close.

Part two of the question does NOT give away part one because part two is NOT specific with regard to where the DG was started from, only that it was started. This is plausible for those that think there are other places outside of the DG room and control room to start the DG from (eg other panels remote from the Control Room).

Distractors:

Those distractors with 'can NOT' are plausible for examinees that are not aware of the controls available at the EDG local panel, and how they work during abnormal conditions. There is a local start / stop switch, and governor controls, and voltage controls, but no output breaker control switch.

Those distractors with 'will' are plausible for examinees that remember the UV signal needing to be remedied by the closure of the output breaker but forget that the overcurrent lockout prevents it.

PLP - 2018 NRC EXAM Page: 81 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 KA: 068 Control Room Evacuation AK2. Knowledge of the interrelations between the Control Room Evacuation and the following:

(CFR 41.7 / 45.7)

AK2.07 ED/G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3 3.4 Test item meets KA. Question poses a Control Room Evacuation and requires the examinee to ascertain whether or not an EDG can be started locally under bus lockout conditions, and how the EDG responds to the UV on the bus. The question does NOT give away any relationship between EDG 1-1 and 2400V Bus 1C.

HIGH COG - multiple mental steps to determine how the EDG is operated / operates from outside the control room under abnormal conditions.

Objective: Given the status of the Plant Electrical Distribution System, determine if the Diesel Generators will be powering Bus 1C/1D in accordance with DBD Section 5.02 Question 23 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848600 User-Defined ID: 068AK2.07 Cross Reference Number: ISBB_CK11.0 The plant was just tripped from 100% power, Conditions in the Topic:

Main Control Room degrade to the poi Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 82 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 83 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 24 ID: 848590 Points: 1.00 Conditions have arisen that may require entry and performance of AOP-32, Loss of Containment Integrity.

The following are excerpts from steps 3 and 8 of the procedure:

ACTIONS\EXPECTED RESPONSE RESPONSE NOT OBTAINED

3. DETERMINE AND PERFORM required procedure steps

Inoperable containment isolation valve(s) in a path with two containment isolation valves (Steps 4 through 5)

Inoperable containment isolation valve in a path with one containment isolation valve and a closed system (Step 6)

Failed containment air lock (Steps 7 through 9)

An INOPERABLE Main Steam Isolation Valve (MSIV) ___(1)___ addressed by the actions of this procedure.

ACTIONS\EXPECTED RESPONSE RESPONSE NOT OBTAINED

8. VERIFY the interlock mechanism is 8.1 ENSURE an OPERABLE door is closed on the OPERABLE in each Containment Air Lock. affected air lock within...

8.2 CAUTION TAG the outer door handwheel of the affected air lock ...

When the interlock mechanism of a Containment Air Lock is NOT OPERABLE, the reason for hanging a CAUTION Tag on the outer door handwheel of the affected air lock in Step 8.2 is to ensure ___(2)___.

A. (1) is NOT (2) ONLY one door on the affected air lock is opened at a time B. (1) is NOT (2) an OPERABLE door is verified locked closed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C. (1) is (2) ONLY one door on the affected air lock is opened at a time D. (1) is (2) an OPERABLE door is verified locked closed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Answer: A PLP - 2018 NRC EXAM Page: 84 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Answer Explanation Answer: A (1) is NOT (2) ONLY one door on the affected air lock is opened at a time; is correct Explanation:

The step 3 basis document for this AOP-32 states:

Technical Basis Intent of this step is to provide a branching point for other sections of the procedure based on the type of event that is in progress. Short titles are used to assist the operator in making a determination of the procedure section/sections that would apply.

Training Emphasis:

Training should take the opportunity to review paths that have two Containment Isolation Valves and those that have one Containment Isolation Valve and a closed system (reference FSAR Table 6.14).

The MSIVs have their own tech spec and actions for inoperabilities in LCO 3.7.2 and are not included as Containment Isolation Valves.

The intent of Step 8.1 is to ensure LCO 3.6.2 B.1 is taken when appropriate (i.e. one or more mechanisms inoperable). This is a one hour action.

Per the procedure basis Document:

The reason for step 8.2 is expressed in the NOTE in the procedure between Steps 8.1 and 8.2 which is NOT included as part of the excerpt. It states:

NOTE: The following actions are to ensure only one door on the affected air lock is opened at a time.

Part two of the Distractors are plausible but incorrect as they are the bases for other steps of the procedures.

B. (1) is NOT; (2) an operable door is verified locked closed every 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s: is plausible but incorrect.

Part one is correct as explained in the answer. Part two is plausible for the examinee that doesn't know the substance of the NOTE, and employs his understanding of TS 3.6.2, RA B.B.2 and 3, which has the door locked and verified locked.

C. (1) is, (2) only one door on the affected air lock is opened at a time; is plausible but incorrect. Part one is plausible for the examinee that believes the MSIVs are CIVs addressed by LCO 3.6.3 (which they are NOT) and are covered by either steps 4-6, or 7-9. Part two is correct.

D. (1) is, (2) an operable door is is verified locked closed every 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s: is plausible but incorrect. Part one is plausible but incorrect as explained in C. Part two is plausible but incorrect as explained in B.

KA:

069 Loss of Containment Integrity AK3. Knowledge of the reasons for the following responses as they apply to the Loss of Containment Integrity:

(CFR 41.5,41.10 / 45.6 / 45.13)

AK3.01 Guidance contained in EOP for loss of containment integrity . . . . . . . . . . . 3.8* 4.2 Although PLP uses an AOP for the loss of containment integrity, this test item meets this KA. Question specifically asks for the reasons / bases for specific steps and notes of the governance for the loss of containment integrity.

MEMORY level.

PLP - 2018 NRC EXAM Page: 85 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Objective:

Given Abnormal Operating plant conditions and control room references, determine the subsequent actions/operator actions to mitigate the event and stabilize the plant in accordance with the applicable Abnormal Operating Procedure(s).

Question 24 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 2.00 System ID: 848590 User-Defined ID: 069AK3.01 Cross Reference Number: IOTF_CK08.0 Conditions have arisen that may require entry and performance Topic:

of AOP-32, Loss of Containment Integ Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 86 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 87 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 25 ID: 848614 Points: 1.00 The plant has been tripped from 100% power due to a LOCA.

An Inadequate Core Cooling condition exists and the operators have implemented Once Through Cooling in accordance with EOP-9.0 Functional Recovery.

BOTH HPSI pumps (P-66A and P-66B) are available and operating

ALL Charging Pumps (P-55A, P-55B, and P-55C) are operating

ALL PCPs are stopped Given these conditions, complete the following statements:

The position of the PZR PORV BLOCK valves for these conditions is ___(1)___ OPEN.

If P-66B were to trip, the operator will ___(2)___.

(1) (2)

A. BOTH maintain any open PORV OPEN B. BOTH CLOSE at least one PORV C. ONLY ONE maintain any open PORV OPEN D. ONLY ONE CLOSE at least one PORV Answer: A Answer Explanation Answer: A.

A. BOTH, maintain any open PORV OPEN is correct.

Explanation:

The step(s) in EOP-9.0 is a CONTINUOUS Action step, meaning that the requirements of the step always apply. With both HPSI pumps operating the venting of the PCS through BOTH PZR PORV BLOCK Valves and PORVs is implemented. The RNO of the step also requires the closure of both PORVs if all HPSI pumps are NOT operating. After the trip of P-66B, one HPSI pump is still available and operating.

BOTH PORVS are left open. Thus 'maintain any open PORV OPEN' is correct.

Distractors:

The distractors with 'ONLY ONE' are plausible for those that think one PORV and BLOCK Valve flowpath is sufficient to provide venting capability that is adequate to restore core cooling. It is not.

The distractors with 'CLOSE at least ONE PORV' are plausible for those that think that with one less HPSI pump, only one vent path is necessary. It is not. The object of OTC is to depressurize the PCS and maximize injection flow.

Reference:

Procedure Basis document and EOP-9.0 step 1 of HR-3 success path.

KA: 074 Inadequate Core Cooling PLP - 2018 NRC EXAM Page: 88 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 EA1 Ability to operate and monitor the following as they apply to an Inadequate Core Cooling:

(CFR 41.7 / 45.5 / 45.6)

EA1.23 PORV block valve indicators, switches, controls (for both RCS and S/G) . . 3.9 4.0 Test item meets KA. Palisades takes a deviation from the EOP Basis document in that the PZR PORVs have a large enough relief capacity together that the secondary depressurization through the SG is not needed. The question poses an ICC situation that is addressed via Once Through Cooling (OTC), and requires the knowledge of what the PZR PORV and Block valves should be for various pump availabilities.

This is RO level because it is basic flowpath requirements for OTC.

MEMORY - recall procedure steps for OTC.

Objective: Given plant conditions and Control Room references, determine the in-use Success Paths and their status in accordance with EOP-9.0, Placekeeper and the Resource Assessment Trees.

Question 25 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848614 User-Defined ID: 074EA1.23 Cross Reference Number: TBAH_E01.01 The plant has been tripped from 100% power due to a LOCA.

Topic:

An Inadequate Core Cooling condition exi Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 89 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 90 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 26 ID: 848645 Points: 1.00 Given the following conditions:

The Plant is at 100% power

T AVE is stable at 560°F

PCS pressure is stable at 2060 psia

NO dilution or boration activities are in progress The following data has been collected:

Parameter 1000:00 1012:00 PZR Level 57.9% 58.5%

VCT Level 71.0% 68.5%

RIA-0631, Off Gas monitor 1E-06 1E-06 RIA-0707, S/G BD monitor 2E-06 2.03E-06 Containment Sump Level Stable Rising Given these conditions, complete the following statements:

The estimated PCS leak rate is ___(1)___ gpm.

The NCO-T should recommend entering ___(2)___.

(1) (2)

A. >5 AOP-24, Steam Generator Tube Leak B. >5 AOP-23, Primary Coolant Leak C. <4 AOP-23, Primary Coolant Leak D. <4 AOP-24, Steam Generator Tube Leak Answer: C Answer Explanation Answer: C C. <4, AOP-23, Primary Coolant Leak; is correct.

Explanation:

Rising containment sump level is an indication of a system leaking to the containment sump.

ARP-8 for EK-1351, Containment Sump Hi level directs operator action to refer to AOP-23 Change in PZR level (58.5 - 57.9)/12 min X 66 gal/% = 3.3 gal/min Change in VCT level (68.5 - 71.0)/12 min X 34 gal/% = -7.08 gal/min Sum of PZR & VCT = 3.3 + (-7.08) = -3.78 gal/min AOP-23 is correct because PCS leak rate exceeds 0.15 gpm unidentified leakage per the calculation is an entry condition.

PLP - 2018 NRC EXAM Page: 91 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Distractors:

Those with >5 gpm are plausible for those who forget to account for the change in PZR level.

Those with AOP-24, SGTL, are plausible for those who think only one rad monitor showing a minor rise is enough to meet the entry conditions of the SGTL AOP. The first place any kind of SGTL shows up is in the Off Gas Monitor, which has remained steady over the 12 minute time frame of data collection. The 0.03E-6 rise in the SG Blowdown Monitor, is NOT an Adverse trend, thus entry conditions to AOP-24 are not met. Step 2 of AOP-24 requires the verification of a tube leak by observing at least two diverse adverse trends in rad monitors. The stem states all other rad monitors are stable. Part of selecting the correct procedure is knowledge of what it requires.

Reference:

AOP-23, Primary Coolant Leak, Attachment 1, Determine PCS Leak Rate KA: A16 Excess RCS Leakage AA2. Ability to determine and interpret the following as they apply to the (Excess RCS Leakage)

(CFR: 43.5 / 45.13)

AA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

IMPORTANCE RO 2.7 SRO 3.5 Test item meets KA. Question poses a situation with excess RCS leakage and requires knowledge to select appropriate procedure based on entry conditions.

HIGH COG - calculate leak rate. Meaningful calculation required to determine which procedure to enter.

Objective: Given AOP-23 and applicable plant parameters, determine the PCS leak rate in accordance with AOP-23.

PLP - 2018 NRC EXAM Page: 92 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 26 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848645 User-Defined ID: A16AA2.1 Cross Reference Number: IOTF2_E11.01-7 Given the following conditions: The Plant is at 100% power Topic:

TAVE is stable at 560?F PCS pressure Num Field 1:

Num Field 2:

Text Field:

Comments: Modified from PAL-LOI-7564 shown below. The leak rate is different, and is calculated from different numbers, including over a different time frame. The modified question also requires knowledge of AOP entry conditions.

Given the following conditions:

The Plant is at full power

T AVE is stable at 560°F

PCS pressure is stable at 2060 psia

No dilution or boration activities are in progress The following data has been collected:

Parameter 1445 1454 PZR Level 58.0% 58.4%

VCT Level 74.0% 70.5%

The estimated PCS leak rate is:

PLP - 2018 NRC EXAM Page: 93 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 94 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 27 ID: 851771 Points: 1.00 A transient has occurred resulting in the implementation of EOP-9.0, Functional Recovery Procedure.

The following procedure step is in progress:

Safety Function: Maintenance of Vital DC Power Success Path: Battery Chargers/Station Batteries Resource Tree: B

4. VERIFY the following on 4.1 PUSH 72-01, Isolation Breaker To DC Battery ED-10L/ED-10R DC Bus: NO. 1 ED-01 (Shunt Trip Breaker.)

LOCATION: ED-11A behind 1C Bus

Any faults are isolated 4.2 REFER TO AOP-41, "Alternate Safe Shutdown Procedure."

Voltage greater than 105 volts If ED-10L/ED-10R DC Bus is 100 volts, complete the following statements:

After completing step 4.1

DG 1-1, Emergency Diesel Generator, ___(1)___ able to be manually started.

DG 1-1, ___(2)___ able to field flash.

(1) (2)

A. is is B. is is NOT C. is NOT is D. is NOT is NOT Answer: A Answer Explanation Answer: A.

A. is, is; is correct.

Explanation:

The reference procedure has a caution just prior to step 4 in the RNO column that states:

The following steps result in separating the respective DC Bus from the Station Battery. Buses ED-11A and ED-21A will still be supplied from the Station Batteries.

PLP - 2018 NRC EXAM Page: 95 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 EOP Supplement 7, strips Battery #1 loads in groups. The last group is most important, and contains EDG 1-1 Start Circuit power, and EDG 1-1 Field Flash circuit power which are sourced from ED-11A. The EDG can NOT be started manually without the circuit power, and its field can NOT flash without the circuit power. The Caution prior to step 4 implies there is no loss of starting power, nor field flash power. Thus, part one and part two are correct. Comprehension of what tripping breaker 72-01 results in is necessary to get the correct answer.

Distractors:

B. is, is NOT is plausible but incorrect. Part one is correct as explained in the answer. Part two is plausible for the examinee that doesn't know the loads being stripped and how they affect the EDG start, and field flash circuits, but is incorrect as explained in the answer.

C. is NOT, is is plausible for the same reason as described in 'A', but the examinee believes the EDG can be started LOCALLY, or AUTOMATICALLY. Part two is correct as explained in A.

C. is NOT, is NOT; is plausible for the examinee that agrees that the EDG will NOT start manually but thinks it can be started LOCALLY but the field will not flash. Part one is incorrect, and part two is incorrect.

KA: E09 Functional Recovery 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

(CFR: 41.10 / 43.5 / 45.13)

IMPORTANCE RO 3.8 SRO 4.0 Test item meets KA. Question tests on knowledge of local operator actions performed during a functional recovery (which loads are affected by opening a breaker on a DC bus, and how that affects EDG operations (resultant operational effects on the EDG).

MEMORY level - recall breaker arrangement for DC buses.

Objective: Given plant conditions or an event involving Emergency Operating Procedures, describe the expected plant or instrument response in accordance with the in-use Emergency Operating Procedure.

Question 27 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 851771 User-Defined ID: E09G2.4.35 Cross Reference Number: TBCORE_CK05.0 A transient has occurred resulting in the implementation of Topic:

EOP-9.0, Functional Recovery Procedure Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 96 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 97 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 98 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 28 ID: 848820 Points: 1.00 Control rod drop testing is being performed.

ZERO Power Mode Bypass is in operation

4 PCPs are operating Consider each of the following situations separately:

Situation 1:

If the impeller on P-50A, Loop 1A PCP degrades and lowers Loop 1A flow to 94%, a reactor trip

___(1)___ occur.

Situation 2:

If the impeller on P-50A, Loop 1A PCP degrades and lowers Loop 1A flow to 94%, AND ONE of the Wide Range Nuclear Instruments drifts over ten minutes to an output of 0.01% power, the reactor ___(2)___ trip.

(1) (2)

A. does does B. does does NOT C. does NOT does D. does NOT does NOT Answer: C Answer Explanation Answer: C C. does NOT, does; is correct Explanation:

The Low PCS Flow RPS Trip setpoint is 95% in a loop.

Control rod drop testing is performed while the reactor is shutdown.

The ZERO Power Mode Bypass allows the manual bypassing of four reactor trips (Low PCS Flow, Low SG level, Low SG Press, and the TM/LP low pressure trips).

The plant must be less than 10e-4% power to place ZERO Power mode bypass in operation.

This protection from unwanted trips is automatically removed when power exceeds 10e-4% on either of two IR NIs.

When flow dropped to 94%, a trip would have occurred but did not because of the ZERO Power Mode Bypass being in operation. Thus, part one is correct.

When one IR NI slowly drifted up (at a rate that does not exceed the High Start up Rate RPS trip that is NOT blocked), the ZERO Power Mode Bypass was automatically removed, reinstating the PCS low flow trip, and the reactor trips. Thus, part two is correct.

Distractors:

PLP - 2018 NRC EXAM Page: 99 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 A. does, does: is plausible but incorrect. Part one is plausible for the examinee that doesn't understand what trips are blocked by the ZERO Power Mode Bypass being in operation. Part one is incorrect as explained in the answer. Part two is correct as explained in C.

B. does, does NOT is plausible but incorrect. Part one is plausible as explained in A. Part two is plausible for the examinee that misunderstands how the bypass is auto removed.

D. does NOT, does NOT; is plausible but incorrect. Part one is correct as explained in C. Part two is plausible but incorrect as explained in B.

KA: 003K3.04 K3 Knowledge of the effect that a loss or malfunction of the RCPS will have on the following:

(CFR: 41.7 / 45.6)

K3.04 RPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9/4.2 Test item meets KA. Answer requires knowledge of the effect that a loss of an PCP has on the operation of RPS (of which the operation of Zero Power Mode Bypass is an integral part.)

HIGH COG - prediction of outcome, based on how the RPS is enabled and blocked and auto reinstated.

Objective:

From memory, and given a loss or malfunction of the PCPs, describe the effects on the following:

- PCS (003 K3.01)

- S/G (003 K3.02)

- Feedwater and Auxiliary feedwater system (003 K3.03)

- Reactor Protection System (003 K3.04) in accordance with GFES principles.

Question 28 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 2.00 System ID: 848820 User-Defined ID: 003K3.04 Cross Reference Number: PCP_CK11.0-5 Control rod drop testing is being performed. ZERO Power Topic:

Mode Bypass is in operation 4 PCPs are ope Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 100 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 101 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 29 ID: 848710 Points: 1.00 The plant is at 100% power with normal single charging and letdown in service.

Consider each of the following situations independently.

Situation 1:

Opening MO-3064, HPSI Train 2 Loop Injection, results in TICA-0201, Regen Heat Exchanger Temperature Indicator ___(1)___.

Situation 2:

Opening CV-2117, PZR Aux Spray valve, results in ___(2)___ flow being returned to the PCS Cold leg.

(1) (2)

A. remaining stable less B. remaining stable more C. rising less D. rising more Answer: A Answer Explanation Answer: A.

A. remaining stable, less; is correct.

Explanation:

From the discharge of the charging pumps prior to going through the Regen Hx, there is a line that goes to MO-3064, HPSI Train 2 Loop Injection valve which is in series with MO-3072, HPSI Train 2 Charging X Connection valve. Opening these valves allows some of the flow that would otherwise cool the regen Hx letdown flow, to be diverted directly to the loop. Less cooling flow, results in Regen Hx Outlet Temperature Rising. This is indicated on TICA-0201. However, in the stem of this question only one of these two valves is opened. The other is still closed. The net impact on the charging flow through the regenerative heat exchanger is nil, and the outlet temperature of charging water being returned to the loops is unchanged. Thus, part one is correct.

The PZR Aux Spray flow from charging, is from a pipe connecting downstream of the Regen Heat Exchanger and connects to the normal PZR Spray line, using the same spray nozzle. Opening this flowpath to the PZR Spray line reduces the amount of flow to the PCS Cold legs which is where the positive displacement charging pumps return flow to the PCS. Thus, part two is correct.

Distractors:

B. remaining stable, more is plausible but incorrect. Part one is correct as explained in the answer. Part two is plausible for the examinee that believes the line for aux spray goes to the PCS Cold legs too and makes the charging return flow to the loops greater. Part two is incorrect as explained in A.

PLP - 2018 NRC EXAM Page: 102 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 C. rising, less is plausible but incorrect. Part one is plausible for the examinee that believes the line for HPSI has only one valve isolation between the pumps and the loops. Part one is incorrect as explained in A. Part two is correct as explained in A.

D. rising, more is plausible but incorrect. Part one is plausible but incorrect as explained in C. Part two is plausible but incorrect as explained in B.

References:

PL-SIS rev 7 lp.

PL-CVCS r8 lp.

Drawings: M-201 Sht 2, M-202 sht 1B, 203 sht 1, and 2, and 204 sht 1A KA: 004 CVCS K4 Knowledge of CVCS design feature(s) and/or interlock(s) which provide for the following:

(CFR: 41.7)

K4.05 Interrelationships and design basis, including fluid flow splits in branching networks (e.g., charging and seal injection flow) . . . . . . . . . . . . . . 3.3 3.2 Test item meets KA. Answer to question requires design knowledge of the flow split from charging to HPSI loops, and PZR Aux spray and how they are affected by changing the flow split.

Palisades does NOT have Seal Injection from Charging to RCPs.

HIGH COG - using knowledge of design features / interlocks, predict impacts on temperatures.

Objective: From memory, explain the purpose/functions of the following CVCS components:

b. Regenerative Heat Exchanger

l. Charging Pumps without error.

From memory, for the following Safety Injection System major components, HPSI Header Loop Injection Valves Train 1 (MO-3007, 3009, 3011, 3013)

Train 2 (MO-3062, 3064, 3066, 3068)

a. describe the operational design of each component

b. describe normal operating range of the component in accordance with the FSAR/DBDs. (006 G2.1.28).

SIS_CK02.0 PLP - 2018 NRC EXAM Page: 103 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 29 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848710 User-Defined ID: 004K4.05 Cross Reference Number: SIS_CK02.0 The plant is at 100% power with normal single charging and Topic:

letdown in service. Consider each of t Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 104 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 105 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 30 ID: 848810 Points: 1.00 The plant is at 100% power with double charging aligned using P-55A, and P-55C Charging pumps.

CV-2003, Letdown Orifice Stop valve is OPEN with its handswitch in AUTO

CV-2004, Letdown Orifice Stop valve is OPEN with its handswitch in OPEN

PIC-0202, Intermediate Letdown Pressure Controller is failed 'AS IS' Which of the following operations will result in exceeding the maximum design letdown flow through the demineralizers as indicated on FIC-0202, Letdown Flow, while operating at 100% power?

A. Opening CV-2005, Letdown Orifice Stop valve B. Opening CV-2002, and CV-2202, Letdown Orifice Bypass valves C. RV-2006, Letdown Relief valve fails open D. RV-2013, Low Pressure Letdown line Safety Relief valve, fails open Answer: B Answer Explanation Answer: B.

B. Opening CV-2002 and CV-2202, Letdown Orifice Bypass Valves is correct Explanation:

With 2 orifices already in service, letdown flow is 80 gpm. Opening the Letdown Orifice Bypass valves opens an additional flowpath that is not restricted by the orifices. Letdown flow rises significantly. The Back-pressure Controller can not respond to the increase in pressure due to the rise in flow because the controller is failed 'as is'. The limit on flow through the ion exchangers is 120 gpm per FSAR Chapter 9, section 10. This is the limit that this condition may exceed.

Distractors:

A. Opening CV-2005, Letdown Orifice Stop valve is plausible because it does increase letdown flow. But is incorrect because the orifice limits the flow to 40 gpm, making the total letdown flow 120 gpm, which does not exceed the maximum allowed.

C. RV-2006, Letdown Relief valve fails open is plausible because it would increase letdown flow. But it is incorrect because the valve takes water out of the system prior to the components of concern (demineralizers). This valve taps in just upstream of the Letdown Heat Exchanger.

D. RV-2013, Low Pressure Letdown Relief valve fails open is plausible because this would raise letdown flow. But the flow is diverted to the VCT prior to the demins.

KA: 004 CVCS A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including:

(CFR: 41.5 / 45.5)

A1.07 Maximum specified letdown flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.7 3.1 PLP - 2018 NRC EXAM Page: 106 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Test item meets KA. Question asks to predict or monitor response of system to operation of controls and

/ or failures, and answer requires knowledge of maximum specified letdown flow.

HIGH COG - answer requires comprehension of how system responds to addition of flow paths.

Reference:

PL-CVCS rev 8, page 15; Objective: From memory, predict how the following conditions will impact operation of the CVCS:

a. Increasing PCS Activity (004 A1.01)

b. Changes is Tave (004 A1.02)

c. PCS pressure and temperature (004 A1.03)

d. PZR pressure and level (004 A1.04)

e. Maximum specified letdown flow (004 A1.07)

f. Reactor power (004 A1.10) without error.

Question 30 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848810 User-Defined ID: 004A1.07 Cross Reference Number: CVCS_CK13.0 The plant is at 100% power with double charging aligned using Topic:

P-55A, and P-55C Charging pumps. CV Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 107 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 108 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 31 ID: 847827 Points: 1.00 The Plant is in a solid condition on Shutdown Cooling:

CV-3006, SDC HX Bypass Valve, is throttled as needed.

The air line to CV-3006 becomes damaged such that NO operating air is supplied.

Given these conditions, complete the following statements:

With NO operator action,

CV-3006 ___(1)___.

ONE minute later, compared to before the air line problem, PCS pressure has ___(2)___.

(1) (2)

A. OPENS risen B. OPENS lowered C. CLOSES risen D. CLOSES lowered Answer: A Answer Explanation Answer: A A. OPENS, risen: is correct.

When air is removed from the operator for the CV-3006, the valve fails open, and is listed as such in AOP-37 Loss of Instrument Air, Attachment 2, page 2/3. The direction is to refer to loss of SDC (AOP-30) because with this valve open, more flow is bypassing the HX, and less cooling of the PCS is occurring.

This lets more PCS flow bypass the heat exchanger, thus PCS temperature rises. When heat is added to a solid system, pressure rises. Thus, part one and part two are correct.

Reference:

Palisades Design Basis Document 2.01 page 210:

The analysis revealed one failure which would prevent the LPSI System from delivering water to the reactor. Closing of CV-3006 would block the normal path for safety injection. Operator action can be taken to bypass this valve through the SDCHXs to the reactor. Administrative controls are in place for this valve; the valve is electrically locked open via a key operated switch on the console.

AOP-37, Loss of Instrument Air.

Distractors:

B. Opens, lowered. Part one is correct as explained above. Part two is plausible for the examinee that doesn't understand the location of the bypass valve and the impact its opening has on PCS temperature.

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EXAMINATION ANSWER KEY Daft N 1 C. Closes, lowered. Part one is plausible because some valves fail closed on loss of air (like the PZR Aux Spray Valve) or are required to close on RAS (like the SIRWT Suction valves). Part two is plausible but incorrect as explained in B D. Closes, risen: is incorrect. Part one is plausible as stated in C, part two is correct.

KA:

005 RHRS K5 Knowledge of the operational implications of the following concepts as they apply the RHRS:

(CFR: 41.5 / 45.7)

K5 05 Plant response during "solid plant": pressure change due to the relative incompressibility of water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.7* 3.1*

Test item meets KA. The operational implication being tested is what happens to the pressure in a solid PCS when an air operated temperature control valve has no air.

HIGH COG - answer requires prediction of valve response and impact on PCS pressure.

Objective: From memory, explain the basis for the failure mode (loss of air/power) of CV-3006, SDC Heat Exchangers E-60A/B Bypass, and CV-3025, SDC Heat Exchangers E-60A/B Bypass, in accordance with the FSAR.

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EXAMINATION ANSWER KEY Daft N 1 Question 31 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847827 User-Defined ID: 005K5.05 Cross Reference Number: SDC_CK24.0-2 The Plant is in a solid condition on Shutdown Cooling: CV-Topic:

3006, SDC HX Bypass Valve, is throttled Num Field 1:

Num Field 2:

Text Field:

Comments: Modified from PAL-LOR-3188, which tests on the reason for the fail position of the valve.

The Plant is on Shutdown Cooling with CV-3006, SDC HX Bypass Valve, throttled as needed. If the air line to CV-3006 is damaged such that NO operating air is supplied, what would be the resulting valve position and the reason for this design?

CV-3006 would fail A. OPEN, to prevent overcooling the PCS.

B. CLOSED, to prevent runout of the operating LPSI pump.

C. OPEN, since that is the fail safe position for an SIAS.

D. CLOSED, since that is the fail safe position for a RAS.

Answer: C PLP - 2018 NRC EXAM Page: 111 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 112 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 32 ID: 849650 Points: 1.00 The plant has been tripped from 100% power due to a LOCA.

EOP-4.0, LOCA, is in progress and the crew is trying to determine if SI Throttling Criteria are met. They note the following:

TWO PCPs running

Pressurizer Pressure is 1200 psia

Average Qualified CETs 539°F

TI-0112HC, Loop 1 Th 536°F

TI-0112HD, Loop 1 Th 535°F

TI-0112CC, Loop 1 Tc 531°F

TI-0112CD, Loop 1 Tc 531°F

Containment Pressure 1.0 psig and steady

Containment Temperature 150°F and steady

PZR Level 50% (corrected)

BOTH S/Gs 40% (corrected) and slowly rising

RVLMS ALL Green Lights SMM-0114, Subcooled Margin Monitor is indicating the following margins:

20°F

242 psi Given these conditions, complete the following statements:

HPSI flow ___(1)___ be throttled.

The indication on SMM-0114 is ___(2)___.

(1) (2)

A. can accurate B. can too low C. can NOT too low D. can NOT accurate Answer: B Answer Explanation Answer: B.

B. can, too low; is correct.

Explanation:

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EXAMINATION ANSWER KEY Daft N 1 1200 psia, has a Tsat of 567°F. The Subcooling Margin Monitor (SMM-0114) uses the highest (from the TE-0112CC, CD, HC, HD instruments) Wide Range Temperatures and subtracts the highest from PZR Tsat. 567°F minus 536°F is 31°F. Thus, part one is correct even though the SMM indication is too low for actual subcooling temperature margin.

SI Throttling requires at least 25°F subcooling, which exists by manual determination. The other criteria required for SI throttling are met (PZR level, SG level, RVLMS). The given Containment pressure and temperature ensure the containment is in a NON degraded condition.

The SMM uses that same auctioneered high wide range temperature and Pressurizer Pressure (from PT-0105). From that it calculates the Saturation Pressure for that temperature. Using the Steam Tables, 536°F has a Psat of 930 psia. Pressurizer Pressure is given as 1200 psia, the difference is the margin.

That is 1200-930 = 270 psia. Since the SMM indicates 242 psia, it is inaccurate because it is indicating too low. Thus, part two is correct.

Distractors:

Each distractor is plausible for the examinee that does not know how the subcooling margin monitors work, or does not know that for non-degraded containment conditions, the minimum subcooling required for throttling SI is 25°F.

The 'can NOT' distractors are plausible for the examinees that only look at the SMM indication and see 20°F without calculating it.

The 'accurate' distractors are plausible for the examinee that can't figure out how to determine Tsat / Psat relationship. 1200 psia -242 psia =958 psia. Tsat for that is about 539°F which is given as the Average Qualified CETs.

Reference:

From DBD-2.04 PCS:

Primary coolant protective channels C and D also furnish wide range hot and cold leg temperature signals to the subcooled margin monitors (through I/I isolation devices) and to recorders PTR-0112 (channel C) and PTR-0122 (channel D) from the wide range outputs of the appropriate temperature transmitters.

KA: 006 ECCS K6 Knowledge of the effect of a loss or malfunction on the following will have on the ECCS:

(CFR: 41.7 / 45.7)

K6.18 Subcooling margin indicators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6 3.9 Test item meets KA. Question poses a situation where ECCS is in service and the SMM indications are suspect. Answer requires knowledge of the effect of the malfunctioning SMM on throttling ECCS.

HIGH COG - meaningful calculation required to determine if instrumentation is functioning properly in order to throttle HPSI flow.

Objective: Given plant conditions involving Emergency Operating Procedure, describe the mitigating strategy of the in-use Emergency Operating Procedure in accordance with the Emergency Operating Procedure Bases Document.

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EXAMINATION ANSWER KEY Daft N 1 Question 32 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 3.00 System ID: 849650 User-Defined ID: 006K6.18 Cross Reference Number: TBCORE_CK01.0 The plant has been tripped from 100% power due to a LOCA.

Topic:

EOP-4.0, LOCA, is in progress and the cr Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 115 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 116 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 33 ID: 847829 Points: 1.00 The plant was tripped from 100% power due to a LOCA.

The Safety Injection Tanks (SIT) are expected to start injecting at a PCS Pressure of ___(1)___

psia.

In the recovery actions for establishing a controlled PCS cooldown, the SITs should be isolated or vented when PZR Pressure FIRST reaches ___(2)___ psia.

(1) (2)

A. 320 250-200 B. 320 350-300 C. 250 250-200 D. 250 350-300 Answer: D Answer Explanation Answer: D D. 250, 350-300 Explanation / Reference PL-SIS rev 7 page 24:

If PCS Pressure drops below approximately 250 psia (sum of gas pressure and elevation head) the check valves open and the tanks discharge to the PCS.

EOP-4.0 LOCA (Step 59)

If PZR Pressure is between 350 and 300 psia and controlled, and a controlled cooldown is in progress, then Isolate the SITs. If any SIT could not be isolated, then vent it to containment.

This prevents an undesired injection of water or nitrogen into the PCS as PCS pressure is reduced during a controlled cooldown.

Distractors:

A. / B. Normal operating pressure is 210-215 psia, making 320 a plausible choice for injection commencement. But this pressure over estimates the elevation head of the tank of water.

C. 250-200 is a logical choice, but there is no margin to prevent the undesirable effect of injecting water or nitrogen when a controlled cooldown and depressurization is being attempted.

KA:

006 A3 Ability to monitor automatic operation of the ECCS, including:

(CFR: 41.7 / 45.5)

A3.01 Accumulators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.0* 3.9 PLP - 2018 NRC EXAM Page: 117 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Test item meets KA. Monitoring the operation of the SI Accumulators includes knowing when ECCS accumulators should inject and should be isolated (basic operation).

MEMORY Objective: From memory, for the following Safety Injection System major components,

- SIRWT (T-58)

- SI Pumps suction from SIRWT valves (CV-3031, CV-3057)

- SI Pump minimum flow stop valves (CV-3027, CV-3056)

- SI Pumps suction from Containment Sump valves (CV-3029, CV-3030)

- HPSI Pumps (P-66A/B)

- LPSI Pumps (P-67A/B)

- Safety Injection Tanks (SIT-T-82A/B/C/D)

- Safety Injection Tank Outlet Valves (MO-3041, MO-3045, MO-3049, MO-3052)

- HPSI Header Loop Injection Valves

- Train 1 (MO-3007, 3009, 3011, 3013)

- Train 2 (MO-3062, 3064, 3066, 3068)

- LPSI Header Loop Injection Valves (MO-3008, 3010, 3012, 3014)

- HPSI Train 1 & 2 Cold Leg Injection Valves (MO-3080, MO-3081)

- HPSI Train 1 & 2 Hot Leg Injection Valves, (MO-3082, MO-3083)

a. describe the operational design of each component

b. describe normal operating range of the component in accordance with the FSAR/DBDs. (006 G2.1.28)

Question 33 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 847829 User-Defined ID: 006A3.01 Cross Reference Number: SIS_CK02.0-3 The plant was tripped from 100% power due to a LOCA. The Topic:

Safety Injection Tanks (SIT) are expected Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 118 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 119 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 34 ID: 846870 Points: 1.00 At time 1000:00, with the Plant operating at 100% power,

LIA-0116, T-73 Quench Tank level is 72%

PIA-0116, T-73 Quench Tank Pressure is 3 psig and rising at 1 psi / minute

SS-0116, Pressurizer Quench Tank PI Range Selector is in NARROW position At time 1005:00:

Venting T-73 to T-76, Waste Gas Surge Tank commences, per SOP-1A, Primary Coolant System

T-73 pressure STABILIZES Given these conditions complete the following statements:

At 1005:00, EK-0732, QUENCH TANK HI PRESS, ___(1)___ in alarm.

If, at 1006:00, T-73 pressure starts rising at 5 psig / minute, it will take ___(2)___ 15 minutes until T-73 pressure is equal to Containment Pressure.

(1) (2)

A. is less than B. is NOT less than C. is NOT more than D. is more than Answer: C Answer Explanation Answer: C C. is NOT; more than; is CORRECT - at 1005, QT pressure went from 3 psig, to 8 psig. The alarm is at 10 psig. At 1006, pressure starts at 8 psig and the rupture disc actuates at 100 psig, 92 psig/5 psi per minute = 18.4 minutes.

Distractors A. is; less than; is plausible but incorrect. Part one is plausible for the examinee that doesn't know the hi pressure alarm setpoint. Part two is plausible if student starts at 9 psig in the QT and utilizes 25 psig for the rupture disc setpoint. This pressure corresponds to the top of the narrow range pressure range. 25-9=16. 16/5= 3.2 minutes which is less than 10.

B. is NOT; less than; is plausible but incorrect. Part one is correct as explained in the answer. Part two is plausible if the student utilizes 25 psia and adds 15 for the rupture disc setpoint (25+15=40). 40/5

=8 minutes, which is less than 10.

D. is; more than; is plausible but incorrect. Part one is plausible as explained in A. Part two is correct as explained in the answer.

PLP - 2018 NRC EXAM Page: 120 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1

Reference:

ARP-4, EK-0732 Quench Tank Hi Press Setpoint 10 psig.

Automatic Function: Quench Tank Rupture Disc ruptures at approximately 100 psig.

KA: 007 PRTS / Pressurizer Quench Tank A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including:

(CFR: 41.5 / 45.5)

A1.02 Maintaining quench tank pressure . . . . 2.7 2.9 Test item meets KA. Stem poses a situation where quench tank pressure is being monitored during a rising pressure event, and controls are being operated to control quench tank pressure in accordance with a procedure. The answer requires knowledge of the relief setpoint that occurs at the design pressure of the quench tank, and a calculation to determine how long before this relief device actuates.

HIGH COG - analysis of rate of pressure rise vs high pressure alarm setpoint, then a change in rate of rise vs rupture disk setpoint.

Objective: Objective:

From memory, describe the consequences of operating the Primary Coolant System under the following conditions:

- Operation with excessive primary to secondary differential pressure. (G2.1.32)

- Differential boron concentration between PCS and Pressurizer equal to or greater than 50 ppm (G2.1.32)

- Operating with Pressurizer level greater than 62.8% (G2.1.32)

- Exceeding motor starting limitations for PCPs (G2.1.32)

- Abnormal pressure in Quench Tank (007 A2.02)

- Exceeding Quench Tank high pressure limits G2.1.32)

- Exceeding PCS Heatup and Cooldown Rates (G2.1.32) in accordance with SOP-1A, 1B, and 1C. (G2.1.32)

PLP - 2018 NRC EXAM Page: 121 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 34 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 846870 User-Defined ID: 007A1.02 Cross Reference Number: PCS_CK16.0 At time 1000:00, with the Plant operating at 100% power, LIA-Topic:

0116, T-73 Quench Tank level is 72% P Num Field 1:

Num Field 2:

Text Field:

Comments: modified from 710505 which asked for a calculation to get from the hi pressure alarm setpoint to rupture disc setpoint at 9 psi/min rate of rise and is shown below:

With the Plant operating at full power,

T-73, Quench Tank pressure is slightly lower than its maximum allowable (alarm setpoint)

RV-1041, Pressurizer Relief Valve, begins to lift partially

T-73, Quench Tank, pressure is then noted to be rising at a rate of 9 psi/min How long will it take for Quench Tank pressure to be equal with Containment pressure?

A. less than 2 minutes B. more than 2 minutes, but less than 9 minutes C. 10 minutes D. more than 11 minutes Answer: C PLP - 2018 NRC EXAM Page: 122 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 123 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 35 ID: 846871 Points: 1.00 The plant is at 100% power

Breaker 152-107, Diesel Generator (DG) 1-1 to Bus 1C is racked to disconnect for cleaning and repair At time 1000:00

A differential current actuation condition at the Safeguards Transformer causes breaker 152-401, Safeguards Transformer 1-1 Incoming Breaker, to open

Breaker 152-202, Startup Transformer 1-2 to Bus 1D, fails to close Given these conditions, with NO other operator actions, which of the following, if any, is energizing P-52A, Component Cooling Pump?

A. Diesel Generator 1-2 B. Startup Transformer 1-2 C. Station Power Transformer 1-2 D. P-52A is de-energized Answer: B Answer Explanation Answer: B B. Startup XFMR 1-2.

Explanation:

When the differential current condition causes breaker 152-401, SG XFMR 1-1 Incoming Breaker, to open, the Safeguards bus de-energizes. A fast transfer to Start-up Transformer 1-2 occurs, keeping 2400V Buses 1C, and 1E energized, but breaker 152-202 fails to close, causing an undervoltage on bus D.

DG 1-1 is the emergency source for 2400V Bus 1C. With DG 1-1 output breaker 152-107 not available, 2400V Bus 1C will be energized from S/U XFMR 1-2. Thus P-52A will be powered from S/U XFMR 1-2.

Per P &ID, Single Line Meter and Relay Diagram 480 Volt Motor Control Center, E-1 Sheet 1 Rev 87.

Distractor Analysis:

A. Diesel Generator 1-2 is plausible for the examinee that believes P-52A is powered from 2400V Bus 1D, which is being powered by DG 1-2 due to breaker 152-202 failing to close. It is incorrect because P-52A is a load on 2400V Bus 1C. However, there are other 'A' loads on Bus 1D (P-67A, P-7A, P-66A)

PLP - 2018 NRC EXAM Page: 124 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 C. Station Power Transformer 1-2 is plausible for the examinee that believes the loss of the Safeguards Transformer via 152-401 will result in a fast transfer via 152-402 to Station Power Transformer 1-2. It is incorrect, because this alignment would require manual actions, which are overridden by the stem of the question.

D. P-52A is de-energized is plausible for the examinee that believes the unavailability of DG 1-1 due to the output breaker being racked out, and a failure of fast transfer would result in P-52A being de-energized. It is incorrect because Bus 1C is energized by S/U XFMR 1-2.

References:

P &ID, Single Line Meter and Relay Diagram 480 Volt Motor Control Center, E-1 Sheet 1 Rev 87.

KA:

008 CCW K2 Knowledge of bus power supplies to the following:

(CFR: 41.7)

K2.02 CCW pump, including emergency backup . . .. . . . 3.0* 3.2*

Test item meets KA. Question poses a loss of normal power supply to CCW pump and answer requires knowledge of emergency backup alignments for power, and the operation of the fast transfer equipment.

HIGH COG - Comprehension of what transformers feed what buses, and the impact of some breakers not available or not functioning on a fast transfer, make this high cog to determine if the equipment has power available or not.

Objective(s):

Predict how the following conditions will impact operation of the Electrical Distribution system:

- Bus undervoltage (062 A1.05)

- Bus overcurrent (062 A1.06) SPS_CK13.0 Describe the effects of a loss or malfunction of the Electrical Distribution system on the following:

- Major system loads (062 K3.01) SPS_CK11.0 Question 35 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 846871 User-Defined ID: 008K2.02 Cross Reference Number: SPS_CK11.0 The plant is at 100% power Breaker 152-107, Diesel Generator Topic:

(DG) 1-1 to Bus 1C is racked to disco Num Field 1:

Num Field 2:

Text Field:

Comments: Bank not used on 2014 or 2017 NRC exams PLP - 2018 NRC EXAM Page: 125 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 126 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 127 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 36 ID: 846918 Points: 1.00 During a power escalation a transient occurs which causes Pressurizer pressure to change to 2075 psia.

If all equipment is functioning properly, the expected indications for the in service Pressurizer Pressure Controller are:

Process Indication is ___(1)___ Setpoint Indication

Output Signal is ___(2)___ 66%

(1) (2)

A. less than less than B. less than greater than C. greater than less than D. greater than greater than Answer: D Answer Explanation Answer: D D. greater than, greater than is correct.

Explanation:

Normal pressurizer pressure is controlled at a setpoint of 2060 psia. With actual pressure (process pressure) at 2075 psia, the process indication is greater than the setpoint.

Reference From PL-PPCS:

a. The pressure indicating controllers (PIC) provide an output signal to control the Pressurizer Spray Valves and the Pressurizer Heaters.

1) Proportional Heaters vary from full on to off as PIC output varies from 0% to 33.3%.

2) Main Spray Valves are controlled from shut to fully open as PIC output varies from 66.6% to 100%.

When PIC output is between 33.3% and 66.6%, the Proportional Heaters and Main Sprays Valves are off and closed, respectively.

Since pressure is high (2075 vice 2060), output signal is high.

Distractor Analysis All distractors are plausible for the examinee that doesnt know how the pressurizer pressure controllers work.

A. Part one and part two are plausible for the examinee that gets the relationship between actual pressure and the controller inputs/outputs backwards.

B. Part one is plausible for the examinee, that either doesn't know the setpoint, or doesn't know what the process indication is representing. Part two is correct as explained in the answer.

PLP - 2018 NRC EXAM Page: 128 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 C. Part one is correct as explained in the answer. Part two is plausible for the examinee that doesn't know what the output signal from the controller does when pressure changes.

KA:

010 PPCS A3 Ability to monitor automatic operation of the PZR PCS, including:

(CFR: 41.7 / 45.5)

A3.02 PZR pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6 3.5 Test item meets KA. Question poses monitoring PCS Pressure, which is NOT at the setpoint, during automatic operations. Answer requires comprehension /comparison of actual pressure vs programmed pressure, and what should the controller be indicating (the Controller indications are part of the PZR Pressure Control System - PCS).

MEMORY - recall controller setpoints, indications Objective:

Describe the operation of the PZR Level and Pressure controllers including automatic and backup control bands. PPCS_CK02.0 PLP - 2018 NRC EXAM Page: 129 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 36 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 846918 User-Defined ID: 010A3.02 Cross Reference Number: PPCS_CK02.06 During a power escalation a transient occurs which causes Topic:

Pressurizer pressure to change to 2075 p Num Field 1:

Num Field 2:

Text Field:

Comments: modified from 710677, shown below:

Stem changed to add a second part, and such that a distractor is now the answer.

During a power escalation a transient occurs which causes Pressurizer pressure to change to 2045 psia.

If all equipment is functioning properly, the expected indications for the in service Pressurizer Pressure Controller are:

Process Indication ___(1)___ Setpoint Indication

Output Signal ___(2)___ 66%

(1) (2)

A. Greater than greater than B. Greater than less than C. Less than greater than D. Less than less than Answer: D PLP - 2018 NRC EXAM Page: 130 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 131 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 37 ID: 847153 Points: 1.00 The Plant is at 60% power,

EY-10, Preferred AC Bus #1, develops a problem and de-energizes

AOP-12, Loss of Preferred Bus EY-10, is entered Given these conditions, complete the following statements:

BEFORE any operator action, the MINIMUM number of additional RPS channels that must reach a trip setpoint on RPS to cause a Reactor Trip is ___(1)___.

AFTER the affected RPS Channel is addressed per the AOP, the MINIMUM number of RPS channels that must reach a RPS trip setpoint to cause a Reactor Trip is ___(2)___.

(1) (2)

A. one one B. one two C. two one D. two two Answer: B Answer Explanation Answer: B B. one, two: is CORRECT. - A loss of EY-10 will cause all 'A' channel RPS trips to trip. Only one other channel needs to trip to cause a reactor trip. After the affected channel has been bypassed, the logic reverts to 2 out of the remaining 3 non bypassed channels.

Explanation /

References:

ARP-3, window 43; AOP-12, Rev-2; TS Basis 3.3.1; PL-RPS rev 6 pages 57-58:

PLP - 2018 NRC EXAM Page: 132 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 (1) Failure of a Preferred AC Bus (a) Failure of a preferred instrument bus will de-energize all parts of the RPS in the affected channel:

i. All field instruments ii. All eleven trip units iii. All three 28 VDC logic matrix power supplies iv. If Y-30 fails, then clutch power supplies 1 and 2 will lose their power, causing a half-trip.

v. If Y-40 fails, then clutch power supplies 3 and 4 will lose their power, causing a half-trip.

(b) This failure will not cause a reactor trip.

i. Field instrument failures will affect the RPS as described above.

ii. Trip unit failures will actuate a trip signal in each associated logic matrix; however, the trip logic requires two trip units in separate channels in the trip condition, and only one trip unit per reactor trip is present.

iii. The affected logic matrices will remain energized by the opposite channels 28 VDC power supply, which will keep the matrix relays energized.

iv. For both half-trip signals (caused by loss of Y-30 or Y-40), the clutch power supplies will remain energized by the opposite halfs power supply.

? If Y-30 fails, then clutch power supplies 1 and 2 de-energize, but the clutch DC power buses remain energized by clutch power supplies 3 and 4.

? Conversely, if Y-40 fails, then clutch power supplies 3 and 4 de-energize, but the clutch DC power buses remain energized by clutch power supplies 1 and 2.

(c) This will not prevent a reactor trip because three operable channels remain for each of the eleven trip signals.

(d) The trip logic is changed to 1/3 for each reactor trip signal until that channels trips are bypassed. At that time the RPS trip logic reverts to 2/3. The loss of the +15 VDC P/S results in a failure of the Channel Trip Bypass light to be lit when bypassed. The trip is bypassed by closing contacts in the logic ladders in parallel with the failed open contacts, at which point RPS overall trip logic reverts to 2/3 (from 1/3).

i. If a second Preferred AC bus de-energizes the reactor will trip. Each of the six logic ladders is supplied by an auctioneered 28 VDC power supply from two of the Preferred AC buses (AC, AB, AD, BC, BD and CD). e.g. If Y-10 is lost then the AB, AC and AD ladders each would have only one P/S keeping their matrix relays energized. If Y-40 subsequently de-energizes then the AD logic ladder matrix relays will de-energize generating a full reactor trip.

(e) If Y-30 and then Y-40 become de-energized the reactor will trip for two reasons even if the Y-30 trips are bypassed prior to the Y-40 loss. Firstly, the reactor will trip due to a complete loss of power to the clutch power supplies. Secondly, the reactor would trip anyway due to the loss of power to the CD logic ladder and its matrix relays.

DISTRACTOR ANALYSIS A. one, one is Incorrect - Part one is correct as explained in the answer above. Part two is plausible for those who don't understand the difference between 'trip' and 'bypass' or don't know what actions the AOP directs to address the loss of the bus.

C. two, one is plausible but incorrect. Part one is plausible for those who think the trip coincidence takes 2 of 4, and doesn't realize the loss of power to one channel produces a trip on that channel. Part two is plausible for those who think the procedure actions are to insert a trip on the de-energized channel, and then only need one more channel to trip.

D. two, two; is plausible but incorrect. Part one is plausible as explained in C. Part two is plausible with part one for those who don't know the impact of the actions directed by the procedure.

KA: 012 RPS K6 Knowledge of the effect of a loss or malfunction of the following will have on the RPS:

(CFR: 41.7 / 45/7)

K6.03 Trip logic circuits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 3.5 PLP - 2018 NRC EXAM Page: 133 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Test item meets KA. Question poses a situation where a loss of power affecting the trip logic /

coincidence occurs in RPS. Answer requires knowledge of the impact of procedurally directed actions on the Trip Logic / coincidence circuitry.

HIGH COG - Multiple mental steps required, including determining impact of procedurally directed action on trip logic. The 'before' statement might be recall of an automatic action, but the 'after' statement requires knowledge of the action and its impact.

Objective: From memory, predict how the following conditions will impact operation of the Reactor Protective System:

- low failure of monitored parameter input signal (012 K6.06)

- high failure of monitored parameter input signal (012 K6.06)

- erratic or faulty trip unit operation (012 A2.01)

- erratic power supply operation (012 A2.04)

- loss of Preferred Bus Y10, Y20, Y30, or Y40 (012 A2.02)

- failure of the clutch power supply trip relay arc suppression network (012 A2.06)

- electrical ground in Clutch power supply control power circuits (012A2.02)

- electrical ground in Clutch AC power supply circuits (012A2.02)

- electrical ground in Matrix logic ladders (012 K6.03) without error.

Question 37 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 847153 User-Defined ID: 012K6.03 Cross Reference Number: RPS_CK13.0-18 The Plant is at 60% power, EY-10, Preferred AC Bus #1, Topic:

develops a problem and de-energizes AOP-1 Num Field 1:

Num Field 2:

Text Field:

Comments: Bank 2008 NRC RO EXAM PAL-LOR-4160 PLP - 2018 NRC EXAM Page: 134 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 135 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 38 ID: 847154 Points: 1.00 (Reference provided)

The plant is at 30% power,

The white high power rate pre-trip bistable unit light on the RPS panel for channel 'A' has illuminated.

Given these conditions complete the following statements:

A cause for the bistable trip unit light to illuminate at this power level is a(n) ___(1)___.

After the condition clears, the bistable light is reset by ___(2)___.

(1) (2)

A. Start-up rate = 1.5 DPM bypassing the channel with the keyswitch B. Start-up rate = 1.5 DPM depressing the two lamp pushbutton C. ASI = -0.56 depressing the two lamp pushbutton D. ASI = -0.56 bypassing the channel with the keyswitch Answer: C Answer Explanation Provide the following page from the Palisades COLR, rev 20:

page 6 of 10, showing TABLE 2.4-2 Power Distribution Measurement Uncertainty Factors.

Answer: C C. ASI = -0.56, depressing the two lamp pushbutton; is CORRECT.

Explanation/

References:

PL-RPS r6 page 18 states:

c. When a trip unit has tripped, the following lamps are illuminated:

(1) White pre-trip lamp (top half of two-lamp split-screen pushbutton) immediately below the trip unit label (a) The white pre-trip lamp seals in. It must be manually reset by momentarily depressing the two-lamp pushbutton.

Tech Spec LCO 3.2.4 Axial Shape Index states the ASI must be within the limits specified within the COLR above 25% RTP. There are both positive and negative values of ASI that are acceptable. The sub-divisions of the X axis are in 0.04 increments.

The provided table identifies the given conditions in the stem as in the Unacceptable Operation region.

There is no power level above 25% where an ASI of -0.56 is allowed.

PLP - 2018 NRC EXAM Page: 136 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 ARP-21 Rack A Window 6 - HIGH POWER RATE CHANNEL PRE-TRIP/ASI states the setpoint of the alarm is 0.011 ASI units prior to COLR Limit (effective above 15% power). One of the actions directed if power is >25%, is to check the power density status of each Thermal Margin Monitor channel on the System Status screen. (thus the ability to monitor ASI in the Control Room). The alarm actuates on 1/4 Bistable Trip Units input.

Distractors:

A. Start-up Rate = 1.5 DPM is plausible, because it is at the setpoint for the alarm, but it is only effective between 10e-4% and 15% power. Part two is plausible for those who have 'bypass' confused with 'reset'.

B. Start -up rate =1.5 DPM is plausible but incorrect as explained in A. Part two is correct.

D. ASI = -0.56 is correct, but part two is incorrect as explained in A.

KA: 012 RPS A4 Ability to manually operate and/or monitor in the control room:

(CFR: 41.7 / 45.5 to 45.8)

A4.02 Components for individual channels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3 3.4 Test item meets KA. Question tests what causes a specific channel indication (white high power rate pre-trip bistable unit light) and how is it operated to reset it. Answer requires ability to determine cause, and how to reset it.

MEMORY - recall setpoint, and method to reset.

objective: From memory, for the Reactor Protective System:

a. list the Control Room indications (G2.1.31)

b. describe the Control Room controls (G2.1.28) without error.

PLP - 2018 NRC EXAM Page: 137 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 38 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847154 User-Defined ID: 012A4.02 Cross Reference Number: RPS_CK08.0-20 (Reference provided) The plant is at 30% power, The white Topic:

high power rate pre-trip bistable unit Num Field 1:

Num Field 2:

Text Field:

Comments: Modified from 102860 (PAL-LOR-994) to include a two part question/answer, shown below:

The white high power rate pre-trip bistable unit light on the RPS panel for channel 'A' has illuminated. The power level is 27%.

Which ONE of the following could cause the bistable trip unit light to illuminate at this power level?

A. ASI = -0.55 B. Start-up rate = 1.5 DPM C. Start-up rate = 0.5 DPM D. ASI = -0.55 OR startup rate = 1.5 DPM Answer: A Changed the values of the ASI distractors and made the question 2 parts.

PLP - 2018 NRC EXAM Page: 138 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 139 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 39 ID: 847828 Points: 1.00 While the plant was at 100% power, PT-0102A, Pressure Transmitter that provides input to RPS and ESF for Pressurizer Pressure Channel A failed.

RPS Channel 'A' has been bypassed Now, the Plant is being cooled down and depressurized in preparation for a refueling outage:

The Actuation Logic Channels for Safety Injection Signal (SIS) have been manually BLOCKED per GOP-9, MODE 3 > 525°F to MODE 4 or MODE 5 CHECKLIST.

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

- A Channel - Failed Low

- B Channel - 1685 psia

- C Channel - 1685 psia

- D Channel - 1700 psia Assuming NO accidents, as the PCS cool down and depressurization continues and pressure is intentionally lowered to 100 psia, Safety Injection...

A. can NOT be actuated manually.

B. will automatically actuate on BOTH trains.

C. will automatically actuate on ONLY one train.

D. will NOT automatically actuate on EITHER train.

Answer: D Answer Explanation Answer: D.

D will NOT automatically actuate on EITHER train; is correct.

When 3 of 4 pressure channels rise to >1690 psia, the SIS rearms itself (unblocks). When 2 of 4 channels lower to <1605 psia, SIS will actuate. Actuating the trip channel bypass only bypasses the RPS signal and does NOT affect the ESF signal. Since only one channel (D) exceeded 1690, SIS did not auto rearm. Thus, it is still blocked, and pressure can be lowered without causing an actuation.

Reference PL-SIS r7 page 28-29 Distractor Analysis:

A. Incorrect. Manually actuating would work, a high containment pressure would automatically actuate SIS too. But with NO accidents stated in the stem, auto actuation is still blocked.

Reference:

E-17, Sheet 3, PL-SIS lesson plan.

B. Incorrect. Plausible for examinee who believes the bypassed channel and the channel that reached 1700 psia, make up the coincidence to let SIS rearm, and concludes both trains actuate.

PLP - 2018 NRC EXAM Page: 140 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 C. Incorrect. Plausible for examinee who believes the bypassed channel and the channel that reached 1700 psia, make up the coincidence to let SIS rearm, but thinks since only two channels were able to allow rearming, that only one train actuates.

KA:

013 ESFAS K4 Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following:

(CFR: 41.7)

K4.09 Spurious trip protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.7 3.1*

Test item meets KA. Question poses situation where examinee has to determine if SIS unblocks and rearms for a situation where one channel is bypassed. This knowledge is of a design feature or interlock that has specific coincidence requirements in order to clear the blocking signal that is preventing a spurious trip of the ESFAS system (actuation of SIS when it is not needed).

HIGH COG - answer requires comprehension of how SIAS is blocked and re-armed, and actuated, and a prediction of response given specific circumstances.

Objective:

Given that SIAS has been properly blocked, describe the consequences of PCS pressure rising above the block permissive setpoint without error.

PLP - 2018 NRC EXAM Page: 141 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 39 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847828 User-Defined ID: 013K4.09 Cross Reference Number: SIS_E04.03 While the plant was at 100% power, PT-0102A, Pressure Topic:

Transmitter that provides input to RPS and E Num Field 1: 006K4.11 Num Field 2:

Text Field:

PLP - 2018 NRC EXAM Page: 142 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Comments: Modified from LO-2017N33 2017 NRC Exam (745390).

changed answer, and distractors, and stem of question. Shown below:

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 risen 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 risen 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 risen 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.

Answer: A PLP - 2018 NRC EXAM Page: 143 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 144 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 40 ID: 849651 Points: 1.00 Which of the following Engineered Safeguards Features (ESF) Instrument Channels has a built-in bypass capability that uses a key operated switch (similar to the RPS Trip Channel Bypasses)?

A. Containment High Pressure (CHP)

B. Containment High Radiation (CHR)

C. Main Steam Isolation Signal (MSIS)

D. Auxiliary Feedwater Actuation Signal (AFAS)

Answer: D Answer Explanation Answer: D D. AFAS is correct Explanation:

The only ESF Instrument Channels with built in bypass capability listed to choose from are the Low SG Level AFAS bistables. Those bypasses are affected by a key operated switch, similar to the RPS Trip Channel Bypasses. A bypassed Low SG Level AFAS bistable can NOT perform its specified function and must be declared inoperable.

From SOP-12 Attachment 2 AFAS - SUBSYSTEM OPERATIONS Page 5 of 5 1.3 TO BYPASS A SENSOR CHANNEL NOTE: Normal AFAS actuation is 2 out of 4 logic. Only 1 Sensor Channel Bistable may be bypassed at any one time to make 2 out of 3 logic.

NOTE: 12V DC Power is required in the Sensor Channel to be bypassed. Bypass cannot be performed if Sensor's 12V DC breaker is OPEN, or 12V DC Power is unavailable.

a. At Panel EC-187, IDENTIFY which Sensor Channel Bistable is to be bypassed.

b. PLACE key switch in sensor cabinet for bistable to be bypassed to BYPASS AND CHECK bypass light above key switch on.

From TS Bases B.3.3.3:

ESF Instrument Channel Bypasses - the only ESF instrument channels with built-in bypass capability are the Low SG Level AFAS bistables. Those bypasses are effected by a key operated switch, similar to RPS Trip channel Bypasses.

Distractors:

All plausible as they are other ESF Functions. Some with unique features, but not having built in bypass capability.

A. CHP has two trains. Each train has an independent set of 4 pressure switches, and each train is independently actuated, and reset.

B. CHR has 4 rad monitors. 2/4 high rad signals actuates BOTH trains of CHR. There are separate selectable rad monitors for CHR during refueling.

C. MSIS cannot be bypassed.

PLP - 2018 NRC EXAM Page: 145 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 KA: 013 ESF 2.1.28 Knowledge of the purpose and function of major system components and controls.

(CFR: 41.7)

IMPORTANCE RO 4.1 SRO 4.1 Test item meets KA. The function of the key operated bypass switch for a specific ESF function must be known to answer the question.

MEMORY level

Reference:

Tech Spec bases, system level knowledge.

Objective:

From memory, explain how the Reactor Protective System (RPS) functionality is protected from a single failure of one of the following:

- monitored input parameter signal

- power supply

- RPS internal component in accordance with FSAR Chapter 7.2. (012 K4.01; 012 K4.04; 012 K4.05; 012 K6.01; 012 K6.02; 012 K6.03; 012 K6.06)

Question 40 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 849651 User-Defined ID: 013G2.1.28 Cross Reference Number: RPS_CK12.0 Which of the following Engineered Safeguards Features (ESF)

Topic:

Instrument Channels has a built-in byp Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 146 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 147 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 41 ID: 847311 Points: 1.00 Plant is at 100% power in mid-summer.

Outside air temperature is 98°F

Service Water Pumps P-7A and P-7B are in service

Service Water Pump P-7C is in Standby

EK-1132, SERVICE WATER PUMP P-7A BASKET STR HI dP, has annunciated

B' Critical Service Water header pressure is 49 psig

A Critical Service Water Header is 50 psig

Containment Cooling system components are in normal operating alignment with ALL components OPERABLE and available What action(s) should be taken?

A. Start the third Service Water Pump, P-7C B. Close CAC VHX-4 outlet Bypass, CV-0843 C. Open CAC VHX-4 outlet control valve, CV-0867 D. Start Containment Cooling Fans V-4A and V-4B Answer: A Answer Explanation Answer: A A. Start the third Service Water Pump, is correct.

Explanation:

A. CORRECT. If either critical header pressure is < 50 psig, starting a 3rd service Water pump is directed by SOP-5, 5.2.3.

Reference:

SOP-5, Containment Air Cooling and Hydrogen Recombining System, ARP-7 Distractors:

B. Plausible because it is in the procedure, but INCORRECT. Per SOP 5, 7.1.1 normal lineup, CAC VHX-4 outlet bypass, CV-0843, is failed closed.

C. Plausible because it is an action, but INCORRECT. Per SOP 5, 7.1.1 normal lineup, CAC VHX-4 outlet control valve, CV-0867, is already open.

D. Plausible because it is an action, but INCORRECT. Per SOP-5, V-4A and V-4B are already running as are all other CAC fans per SOP-5 normal lineups (SOP 5, 7.1.1).

022 CCS (Containment Cooling system)

K1 Knowledge of the physical connections and/or cause effect relationships between the CCS and the following systems:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.01 SWS/cooling system . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5 3.7 PLP - 2018 NRC EXAM Page: 148 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Test item meets KA. SWS is the cooling medium for the systems that cool containment, (that's the interconnection). Increasing the number of operating pumps when conditions require is the cause effect relationship.

HIGH COG - assess situation, determine action going forward.

Objectives:

From memory, explain the consequences of and how to mitigate operating the Containment Cooling System under the following conditions:

- Opening VHX-1, 2, 3, 4 inlet valves prior to opening the outlet valves

- Opening VHX-1, 2, 3 , 4 inlet valves prior to opening the containment isolation valves (CV-0847, CV-0824)

- Operating with containment temperature outside the maximum or minimum allowed values (022 A1.01)

- All high capacity outlet valves open and operating with Critical SW Header pressure less than 50 psig (mitigation only)

- All high capacity outlet valves open and operating with both Critical SW Header pressures greater than 70 psig (mitigation only) in accordance with SOP-5. (G2.1.32). CAIR_CK16.0 Question 41 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847311 User-Defined ID: 022K1.01 Cross Reference Number: CAIR_CK16.0-4 Plant is at 100% power in mid-summer. Outside air Topic:

temperature is 98?F Service Water Pumps P-7A Num Field 1: 076G2.1.32 Num Field 2:

Text Field:

Comments: Bank 99467 , Not used on 2014 or 2017 NRC Exams PAL-LOR-3776 PLP - 2018 NRC EXAM Page: 149 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 150 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 42 ID: 847174 Points: 1.00 The plant is at 100% power, when a transient occurred resulting in an SIAS and a Loss of Offsite Power.

At time 1000:00,

ONLY the DBA Sequencer from the SIS RIGHT channel starts operating

Containment Pressure is 10 psig With NO operator action, which Containment Spray (CS) Pump(s) is / are operating at 1006:00?

A. CS pump P-54A ONLY B. CS pumps P-54B and P-54C ONLY C. CS pumps P-54A and P-54C ONLY D. CS pumps P-54A, P-54B and P-54C Answer: A Answer Explanation Answer: A CS pump P-54A ONLY.

Explanation /

References:

The loss of offsite power causes the load shed of the 1C and 1D 2400 volt buses. Even if the CS pumps were running before that, they have been shed from their buses before the DBA sequencers can operate.

At 1000:00 the SIS RIGHT Channel DBA sequencer is working. Since it is working, that means DG 1-2 is carrying Bus 1D.

With a DBA sequencer operating, and a CHP (4.0 psig in Cnmt), P-54A gets a start signal from the operating DBA sequencer at the 2 second point. P-54A is loaded onto 2400 Volt Bus 1D.

P-54 B and P-54C, are powered from bus 1C, but that sequencer is not working, and they are NOT sequenced on the bus. Since the LEFT sequencer never started working P-54B and P-54C are NOT running at 1006:00.

Logic Diagram Safety Injection Actuation E-17, sht 4 provides the timing of the sequencers.

Single Line Meter & Relay Diagram E-1 Sht 1 shows which buses carry which loads.

Distractors:

B. P-54B and P-54C only is plausible for those who get their buses and channels reversed. There are 'A',

'B' and 'C' loads on each of the 2400V buses.

C. P-54A and P-54C only is plausible for those who get their buses wrong.

D. P-54A, P-54B, and P-54C is plausible for those who get their buses, channels and sequencer actuations wrong (thinking either sequencer will cause either train to go).

KA: 026 K2.01 PLP - 2018 NRC EXAM Page: 151 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 K2 Knowledge of bus power supplies to the following:

(CFR: 41.7)

K2.01 Containment spray pumps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4* 3.6 Test item meets KA. Question poses a situation where a loss of power occurs. Answer requires knowing which emergency power is provided, and how it is made available.

MEMORY - recall which CS pumps on which sequencer Objective: From memory, for the following Containment Spray System major components

- Containment Spray Pumps (P-54 A/B/C)

- Containment Spray Valves (CV-3001/3002)

- Containment Spray Headers

- Containment Sump

a. describe the operational design of each component

b. describe the normal range of the component (n/a for valves, headers) in accordance with the FSAR and DBDs. (G2.1.28)

Question 42 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 847174 User-Defined ID: 026K2.01 Cross Reference Number: CSS_CK02.0 The plant is at 100% power, when a transient occurred Topic:

resulting in an SIAS and a Loss of Offsite P Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 152 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 153 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 43 ID: 847312 Points: 1.00 The Plant is at 100% power.

Due to a failure of the input to PIC-0511, Turbine Bypass Valve controller, CV-0511, Turbine Bypass Valve (TBV), has been closed by placing PIC-0511 in MANUAL with a zero output signal

HIC-0780A, Steam Dump controller is in AUTO

The TBV is NOT isolated yet If a Reactor Trip were to occur now, the quick open feature would cause ___(1)___ to open.

A. ONLY the TBV B. ONLY the ADVs C. BOTH the ADVs and the TBV D. NEITHER the ADVs NOR the TBV Answer: C Answer Explanation Answer: C C. BOTH the ADVs and the TBV Quick Open.

Explanation:

Reference(s): DBD 1.09, FSAR 10.2.1, FSAR 7.5, Even with the PIC-0511 controller in Manual, quick open is still available. If T AVE 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. Since the TBV has not yet been isolated, it will respond to the trip by functioning in quick open.

Distractors:

Distractors are all plausible for examinees that don't know how the controllers work, or respond in auto or manual.

Distractor A: Plausible for the examinee that does not know which controller controls which valve(s)

Distractor B: Plausible for the examinee that does not know which controller controls which valve(s)

Distractor D: Plausible for the examinee that does not know which controller controls which valve(s), nor how the control circuit is set up.

KA: 039 Main and Reheat Steam System (MRSS)

K3 Knowledge of the effect that a loss or malfunction of the MRSS will have on the following:

(CFR: 41.7 / 45.6)

K3.06 SDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.8* 3.1 Test item meets KA. Question poses a loss or malfunction of a component of the Main Steam System (TBV Controller) and the answer requires knowledge of how that loss affects operation of the Steam Dump system after a reactor trip from full power.

PLP - 2018 NRC EXAM Page: 154 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 HIGH COG - predict how a failure impacts operation.

Objective:

From memory, describe the consequences of operating the Main Steam system under the following conditions:

- Loss of power

- High steamline radiation (039 A2.03)

- Safety valve leakage (039 A2.02)

- Malfunctioning TBV or ADV (039 A2.04)

- Loss of instrument air (041 A2.03) without error.

Question 43 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847312 User-Defined ID: 039K3.06 Cross Reference Number: MSS_CK16.0-2 The Plant is at 100% power. Due to a failure of the input to Topic:

PIC-0511, Turbine Bypass Valve contr Num Field 1:

Num Field 2:

Text Field:

Comments: Modified from Bank to reduce to one part - not used on 2014 or 2017 NRC exam PAL-LOR-2710 (709614, 98307)

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EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 156 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 44 ID: 847313 Points: 1.00 Given the following:

The Plant is at 100% power

A steam line break occurs OUTSIDE the Containment Building UPSTREAM of CV-0510, 'A' S/G MSIV

'A' S/G Pressure indicates 475 psia

'B' S/G Pressure indicates 850 psia Which of the following describes the expected response of the MSIVs and Feed Regulating Valves (FRVs) to this event?

A. BOTH 'A' and 'B' S/G MSIVs close.

BOTH 'A' and 'B' S/G FRVs close.

B. BOTH 'A' and 'B' S/G MSIVs close.

ONLY 'A' S/G FRV closes.

C. ONLY 'A' S/G MSIV closes.

BOTH 'A' and 'B' S/G FRVs close.

D. ONLY 'A' S/G MSIV closes.

ONLY 'A' S/G FRV closes.

Answer: B Answer Explanation Answer: B B. BOTH 'A' and 'B' S/G MSIVs close.

ONLY 'A' S/G FRV Closes.

Reference)(s):

M-207, sheet 1 Pl-MSS r7 page 12:

PLP - 2018 NRC EXAM Page: 157 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1

b. Main Steam Isolation Signal (MSIS)

1) 2/4 low steam generator pressure at 500 psia on one S/G close both MSIVs.

2) Four channels of Steam Generator Pressure provide a signal for an automatic closure to the MSIVs on low S/G pressure (MSIS).

3) 2/4 low steam generator pressure at 500 psia on one S/G close the FW Regulating Valve and the FW Reg Valve Bypass Valve for the affected S/G (affected S/G only).

a) E-50A, Steam Generator (1) CV-0701, E-50A FW Regulating Valve (2) CV-0735, E-50A FW Reg Valve Bypass Valve b) E-50B Steam Generator (1) CV-0703, E-50B FW Regulating Valve (2) CV-0734, E-50B FW Reg Valve Bypass Valve 2008 NRC RO EXAM DISTRACTOR ANALYSIS

a. Plausible if the student believes that all valves close similar to a CHP event.

b. CORRECT - S/G pressure is a diagnostic indication in determining entry into EOP-6.0, ESDE.

c. Plausible if the student misapplies the concept that only the "A" FRV closes.

d. Plausible if the student believes that the goal of a MSIS on low S/G pressure is to isolate only the affected S/G.

KA:

039 MRSS A3 Ability to monitor automatic operation of the MRSS, including:

(CFR: 41.5 / 45.5)

A3.02 Isolation of the MRSS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 3.5 Test item meets KA. Question poses a steam pressure, and answer requires knowledge of MSIS and FWIS.

MEMORY Objective: From memory, describe the design features and interlocks that provide the following Main Steam system functions: - MSIV auto closure (039 K4.05) without error.

PLP - 2018 NRC EXAM Page: 158 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 44 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 847313 User-Defined ID: 039A3.02 Cross Reference Number: MSS_CK09.0-22 Given the following: The Plant is at 100% power A steam line Topic:

break occurs OUTSIDE the Containment Num Field 1:

Num Field 2:

Text Field:

Comments: Bank NOT used on 2014 or 2017 NRC exam PAL-LOR-4147 (99854, 709571, 739411)

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EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 160 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 45 ID: 847310 Points: 1.00 A plant power escalation is in progress, with the following alignments:

LIC-0701/LIC-0703, S/G Level Indicating Controllers are in AUTO

BOTH MFW pumps are at 3900 RPM each

HIC-0525, Feedwater Control Mode Selector, is in MANUAL

HIC-0526/HIC-0529, MFW Pump Speed Controllers are in MANUAL (Controller Span 0-6000 RPM)

The red pointers and the blue pointers on the HIC-0526 and HIC-0529 are matched

SG levels are stable at program value

Feedwater pump discharge pressure is 160 psi greater than Steam Generator (S/G) pressure On an Individual Speed Controller, should the PF light be flashing? If NOT, what action is necessary to cause it to start flashing?

A. YES; The PF light should be flashing.

B. NO; LOWER Main Feed Pump Speed C. NO; RAISE Main Feed Pump Speed D. NO; RAISE FW/SG Delta Pressure Answer: A Answer Explanation Answer: A.

A. YES; The PF light should be flashing.

Explanation:

Got to have actual feed pump speed within 0.5% of demanded speed. Demanded speed and actual speed are matched as indicated by the red and blue pointers being matched.

Being within 0.5% is the minimum speed before the PF light will illuminate, and the transfer to Cascade can be accomplished by depressing the A button on the respective individual speed controller.

Do not have to change speed. The PF light should be flashing. No action required to cause it to start flashing.

Distractors:

B. Lowering speed is plausible because it is an action in the applicable procedure, but it is for matching speeds, not getting the PF light to flash.

C. Raising speed is plausible because it is also an action that is normally needed, but not in this case given the indications in the stem.

D. STM/FW DP is NOT a requirement to transfer to CASCADE, but it is a requirement for transferring from FR b/p V to FRV controllers. (common misconception)

Reference:

SOP-12,

Title:

Feedwater System, Rev 80 7.6.4 To Transfer Between Feedwater Pump Speed Control Modes PLP - 2018 NRC EXAM Page: 161 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1

a. General Information

1. HIC-0525, Feedwater Control Mode Selector has two operating modes: MANUAL and CASCADE.

(a) In MANUAL, the HIC operating button provides the cascade set point to HIC-0526, K-7A Speed Control, and HIC-0529, K-7B Speed Control this is indicated on the process (red) pointer and the output indicator.

(b) To transfer into CASCADE, the output must be moved to be within 0.5% of speed demand signal from LIC-0701 and LIC-0703. Then AUTO Button is depressed and controller transfers directly to CASCADE and speed demand from LIC-0701 and LIC-0703 is passed directly to the Individual Speed Controllers.

(c) MANUAL signal on HIC-0525 tracks the CASCADE signal and allows bumpless transfer from CASCADE to MANUAL to be performed without balancing.

(d) The following are HIC-0525 signal adjustment characteristics:

HIC-0526 (HIC-0529) has a very slow response time in CASCADE. Sufficient time for the system to respond must be allowed between adjustments on HIC-0525.

HIC-0525 signal only has an effect on Main Feedwater Pump speed between 54.2% (minimum speed, 3250 rpm) and 87.5% (maximum speed, 5250 rpm).

(e) Placing HIC-0525, Feedwater Control Mode Selector, in CASCADE with only one Main Feedwater Pump in service is only recommended if power will be raised to greater than or equal to 55% on one Main Feedwater Pump.

2. HIC-0526, K-7A Speed Control, and HIC-0529, K-7B Speed Control, have three operating modes: MANUAL, AUTOMATIC, and CASCADE.

(a) In MANUAL, the manual operating button will provide direct control of pump speed.

(b) AUTOMATIC mode is used for the ramp down feature of the controllers.

(c) In CASCADE, speed demand from HIC-0525, Feedwater Control Mode Selector, will be passed to the individual pumps.

(d) The MANUAL signal on HIC-0526 (HIC-0529) tracks the CASCADE signal and allows a bumpless transfer from CASCADE to MANUAL to be performed without balancing.

(e) HIC-0526 (HIC-0529), K-7A (K-7B) Speed Control, will NOT transfer to CASCADE from MANUAL unless the setpoint and the actual turbine speed are within 0.5% of each other and PF light is flashing.

(f) HIC-0526 (HIC-0529), K-7A (K-7B) Speed Control, shows actual turbine speed on the process (Red) pointer and the speed demand from HIC-0525, Feedwater Control Mode Selector, is displayed on the setpoint (Blue) pointer.

059 Main Feedwater K4 Knowledge of Main Feedwater System design feature(s) and/or interlock(s), which provide for the following:

(CFR: 41.7)

K4.05 Control of speed of MFW pump turbine 2.5* 2.8*

Question tests objective and meets KA. Answer requires comprehension of how the FWP and SGWLC systems work, and how shifting from one mode of operation to another which requires knowledge of how speed is controlled in the various modes of operation is accomplished.

HIGH COG - answer requires comprehension of how the circuit works (inputs and expected outputs and indications).

PLP - 2018 NRC EXAM Page: 162 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Objectives:

1. Describe the design features and interlocks that provide the following Steam Generator Water Level Control system functions:

_ Steam Generator Level Control (035 K4.01)

_ Main Feedwater Reg Valve (059 K4.08)

_ Main Feedwater Pump Turbine speed control (059 K4.05)

_ Automatic Feedwater reduction on a plant trip (059 K4.18) without error. SGWL_CK09.0

2. For the following Steam Generator Water Level Control system major components, (035 K4.01, 035 K4.02, 059 K4.02, 059 K4.05, 059 K4.18, G2.1.28)

_ Main Feed Reg Valve Level Indicating Controllers, LIC-0701, LIC-0703

_ Main Feed Reg Bypass Valve Level Indicating Controllers, LIC-0734, LIC-0735

_ Combined Speed Hand Indicating Controller, HIC-0525

_ Main Feed Pump Speed Hand Indicating Controllers, HIC-0526, HIC-0529

_ Main Feed Reg Valves, CV-0701, CV-0703

_ Main Feed Reg Bypass Valves, CV-0734, CV-0735

_ Wide Range Level Indicators LI-0758A, LI-0758B, LI-0757A, LI-0757B a) describe the operational design of each component b) describe the normal operating range of the component without error SGWL_CK02.0 Question 45 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847310 User-Defined ID: 059K4.05 Cross Reference Number: SGWL_CK02.0 A plant power escalation is in progress, with the following Topic:

alignments: LIC-0701/LIC-0703, S/G Lev Num Field 1: 059A4.03 Num Field 2: 3.9 Text Field: 1122 rev 3 Comments: Bank LO-MFW CONTROLLERS 4 PLP - 2018 NRC EXAM Page: 163 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 164 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 46 ID: 847236 Points: 1.00 Given the following conditions:

EOP-7.0, 'Loss of All Feedwater', actions are in progress

E-50A, 'A' S/G level is -130% and lowering at 2% / minute

E-50B, 'B' S/G level is -80% and lowering at 2% / minute Feedwater flow has just become available, but due to valve control issues, is available ONLY at flow rates of 350 gpm or more.

Complete the following statements:

Feeding at flow rates of 350 gpm should be commenced to S/G ___(1)___.

The impact of feeding the OTHER S/G at 350 gpm is ___(2)___.

(1) (2)

A. A causing significant tube bundle damage B. B causing significant tube bundle damage C. A challenging the PCS cooldown limits D. B challenging the PCS cooldown limits Answer: B Answer Explanation Answer: B B. B, causing significant tube bundle damage; is correct.

Explanation /

Reference:

When S/ G level is lower than -125%, the S/G is considered dry. Feeding is allowed to a dry S/G, but at lower flowrates to avoid significant tube bundle damage.

EOP-7.0, LOAF, just prior to step 8, has a CAUTION that states:

Feedwater flow restoration to a dry S/G (level less than -125%) may cause significant tube bundle damage. Limit feed flow to less than 300 gpm for any S/G with level less than -84%.

The EOP-7.0 Bases document for steps to restore feed state that the S/Gs are rated for a number of cycles of feeding 32°F feedwater with the secondary side dry and at 600°F, provided initial feedwater flow is less than 300 gpm per S/G for up to eight cycles. (FSAR 4.3.4)

Thus the 'B' S/G can be fed at 350 gpm, but the 'A' S/G can NOT, without the risk of significant tube bundle damage resulting from feeding the 'A' S/G.

Distractors:

PLP - 2018 NRC EXAM Page: 165 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 A. A, causing tube bundle damage is plausible but incorrect. Part one is plausible for those who don't know the limit stated in the Caution and believe recovering level in the S/G with the lowest level would help the most. Part two is correct.

C. A, challenging the PCS cooldown rates is plausible but incorrect. Part one is plausible but incorrect as explained in A. Part two is plausible for those who think that a large flow rate would cause a large cooldown of the PCS. Part two is incorrect because the PCS is heating up due to the loss of S/G level and its ability to remove heat from the PCS.

D. B, challenging the PCS cooldown rates is plausible but incorrect. Part one is correct as explained in the answer. Part two is plausible but incorrect as explained in C.

KA: 059 Main Feedwater A2 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:

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.04 Feeding a dry S/G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.9* 3.4*

Test item meets KA. Question poses loss of feedwater (malfunction of MFW) and impact of feeding a dry S/G. Answer requires knowledge of impact of feeding dry S/G, procedure dictates allowable feeding rates.

HIGH COG - requires comparison of available feed flow rate to what's allowed and to which S/G (application).

Objective:

Given emergency conditions, determine if there is adequate heat removal from the PCS via a Steam Generator in accordance with EOP-7.0 or EOP-9.0, HR-3.

Question 46 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847236 User-Defined ID: 059A2.04 Cross Reference Number: TBAE_E01.01 Given the following conditions: EOP-7.0, 'Loss of All Topic:

Feedwater', actions are in progress E-50A, '

Num Field 1:

Num Field 2:

Text Field:

Comments: NEW PLP - 2018 NRC EXAM Page: 166 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 167 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 47 ID: 847257 Points: 1.00 The Plant is in MODE 1.

The Technical Specification volume requirement of LCO 3.7.6 Condensate Storage and Supply is met if there is a minimum of at least ___(1)___ gallons of useable volume in the designated tank(s).

Meeting this volume requirement ensures the AFW system can supply the necessary water flow at a rate sufficient to cooldown the plant after a trip from MODE 1 with a loss of offsite power to SDC entry requirements within ___(2)___ hours.

(1) (2)

A. 100,000 4 B. 100,000 8 C. 250,000 4 D. 250,000 8 Answer: B Answer Explanation Answer: B B. 100,000; 8 is correct.

Explanation /

References:

Tech Spec LCO 3.7.6, Condensate Storage and Supply states:

The combined useable volume of the CST and PMST (T-81) shall be > 100,000 gallons.

The Tech Spec Bases for B3.7.6 states:

To satisfy accident analysis assumptions, the CST and T-81 must contain sufficient cooling water to remove decay heat for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following a reactor trip from 2580.6 MWth. This amount of time allows for cool down of the PCS to SDC entry conditions, assuming a coincident loss of offsite power and the most adverse single failure. In doing this the CST and T-81 must retain sufficient water to ensure adequate NPSH for the AFW pumps and make up for steaming required to remove decay heat.

Distractors:

A. part one is correct. Part two is plausible because it is the completion time for not meeting the TS volume requirements and less than the most restrictive completion time of actions requiring the plant to be moved below mode 3 (which is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ).

C. 250,000 is plausible as that is a volume requirement for actions to be taken regarding the SIRW tank.

D. Plausible as explained above.

KA: 061 AFW K5 Knowledge of the operational implications of the following concepts as the apply to the AFW:

PLP - 2018 NRC EXAM Page: 168 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 (CFR: 41.5 / 45.7)

K5.02 Decay heat sources and magnitude . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2 3.6 Test item meets KA. Question asks about a TS required volume to support decay heat removal assumptions (an operational implication) via the source of water for AFW.

MEMORY level.

Objective:

From memory, describe the following for the Main Condenser, Condensate and Feedwater system in accordance with Technical Specification 3.7.3 and 3.7.6. (G2.2.22)

a. LCO Statement (3.7.3 and 3.7.6)

b. Applicability (3.7.3 and 3.7.6)

Question 47 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 847257 User-Defined ID: 061K5.02 Cross Reference Number: CDFW_CK20.0 The Plant is in MODE 1. The Technical Specification volume Topic:

requirement of LCO 3.7.6 Condensate St Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 169 of 369 15 January 2020

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EXAMINATION ANSWER KEY Daft N 1 48 ID: 848150 Points: 1.00 The plant is at 90% power.

Preferred AC Bus EY-30 is to be placed on the Bypass Regulator in accordance with SOP-30, Station Power.

For this evolution, complete the following statements:

In accordance with this procedure Emergency Diesel Generator ___(1)___is to be considered INOPERABLE.

Although NO power interruption to EY-30 is expected to occur, if EY-30 did momentarily lose power then RPS Breaker 42-2 ___(2)___ trip.

(1) (2)

A. 1-1 would NOT B. 1-1 would C. 1-2 would NOT D. 1-2 would Answer: A Answer Explanation Answer: A A. 1-1, would NOT; is correct Explanation:

Per the reference, the associated Sequencer and EDG are to be considered INOPERABLE.

Reference:

SOP-30 rev 91, page 68, and Attachment 2. page 5/11:

Possible Effects of De-energizing and subsequently re-energizing/transferring a Preferred AC Bus:

Actions to take Prior to /After De-energizing:

If closed, THEN RPS Breaker 42-1 should go to the TRIPPED position when EY-30 is de-energized, and subsequently re-energized creating a half trip condition in the RPS, but NOT a reactor trip.

A momentarily loss of EY-40 would cause RPS Breaker 42-2 to trip.

Distractors:

All plausible for those that don't know the power supplies, or the notes in the procedures. All incorrect because at least one part is from the unaffected train (except D - both parts are from the unaffected train.)

1-2 is plausible for the examinee that doesn't recall which EDG is the emergency power supply for the division having EY-30.

'would' is plausible for the examinee that doesn't know which RPS Breaker is powered from which preferred AC bus; or doesn't comprehend how a single power source would not affect both RPS breakers.

PLP - 2018 NRC EXAM Page: 171 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 KA: 062 AC Distribution A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including:

(CFR: 41.5 / 45.5)

A1.03 Effect on instrumentation and controls of switching power supplies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5 2.8 Test item meets KA. Answer requires system knowledge and knowledge of the procedure used to change sources of power to a vital AC instrument bus. The NSD and DBA Sequencers for loading the EDGs need the instrument bus power to be operable. The Sequencers are 'controllers' for sequencing the loads on to the safety buses. Even though the question asks about an 'inoperability', it is RO level, because the inoperability is stated in a NOTE in the procedure. The RO is not making the operability determination, its already made.

MEMORY Objective:

From memory, describe the consequences of operating the 125V DC and AC Power system under the following conditions:

- Grounds (062 A2.02, 063 A2.01)

- Improper inverter manipulations (062 A2.03)

- Exceeding voltage limitations (062 A2.08)

- Exceeding current limitations (062 A2.09)

- Restoration of power to a system with a fault on it (062 A2.12)

- Switching power supplies on instruments and controls (062 A2.10)

- Degraded system voltages (062 A2.16)

- Loss of ventilation during battery charging (063 A2.02)

- DC Bus tie breakers open (062 A2.16)

- Abnormal cable spreading room temperature (063 A2.02) without error.

Question 48 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848150 User-Defined ID: 062A1.03 Cross Reference Number: EPS_CK16.0 The plant is at 90% power. Preferred AC Bus EY-30 is to be Topic:

placed on the Bypass Regulator in accor Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 172 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 173 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 174 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 49 ID: 851907 Points: 1.00 The Plant has just been synchronized to the grid with power at 10% when a DC Bus ground occurs.

The 250 amp fuses supplying ED-21-2, 125 VDC Distribution Panel, blow and de-energize ED-21-2.

Given these conditions, complete the following statements:

___(1)___ CLOSES.

The required action(s) is/are to ___(2)___.

A. (1) CV-1359, Non-critical Service Water Isolation (2) Trip the Reactor B. (1) CV-1359, Non-critical Service Water Isolation (2) Trip the Turbine ONLY C. (1) CV-2099, PCP Controlled Bleedoff Containment Isolation (2) Trip the Reactor and all Primary Coolant Pumps D. (1) CV-2099, PCP Controlled Bleedoff Containment Isolation (2) Verify Controlled Bleedoff pressure is < 140 psig Answer: A Answer Explanation Answer: A A. CV-1359 fails closed, Trip the Reactor Explanation:

AOP-17, Loss of 125V DC Panel(s) Reactor Trip Criteria includes for a Loss of ED-21-2:

CV-1359 fails closed.

Ensuring the Reactor is Tripped, and Ensuring the Turbine is tripped, are stated as procedure step 2.

a. CORRECT - CV-1359 is a fail safe on loss of power valve and it receives power from D-21-2 and AOP-35 trip criteria state to trip the reactor if CV-1359 closes.

b. Plausible because for most off-normal events when plant power is <15%, only a turbine trip is required, however, for a loss of service water, a reactor trip is required. Since the reactor is manually tripped, that causes the Turbine to trip too. So there is no additional action to take to 'trip the turbine.' It is also plausible because the closure of the CV-1359 basically stops cooling to vital parts of the main generator. Tripping the turbine is necessary per AOP-35, Loss of Service Water.

c. Plausible because this is the plant response if DC panel D21-1 is lost. AOP-29, PCP Abnormal Conditions step 14 directs stopping all PCPs if Controlled Bleedoff not available.

d. Plausible because this is the plant response if DC panel D21-1 is lost. AOP-29, PCP Abnormal Conditions directs maintaining bleedoff pressure at 50-100 psig, if CV-2099 is closed.

PLP - 2018 NRC EXAM Page: 175 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1

References:

AOP-17 reactor trip criteria KA: 063 DC Electrical Distribution A2 Ability to (a) predict the impacts of the following malfunctions or operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.01 Grounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5 3.2*

Test item meets KA. Question poses a loss of a DC bus due to a ground, and the answer requires procedurally driven actions to address the loss of the bus. Selecting the correct answer requires prediction of the impact of a loss of DC due to grounds and using the procedure direction to mitigate the consequences.

MEMORY level Question 49 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 851907 User-Defined ID: 063A2.01 Cross Reference Number: IOTF_CK05.0-38 The Plant has just been synchronized to the grid with power at Topic:

10% when a DC Bus ground occurs. T Num Field 1:

Num Field 2:

Text Field:

Comments: Bank Not used on 2014 or 2017 NRC exams PAL-LOI-7719 PLP - 2018 NRC EXAM Page: 176 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 177 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 50 ID: 848151 Points: 1.00 The Plant is at full power:

Fuel Oil Transfer Pump P-18B is tagged out for maintenance

Due to Chemistry sampling, D/G 1-2 Fuel Oil Day Tank T-25B level has lowered

Fuel Oil Transfer Pump P-18A has automatically started and is pumping fuel oil to T-25B While this fuel oil transfer is in progress the following occurs:

A Loss of All Offsite Power occurs

D/G 1-1 FAILS to automatically start and will NOT start manually

D/G 1-2 starts and loads per design Given these conditions, complete the following statement:

Fuel oil transfer to T-25B A. Is occurring by gravity feed ONLY.

B. Occurs ONLY after manually re-energizing MCC-2 C. Occurs ONLY after manually re-energizing MCC-8.

D. Automatically occurs as part of the D/G 1-2 loading sequence.

Answer: C Answer Explanation Answer: C Occurs ONLY after manually re-energizing MCC-8.

Per the reference, neither fuel transfer pump auto sequences on, which makes D incorrect.

MCC-EB01 is the power source for P-18B, MCC-8 is the power source for P-18A.

Reference:

DBD-5.06 page 21:

The fuel oil transfer pump motors and their power and control circuits are non-Class 1E. These circuits are supplied from redundant channels but are not separate and independent. Neither transfer pump's control logic gives the emergency generator day tank makeup systems preference over or isolates those non-essential makeup systems which are also served. The arrangement of the fuel oil transfer pump power and control circuits requires operator action to establish fuel transfer via either pump during loss of off-site power conditions. This arrangement minimizes the potential for inadvertent fuel oil transfer to system ruptures caused by natural phenomena.

Transfer pump P-18A can be controlled either manually, or automatically via level controls. It is powered from non-Class 1E MCC-8, which is capable of being manually loaded onto EDG 1-2. Those portions of the P-18A control circuit which carry the demand signals from the two independent emergency generator day tank makeup control systems are not electrically isolated or separated from each other; nor are they isolated or separated from circuits for non-essential makeup systems.

PLP - 2018 NRC EXAM Page: 178 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Transfer pump P-18B can only be controlled manually. It is powered from Class 1E MCC EB01, which is automatically loaded onto EDG 1-1. Those portions of the P-18B control circuit which carry the demand signals from the two independent emergency generator day tank makeup control systems are not electrically isolated or separated from each other; nor are they isolated or separated from circuits for non-essential makeup systems.

ARP-3 EK-0554:

DIESEL GEN DAY TANK T-25A HI-LO LEVEL NOTE: MCC-8 (P-18A) is load shed and requires manual reclosing of breaker 52-1201 (Panel EC-04) to reenergize MCC-8.

Distractor Analysis:

A. Is occurring by gravity feed ONLY is plausible because another system (Boric Acid) can gravity feed under some circumstances but is incorrect because the distance from the storage tank (T-10) to the day tank (T-25) is to far for gravity fill ONLY to work.

B. Occurs ONLY after manually re-energizing MCC-2 is plausible for the examinee that doesn't know the power supplies to the FO Xfer pumps. It is incorrect, because MCC-2 cannot power the P-18A.

C. Correct D. Automatically occurs as part of the DG 1-2 loading sequence is plausible because many loads do auto-sequence on the DG but is incorrect because the MCC-8 doesn't.

KA 064 EDG K2 Knowledge of bus power supplies to the following:

(CFR: 41.7)

K2.02 Fuel oil pumps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.8* 3.1 Test item meets KA. Knowledge of the power supply for fuel oil transfer pump for the EDG is tested by the question.

MEMORY - recall power supply and sequenced loads for P-18A during LOOP.

Objective: From memory, explain the purpose of the interfaces (physical connections) between the Emergency Diesel Generator System and other plant systems in accordance with DBD-5.01. (064 K1.01, 064 K1.02, 064 K1.03, 064 K1.04)

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EXAMINATION ANSWER KEY Daft N 1 Question 50 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848151 User-Defined ID: 064K2.02 Cross Reference Number: EDG_CK07.0-2 The Plant is at full power: Fuel Oil Transfer Pump P-18B is Topic:

tagged out for maintenance Due to Chem Num Field 1:

Num Field 2:

Text Field:

PLP - 2018 NRC EXAM Page: 180 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Comments: PAL-LOR-670 Modified from question 50 on 2017 NRC exam shown below:

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 and similar to 843616 (LOI2020 Audit):

Given the following conditions:

Plant is at 100% power

Fuel Oil Transfer Pump P-18A is tagged out for maintenance

Due to Chemistry sampling, T-25B, D/G 1-2 Fuel Oil Day Tank, level has lowered

Fuel Oil Transfer Pump P-18B is pumping fuel oil to T-25B While this fuel oil transfer is in progress the following occurs:

A Loss of All Offsite Power

D/G 1-1 FAILS to start automatically or manually

D/G 1-2 starts and loads per design.

When D/G 1-2 loading sequence is complete, fuel oil transfer to T-25B will resume A. as a result of D/G 1-2 loading as designed.

B. ONLY if MCC-2 is reenergized by manual action.

C. ONLY if MCC-8 is reenergized by manual action.

D. ONLY as a result of gravity feed from T-10A, Fuel Oil Tank.

Answer: D PLP - 2018 NRC EXAM Page: 181 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 182 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 183 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 51 ID: 850890 Points: 1.00 Emergency Diesel Generator (EDG) 1-1, is operating in PARALLEL with its bus for the monthly surveillance.

ALL the switches are appropriately aligned in accordance with the governing procedure(s).

One of the loads on the bus trips Given these conditions, complete the following statements:

Immediately after the load trips,

the Speed of EDG 1-1 __(1)___.

the Voltage of EDG 1-1 ___(2)___.

(1) Speed (2) Voltage A. rises rises B. rises lowers C. lowers rises D. lowers lowers Answer: A Answer Explanation Answer: A A. rises, rises; is correct Explanation:

The purpose of the Speed Droop circuit is to automatically adjust the speed setting to a preset value based on generator load. This prevents excessive loading due to fluctuations in grid frequency. The more load the EDG carries, the higher the base speed setting for no load conditions. The reduction of load on the EDG causes the EDG to speed up, until the speed droop takes over to lower the speed. (frequency)

The purpose of the Voltage Droop circuit is to prevent the EDG from carrying too much reactive load. If it doesn't work, then a loss of load would raise the reactive load being carried because the EDG would become more overexcited. (VARS)

Distractors:

Plausible for those who don't understand the control circuits.

B. Part one is correct as explained in A. Part two is plausible for the examinee that thinks the loss of load would cause the excitation voltage to lower, resulting in lower VARS. Part two is incorrect as explained in the answer.

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EXAMINATION ANSWER KEY Daft N 1 C. Lowers is plausible for the examinee that thinks the reduction in load, with no speed droop results in lower frequency. Part one is incorrect as explained in the answer. Part two is correct as explained in the answer.

D. Part one is plausible but incorrect as explained in C. Part two is plausible but incorrect as explained in B.

Reference:

PL-EDG r11 (3) To prevent excessive real loading and add stability to the control system, the electronic governor operates with speed droop while in the Parallel Control Mode.

(a) Speed droop is set internally in the EGA.

(b) Speed droop automatically adjusts the speed setting to a preset value based upon generator output (load).

(c) This prevents excessive loading (or motoring) due to fluctuations of grid frequency.

(4) To prevent excessive reactive loadings and add stability to the control system, the voltage regulator operates automatically with voltage droop while in the Parallel Control Mode.

The voltage droop mode prevents the generator from carrying too much reactive load in the over-excited or under-excited mode with respect to the infinite bus (grid).

KA: 064 EDG A3 Ability to monitor automatic operation of the ED/G system, including:

(CFR: 41.7 / 45.5)

A3.05 Operation of the governor control of frequency and voltage control in parallel operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.8 2.9 Test item meets KA. Answer requires knowledge of what happens to control voltage and frequency of an operating loaded EDG.

HIGH COG - requires comprehension of the speed and voltage droop circuits work, and a prediction of outcome.

Objective: From memory, describe the design features and interlocks that provide the following Emergency Diesel Generator functions

- Emergency Diesel Generator Output Breaker Interlocks and Trips (064 K4.02)

- Governor Operation (064 K4.03)

- Overload ratings (064 K4.04)

- Incomplete or Cranking Failure Start (064 K4.05)

- Speed Droop control (064 K4.06)

- NSD Sequencer (064 K4.10)

- DBA Sequencer (064 K4.11) in accordance with DBD-5.01 & FSAR 8.4.1, Drawings E-17 Sheets 4, 9, 10, 11, 12, 13, and E-209 Sheet 2.

PLP - 2018 NRC EXAM Page: 185 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 51 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 850890 User-Defined ID: 064A3.05 Cross Reference Number: EDG_CK09.0 Emergency Diesel Generator (EDG) 1-1, is operating in Topic:

PARALLEL with its bus for the monthly survei Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 186 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 187 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 52 ID: 848152 Points: 1.00 Digital Rate meters in the Process Liquid Monitor and Gaseous Process Monitoring Systems ___(1)___

have a CHECK SOURCE Function.

When the HIGH Pushbutton on the operating Analog Linear Rate Meter, RIA-1809, Waste Gas Plenum Radiation Monitor, is illuminated it means ___(2)___.

(1) (2)

A. do depressing this button RESETS the High alarm B. do NOT depressing this button RESETS the High Alarm C. do the monitor reading is (or was) above the HIGH Setpoint D. do NOT the monitor reading is (or was) above the HIGH Setpoint Answer: C Answer Explanation Answer: C.

C. do, the monitor reading is (or was) above the High Setpoint; is correct Explanation:

Per SOP-37, Process Liquid Monitor System, the process monitors are all digital rate meters, and they each have a check source function.

Per SOP-38, Gaseous Process Monitoring system, there are digital rate meters and analog rate meters.

Only the digital rate meters have a check source function. Thus, part one is correct.

Per SOP-38, when the HIGH Pushbutton on RIA-1809 is illuminated, it means the indication on the monitor exceeded the high alarm setpoint.

The RESET pushbutton is used to reset alert and high alarms. Thus, part two is correct.

Reference:

SOP-37, and SOP-38 Distractors All are plausible as they are combinations of the opposite actions.

The 'do NOT' distractors are plausible for the examinee that doesn't know the attributes of how to check functionality / test the different types of meters.

The 'depressing this button RESETS the High Alarm' distractor is plausible for the examinee that hasn't mastered operations of the meter.

KA:

073 Process Radiation Monitoring System A4 Ability to manually operate and/or monitor in the control room:

(CFR: 41.7 / 45.5 to 45.8)

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EXAMINATION ANSWER KEY Daft N 1 A4.02 Radiation monitoring system control panel . 3.7 3.7 Test item meets KA. While Palisades does not use check sources for 'operability', they are used to determine 'functionality'. The test item asks which type of monitors have 'check source function', and also tests how some monitors work (ability to operate and / or monitor system control panel).

MEMORY Objective:

Given references, manually operate the following Radiation Monitoring System components

- Alarm and interlock setpoint checks and adjustment (072 A4.01)

- Check source for operability demonstration (072 A4.02, 072 A4.03, 073 A4.03)

- Effluent releases (073 A4.01)

- Radiation Monitoring System Controls (073 A4.02) without error.

Question 52 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 848152 User-Defined ID: 073A4.02 Cross Reference Number: RMS_CP03.0 Digital Rate meters in the Process Liquid Monitor and Gaseous Topic:

Process Monitoring Systems ___(1)___

Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 189 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 190 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 53 ID: 848153 Points: 1.00 The plant is at 100% power.

P-7A, Service Water System (SWS) pump trips on an overcurrent condition Given these conditions, complete the following statements:

The Service Water System Train associated with the Right Safeguards Electrical Distribution Train includes the SWS pump ___(1)___.

With P-7A tripped as described, entering LCO 3.7.8 Service Water System, ___(2)___ required.

(1) (2)

A. 7C is B. 7B is C. 7C is NOT D. 7B is NOT Answer: A Answer Explanation Answer: A.

A. 7C, is; is CORRECT Explanation:

Per the reference (TS) Pump 7B is in the LEFT Train, and Pumps 7A and 7C are part of the RIGHT Train.

When a pump trips on overcurrent, there is a short somewhere. The pump is inoperable and can NOT meet its Safety Design Function. Because each pump is on a list of components that are required to be operable for the two trains of Safeguards, having any one pump inoperable makes one train inoperable and requires entry into TS 3.7.8, Condition A, (at the very least).

The ARP for Low Hdr pressure refers to LCO 3.7.8.

Distractors All plausible for those that are mistaken about this system knowledge. Another Pair of 'A' and 'C' pumps (CCW) are loads on Bus 1C. These may be confused with the SWS pumps (or the CS pumps which have the B and C CS pump on one bus and the A CS pump on another), lending plausibility to the distractors. The 'is NOT' distractors are plausible for the examinee that is not familiar with what is required for each train to be operable and may assume having any two SWS pumps operable is 'good enough'.

For example: P-52C, CCW Pump, is not required for operability of either CCW Train. The specific equipment for each train is covered in the training material and is NOT designated as SRO ONLY.

KA: 076 Service Water System (SWS)

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EXAMINATION ANSWER KEY Daft N 1 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

(CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3)

IMPORTANCE RO 3.9 SRO 4.6 Test item meets KA. Answer requires knowledge of Tech Spec entry into LCO 3.7.8 SWS.

MEMORY Objective:

From memory describe the following for the Service Water System and Ultimate Heat Sink in accordance with Technical Specification 3.7.8 and 3.7.9. (G2.2.22, G2.2.39)

a. LCO Statement (3.7.8 and 3.7.9)

b. Applicability (3.7.8 and 3.7.9)

c. Action statements requiring action in less than one hour (3.7.8.C)

Question 53 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 848153 User-Defined ID: 076G2.2.42 Cross Reference Number: SWS_CK20.0 The plant is at 100% power. P-7A, Service Water System Topic:

(SWS) pump trips on an overcurrent conditi Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 192 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 193 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 54 ID: 848156 Points: 1.00 The plant is at 100% power

Feedwater Purity Air has been isolated from the Instrument Air system due to a leak on the cross tie valve.

A loss of instrument air has occurred, and the following annunciators are in alarm:

EK-1101, CONTAIMENT INSTR AIR LO PRESS

EK-1102, INSTRUMENT AIR LO PRESS

EK-1103, SERVICE AIR LO PRESS AOP-37, Loss of Instrument Air has been entered

All Instrument Air Compressors are running,

Air header pressure is 30 psig and lowering Until air is restored to containment,

Pressurizer Pressure will be maintained by __(1)___.

Pressurizer level will be maintained by intermittent automatic operation of ___(2)___.

(1) (2)

A. Manual control of heaters ONLY P-55A, Charging Pump B. Manual control of heaters ONLY P-55B or P-55C, Charging Pump C. Aux Spray and manual control of heaters P-55A, Charging Pump D. Aux Spray and manual control of heaters P-55B or P-55C, Charging Pump Answer: B Answer Explanation Answer: B B. Manual control of the heaters ONLY, P-55B or P-55C, Charging Pump is correct.

Explanation:

The loss of air to containment prevents any /all PZR Spray from being available because the valves fail closed, per AOP-37 Attachment 1, page 1.

The loss of air to containment causes all the letdown orifice stop / bypass valves (CV-2002, 03, 04, 05, and 2203) to all fail closed too. The required action is to stop P-55A and let P-55B or C automatically respond to level deviations as they occur intermittently.

Distractors Distractors with 'Aux Spray' (and manual control of heaters) are plausible for the examinee that believes the aux spray either have an accumulator for operation (16 other valves do per AOP-37 Loss of Instrument Air Attachment 4, Valves with N2 backup), or don't need air. These distractors are incorrect, because the aux spray also fail closed and does NOT have an accumulator.

PLP - 2018 NRC EXAM Page: 194 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 P-55A is plausible because it is normally aligned at 100% power. It is incorrect because it is explicitly stopped by the procedure, AOP-37, which has been entered per the stem.

KA: 078 IAS K1 Knowledge of the physical connections and/or cause-effect relationships between the IAS and the following systems:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.03 Containment air . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3* 3.4*

Test item meets KA. Answer requires knowledge of the cause-effect relationship between IAS and Containment air (and procedural requirements for loss of air). When they are tied together, a loss of one affects the operation of the loads of both.

MEMORY - recall procedure guidance

Reference:

AOP-37, Loss of Instrument Air, Attachment 1, rev 3, valves which fail closed.

Objective:

Given Abnormal Operating plant conditions and control room references, determine the subsequent actions/operator actions to mitigate the event and stabilize the plant in accordance with the applicable Abnormal Operating Procedure(s).

Question 54 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848156 User-Defined ID: 078K1.03 Cross Reference Number: IOTF_CK08.0 The plant is at 100% power Feedwater Purity Air has been Topic:

isolated from the Instrument Air system d Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 195 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 196 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 55 ID: 848163 Points: 1.00 The plant was shutdown for refueling 10 days ago.

The plant is in MODE 6

CORE ALTERATIONS are in progress

Irradiated Fuel is being moved in the Spent Fuel Pool

Equipment Hatch is closed

ONLY ONE door of the personnel air lock is closed

ALL other Containment Penetrations meet the requirements of LCO 3.9.3, Containment Penetrations The ONLY running Fuel Handling Area Fan has just tripped.

Given these conditions, complete the following statements:

Tech Spec LCO 3.9.3, Containment Penetrations ___(1)___ met.

Irradiated fuel movements in the Spent Fuel Pool ___(2)___.

(1) (2)

A. is can continue B. is must be suspended C. is NOT must be suspended D. is NOT can continue Answer: B Answer Explanation Answer: B B. is, must be suspended; is correct.

Explanation:

The requirement to close at least one door of the personnel air lock comes from the NOTES of LCO 3.9.3.

With one or more penetrations not in the required status, suspend core alterations immediately.

LCO 3.7.12 requires the ventilation to be aligned, and that fan tripped. With the FH area ventilation not aligned or in operation the immediate action required is to suspend movement of fuel assemblies (irradiated fuel) if decay times are < 30 days on the irradiated fuel.

All of this information is above the double line in Tech specs, except the action, which is an 'immediate'.

Thus, it is RO level.

Distractors:

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EXAMINATION ANSWER KEY Daft N 1

'IS' and 'can continue' is plausible for those who don't understand the ventilation alignment requirement for LCO 3.9.3. Part one is correct, and Part two is incorrect.

'IS NOT' and 'must be suspended' is plausible together for the examinee that believes the requirements of LCO 3.9.3 Containment Penetrations' are not met because only one door is closed and that there are ventilation alignements that if not met requires fuel movement to be suspended in the Fuel Handling Building.

'IS NOT' and 'can continue' are plausible together for the examinee that believes there are requirements of LCO 3.9.3 Containment Penetrations' that are not met, and that there are no ventilation alignemnts that require fuel movement to be suspended in the Fuel Handling Building.

'can continue' is plausible for those who don't understand how 3.9.3 and 3.7.12 are linked and misunderstand the applicabilities listed for 3.7.12.

KA: 103 Containment K3 Knowledge of the effect that a loss or malfunction of the containment system will have on the following:

(CFR: 41.7 / 45.6)

K3.03 Loss of containment integrity under refueling operations. . . . . . . . . . . . . . . 3.7 4.1 Test item meets KA. Question poses a loss of containment integrity during refueling ops through the loss of a ventilation alignment. Answer requires TS LCO knowledge for containment penetrations (which relate directly to integrity) and FH ventilation.

HIGH COG - multiple mental steps to align notes, actions and applicabilities of 3.9.3 with 3.7.12.

Objective:

Given Abnormal Operating plant conditions and control room references, SELECT the applicable Technical Specification LCO REQUIRED ACTIONS and COMPLETION TIMES in accordance with Technical Specifications. (K/A G2.2.42)

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EXAMINATION ANSWER KEY Daft N 1 Question 55 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848163 User-Defined ID: 103K3.03 Cross Reference Number: IOTF_CK09.0 The plant was shutdown for refueling 10 days ago. The plant is Topic:

in MODE 6 CORE ALTERATIONS are in p Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 199 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 200 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 56 ID: 851790 Points: 1.00 The plant is at 100% power.

The following alarm annunciates:

EK-0954, Rod Drive Seal Leak Off HI Temp The indication on TRA-150 for CRD-23 shows 201°F.

Given these conditions, complete the following statement:

When a CRDM has excessive leak off, ___(1)___ is leaking into the ___(2)___.

(1) (2)

A. CCW Quench Tank B. CCW Containment Sump C. PCS Containment Sump D. PCS Quench Tank Answer: C Answer Explanation Answer: C.

C. PCS, Containment Sump; is correct.

Explanation:

Per the reference (PL-CRD rev 9),

A. A seal leakage cup, located above seal, collects any leakage past the seal and contains a thermocouple to monitor for cooling water (CCW) or seal failure. CRDM leak off is used to identify seal failure, which could be indicated by excessive leakage, and/or high seal leak off temperature. High seal leak off temperature could also result from a loss of CCW or loss of CRDM fan cooling.

B. CRDM leak off is directed to the containment sump, thereby preventing pressure buildup above the seal.

Distractors:

The distractors with CCW are plausible because the CCW system uses a jacket of cooling for the CRDMs, but is incorrect because excess seal leakage is more than normal seal leak off (typically around 3cc/hr per CRDM).

The distractors with quench tank, are plausible for the examinee that doesn't know the leakoff path, and the quench tank is located in containment and collects various valve leakoff, and relief valve effluents (Shutdown cooling reliefs, SIT drain reliefs, Letdown Line Reliefs, PW),

KA: 002 RCS PLP - 2018 NRC EXAM Page: 201 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 K1 Knowledge of the physical connections and/or cause-effect relationships between the RCS and the following systems:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.02 CRDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.9* 3.0*

Test item meets KA. answer requires knowledge of the physical connection between RCS and CRD.

MEMORY level Objective:

From memory, identify redundant indications that can be used to validate the following Control Room alarms:

- EK-0954, "ROD DRIVE SEAL LEAK OFF HI TEMP" in accordance with ARP-5.

Question 56 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 851790 User-Defined ID: 002K1.02 Cross Reference Number: CRD_CK14.0 The plant is at 100% power. The following alarm annunciates:

Topic:

EK-0954, Rod Drive Seal Leak Off HI T Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 202 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 203 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 57 ID: 848171 Points: 1.00 The reactor has been tripped from 60% power.

During performance of EOP 1.0, Standard Post Trip Actions, the Reactor Operator notes the following:

At time 1000:00,

PCS pressure is 1850 psia and is rising at 10 psia per minute.

Pressurizer Level is 30% and rising at 1% per minute Given these conditions, and assume NO manipulation of any PZR Controllers, complete the following statements:

At time 1003:00,

There are ___(1)___ PZR Heater Groups ON / Energized.

At time 1015:00,

There are a MAXIMUM of ___(2)___ PZR Heater Groups ON / Energized.

(1) 1003:00 (2) 1015:00 A. NO 3 B. NO 6 C. SOME 3 D. SOME 6 Answer: A Answer Explanation Answer: A Explanation:

A. NO, 3, is correct. The PZR heaters are split into 6 groups, with 3 groups ultimately powered from bus 1E, and 3 groups from bus 1D. All heaters are de-energized automatically below 36% PZR Level. As level recovers to >36%, the 3 groups from bus 1E will auto re-energize. The 3 groups from bus 1D will NOT auto re-energize until their breakers are reset. At 1000, there is a demand for heaters because of low PZR pressure, but low level keeps the heaters from re-energizing. At 1003, pressure has risen to 1880 psia, and still demands heaters, but level has only risen to 33%, so the heaters stay de-energized. Thus, Part one is correct. At 1015, pressure has risen to 2000 psia, and still demanding heaters, and level has risen to 45%. Only the 3 groups from bus 1E auto re-energize.

Thus, part two is correct.

Distractor Analysis:

B. NO, 6, is plausible but incorrect. Part one is correct as explained in A. Part two is plausible for the examinee that realizes level has recovered enough but doesnt understand the auto re-energization only applies to half the heater groups.

PLP - 2018 NRC EXAM Page: 204 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 C. SOME, 3 is plausible but incorrect. Part one is plausible for the examinee that believes the low pressure demand for heaters overrides the low level cutout. Part two is correct as explained in A.

D. SOME, 6 is plausible but incorrect. Part one is plausible but incorrect as explained in C. Part two is plausible but incorrect as explained in B.

References:

PL-PPCS Rev 6 Pressurizer Level Heater Trips:

Low Level Heater Cutoff. All Pressurizer heaters de-energize on low-low Pressurizer level (36%) to prevent heater burnout.

4) Heater Control Selector switch:

A separate control circuit trips the circuit breakers of the Pressurizer heater 480 V transformers on a low-low Pressurizer level signal (36%).

a) Channel A, PZR Level LT-0101A input to heater interlocks.

b) Channel A and B - PZR Level LT-0101A and LT-0101B input to heater interlocks.

c) Channel B - PZR Level LT-0101B input to heater interlocks.

5) For heaters supplied by Bus 1E (LCC-15), the LC-15 breaker on Bus 1E will reclose when Pressurizer level returns > 36%.

For heaters supplied by Bus 1D (LCC-16), the LCC-16 breaker will NOT reclose when Pressurizer level returns > 36%. This is due to the possibility that the Diesel Generator sequencer could be loading the Diesel Generator and the Pressurizer Heaters are NOT part of the sequence but could energize during the sequencing and overload the Diesel Generator.

KA: 011 PZR Level Control K2 Knowledge of bus power supplies to the following:

(CFR: 41.7)

K2.02 PZR heaters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 3.2 Test item meets KA. Answer requires knowledge of power supply to the various banks of PZR Heaters.

HIGH COG - multiple mental steps to determine if the various permissives are met to allow heaters to turn on.

Objective:

From memory, list the power supplies for the following Pressurizer Pressure Control System components:

a. Backup heaters

b. Proportional heaters

c. PIC-0101A and PIC-0101B

d. LTOP Control Units PY-0105A and PY-0105B in accordance with SOP-1A, AOP-12, AOP-13, AOP-14 and AOP-15. (010 K2.01, 010 K2.02)

PLP - 2018 NRC EXAM Page: 205 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 57 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848171 User-Defined ID: 011K2.02 Cross Reference Number: PPCS_CK07.0 The reactor has been tripped from 60% power. During Topic:

performance of EOP 1.0, Standard Post Trip Act Num Field 1:

Num Field 2:

Text Field:

Comments: bank (see 710681)

Not used on 2014 or 2017 NRC exams PLP - 2018 NRC EXAM Page: 206 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 207 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 58 ID: 852019 Points: 1.00 On ALL Power Range NIs (NI-5, 6, 7, and 8), Reactor power is STABLE at 6%.

The Main Turbine is TRIPPED

WR NI-1/3 starts drifting high at 1% per minute from 6%

In 10 minutes, with NO operator action,

the Loss of Load Reactor Trip ___(1)___ BYPASSED.

the Reactor ___(2)___ automatically TRIPPED.

(1) (2)

A. is has B. is has NOT C. is NOT has D. is NOT has NOT Answer: B Answer Explanation Answer: B B. is, has NOT is correct Answer Explanation:

With power at 6% and stable, actual power is not changing, and SUR is 0.0.

The Loss of Load reactor trip depends on the input from the turbine trip system and the 4 power range nuclear instruments. The setpoint of the loss of load trip is nuclear power >15%, AND the turbine Tripped.

the coincidence for the trip is 2/4. The Loss of Load Reactor Trip is automatically bypassed when 3/4 PRNIs <15%. The Loss of Load Reactor Trip is enabled on a per channel basis above 15% power from the power range Nuclear Instruments. Power is stable at 6% on the power range NIs, and the Turbine is Tripped. At the end of 10 minutes, the power ranges have NOT changed and since they are less than 15%, the loss of load trip is still bypassed. Thus, part one is correct.

The WR NIs input into RPS for the Hi SUR trip. Each of the two WR NIs input into two PR NI inputs in RPS. However, at a rate of change of 1% per minute for 10 minutes, results in only a change of 10%,

which is less than one decade per minute. The setpoint for the HI SUR RPS Trip is 2.6 decades per minute (dpm). Since power did not change fast enough, the reactor did NOT automatically trip.

Thus, part two is correct.

The High Power Rate reactor trip is automatically enabled on a per channel basis when Wide Range Nuclear Instruments are 10e-4% or higher, and Power Range Nuclear Instruments are less than 15%

power. It takes 2/4 channels to generate the reactor trip signal when the rate of power change is 2.6 dpm or greater.

PLP - 2018 NRC EXAM Page: 208 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Distractor Analysis:

A. Is, has; is plausible but incorrect. Part one is correct as explained in the answer. Part two is plausible for the examinee that doesnt comprehend how the RPS HI SUR trip is developed.

C. Is NOT, has; is plausible but incorrect. Part one is plausible for the examinee that doesn't comprehend how the loss of load trip is developed, but incorrect as explained in B. Part two is plausible but incorrect as explained in A.

D. Is NOT, has NOT; is plausible but incorrect. Part one is correct as explained in B. Part two is plausible but incorrect as explained in A.

Reference:

ARP-21, Reactor Protection Scheme EK-06 A-02 HIGH POWER RATE CHANNEL TRIP Sensor: Bistable Trip Unit (Any 1 of 4)

Trip Setpoints: 2.6 DPM (Effective between 10-4% and 15% Power)

Either SR/WR Instrument in TEST or CAL Reactor Trip on 2/4 coincidence ARP-21, Reactor Protection Scheme, EK-06 C-06 LOSS OF LOAD CHANNEL TRIP Sensor: Bistable Trip Unit (Any 1 of 4)

Trip Setpoints: Turbine Auto Stop Oil Pressure 45 psi (Effective above 15% power)

Reactor Trip on 2/4 coincidence Tech Spec Bases RPS Instrumentation 3.3.1 High Startup Rate Trip The High Startup Rate trip uses the wide range Nuclear Instruments (NIs) to provide an input signal.

There are only two wide range NI channels.

Loss of Load Trip The Loss of Load Trip is initiated by two-out-of-three logic from pressure switches in the turbine auto stop oil circuit that sense a turbine trip for input to all four RPS auxiliary trip units.

Operating Bypasses The Operating Bypasses are initiated and removed automatically during startup and shutdown as power level changes. An Operating Bypass prevents the associated RPS auxiliary trip unit from receiving a trip signal from the associated measurement channel. With the bypass in place, neither the pre-trip alarm nor the trip will actuate if the measured parameter exceeds the set point. An annunciator is provided for each Operating Bypass.

The RPS trips with Operating Bypasses are:

a. High Startup Rate Trip bypass. The High Startup Rate trip is automatically bypassed when the associated wide range channel indicates below 1E-4% RTP, and when the associated power range excore channel indicates above 13% RTP. These bypasses are automatically removed between 1E-4% RTP and 13% RTP.

b. Loss of Load bypass.

The Loss of Load trip is automatically bypassed when the associated power range excore channel PLP - 2018 NRC EXAM Page: 209 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 indicates below 17% RTP. The bypass is automatically removed when the channel indicates above the set point. The same power range excore channel bistable is used to bypass the High Startup Rate trip and the Loss of Load trip for that RPS channel. Each wide range channel contains two bistables set at 1E-4% RTP, one bistable unit for each associated RPS channel. Each of the two wide range channels affect the Operating Bypasses for two RPS channels; wide range channel NI-1/3 for RPS channels A and C, wide range channel NI-2/4 for RPS channels B and D. Each of the four power range excore channel affects the Operating Bypasses for the associated RPS channel. The power range excore channel bistables associated with the Operating Bypasses are set at a nominal 15%, and are required to actuate between 13% RTP and 17% RTP.

Objective: From memory, describe the design features and interlocks that provide for the following NI System functions:

- High Power and High Start-Up Rate (SUR) Reactor Trip (015 K4.05)

- High SUR and Loss of Load Trips disabling (015 K4.06)

- High SUR and Loss of Load Trips enabling (015 K4.07) in accordance with FSAR Chapter 7 and DBD-2.05.(NI_CK09.0)

From memory, for automatic actions associated with the NI System,

a. state the parameter and value (setpoint) at which each automatic action occurs

b. explain the purpose of each automatic actuation in accordance with SOP-35, FSAR Chapter 7 and DBD-2.05. (015 K4.02, 015 K4.05, 015 K4.06, 015 K4.07). (NI_CK10.0)

From memory, explain the purpose of interfaces (physical connections) between the Nuclear Instrumentation System and the following:

- Reactor Protection System (015 K1.01) in accordance with DBD-2.05.(NI_CK06.0)

From memory, and given a loss or malfunction of the NI System describe the effects on the following:

- Reactor Protection System (015 K3.01) in accordance with SOP-35. (NI_CK11.0)

From memory, explain how to determine if a Reactor Protective System input parameter is satisfying the two out of four reactor trip logic without error. (012 K4.03) (RPS_E01.03)

From memory, for each reactor trip signal,

a. state the parameter and value at which the trip occurs

b. explain the purpose of the trip bypass in accordance with ARP-21 and DBD-2.05. (012 K4.02) (RPS_CK10.0)

From memory, explain how the Reactor Protective System trip coincidence logic is affected when a trip unit is placed in the tripped condition in accordance with Technical Specification Basis 3.3.1. (012 K4.01) (RPS_E03.05)

From memory, predict how the following conditions will impact operation of the Reactor Protective System:

- high failure of monitored parameter input signal (012 K6.06) without error. (RPS_CK13.0)

KA:

015 NIS K3 Knowledge of the effect that a loss or malfunction of the NIS will have on the following:

PLP - 2018 NRC EXAM Page: 210 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 (CFR: 41.7 / 45.6)

K3.01 RPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9 4.3 Test item tests knowledge specified in objective. Correct response requires knowledge of how the WR/PR NIs are arranged, how they provide input into RPS trips and bypasses and permissives, and how malfunctions of the NIs impact the RPS.

Test item meets KA . Correct response requires knowledge of how the WR/PR Nis are arranged, how they provide input into RPS trips and bypasses and permissives, and how malfunctions of the NIs impact the RPS.

HIGH COG - answer requires application of knowledge of coincidence and logic for low power trips to predict an outcome.

Question 58 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 852019 User-Defined ID: 015K3.01 Cross Reference Number: RPS_CK13.0 On ALL Power Range NIs (NI-5, 6, 7, and 8), Reactor power is Topic:

STABLE at 6%. The Main Turbine is TRI Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 211 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 212 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 59 ID: 848173 Points: 1.00 The plant is at 100% power:

ALL control switches for the CRHVAC system on panel C-11A are in the AUTO position with the exception of V-95, Air Handling Unit Fan, which is in the ON position

Containment Pressure is 1.0 psig

Containment Radiation level is 1 R/hr At time 1000:00, an event occurs inside containment

Containment Pressure is rising at 0.15 psig per minute

Containment Radiation is rising by 0.7 R/hr every minute Given these conditions, and assuming equipment operates as designed, complete the following statements:

At time 1015:00,

Charcoal Filter Unit 26A ___(1)___ in service.

CRHVAC Compressor VC-11 ___(2)___ running.

(1) (2)

A. IS IS B. IS IS NOT C. IS NOT IS D. IS NOT IS NOT Answer: B Answer Explanation Answer: B B. IS, IS NOT Explanation B. IS, IS NOT is correct. The CRHVAC lineup will change to the Emergency Mode on either a CHP or a CHR actuation signal. The setpoint for CHP is 3.7-4.3 psig. The setpoint for CHR 10 R/Hr. At 1015 with the given rates of change, Containment Pressure will be 3.25 psig (which is NOT high enough for CHP),

and Containment Radiation will be 11.5 R/hr. The re-alignment to the Emergency mode includes automatically aligning the Charcoal Filter unit 26A, and Tripping VC-11.

PLP - 2018 NRC EXAM Page: 213 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Distractor Analysis:

A. IS, IS, is plausible but incorrect. Part one is correct as explained in the answer. Part two is plausible but incorrect for the examinee that does NOT know the alignment of the system in the Normal Mode (VC-11 running) or in the Emergency mode (VC-11 Trips on CHP or CHR); or doesn't know the actuation setpoints.

B. Correct C. IS NOT, IS, is plausible but incorrect. Part one is plausible for the examinee that either doesnt know the actuation to emergency mode occurs from either CHR, or CHP; or does NOT know the setpoints for the actuation, or does not know what equipment is re-aligned (like the response of the V-95). Part two is correct as explained in the answer and is plausible for the examinee that doesnt know the normal alignment.

D. IS NOT, IS, NOT, is plausible but incorrect. Part one is plausible but incorrect as explained in C.

Part two is plausible but incorrect as explained in B.

References:

PL-CRHV r08-lp Control Room HVAC:

Emergency Mode is started automatically upon receipt of a Containment High Pressure (CHP) AND/OR Containment High Radiation (CHR)

Flow Paths For Various Modes of Operation

Normal Flow Path

Outside air is drawn in through a normal outside air intake

Tornado Damper TD-1 (TD-2)

Normal Air Intake Damper D-1 (D-8)

Modulating Damper D-2 (D-9)

Recirculation Damper D-3 (D-10) open, Recirculation Damper D-6 (D-13) closed

Emergency Flow Path

Verify Air Handling Unit V-95 (V-96) in service

Air Filter Unit Fan V-26A (V-26B) in service

Outside air is drawn in through emergency air intake through Tornado Damper TD-4 (TD-5) and Emergency Air Inlet Damper D-7 (D-14)

Discharge Damper D-5 (D-12) opens

VHX-26A (VHX-26B) heater energizes

Dampers D-20 (D-21) modulate to control flow

Normal Air Inlet Dampers D-1 & D-2 (D-8 & D-9) close

Recirc Dampers D-3 & D-6 (D-10 & D-13) open

Exhaust Fan V-16 trips if running, Dampers D-17 & D-18, TD-6 close

Purge Fan V-94 trips if running, Dampers D-15 & D-16 close

Emergency mode of operation is initiated manually or automatically by a Containment High Pressure (CHP) signal of 4.0 psig or Containment High Radiation (CHR) signal of 10 R/hour.

CHP/CHR shuts down Condensing Units. (VC-10 and VC-11)

If cooling is required, control switch must be taken to CHP/CHR OVRD position.

KA: 016 NNIS K4 Knowledge of NNIS design feature(s) and/or interlock(s) which provide for the following:

(CFR: 41.7)

PLP - 2018 NRC EXAM Page: 214 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 K4.03 Input to control systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.8* 2.9*

Test item meets KA. The Containment pressure and containment radiation signals are design features that provide inputs to control systems that result in CIS, and ventilation re-alignments.

HIGH COG - predict response Objectives:

Describe the design features and interlocks that provide for the following Control Room HVAC System functions: Shifting of ventilation operating modes, starting of emergency filtration units, preventing depressurization of control room during a tornado, minimizing exposure to radioactive materials during accident conditions in accordance with M-218, Sheets 6 & 6A.

For the following Control Room HVAC major components:

- AHU Fan(s) V-95, V-96

- Condensing Unit(s) VC-11, VC-10

a. describe the operational design of each component

b. describe the normal operating range of the component in accordance with the FSAR. (G 2.1.28)

Given Control Room HVAC operating conditions, analyze the conditions and judge if the Control Room HVAC System is operating in accordance with SOP-24. CHRV_CK09.0 PLP - 2018 NRC EXAM Page: 215 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 59 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848173 User-Defined ID: 016K4.03 Cross Reference Number: CRHV_CK09.0-1 The plant is at 100% power: ALL control switches for the Topic:

CRHVAC system on panel C-11A are in the Num Field 1: 050K4.01 Num Field 2:

Text Field:

PLP - 2018 NRC EXAM Page: 216 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Comments: Modified from PAL-LOI-74 (706893) changed part 2:

The plant is at 100% power:

ALL control switches for the CRHVAC system on panel C-11A are in the AUTO position with the exception of V-95, Air Handling Unit Fan, which is in the ON position

Containment Pressure is 1.0 psig

Containment Radiation level is 1 R/hr At time 1000:00, an event occurs inside containment

Containment Pressure is rising at 0.15 psig per minute

Containment Radiation is rising by 0.7 R/hr every minute Given these conditions, and assuming equipment operates as designed, complete the following statements:

At time 1015:00,

Charcoal Filter Unit 26A ___(1)___ in service.

Recirculation Damper, D-6 ___(2)___ open.

(1) (2)

A. IS IS B. IS IS NOT C. IS NOT IS D. IS NOT IS NOT Answer: A PLP - 2018 NRC EXAM Page: 217 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 218 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 60 ID: 851872 Points: 1.00 Given the following conditions:

A loss of coolant accident occurs inside containment

EOP-4.0, "Loss of Coolant Accident Recovery" is being implemented

ALL PCPs have been tripped

CETs temperatures are 550°F and rising on TRI-0101A1, A2, and TRI-0101B1, B2

PCS Thots are 560°F

PCS Tcolds are 518°F

PCS Tave is 539°F

PZR level is at 45% and rising

PCS pressure is at 1100 psia and lowering

BOTH channels of Reactor Vessel Level Monitoring System (RVLMS) have five red lights LIT These conditions indicate that the fluid in the vessel is ...

A. superheated.

B. more than 1°F, but less than 10°F subcooled.

C. more than 10°F, but less than 25°F subcooled.

D. more than 25°F subcooled.

Answer: B Answer Explanation Answer: B.

B, more than 1°F, but less than 10°F subcooled; is correct.

Explanation:

At 1100 psia PCS pressure, Tsat is 556°F.

The highest CETs are 550°F, which is within 10°F of superheated. The indication TRI-0101A1, A2, B1, B2 is the QCETs.

DISTRACTER ANALYSIS A. Superheated is plausible for the examinee that incorrectly uses Thot. With NO PCPs operating, the Thots are not indicating conditions in the vessel.

C. >10 but less than 25°F subcooled is plausible for the examinee that incorrectly uses Taves. With NO PCPs operating, the Taves have bad inputs.

D. more than 25°F subcooled is plausible for the examinee incorrectly using the Tcolds. With NO PCPs operating, the Tcolds are not indicating conditions in the vessel.

Reference:

Steam Tables PLP - 2018 NRC EXAM Page: 219 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 EOP-4.0 Step 25-26, and EOP Supplement 1 KA: 017 In-Core Temperature Monitoring K5 Knowledge of the operational implications of the following concepts as they apply to the ITM system:

(CFR: 41.5 / 45.7)

K5.03 Indication of superheating . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7 4.1 Test item meets KA. Question requires determination of thermodynamic condition of fluid in vessel with respect to superheating/ subcooling.

HIGH COG - Use of Steam Tables.

Objective: Given plant indications or an event involving Emergency Operating Procedures, describe the indications used to determine if voiding is occurring in the PCS in accordance with EOP-9.0, Success Path HR-1.

PLP - 2018 NRC EXAM Page: 220 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 60 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 851872 User-Defined ID: 017K5.03 Cross Reference Number: TBAH_E04.01-3 Given the following conditions: A loss of coolant accident Topic:

occurs inside containment EOP-4.0, "Lo Num Field 1:

Num Field 2:

Text Field:

PLP - 2018 NRC EXAM Page: 221 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Comments: modified from PAL-LOR-4315 which is an SRO only question that assessed the condition and selected a procedure.

SRO ONLY Given the following conditions:

A small-break loss of coolant accident occurs inside containment

EOP-4.0, "Loss of Coolant Accident Recovery" is being implemented

Bus 1C is unavailable due to an overcurrent condition

A controlled cooldown was commenced

PZR level is at 45% and rising

PCS pressure is at 980 psia and rising

CETs temperatures are 584°F and rising

Both channels of Reactor Vessel Level Monitoring System (RVLMS) have eight red lights LIT What do the above conditions indicate and what action(s) is required to address the condition?

A. Core is possibly uncovered and superheat conditions exist, CRS will remain in EOP-4.0 and direct implementation of EOP Supplement 26, "PCS Void Removal."

B. Core is possibly uncovered and superheat conditions exist, CRS will direct transition from EOP-4.0, to EOP-9.0, "Functional Recovery Procedure."

C. Core is not uncovered and saturations conditions exist, CRS will remain in EOP-4.0 and direct implementation of EOP Supplement 26, "PCS Void Removal".

D. Core is not uncovered and saturations conditions exist, CRS will direct transition from EOP-4.0, to EOP-9.0, "Functional Recovery Procedure."

Answer: B PLP - 2018 NRC EXAM Page: 222 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 223 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 61 ID: 848193 Points: 1.00 (Reference Provided)

In accordance with SOP-27, Fuel Pool System, and assuming the transfer tube flange is installed:

When filling the Spent Fuel Pool from the SIRWT, the reason the maximum level should be NO higher than 5 inches below the Southeast Skimmer is to ___(1)___.

When transferring water in the South Tilt Pit to the SIRWT from a level of approximately 1 foot above the transfer tube to 1 foot below the transfer tube, the operator can expect a change in SIRWT level of ___(2)___.

(1) (2)

A. prevent unwanted SFP liner leakage >10%

B. prevent unwanted SFP liner leakage <10%

C. conserve SIRWT inventory <10%

D. conserve SIRWT inventory >10%

Answer: B Answer Explanation Answer: B B. prevent unwanted SFP liner leakage, <10%; is correct.

Explanation:

Per the reference (SOP-27, Fuel Pool System, sections 7.2.1 and 7.2.7:

7.2.1: To Fill the SFP from the SIRWT:

The SFP level at the completion of filling operations should not be higher than 5" below the southeast skimmer. This prevents SFP liner leakage.

7.2.7 Drain South Tilt Pit to SIRWT:

NOTE: South Tilt Pit volume is approximately 30,800 gallons. This is approximately 10% of SIRWT Tank volume. Transfer of the South Tilt Pit from a level of approximately one foot above the transfer tube to one foot below the transfer tube have resulted in an approximate 3% level change in the SIRWT.

Distractors:

All plausible. While leaving the water in the SIRWT does conserve SIRWT inventory, it is not the reason SFP level should be no higher than 5" below the southeast skimmer because there are known liner leaks above that point as evidenced by finding water in the aux building stairwell.

>10% is plausible for those who don't realize how much water that really is and confuse the volume of the entire South Tilt Pit with only the volume being asked for.

KA: 033 Spent Fuel Pool Cooling A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including:

PLP - 2018 NRC EXAM Page: 224 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 (CFR: 41.5 / 45.5)

A1.01 Spent fuel pool water level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.7 3.3 Test item meets KA. Question asks for expected changes in levels of tanks when operating SFP controls to move / transfer water volumes.

MEMORY Objective:

From memory for the following systems major components:

1] Spent Fuel Pool System Major Components

- The Spent Fuel Pool

- P-51A/B, Spent Fuel Pool Cooling Pumps

- P-82, SFP Recirculation Booster Pump

- T-50, Spent Fuel Pool Demineralizer

- Spent Fuel Pool Side Tilt Pits

- Spent Fuel Pool Skimmers

- Spent Fuel Pool Liner and Leak Detection System

- E-53A/B, Spent Fuel Pool Heat Exchanger 2] Dry Fuel Storage Cask System Major Components

- Concrete Ventilated Storage Cask (VSC)

- Multi-Assembly Sealed Basket (MSB)

- MSB Internal Atmosphere

- MSB Transfer Cask

- Horizontal Storage Module (HSM)

a. describe the operational design of each component

b. describe the normal operating range of the component in accordance with FSAR 9.4, DBD 7.09 and/or Dry Fuel Storage Certificate of Compliance.

Question 61 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 848193 User-Defined ID: 033A1.01 Cross Reference Number: ISHB_CK02.0 (Reference Provided) In accordance with SOP-27, Fuel Pool Topic:

System, and assuming the transfer tube f Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 225 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 226 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 227 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 62 ID: 848195 Points: 1.00 The plant was at 100% power, when a total loss of Instrument Air occurs

The reactor was manually tripped

386AST, Turbine Trip Relay, does NOT actuate when the turbine trips

Instrument Air header pressure to the ASDVs and TBV is near zero psig

Tave is 542°F and STABLE

HIC-0780A, Steam Dump Controller demand is Steady at 2-3%

ALL other equipment functions as designed Given these conditions, complete the following statements:

CV-0511, Turbine Bypass Valve, (TBV) ___(1)___ open and controlling Tave.

To bring Tave to the requirements of EOP-1.0, Standard Post Trip Actions, (525°F-540°F) the

___(2)___ must be opened in MANUAL.

(1) (2)

A. is TBV B. is ASDVs C. is NOT TBV D. is NOT ASDVs Answer: D Answer Explanation Answer: D D. is NOT, ASDVs; is correct.

Explanation:

Without the AST relay getting energized when the turbine trips, the quick open mode of operation is not armed for either ASDVs or TBV. The modulate function works from the ASTX1 relay.

On a loss of IAS, the ASDVs have Nitrogen backup for operation, but the TBV does NOT. The TBV is unable to respond to the signals from its controllers post trip for Tave-Tref error. Thus part one is correct.

With Tave at 542° and STABLE, the ASDVs have used the modulate mode to lower temperature, but because the value is stable, and above EOP-1.0 step 7 requirements, manually opening the ASDVs more is required to lower Tave to between 525 and 540.

References:

M-207, sheet 1; EOP-1.0, step 7 DISTRACTERS:

a. TBV response is plausible for the examinee that reverses the availability of N2 backup. The preferred method of lowering Tave per the EOP is the TBV, making the second part plausible too.

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EXAMINATION ANSWER KEY Daft N 1

b. TBV response is plausible as stated in A. ASDVs is correct as stated in the answer.

c. is NOT is correct as stated in the answer. TBV is plausible as stated in A.

d. Correct as explained above.

KA: 041 Steam Dump Turbine Bypass Control A2 Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; and (b) based on those predictions or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.03 Loss of IAS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.8 3.1 Test item meets KA. Question poses loss of IAS, and answer requires impact on SDS and procedural direction.

HIGH COG - determine response of components to a trip when there is a loss of air. Determine action to take to accomplish procedure direction.

Objective:

From memory, predict how the following conditions will impact operation of the Main Steam system:

- Raising Steam Demand (039 A2.05)

- Stuck open steam valve (041 A2.02)

- Loss of Instrument Air (041 A2.03)

- High Steamline Radiation (039 A2.03)

- Malfunctioning TBV or ADV (039 A2.04)

- EK-0130, "STEAM AND FEEDWATER PENETRATION HI TEMP"

- EK-0166, "MAIN STEAM ISO VALVE ACCUMULATOR LO PRESS" (041 A2.03)

- EK-0965, "STEAM GEN LO PRESS CONTROL CKT UNDERVOLTAGE" (039 K2.01)

- EK-0966, "MAIN STEAM ISOLATION VALVE UNDERVOLTAGE"

- EK-0970, "STEAM GEN VALVES ISOLATION LOCKOUT"

- EK-3405, "MAIN STEAM PEN ROOM HIGH TEMP" without error.

PLP - 2018 NRC EXAM Page: 229 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 62 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848195 User-Defined ID: 041A2.03 Cross Reference Number: MSS_CK13.0-5 The plant was at 100% power, when a total loss of Instrument Topic:

Air occurs The reactor was manually Num Field 1:

Num Field 2:

Text Field:

Comments: modified from PAL-LOR-4335 (709611) which does not include a loss of air:

The following conditions exist:

A Plant trip from full power occurs

386AST, Turbine Trip Relay, does not actuate when the turbine trips

All other equipment functions as designed Which one of the following describes the effect on the Plant and the appropriate action?

A. Atmospheric Dump Valves ONLY do NOT open. Verify Steam Generator Code Safety Valves are controlling Steam Generator pressure.

B. Atmospheric Dump Valves ONLY do NOT open. Ensure PIC-0511, Turbine Bypass Valve Controller, is controlling Steam Generator pressure.

C. Atmospheric Dump Valves and Turbine Bypass Valve do NOT open. Verify Steam Generator Code Safety Valves are controlling Steam Generator pressure.

D. Atmospheric Dump Valves and Turbine Bypass Valve do NOT open. Place HIC-0780A, Steam Dump Controller, in MANUAL to control Steam Generator pressure.

Answer: B PLP - 2018 NRC EXAM Page: 230 of 369 15 January 2020

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EXAMINATION ANSWER KEY Daft N 1 63 ID: 848215 Points: 1.00 The plant is in MODE 1, preparing for synchronization,

Main Turbine is at 1800 RPM

The Primary, Backup, and Coastdown Relays (386P, B, and C) have NOT been reset

Then,

Actual main turbine speed reaches 1900 RPM

Reactor Power is 13% by Nuclear Instrumentation At the turbine controls in the Main Control Room, the operator observes ___(1)___ close.

As the turbine coasts down to 1700 RPM, the reactor operator withdraws rods to raise the demand on the Turbine Bypass Valve controller.

Reactor Power reaches 16%

At the turbine controls in the Main Control Room, the operator observes ___(2)___ close.

A. (1) ONLY the Governor and Intercept valves (2) ONLY the Governor and Intercept valves B. (1) ONLY the Governor and Intercept valves (2) ALL the Main Stop, Governor, Intercept, and Reheat Stop valves C. (1) ALL the Main Stop, Governor, Intercept, and Reheat Stop valves (2) ONLY the Governor and Intercept valves D. (1) ALL the Main Stop, Governor, Intercept, and Reheat Stop valves (2) ALL the Main Stop, Governor, Intercept, and Reheat Stop valves Answer: B Answer Explanation Answer: B.

B. (1) ONLY the Governor and Intercept valves (2) ALL the Main Stop, Governor, Intercept and Reheat Stop valves is Correct.

Explanation:

During an actual speeding up of the main turbine the OPC circuits protect from overspeed at 103% of 1800 RPM (1854 RPM) by closing the GVs and IVs, until the overspeed condition clears. It is NOT a Turbine Trip.

On a Turbine Trip, the ASO depressurizes and ALL the valves close by Springs (MSS, GV, IV, RS).

When power is greater than 15% and the generator output breakers are both open (with 386 P, B, and C devices are NOT reset), a reactor trip and turbine trip occurs. (Loss of Load) ALL the valves close.

The Overspeed Trip Setpoint is 110% of 1800 RPM (1980 RPM) per EK-0102, turbine Overspeed Trip, and the lesson plan PL-EHC r08, pg 39.

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EXAMINATION ANSWER KEY Daft N 1 Distractors:

A. Part one is correct. Part two is plausible for those that don't understand the low load trip.

C. Part one is plausible for those who don't understand the auto actions of OPC and believe that all the valves close on an overspeed condition. Part one is incorrect as explained in the answer. Part two is plausible for those that don't understand the low load trip and believe the overspeed actions occur when power is > 15%. Part two is incorrect as explained in the answer.

D. Both parts are plausible as explained in B and C.

KA:

045 Main Turbine Generator A3 Ability to monitor automatic operation of the MT/G system, including:

(CFR: 41/7 / 45.5)

A3.07 Turbine stop/governor valve closure on turbine trip . . . . . . . . . . . . . . . . . . . . 3.5 3.6 Test item meets KA. Question asks what do you expect to see / monitor indications when a turbine overspeed and trip occur.

HIGH COG - multiple mental step to determine if low load trip will apply. comprehension of inputs (prep for synchronization)

Objective:

From memory, describe the design features and interlocks that provide the following Electro-Hydraulic Control System functions:

a. Trip/Auto Start of EHC Pumps P-19A/B

b. Turbine Runback (045 K4.12)

c. Turbine Trip/Overspeed Trip (045 K4.11, 045 K4.13)

d. DEH Operator Auto Control (045 A3.05, 045 K4.07) in accordance with ARP-1, E-17 Sh. 9 and SOP-8.

Question 63 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 848215 User-Defined ID: 045A3.07 Cross Reference Number: EHC_CK09.0 The plant is in MODE 1, preparing for synchronization, Main Topic:

Turbine is at 1800 RPM The Primary, Ba Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 233 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 234 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 235 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 64 ID: 851770 Points: 1.00 Note to examinee: There is a color image provided as a handout With the Plant at 100% power, the following alarm is received in the Control Room during a Waste Gas Decay Tank batch release:

EK-1364, GASEOUS WASTE MONITORING HI RADIATION The Control Room Operator checks RIA-1113, Waste Gas Discharge Monitor, and sees the following:

Based on the above indication, RIA-1113 (1) and CV-1123, Waste Gas Decay Tank Discharge Valve, will (2) .

A. (1) indication is high out of range (2) remain open B. (1) indication is high out of range (2) automatically close C. (1) has an internal CPU failure (2) remain open D. (1) has an internal CPU failure (2) automatically close Answer: B Answer Explanation Answer: B Answer Explanation / Distractor Analysis:

B. (1) indication is high out of range; (2) automatically close; is correct.

a. Incorrect, plausible for the examinee that believes that the valve will remain open since the monitor does not have a valid indication.

b. CORRECT - For Waste Gas RIA-1113, IF Radiation Field goes above the maximum range of the detector, THEN the Range Indicator will illuminate, and front panel display will indicate EEEEE CPM.

c. Incorrect, plausible for the examinee that believes that the valve will remain open since the monitor does not have a valid indication.

d. Incorrect, plausible for the examinee that believes that the monitor has an internal CPU failure because there is not a valid indication.

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EXAMINATION ANSWER KEY Daft N 1 KA: 071 Waste Gas A4 Ability to manually operate and/or monitor in the control room:

(CFR: 41.7 / 45.5 to 45.8)

A4.27 Opening and closing of the decay tank discharge control valve . . . . . . . . . . . 3.0* 2.7*

Test item meets KA. Question gives indication of the rad monitor that is monitoring a release. Answer requires discrimination of ICU failure or High Rad, and response of system controlling release. There is NO position indication of the discharge valve, CV-1123, in the control room.

HIGH COG - read and interpret monitor indications.

Objective: From memory, for the Radiation Monitoring System, a. list the Control Room indications b.

describe the Control Room controls in accordance with P&IDs series M-223. (G2.1.31)

Reference:

SOP-38 Gaseous Process monitoring System, Attachment 1, System Malfunctions, Section 4, Abnormal Digital Rate Meter Indications:

For Waste Gas RIA-1113 IF Radiation Field goes above the maximum range of the detector, THEN the Range Indicator will illuminate, and front panel display will indicate EEEEE CPM.

Question 64 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 851770 User-Defined ID: 071A4.27 Cross Reference Number: RMS_CK08.0-10 Note to examinee: There is a color image provided as a Topic:

handout With the Plant at 100% power, the Num Field 1:

Num Field 2:

Text Field:

Comments: Bank not used on 2014 or 2017 NRC exam PAL-LOR-4284 PLP - 2018 NRC EXAM Page: 237 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 238 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 65 ID: 848219 Points: 1.00 Placing the cutout switches for the RIA-2316 and RIA-2317, Fuel Handling Area Radiation Monitors, to the "IN" position will:

A. Enable automatic closure of specific Containment Isolation valves.

B. Trip the Fuel Handling Area Supply Fan, V-7, on one out of two logic.

C. Enable automatic closure of Fuel Handling Area exhaust dampers.

D. Trip the Penetration and Fan Room Fans, V-78 and V-79, on high radiation.

Answer: A Answer Explanation Answer: A.

A. Enable the automatic closure of specific Containment Isolation valves.

Explanation /

Reference:

SOP-39, Rev. 18, 2nd NOTE before Step 7.3.2b RIA-2316 and RIA-2317 are each connected to one channel for Containment Isolation upon detection of high radiation in the Reactor Refueling Area (noncoincident).

Distractors:

All plausible choices as they are or could be resulting actions from other signals and controls.

Distractor B is plausible but incorrect. A CHR signal that arises from a high rad from either of the two Containment Refueling Monitors mentioned RIAs (2316, and 2317) will trip the Air Room Purge Fan V-46 but does NOT impact the V-7 fan. (A fan is tripped, but NOT the V-7). The on one out of two logic is plausible because the CHR logic is one out of two on the Containment Refueling Monitors. Stopping the V-7 fan manually is an operator action for a high alarm from either RIA-5709 or 2313. The position of the cutout switches for the RIA-2316/17 do not matter.

Distractor C is plausible but incorrect. An alarm on RIA-5712 trips the supply fan which may trip the exhaust fan. If the exhaust fan trips, the dampers close. The position of the cutout switches for the RIA-2316/17 do NOT matter to the RIA-5712. (ARP-8, Attachment 2 pg 8 -RIA-5712)

Distractor D is plausible but incorrect. A high rad signal from RIA-5710 will automatically trip the Penetration and Fan Room Fans, and close dampers PO-8035, and 8036. The position of the cutout switches for RIA-2316/17 do not matter.

KA: 072 ARM 2.1.27 Knowledge of system purpose and/or function.

(CFR: 41.7)

IMPORTANCE RO 3.9 SRO 4.0 Test item meets KA. Question asks function of cutout switches for ARMs.

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EXAMINATION ANSWER KEY Daft N 1 MEMORY Objective:

From memory, explain the purpose of interfaces (physical connections) between the Radiation Monitoring System and other plant systems:

- Plant Ventilation Systems (072 K1.01)

- Containment Isolation (072 K1.02)

- Fuel Handling Area Vent Isolation (072 K1.03)

- Control Room Ventilation (072 K1.04)

- Main Steam System (072 K1.05)

- Plant systems containing Process Radiation Monitors (073 K1.01) without error.

Question 65 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 848219 User-Defined ID: 072G2.1.27 Cross Reference Number: RMS_CK06.0-4 Placing the cutout switches for the RIA-2316 and RIA-2317, Topic:

Fuel Handling Area Radiation Monitors, Num Field 1:

Num Field 2:

Text Field:

Comments: bank not used on 2017 or 2014 NRC exams PAL-LOR-1471 PLP - 2018 NRC EXAM Page: 240 of 369 15 January 2020

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EXAMINATION ANSWER KEY Daft N 1 66 ID: 847460 Points: 1.00 Conduct of Operations requires:

When reporting to the control room for shift turnover, the oncoming operators ___(1)___ initially request permission to enter the Controls Area of the control room.

The operator designated as the At-The-Controls Operator (ATC) needs to discuss a procedure step with the Control Room Supervisor (CRS). If this discussion takes place at the PPC on the CRS' island, a turnover of the ATC position ___(2)___ occur first.

(1) (2)

A. must must B. must does NOT have to C. do NOT have to must D. do NOT have to does NOT have to Answer: B Answer Explanation Answer: B.

B. must, does NOT have to; is correct.

Explanation /

Reference:

Admin Procedure 4.00 Operations Organization, Responsibilities, and Conduct, Attachment 5 Control Room Conduct, Section 3.0 Control Room Access Control states:

a. During Control Room Watchstander turnover/relief, oncoming operators shall initially request permission to enter the control room Controls Area but will not require additional permission to enter other areas of the control room while actively engaged in turnover/relief. (thus, part one is correct)

b. The ATC operator should remain in the surveillance area at all times, except for brief entries into the Controls Areas near the CRS Desk (CRS Island) to retrieve log clipboards and items from the printer. All other entries into the Controls Area require the ATC operator to be turned over per EN-OP-115-02. Per the map in the EN-OP-115-02 Palisades addendum, the area around 90% of the CRS desk is part of the surveillance area. Thus, part two is correct.

Distractors are plausible as combinations of allowed / not allowed requirements for those not knowledgeable of the procedure requirements as explained in the answer.

'do NOT have to' distractors are plausible for the examinees that believe an official duty like getting turnover, does NOT have to follow the protocols for entry. They are incorrect as explained in the answer.

'must' distractors are plausible for the examinees that may think the CRS desk is not located in the ATC Surveillance area, but it is.

KA:

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EXAMINATION ANSWER KEY Daft N 1 2.1.3 Knowledge of shift or short-term relief turnover practices.

(CFR: 41.10 / 45.13)

IMPORTANCE RO 3.7 SRO 3.9 Test item meets KA as getting the correct answer requires knowledge of how turnover and short term relief occurs.

MEMORY Objective: Given plant conditions requiring a Shift Change and the NCO Shift Turnover Checklist, explain the following:

a. Items to be completed prior to relieving the watch

b. Items to be reviewed during the Shift

c. Items to be completed during the Shift

d. Items to be completed prior to being relieved in accordance with EN-OP-115 and AP-4.00 (K/A G2.1.3)

Question 66 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 847460 User-Defined ID: G2.1.3 Cross Reference Number: APCO_CK13.01 Conduct of Operations requires: When reporting to the control Topic:

room for shift turnover, the oncomi Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 243 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 67 ID: 847657 Points: 1.00 A work group who has been granted access to the surveillance area and the At-The-Controls (ATC) area needs to go behind the control room panels and then return.

Upon return to the surveillance area, the work group ___(1)___ required to ask permission to enter the surveillance area.

If the work group needs to enter the At-The-Controls (ATC) area, permission must be granted

___(2)___.

(1) (2)

A. is ONLY once B. is each time of entry C. is NOT ONLY once D. is NOT each time of entry Answer: D Answer Explanation Answer: D.

D. is NOT, each time of entry; is Correct.

Explanation /

Reference:

From Admin Proc 4.00, rev 63, Attachment 5 Section 3.0.c:

c. Once access has been granted to the surveillance area, a work group can go behind the Control Room panels (into the controls area) and re-enter the surveillance area without asking permission. This does not apply for the At-the-Controls area. Permission shall be requested for each entry into the ATC area.

Distractors are plausible for those that don't know this requirement.

Distractors are combinations of the different variations on the requirements.

'is' distractors are plausible for the examinee that doesn't know the boundary of the areas mentioned, or the requirements of the procedure.

'ONLY once' distractors are plausible for the examinee that don't know the procedure requirement or treats it like being granted access to the surveillance area.

KA:

2.1.13 Knowledge of facility requirements for controlling vital/controlled access.

(CFR: 41.10 / 43.5 / 45.9 / 45.10)

IMPORTANCE RO 2.5 SRO 3.2 Test item meets KA. Knowledge of requirements to enter vital areas it required to answer this question.

MEMORY PLP - 2018 NRC EXAM Page: 244 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 67 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 847657 User-Defined ID: G2.1.13 Cross Reference Number: APCO_E13.01-2 A work group who has been granted access to the surveillance Topic:

area and the At-The-Controls (ATC) ar Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 245 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 246 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 68 ID: 847462 Points: 1.00 EOP-1.0, Standard Post Trip Actions, is in progress.

E-50A, 'A' SG level is 6% and rising slowly

E-50B, 'B' SG level is -5% and lowering slowly

P-8A, 'A' AFW pump is providing 165 gpm to each SG During the Verbal Verifications, the CRS has asked:

Do we have at least one Steam generator with level between 5 and 70% with feedwater available?

In accordance with the Operations Standards Clarifications, the MINIMUM acceptable response to this direction is...

A. NO B. YES C. NO, B SG level is -5% and lowering slowly, being fed by P-8A Aux Feedwater Pump at 165 gpm to each SG D. YES; A SG level is 6% and rising slowly, B SG is -5% and lowering slowly, being fed by P-8A Aux Feedwater Pump at 165 gpm to each SG Answer: D Answer Explanation Answer: D D. YES; A SG level is 6% and rising slowly, B SG is -5% and lowering slowly, being fed by P-8A Aux Feedwater Pump at 165 gpm to each SG is correct.

Explanation /

Reference:

EOP-1.0 SPTA Verbal Verifications Script (from Orange Standards Book in Simulator) Operator Actions Section page 4 step 7.a.

OP-STN-004 (Operations Standards Clarifications) requires values AND trends for responses, as an expectation of operations department personnel when communicating plant parameters.

Admin Procedure 4.06, Rev 26, Attachment 15, page 1. Section 1.1:

1.1 COMMUNICATIONS STANDARDS The Palisades Operations Department Communications Policy shall be followed.

The examinee not familiar with the Clarification, may be familiar with the Additional communications standards which are as follows.

To the extent possible, the individual responses to the verbal verification shall include a statement of compliance ("yes"/"no".) Repeat back of the criterion to include the parameter's value and trend is required for "no" responses. A "yes" response does not require repeat back.

PLP - 2018 NRC EXAM Page: 247 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Since at least one S/G level is greater than -84%, a response of YES, is a plausible distractor.

A. plausible because it conveys an idea that more needs to be said if the receiver of this message desires. Incorrect because it does NOT meet the standard, and one S/G level is between 5 and 70% with AFW available.

B. plausible because one SG level is between 5 and 70%, but incorrect because it doesn't meet the clarification of the standard, or the verifications script that requires values and trends of both SGs and AFW flow from specified pumps.

C. plausible because it includes a repeat back with values and trends, but incorrect because the repeat back of the criterion is not complete (it omits the key piece of info that states one SG is 5-70% and meets the level requirement).

KA:

2.1.17 Ability to make accurate, clear, and concise verbal reports.

(CFR: 41.10 / 45.12 / 45.13)

IMPORTANCE RO 3.9 SRO 4.0 Test item meets KA as the answer requires the ability to select the communication of the reporting of some vital information during EOP usage that meets the standard expected by the facility.

HIGH COG - application of a rule of proper communications.

Objective: Given an example of a verbal communication, determine if the communication is complete and correct in accordance with EN-OP-115, Conduct of Operations.

Question 68 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 847462 User-Defined ID: G2.1.17 Cross Reference Number: ECAA_CK01.04 EOP-1.0, Standard Post Trip Actions, is in progress. E-50A, 'A' Topic:

SG level is 6% and rising slowly E Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 248 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 249 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 69 ID: 847501 Points: 1.00 What part of the procedure change process in Administrative Procedure 10.41, Site Procedure Processes, ensures the usability of the revised procedure by the least experienced qualified worker?

A. Technical Review B. Validation Review C. Cross-Discipline Review D. On-Site Safety Review Committee Answer: B Answer Explanation Answer: B B. Validation Review Explanation /

Reference:

Per Admin Procedure 10.41, rev 51 page 14, section 5.7:

5.7 VALIDATION REVIEWER

a. Reviews procedures to ensure usability by the least experienced qualified worker.

b. Assists the procedure sponsor by providing input related to human performance factors.

c. Performs "Validation Review Checklist" (Attachment 6) (optional but recommended).

Distractors are all plausible because each 'reviewer' has specific responsibilities per the referenced procedure:

5.6 TECHNICAL REVIEWER

a. Ensures procedure is technically and functionally adequate.

b. Determines impact on nuclear safety.

c. Ensures plant configuration and design specifications are maintained.

d. Ensures consistency with the following:

Technical Specifications & Technical Specification Bases

NRC Commitments

Corrective Actions to Prevent Recurrence (CATPR)

INPO Commitments

FSAR/UFSAR

Quality Assurance Program Manual

e. Validates that previously implemented corrective actions and commitments included in procedure are NOT inadvertently removed or nullified.

f. Performs "Technical Review Checklist" (Attachment 7) (optional but recommended).

5.2 CROSS-DISCIPLINE REVIEWER

a. Ensures procedure or change is technically and functionally accurate relative to the reviewer's area of expertise.

b. Validates that the impacted department can adequately support resource requirements identified, added or changed within the procedure.

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EXAMINATION ANSWER KEY Daft N 1

c. Determines if the procedure or change negatively affects the use or operation of equipment under the control of the reviewer's department.

d. Determines if the procedure or change affects response of a system under direct control of reviewer's department.

5.4 ON-SITE SAFETY REVIEW COMMITTEE (OSRC)

a. Independently reviews activities to provide assurance that the plant is operated and maintained in accordance with the operating license and applicable regulations that affect nuclear safety KA: 2.2.6 Knowledge of the process for making changes to procedures.

(CFR: 41.10 / 43.3 / 45.13)

IMPORTANCE RO 3.0 SRO 3.6 Test item meets KA. Answer requires knowledge of procedure change process.

MEMORY PLP - 2018 NRC EXAM Page: 251 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 69 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 847501 User-Defined ID: G2.2.6 Cross Reference Number: APPR_E01.02-2 What part of the procedure change process in Administrative Topic:

Procedure 10.41, ?Site Procedure Proce Num Field 1:

Num Field 2:

Text Field:

Comments: Modified from PAL-LOI-7994, which asked the functions of a different reviewing step. (425688)

The procedure change process in Administrative Procedure 10.41, Site Procedure and Policy Processes, requires the Verification Reviewer to perform all of the following actions except:

A. Ensure usability by the least experienced qualified worker.

B. Determine impact on nuclear safety, where applicable.

C. Ensure consistency with NRC commitments.

D. Ensure consistency with INPO commitments.

Answer: A PLP - 2018 NRC EXAM Page: 252 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 253 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 70 ID: 847470 Points: 1.00 The following is an excerpt from Admin Procedure 4.02, Control of Equipment', Attachment 4:

When performing this checklist for the component listed as MO-1043A Indication, you observe the indication as 'OPEN'.

Given these conditions, complete the following statements:

Without any further instruction, the 'INITIAL' block for MO-1043A Indication, should be ___(1)___.

This Checklist is to be completed ___(2)___.

(1) (2)

A. left blank at any time during each shift B. left blank within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the beginning of shift C. initialed with comments added within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the beginning of shift D. initialed with comments added at any time during each shift Answer: B Answer Explanation Answer: B B. left blank, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the beginning of shift; is correct.

Explanation / Reference Admin Procedure 4.02 rev 81 Attachment 4 page 1:

Items 1 and 6 below support the correct answer.

INSTRUCTIONS

1. Initial ONLY if the component is in the position required by the checklist. Note all exceptions in the COMMENTS/EXCEPTIONS section on the applicable page.

2. Each individual performing the checklist shall PRINT his/her FULL NAME, then INITIAL and DATE in the space provided below.

3. Performance of checklists is addressed in Palisades Administrative Procedure 4.02.

4. Any components found out of position shall not be repositioned without prior approval by the Shift Manager.

PLP - 2018 NRC EXAM Page: 254 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1

5. As part of SRO turnover, the oncoming shift shall verify that the offgoing shift completed this checklist and has resolved any discrepancies per existing procedures.

6. Checklist is to be completed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the beginning of shift.

From Admin Procedure 4.02, attachment 4 page 2: (Frequency supports plausibility of distractors)

PURPOSE Provide a shiftly checklist of safety related components which if not in their normal position would be incapable of performing their safety function, and do not have a readily evident means of detecting them being out of their normal position (ie, alarms, key inserted etc).

Criteria for Checklist

1. Component is safety related or important to safety.

2. The following is criteria for not listing components on this checklist:

Components that would go to their safety position upon an actuation signal regardless of the position of the switch.

Key operated valves (with exceptions).

Any component which if taken out of the position listed would result in the component being in its safety position.

Any component out of position that would result in an alarm.

Any component whose position is checked as part of SHO-1.

FREQUENCY

1. Perform shiftly in Modes 1, 2 and 3.

Distractors are incorrect because the stem asks when does the checklist have to be completed, which is different than the frequency of performance.

'initialed with comments added' distractors are plausible because noting all exceptions in the comments

/exceptions is part of item 1 of the instructions. It is incorrect because the first part of instruction one, directs initialing only if the component is in the position required (which it is NOT).

'at any time during each shift' is plausible because some logs require most of the shift to complete. It is incorrect because instruction 6 specifies 'within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the beginning of shift'.

KA: 2.2.14 Knowledge of the process for controlling equipment configuration or status.

(CFR: 41.10 / 43.3 / 45.13)

IMPORTANCE RO 3.9 SRO 4.3 Test item meets KA. Answer requires knowledge of how to perform a checklist that tracks / controls equipment status.

MEMORY PLP - 2018 NRC EXAM Page: 255 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 70 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 847470 User-Defined ID: G2.2.14 Cross Reference Number: APCO_E06.01-1 The following is an excerpt from Admin Procedure 4.02, Control Topic:

of Equipment', Attachment 4: Whe Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 256 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 257 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 71 ID: 850975 Points: 1.00 Note to examinee: RWP 2020-12345 is provided as a handout Complete the following statements based on the information provided in RWP No. 2020-12345.

You are the operator tasked with performing the manipulations of Valves 'A' and 'B'.

You are signed onto RWP No. 2020-12345 and are now at the location of the valves.

After unlocking and closing valve 'A', some liquid is observed to be dripping from the stem of valve 'A',

which is absorbed by the absorbent material in the bag.

After this manipulation, you notice a tear in the outer glove of your right hand.

You reopen and backseat valve 'A' and the liquid flow stops.

The manipulations you performed ___(1)___ covered by this RWP.

A re-survey of the valve ___(2)___ required.

(1) (2)

A. ARE IS B. ARE NOT IS C. ARE IS NOT D. ARE NOT IS NOT Answer: B Answer Explanation Answer: B B. ARE NOT, IS; is correct.

Provide RWP 2020-12345 Explanation:

The covered tasks were for unlocking and closing valves 'A' and 'B'. NO provision for backseating valve

'A' is listed. the small amount of leakage was easily absorbed by the material in the bag, and the bag could drain to the 5 gallon collection bottle if necessary. Thus, part one is correct.

The RWP called for a resurvey if any moisture was evident. The drip from the stem, (even though it was stopped) required a resurvey. Thus, part two is correct.

Distractors are plausible for the examinee that hasn't mastered compliance with an RWP instructions and think that immediate action was necessary (and within the scope of the task) to stop a drip, and since it stopped, no survey was needed.

KA:

PLP - 2018 NRC EXAM Page: 258 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions.

(CFR: 41.12 / 45.10)

IMPORTANCE RO 3.5 SRO 3.6 Test item meets KA, as the answer requires reading and understanding the RWP in order to comply with its instructions during a normal condition.

HIGH COG - Application of instructions.

Objective: Discuss the methods used to prevent contamination of personnel and areas, including:

a. planning work and conducting prejob briefings

b. using protective clothing (PCs)

c. avoiding potentially contaminated water

d. avoiding skin contact with contam Question 71 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 850975 User-Defined ID: 2.3.7 Cross Reference Number: RWT39 Note to examinee: RWP 2020-12345 is provided as a handout Topic:

Complete the following statements based Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 259 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 260 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 72 ID: 850972 Points: 1.00 The Plant is at 50% power.

A Containment Entry needs to be made.

The limit on the change in Heat Balance Power during a Containment Entry is ___(1)___%

power.

Raising letdown flow by placing an additional letdown orifice in service ___(2)___ violate the constant power requirements of the Containment Entry procedure.

(1) (2)

A. 0.5 does B. 1.0 does C. 0.5 does NOT D. 1.0 does NOT Answer: C Answer Explanation Answer: C.

C. 0.5; does NOT, is correct.

Explanation:

The limit is 0.5% Heat Balance power per the reference.

Adding one letdown orifice does cause heat balance power to rise. But the amount of the addition, depending on the total number of orifices on line, is between 7 and 9 MWT.

To exceed the limit of 0.5%, the change would need to exceed 12.8 MWT.

Distractors:

Those distractors with 'does' are plausible for the examinee that either don't know the limit or can't figure out how to calculate it.

Those distractors with '1.0' are plausible for the examinee that does NOT know the Containment Entry while at power procedure.

References:

DWO-1 for Heat Balance Power Calculations.

PAL HP procedure HP 2.6, Containment Entry with the Reactor Critical KA:

2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(CFR: 41.12 / 43.4 / 45.9 / 45.10)

PLP - 2018 NRC EXAM Page: 261 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 IMPORTANCE RO 3.4 SRO 3.8 Test item meets KA. Entry into Containment while at power is a radiological safety procedure. Question tests knowledge of requirements.

MEMORY Level Objective: From memory, state the authorized steady-state reactor core power level for the Palisades reactor in accordance with the Palisades Plant Facility Operating License.

Question 72 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 850972 User-Defined ID: 2.3.13 Cross Reference Number: ISTS_E01.01 The Plant is at 50% power. A Containment Entry needs to be Topic:

made. The limit on the change in Heat Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 262 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 263 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 73 ID: 847851 Points: 1.00 EOP-1.0, Standard Post Trip Actions, are in progress due to a LOCA

Immediate Actions are complete

During the performance of the first Operator Action, the NCO-R recognizes the need to trip all Primary Coolant Pumps (PCPs)

To meet the EOP Performance Standards, the NCO-R will...

A. immediately trip all four PCPs, and then report the action taken.

B. announce Pressurizer pressure value, then wait for CRS to direct the action.

C. announce the action to be taken, provide opportunity for intervention from the crew, and then trip all four PCPs.

D. advocate the need to trip all four PCPs, obtain permission from the CRS, and then perform the action once permission is given.

Answer: C Answer Explanation Answer: C C.announce the action to be taken, provide opportunity for intervention from the crew, and then trip all four PCPs; is correct.

DISTRACTER ANALYSIS A. Immediately tripping the PCS, and then reporting the action taken, is plausible for the examinee that believes the action is an IMMEDIATE ACTION of the procedure. Per EOP Performance Standards (AP 4.06, Att. 15), the operator shall take the action, AFTER allowing a short pause for crew intervention.

B. Announce Pressurizer Pressure value, then wait for CRS direction is plausible for the examinee who believes he needs to be directed to perform a required action. Candidate correctly recognizes abnormal Pressurizer pressure concern but fails to correctly apply EOP Performance Standards.

C. CORRECT D. Advocate the need to trip all four PCPs, obtain permission from the CRS, and then perform the action is plausible for the examinee that believes he must get permission to perform a required action. It is important for the operator to immediately take the action, without any direction from the CRS.

References:

From EOP-1.0:

All steps that result in tripping the PCPs are written in the IF, THEN format described below:

From Admin Procedure 4.06 section 5.0 definitions:

PLP - 2018 NRC EXAM Page: 264 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Non-sequential Step - A step which directs the performance of action(s) whenever a specified set of conditions exist. While Non-sequential steps are inserted into the procedure at locations where they would likely apply, they should be performed at any point in the procedure where the conditional aspects of the step become true. Non-sequential steps are of the general form "IF..., THEN..." or "WHEN...,

THEN..." Both forward and backward accountability for Non-sequential step performance is assumed.

When the given sequence is altered by moving forward in the procedure to perform a Non-sequential step, the steps in the procedure between the last performed step and the Non-sequential step should be reviewed for applicability prior to performance of the Non-sequential step.

From Admin Procedure 4.06, Attachment 15 EOP Performance Standards:

Non-sequential steps should be utilized whenever the condition specified in the step is met. If a non-sequential step is needed prior to its normal sequence, then the operator shall ensure that all sequential steps required to perform the function specified by the non-sequential step, are completed prior to performing the non-sequential step. Note that this does not require completion of all sequential steps, only steps applicable to the non-sequential step.

EOP-1.0 Adherence to the standard will facilitate consistent performance of EOP-1.0, enabling an accurate understanding of the plant status and of the Plant's response and allowing for a timely diagnosis of the event.

a. Licensed operators are expected to possess a high level of familiarity of EOP-1.0. An operator aid is provided that paraphrases each A\ER and RNO step in EOP-1.0. This facilitates an accurate and efficient performance of the steps by the NCOs.

b. If during the initial performance of EOP-1.0 any of the RNO steps are required to be performed, then they shall be performed without need for direction from the CRS. The RNO action should be announced before taking the action so that the CRS or another crewmember has an opportunity to intervene if they feel the intended action is not appropriate.

c. After the initial performance of EOP-1.0 has been completed, then a verbal verification of the completion of the A\ER and RNO steps shall be conducted by the CRS utilizing the procedure.

Following verification EOP-1.0, the CRS shall proceed with the performance of the Operator Actions section.

d. Additional actions to those stated in the Standard Post-Trip Actions are not permitted, except as directed as an Immediate Action of an applicable AOP, or are otherwise immediately essential for personnel safety, plant safety, equipment protection or safety of the public.

e. RNO actions shall not be taken early (in advance of the stated criteria) except where early performance of the RNO is immediately essential for personnel safety, Plant safety, equipment protection or safety of the public KA:

2.4.13 Knowledge of crew roles and responsibilities during EOP usage (CFR: 41.10 / 45.12)

IMP 4.0 Test item meets KA. Answer requires knowledge of how / when to perform / report important steps of EOPs.

PLP - 2018 NRC EXAM Page: 265 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 MEMORY Question 73 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 847851 User-Defined ID: G2.4.13 Cross Reference Number: TBAA_E05.01-4 EOP-1.0, Standard Post Trip Actions, are in progress due to a Topic:

LOCA Immediate Actions are complete Num Field 1:

Num Field 2:

Text Field:

Comments: Bank PAL-LOI-5975 Not used on 2014 or 2017 NRC exam PLP - 2018 NRC EXAM Page: 266 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 74 ID: 852017 Points: 1.00 Question 74 withheld from public disclosure due to security-related content.

KA:

2.4.28 Knowledge of procedures relating to a security event (non-safeguards information).

(CFR: 41.10 / 43.5 / 45.13)

IMPORTANCE RO 3.2 SRO 4.1 PLP - 2018 NRC EXAM Page: 267 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 268 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 269 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 75 ID: 850991 Points: 1.00 An event has occurred while you are assigned to the Main Control Room.

The Shift Manager has classified the event as an Unusual Event, with no release in progress and directs you to perform the duties of the communicator for the initial notifications of the Offsite Response Organizations (OROs).

If ALL the InForm machines are NOT in service, the FIRST Agency to contact is...

A. Allegan County B. Berrien County C. Van Buren County D. State of Michigan Answer: C Answer Explanation Answer: C C. Van Buren County is correct Explanation:

A note (page 3 of 6) in the reference procedure (EI-3, Communications and Notifications), states:

If backup methods to communicate the notification form are needed, order of contact will depend on which OROs are not reachable via InForm. Convention for order of contact is that if all InForm machines are not in service, the first agency to contact should be Van Buren County, unless the SOM EOC is operational, then SOM should be contacted first.

Distractors:

The distractors are plausible for those not familiar with the requirements of the procedure(s).

Step 1.2 Identifies the OROS (Offsite Response Organizations):

When the Emergency Operations Facility (EOF) is operational, the EOF Offsite Communicator shall assume notification responsibilities including the Offsite Response Organizations (OROs), which consist of Van Buren County, Allegan County, Berrien County and the State of Michigan (SOM).

KA:

2.4.39 Knowledge of RO responsibilities in emergency plan implementation.

(CFR: 41.10 / 45.11) l IMPORTANCE RO 3.9 SRO 3.8 Test item meets KA. Communicators responsibilities are included in that of an RO. Question tests basic notification knowledge.

MEMORY level PLP - 2018 NRC EXAM Page: 270 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Objective: Given EI-1, list the ACTIONS to be taken by the Palisades Emergency Response Organization for a given emergency level, per EI-1.

Question 75 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 850991 User-Defined ID: 2.4.39 Cross Reference Number: PL-NI00113_E05.01 An event has occurred while you are assigned to the Main Topic:

Control Room. The Shift Manager has class Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 271 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 272 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 76 ID: 848220 Points: 1.00 SRO ONLY With the Plant at full power, a reactor trip occurs due to a loss of off-site power.

During the actions of EOP-1.0, "Standard Post Trip Actions," a fire develops in the cable spreading room causing ALL Preferred AC Buses to DEENERGIZE.

Given these conditions, complete the following statements:

The location for controlling the plant is at ___(1)___.

The appropriate procedure to mitigate this event is ___(2)___.

(1) (2)

A. Panel C-150/C-150A EOP-3.0, "Station Blackout Recovery" B. Panel C-33 EOP-3.0, "Station Blackout Recovery" C. Panel C-33 EOP-9.0, "Functional Recovery Procedure" D. Panel C-150/C-150A EOP-9.0, "Functional Recovery Procedure" Answer: D Answer Explanation Answer: D D. Panel C-150/C-150A, EOP-9.0 Functional Recovery Procedure; is correct. CORRECT - EOP-9.0 is entered when diagnosis of an ORP is not apparent and safety functions are not met.

With NO Preferred AC buses energized, the indications at C-33 don't work.

Explanation:

EOP-1.0 attachment 1. Diagnostic flowchart directs the user to EOP-9.0 when there are not at least 3 Preferred AC Buses Energized.

AOP-40, Fire Which Threatens Safety Related Equipment entry conditions are met, as are AOP-41 Alternate Safe shutdown Procedure.

Both AOPs direct the evacuation of the Control Room. The protocol includes establishing control from panels C-150/C-150A.

DISTRACTOR ANALYSIS

a. Part one is correct as explained in the answer. Part two is Plausible because the student may believe that EOP-3.0 is appropriate due to no preferred AC.

b. Part one is plausible as there are some controls at C-33, but incorrect because the Panel C-33 indications are powered from preferred AC. Part two is Plausible because the student may believe that EOP-3.0 is appropriate due to no preferred AC.

PLP - 2018 NRC EXAM Page: 273 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1

c. Part one is plausible as explained in B, but panel C-33 indications are powered from preferred AC.

Part two is correct as explained in the answer.

KA: 007 CE E02 Reactor Trip, Stabilization, Recovery 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

(CFR: 41.10 / 43.5 / 45.2 / 45.6)

IMPORTANCE RO 4.3 SRO 4.4 Test item meets KA. Question asks SROs what procedure gets performed under these post trip loss of power conditions, and where are the actions performed from.

HIGH COG - assess conditions and determine procedure to enter, taking into account the loss of power.

SRO ONLY:

Assessment of Facility Conditions and Selection of Appropriate Procedures during Normal, Abnormal, and Emergency Situations [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both (1) assessing plant conditions (normal, abnormal, or emergency) and then (2) selecting a procedure or section of a procedure to mitigate or recover, or with which to proceed. One area of SRO-level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item. The following are examples:

knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

knowledge of diagnostic steps and decision points in the emergency operating procedures (EOPs) that involve transitions to event-specific sub-procedures or emergency contingency procedures Objective:

Given post reactor trip conditions, determine the proper follow-up EOP in accordance with the Diagnostic Flowchart.

PLP - 2018 NRC EXAM Page: 274 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 76 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 2.00 System ID: 848220 User-Defined ID: 007G2.1.23 Cross Reference Number: TBAB_E01.06-17 SRO ONLY With the Plant at full power, a reactor trip occurs Topic:

due to a loss of off-site power. Du Num Field 1:

Num Field 2:

Text Field:

Comments: 2008 NRC SRO EXAM PAL-LOR-4175 Bank PLP - 2018 NRC EXAM Page: 275 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 276 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 77 ID: 848251 Points: 1.00 SRO ONLY The plant is at 100% power.

At time 1000:00,

Component Cooling Water (CCW) Surge tank level starts lowering

CCW discharge pressure starts lowering

AOP-36, Loss of Component Cooling Water is entered At time 1005:00

AOP-29, Primary Coolant Pump (PCP) Abnormal Conditions is entered due to low CCW flow and rising PCP temperatures causing alarms on P-50A, PCP At time 1010:00,

CCW to P-50A seal cooling has degraded to the point where imminent PCP Failure indications are present for P-50A

P-50A Lower Guide bearing temperature is 184°F and rising

P-50A vibrations are 35 mils and oscillating Given these conditions, complete the following statements:

P-50A should be manually tripped ONLY after the reactor has been tripped and the ___(1)___

Safety Function is satisfied.

After tripping ONLY P-50A, CV-2083, CV-2099, and CV-2191, Controlled Bleedoff valves should be ___(2)___.

(1) (2)

A. Reactivity CLOSED B. Reactivity left OPEN C. Core Heat Removal CLOSED D. Core Heat Removal left OPEN Answer: B Answer Explanation Answer: B.

B. Reactivity, left OPEN; is correct.

Explanation:

The conditions presented place the examinee at step 3 of AOP-29, PCP Abnormal Conditions.

PLP - 2018 NRC EXAM Page: 277 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 When imminent PCP failure is corroborated by other indications, trip the reactor, then when the reactivity Safety Function is satisfied, trip the affected PCP.

Step 5 of the same procedure checks PCP parameters below action limits for the affected PCP. If they are not and a LOSS of CCW exists, a similar direction is stated in the RNO. Additionally, Step 5.2.c states when ALL PCPs are stopped, then close the PCP Controlled Bleedoff valves.

Distractors:

Distractors with Core Heat Removal are plausible as that is a Safety Function that needs to be checked as satisfied in EOP-1.0, but it does NOT include any actions with regard to tripping PCPs. It only verifies at least one running.

Distractors with CLOSED, are plausible, because it is an action that is taken after ALL PCPs are stopped.

The stem states only P-50A was stopped.

Reference:

AOP-29, PCP Abnormal Conditions, rev 5, steps 3 and 8.

KA:

015 RCP Malfunctions AA2. Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

(CFR 43.5 / 45.13)

AA2.10 When to secure RCPs on loss of cooling or seal injection . . . . . . . . . . . . 3.7 3.7 Test item meets KA. Question asks when to secure RCP / PCP on loss of seal cooling.

HIGH COG - comprehension of degraded conditions. Do they exceed allowable values necessitating a trip?

SRO ONLY:

Assessment of Facility Conditions and Selection of Appropriate Procedures during Normal, Abnormal, and Emergency Situations [10 CFR 55.43(b)(5)]. This 10 CFR 55.43 topic involves both (1) assessing plant conditions (normal, abnormal, or emergency) and then (2) selecting a procedure or section of a procedure to mitigate or recover, or with which to proceed. One area of SRO-level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item. The following are examples:

knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

knowledge of diagnostic steps and decision points in the emergency operating procedures (EOPs) that involve transitions to event-specific sub-procedures or emergency contingency procedures

knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures The question requires SRO ONLY level of specific procedure content knowledge on which safety function needs to be satisfied before tripping a PCP, and whether conditions are met to close the controlled bleedoff valves.

PLP - 2018 NRC EXAM Page: 278 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Objective:

Given plant conditions specific to operation of the PCPs, determine if an operating PCP must be shut down under emergency conditions in accordance with in use procedures. (003 A2.02)

Question 77 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848251 User-Defined ID: 015AA2.10 Cross Reference Number: PCP_E03.03 SRO ONLY The plant is at 100% power. At time 1000:00, Topic:

Component Cooling Water (CCW) Surge tank l Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 279 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 280 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 78 ID: 848270 Points: 1.00 SRO ONLY The plant has just established Shutdown Cooling (SDC) when the LPSI pumps were found to be steam bound.

AOP-30, Loss of Shutdown Cooling, has been entered

Highest PCS temperature has risen to 315°F

BOTH Containment Spray (CS) pumps are available

SIRWT is at 80%

Containment is accessible Given these conditions, complete the following statements:

When performing the step to ENSURE PCS Level is restored as high as possible, the step

___(1)___ imply it is permissible to exceed Technical Specification Cooldown Limits.

Using a Containment Spray (CS) pump to remove steam from a steam bound LPSI pump

___(2)___ make the CS pump inoperable.

(1) (2)

A. does NOT would B. does NOT would NOT C. does would D. does would NOT Answer: A Answer Explanation Answer: A.

A. does NOT, would; is correct.

Explanation:

The basis for step 8 of AOP-30, Loss of SDC, explicitly states this step does NOT imply that it is permissible to exceed tech spec limits. Thus, part one is correct.

In AOP-30, Loss of Shutdown Cooling, steps 48-71 are specific directions for the situation of a steam bound LPSI pump.

The basis for step 50, which Verifies PCS temperature less than 300°F, is stated as an operability requirement for the CS pumps. LCO 3.6.6 requires the CCS to be operable in Mode 3 (> 300°F). The realignment of the CS pumps to vent the steam bound LPSI pumps would violate the required alignment of CS for being in Mode 3.

Distractors:

PLP - 2018 NRC EXAM Page: 281 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 The 'does' distractors (C and D) are plausible for the examinee that believes the procedure direction to recover PCS level as high as possible, overrides TS limits. It does NOT.

The 'would NOT' distractor (B) is plausible for the examinee that doesn't know the Mode 3 requirement for CCS.

KA:

025 Loss of RHRS 2.4.6 Knowledge of EOP mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13)

IMPORTANCE RO 3.7 SRO 4.7 Test item meets KA. Question asks specific procedure content knowledge on actions directed by the procedure (mitigating strategy) lineup to address a loss of SDC (RHR), due to a specific cause (steam binding of the LPSI pumps) and how the mitigating strategy actions are governed by tech specs.

SRO ONLY:

Assessment of Facility Conditions and Selection of Appropriate Procedures during Normal, Abnormal, and Emergency Situations [10 CFR 55.43(b)(5)] This 10 CFR 55.43 topic involves both (1) assessing plant conditions (normal, abnormal, or emergency) and then (2) selecting a procedure or section of a procedure to mitigate or recover, or with which to proceed. One area of SRO-level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item. The following are examples:

knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

knowledge of diagnostic steps and decision points in the emergency operating procedures (EOPs) that involve transitions to event-specific sub-procedures or emergency contingency procedures

knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures The is SRO ONLY because specific procedure content knowledge is required to answer the 'how is recovery accomplished?' question. It is detailed knowledge that is beyond the basic mitigating strategy required of an RO.

MEMORY PLP - 2018 NRC EXAM Page: 282 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 78 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848270 User-Defined ID: 025G2.4.6 Cross Reference Number: PL-OPS-AOP-030S SRO ONLY The plant has just established Shutdown Cooling Topic:

(SDC) when the LPSI pumps were found to Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 283 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 284 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 79 ID: 848285 Points: 1.00 SRO ONLY:

A Steam Generator Tube Rupture has occurred on the 'A' Steam Generator. The appropriate EOP(s) are being performed:

A S/G has been isolated

A S/G pressure is 840 psia

B S/G pressure is 560 psia

Pressurizer pressure is 960 psia

Pressurizer level is 42%

PCS temperature is 510°F

'A' S/G Level is 80% and rising For these conditions, which of the following actions, regarding Pressurizer pressure, will the CRS direct?

A. Raise Pressurizer pressure to 980 psia to re-establish adequate subcooling margin B. Raise Pressurizer pressure to 970 psia to prevent backflow dilution of the PCS C. Lower Pressurizer pressure to 750 psia to initiate backflow into the PCS D. Lower Pressurizer pressure to 930 psia to ensure Main Steam Safety Valves remain closed Answer: D Answer Explanation Answer: D D. Correct. Pressurizer Pressure shall be maintained per the following criteria of EOP-5.0 Step 17 (continuously applicable):

-Less than 940 psia

-within limits of EOP supplement 1

- preferably within 50 psid of the isolated SG pressure Maintaining pressure within 50 psid minimizes the flow from the PCS to the SG.

With PCS temperature at 510°F, there is enough energy in the PCS to cause sufficient heat transfer to the ruptured SG and lift its MSSV. Continuing to allow PCS to leak into the ruptured SG could lead to overfill which may result in lifting the MSSV.

Distractor Analysis:

A. Incorrect. Subcooling margin is adequate at 28°F.

B. Incorrect. Raising PZR pressure would minimize backflow and dilution, but the basis document for the EOP states the amount of dilution would not jeopardize SDM, and there'd be no backflow because the affected S/G is already less than that pressure.

C. Incorrect. Backflow is necessary when SG level exceeds 140%. (step 35) It is hardly a concern at the conditions given. However, it is incorrect to lower pressure such that the differential exceeds 50 psid.

Thus this distractor is incorrect.

PLP - 2018 NRC EXAM Page: 285 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1

Reference:

EOP-5.0, EOP-5.0 Basis, EOP supplement 1 SRO Objective:

Given Plant conditions involving a Steam Generator Tube Rupture, describe the operator actions necessary to minimize Primary to Secondary leakage in accordance with EOP-5.0.

KA:

038 SGTR EA2 Ability to determine or interpret the following as they apply to a SGTR:

(CFR 43.5 / 45.13)

EA2.15 Pressure at which to maintain RCS during S/G cooldown . . . . . . . . . . . . . . . 4.2 4.4 Test item meets KA. Answer requires assessment of conditions, and selecting of correct answer with justification of why it is correct, regarding what to do with PCS pressure during a SGTR cooldown.

HIGH COG - decision making involved to assess the conditions and determine a course of action by comparing given parameter values and trends to procedurally required norms.

SRO ONLY:

Assessment of Facility Conditions and Selection of Appropriate Procedures during Normal, Abnormal, and Emergency Situations [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both (1) assessing plant conditions (normal, abnormal, or emergency) and then (2) selecting a procedure or section of a procedure to mitigate or recover, or with which to proceed. One area of SRO-level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item. The following are examples:

knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

knowledge of diagnostic steps and decision points in the emergency operating procedures (EOPs) that involve transitions to event-specific sub-procedures or emergency contingency procedures PLP - 2018 NRC EXAM Page: 286 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 79 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848285 User-Defined ID: 038EA2.15 Cross Reference Number: TBAF_E01.01 SRO ONLY: A Steam Generator Tube Rupture has occurred on Topic:

the 'A' Steam Generator. The appropriate Num Field 1: 038G2.1.20 Num Field 2:

Text Field:

Comments: modified from LO-2017N77 which has different initial conditions and the answer and distractors are different:

SRO ONLY:

A Steam Generator Tube Rupture has occurred on the 'A' Steam Generator. The appropriate EOP(s) are being performed:

A S/G has been isolated

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 508°F For these conditions, which of the following actions, regarding Pressurizer pressure, will the CRS direct?

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 PC Answer: D PLP - 2018 NRC EXAM Page: 287 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 288 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 289 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 80 ID: 848302 Points: 1.00 SRO ONLY The plant is addressing a Loss Of All Feedwater per EOP-7.0:

ALL PCPs have been stopped

Pressurizer Pressure is 1250 psia

PZR Level is 10%

SI Pump Throttling criteria is NOT met The Control Room Supervisor has entered EOP Supplement 39, Alternate Methods of Reducing PCS Pressure.

Given these conditions complete the following statements:

The method the CRS will direct the NCO to use to reduce PCS Pressure is ___(1)___.

The chosen method will result in exceeding ___(2)___.

A. (1) Establishing Auxiliary Spray using HPSI pumps (2) Spray line differential temperature B. (1) Establishing Auxiliary Spray using HPSI pumps (2) Pressurizer cooldown limits C. (1) By controlling PZR level using HPSI injection valve throttling (2) Spray line differential temperature D. (1) By controlling PZR level using HPSI injection valve throttling (2) Pressurizer cooldown limits Answer: A Answer Explanation Answer: A (1) Establishing Auxiliary Spray using HPSI pumps (2) Spray line differential temperature Explanation:

EOP Supplement 39 has 3 options for reducing PCS Pressure using alternate methods.

Method 1 is Pressure Control by Controlling PZR Level Using HPSI Injection Valve Throttling.

Method 2 is Establishing Aux Spray with HPSI pumps Method 3 is Venting the PZR.

To use method 1, SI Throttling criteria must be met. As stated in the stem, it is not.

Using method 3 may result in negative impacts on PCS Pressure Control and Inventory Control safety functions. Method 3 was not provided as an option to choose to use.

The most appropriate choice is method 2. Thus, part one is correct.

PLP - 2018 NRC EXAM Page: 290 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Within the steps of method 2, there are NOTES informing the user of the spray delta T limits and that they will be exceeded.

The normal delta T limit is 200°F, and the abnormal (emergency/extreme) limit is 350°F The HPSI pumps take suction from the SIRWT, the temperature of which is 100°F max.

Both delta T limits are exceeded.

Distractors:

Method 1 is plausible, as it is listed first in the EOP Supplement, but the required conditions to use it are not met. Since the conditions to use method one are not met, then the raising and lowering of PZR level from 20-90% will not occur, and the PZR cooldown rate won't be challenged.

Plausible for those that don't know the limits, or the conditions under which the limits apply.

C is plausible for those that believe that abnormal limit means under an abnormal situation, the normal limit is not applicable.

Part 1 of distractors C and D are plausible because they are methods described in the procedure, but incorrect because SI throttling Criteria are NOT met.

Part 2 of distractors B and D are plausible because they are likely to be challenged if using method 1 to fill and drain the PZR. But they are incorrect because the criteria to use method 1 are not met.

Reference:

EOP-Supplement 39 Section 2.0 and 3.0 and notes; EOP-7.0 Caution at step 38.

KA: 054 Loss of Feedwater 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

(CFR: 41.10 / 43.5 / 45.13)

IMPORTANCE RO 3.8 SRO 4.3 Test item meets KA. Question requires knowledge of a NOTE and Caution and when they apply from a Loss of Feedwater and associated procedures.

SRO ONLY:

Assessment of Facility Conditions and Selection of Appropriate Procedures during Normal, Abnormal, and Emergency Situations [10 CFR 55.43(b)(5)] This 10 CFR 55.43 topic involves both (1) assessing plant conditions (normal, abnormal, or emergency) and then (2) selecting a procedure or section of a procedure to mitigate or recover, or with which to proceed. One area of SRO-level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item. The following are examples:

knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

knowledge of diagnostic steps and decision points in the emergency operating procedures (EOPs) that involve transitions to event-specific sub-procedures or emergency contingency procedures

knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures PLP - 2018 NRC EXAM Page: 291 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 SRO ONLY due to the incorporating the use of a Supplemental EOP, and the specific content knowledge of the notes and cautions and how they are applied.

HIGH COG - assess conditions and select procedure section.

Objective:

Given Emergency plant conditions and from memory, perform immediate actions to mitigate the event and stabilize the plant in accordance with the applicable Emergency Operating Procedure(s) and EOP Performance Standards (AP-4.06).

Question 80 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 2.00 System ID: 848302 User-Defined ID: 054G2.4.20 Cross Reference Number: TBCORE_CP01.0 SRO ONLY The plant is addressing a Loss Of All Feedwater Topic:

per EOP-7.0: ALL PCPs have been stopped Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 292 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 293 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 81 ID: 848311 Points: 1.00 SRO ONLY:

At time 1000:00,

The plant was at 100% power, when a complete loss of Instrument Air (IA) occurred

The plant was tripped due to erratic equipment response from lowering IA pressure

Troubleshooting and repairs have begun At time 1030:00,

A high energy line break inside containment causes containment pressure to reach 4.2 psig At time 1300:00,

IA header pressure is 50 psig and still lowering Given these conditions, complete the following statements:

Closing manual valves MV-CC712, CCW Supply to Containment and MV-CC713, CCW Return from Containment by ___(1)___ at the LATEST, is required by Technical Specifications.

The closure of these manual valves is directed in ___(2)___ .

(1) (2)

A. 1400:00 AOP-37, Loss of Instrument Air B. 1400:00 AOP-36, Loss of Component Cooling Water C. 1700:00 AOP-37, Loss of Instrument Air D. 1700:00 AOP-36, Loss of Component Cooling Water Answer: A Answer Explanation Answer: A A. 1400:00, AOP-37, Loss of Instrument Air, is correct Explanation:

The two valves in question are required to be closed by AOP-37, Loss of instrument Air, Step 10, if IA header pressure is less than 80 psig for greater than 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and CCW Containment Isolation valves are required to be closed. The closure of these manual valves must occur within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of the loss of IA per Tech Spec 3.6.3, Action A.1.

The loss of IA occurred at 1000:00. The CIS occurred on CHP at 1030:00. This established the required to be closed part of the procedure guidance as 1330 to meet the procedure, but 1400 would be the end of the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowance from the loss of air which made the CIVs inoperable.

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EXAMINATION ANSWER KEY Daft N 1 The CCW CIS valves are normally open, and AIR to CLOSE. On a loss of AIR, they will open. The have accumulators, but after a few hours, the accumulators are exhausted. The CIS automatically closes these normally open valves for adequate cooling concerns. Thus the requirement to close the manual CCW valves upstream of the CCW CIVs.

Tech Spec 3.7.7, and Surveillance 3.7.7.2 requires two operable CCW trains. The surveillance requires the valves to perform their safety related function on an actual or simulated signal. When a CIV is inoperable, TS 3.6.3 is entered requiring closure of a manual valve.

Distractors:

1700:00 is plausible because it is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the 1300 time which was listed in the stem as when IA pressure was at 50 psig.

AOP-36, Loss of Component Cooling, is plausible for those who don't know the AOP contents, or believe that the CIS, should've resulted in the isolation of CCW from Containment, and don't know how the loss of Air / and accumulator pressure affects the CIVs.

KA: 065 Loss of IAS AA2. Ability to determine and interpret the following as they apply to the Loss of Instrument Air:

(CFR: 43.5 / 45.13)

AA2.08 Failure modes of air-operated equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.9* 3.3 Test item meets KA. Question poses a loss of air and requires the determination of the failure position of a couple of CCW CIVs. The action to take under specific circumstances is specific procedure content knowledge at the SRO level.

HIGH COG - determining when a time clock starts/ ends, and the actions required to meet procedural requirements requires comprehension of the TS and procedure.

SRO ONLY:

Facility Operating Limitations in the Technical Specifications and Their Bases

[10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic the following:

application of required actions (TS Section 3) and surveillance requirements (SR) (TS Section 4) in accordance with rules of application requirements (TS, Section 1)

application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).

knowledge of TS bases that are required to analyze TS-required actions and terminology

same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM)

SRO ONLY:

Assessment of Facility Conditions and Selection of Appropriate Procedures during Normal, Abnormal, and Emergency Situations [10 CFR 55.43(b)(5)] This 10 CFR 55.43 topic involves both (1) assessing plant conditions (normal, abnormal, or emergency) and then (2) selecting a procedure or section of a procedure to mitigate or recover, or with which to proceed. One area of SRO-level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item.

PLP - 2018 NRC EXAM Page: 295 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 This answer requires specific content knowledge of AOP-37, as the source of the direction to isolate CCW to and from containment manually.

Objective:

Given Abnormal Operating plant conditions and control room references, SELECT the applicable Technical Specification LCO REQUIRED ACTIONS and COMPLETION TIMES in accordance with Technical Specifications. (K/A G2.2.42)

Question 81 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 848311 User-Defined ID: 065AA2.08 Cross Reference Number: IOTF_CK09.0 SRO ONLY: At time 1000:00, The plant was at 100% power, Topic:

when a complete loss of Instrument Air (I Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 296 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 297 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 82 ID: 851536 Points: 1.00 SRO ONLY:

The plant is holding at 24% power during a power escalation after a brief shutdown for maintenance.

Rods are being balanced The following two alarms come in simultaneously:

EK-0905, Shutdown Rod Position Abnormal

EK-0906, Incore Alarm Given these conditions, complete the following statements.

The alarm requiring the soonest action is ___(1)___.

The soonest required action is to ___(2)___.

(1) (2)

A. EK-0905 withdraw the rod to above 124 inches B. EK-0905 determine the LHR is within the COLR limits C. EK-0906 withdraw the rod to above 124 inches D. EK-0906 determine the LHR is within the COLR limits Answer: A Answer Explanation Answer: A A. EK-0905, withdraw the rod to above 124 inches; is correct.

Explanation:

When a Shutdown rod position is abnormal, it is below the TS required height of 128 inches (LCO 3.1.5).

The alarm comes in between 119 and 123.5 inches. It is applicable in Modes 1 and 2 with any regulating rod above 5 inches. At 24% power, the regulating rods are all above 5 inches.

The action for one or more shutdown rods not within the limit, is to immediately declare the rod inoperable AND enter LCO 3.1.4. The applicable condition for a shutdown rod in LCO 3.1.4 is Condition E, which has an action of being in mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The Tech Spec Bases for LCO 3.1.4 states that if one or more s/d rods are inserted beyond the insertion limit, several conditions apply from both LCO 3.1.4 and 3.1.5, but the if the rod can be moved, it should be withdrawn and all conditions exited.

As for the Incore alarm, the LCO addressing the LHR does NOT apply below 25% power.

The ARP for the Incore alarm directs the determination of which Incore(s) are alarming from the PPC.

This alarm is determined by the PIDAL computer program.

PLP - 2018 NRC EXAM Page: 298 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1

References:

TS and Bases for LCO 3.1.4, and 3.1.5 and 3.2.1 ARP-5 EK-0905 and EK-0906 Distractors:

Combined with the answer explanation, All distractors are plausible for the examinee that doesn't know the LCOs, Applicability and the BASES for LCO 3.1.4, and 3.1.5.

B. part one is correct. Part two is plausible for the examinee that believes the rod misalignment has affected the LHR.

C part one is plausible for the examinee that believes the rod misalignment is causing the incore alarm.

Part two is correct.

D. Part one is plausible as in C, part two is plausible as in B.

KA: 005 Inoperable or Stuck / Misaligned Control Rod 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10 / 43.5 / 45.3 / 45.12)

IMPORTANCE RO 4.1 SRO 4.3 Test item meets KA. Question poses situation where a couple of alarms potentially associated with a stuck or misaligned control rod come in and prioritizing the correct alarm is necessary.

HIGH COG - comparison of the attributes of two LCOs, requires comprehension of the application of the LCOs.

SRO ONLY:

Facility Operating Limitations in the Technical Specifications and Their Bases [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic the following:

application of required actions (TS Section 3) and surveillance requirements (SR) (TS Section 4) in accordance with rules of application requirements (TS, Section 1)

application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).

knowledge of TS bases that are required to analyze TS-required actions and terminology The question is SRO ONLY because a thorough understanding of the information on the expected action to take if the rod(s) are not stuck in the TS bases provides the means to determine which LCO applies soonest.

Objective: From memory, describe the operational implications of the following Control Rod Drive system annunciators:

- EK-0971, "SPI TROUBLE"

- EK-0947, "EMERGENCY ROD DRIVE POWER INTERRUPT"

- EK-0905, "SHUTDOWN ROD POSITION ABNORMAL"

- EK-0911, "ROD POSITION 4 INCHES DEVIATION"

- EK-0912, "ROD POSITION 8 INCHES DEVIATION"

- EK-0917, "ROD WITHDRAWAL PROHIBIT"

- EK-0923, EK-0929, EK-0935, EK-0941, "GROUP 1, 2, 3, 4 PREPWR DEPENDENT INSERTION LIMIT"

- EK-0924, EK-0930, EK-0936, EK-0942, "GROUP 1, 2, 3, 4 PWR DEPENDENT INSERTION LIMIT"

- EK-0948, "DROPPED ROD"

- EK-0916, "CONTROL RODS OUT OF SEQUENCE" PLP - 2018 NRC EXAM Page: 299 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1

- EK-0918, "PIP TROUBLE"

- EK-0954, "ROD DRIVE SEAL LEAK OFF HI TEMP" in accordance with ARP-5.

Question 82 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 851536 User-Defined ID: 005G2.4.45 Cross Reference Number: CRD_CK13.0 SRO ONLY: The plant is holding at 24% power during a power Topic:

escalation after a brief shutdown for Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 300 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 301 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 83 ID: 847672 Points: 1.00 SRO ONLY The Plant is at 50% power with the following conditions:

LIC-0101B, B Channel PZR Level Controller, is in service in CASCADE

P-55A, Charging Pump, is operating with P-55B and P-55C secured in Automatic

CV-2003, Letdown Orifice Stop Valve, is in Automatic and open

CV-2004 and CV-2005, Letdown Orifice Stop Valves, are in AUTO and closed

SS-T AVE , Avg Temp Display Select Switch, on Panel C-02 is in the LOOP 2 position Then, the following occurs:

EK-0761, PRESSURIZER LEVEL HI-LO, annunciates

EK-0967, LOOP 1/LOOP 2 TAVE DEVIATION, annunciates

EK-0969, LOOP 2 T AVE / T REF GROSS DEVIATION, annunciates

The setpoint meter (blue pointer) on LIC-0101B indicates 42%

The process meter (red pointer) on LIC-0101B indicates 49%

TI-0110, T AVE Digital Temperature Indicator, on Panel C-02 indicates 515°F

A valid POWER_PIP_DELTA_T reading on the PPC is 45%

Given these conditions, complete the following statements:

For PDIL monitoring, the POWER_PIP_DELTA_T is ___(1)___.

The procedure and action(s) to address these conditions are ___(2)___.

A. (1) OPERABLE (2) alternate PZR Level Controller to LIC-0101A per AOP-22, Pressurizer Level Control Malfunctions B. (1) OPERABLE (2) select valid Tave channel with SS-Tave Selector Switch per AOP-27, Tave/Tref Controller Failure C. (1) INOPERABLE (2) alternate PZR Level Controller to LIC-0101A per AOP-22, Pressurizer Level Control Malfunctions D. (1) INOPERABLE (2) select valid Tave channel with SS-Tave Selector Switch per AOP-27, Tave/Tref Controller Failure Answer: D Answer Explanation Answer: D.

PLP - 2018 NRC EXAM Page: 302 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 D. (1) INOPERABLE, (2) select valid Tave channel with SS-Tave Selector Switch per AOP-27, Tave/Tref Controller Failure; is correct.

Explanation /

Reference:

Part one of the question is asking about the operability of an input into the Power Dependent Insertion Limits (PDIL), after determining that one of the inputs into the calculator is bad and requires knowledge of the operability statements in the AOP.

The second part requires the examinee to determine if there is a failure of an input, what the failure is based on the indications given, and is asking about knowledge of which procedure step takes care of the failure. AOP-27, Tave/Tref Controller Failure, NOTE after step 6.3.2 states:

With PIP delta T Power higher than actual Reactor Power by more than 4%, the NODE is considered OPERABLE for PDIL monitoring, although an overly conservative PDIL is being calculated.

The above NOTE is followed by step 4.1, which states:

IF PIP Delta T Power is lower than actual Reactor Power by more than 4%, then declare POWER_PIP_DELTA_T inoperable for PDIL monitoring.

The fact that there is an allowable band stated in the procedure makes both choices (OPERABLE, and INOPERABLE) plausible. 2% in the other direction would make a distractor correct.

While the PZR Level Hi-lo alarm is listed as an entry condition for AOP-22, the reason the alarm is in is because an RTD in the circuit calculating Tave has failed, providing a signal that generates the lower than normal setpoint for the PZR level controller. In other words, there is nothing wrong with the PZR Level controller, but it is responding to a bad input. Thus, the exit condition for AOP-22 is met too.

The Tave deviation and Gross Deviation alarms are entry conditions for the correct procedure entry into AOP-27. The vaild POWER PIP DT reading from the PPC is more than 4% below the actual Reactor power of 50%. Steps 3 and 4 of the procedure direct the declaration of this input as INOPERABLE. Thus, part one is correct.

Step 2 of the procedure directs the selection of a valid channel of Tave. Thus, part two is correct.

Distractors:

A. part one is plausible for those that don't know the operability requirement. Part two is plausible because it is the action to take if misdiagnosing the problem that is in the AOP-22 procedure.

B. part one is plausible as stated in A. Part two is correct.

C. Part one is correct. Part two is plausible as stated in A.

Reference:

AOP-27, steps 2-4.

KA:

028 PLC Malfunction AA2. Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions:

(CFR: 43.5 / 45.13)

AA2.08 PZR level as a function of power level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 3.5 Test item meets KA. To determine the nature of the failure and the procedure and actions that address it, the examinee must determine if the given PZR level agrees with the given power level.

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EXAMINATION ANSWER KEY Daft N 1 Test item is at the SRO ONLY level per guidance provided in NUREG 1021 ES401, Rev 11, Attachment 2:

Assessment of Facility Conditions and Selection of Appropriate Procedures during Normal, Abnormal, and Emergency Situations [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both (1) assessing plant conditions (normal, abnormal, or emergency) and then (2) selecting a procedure or section of a procedure to mitigate or recover, or with which to proceed. One area of SRO-level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item. The following are examples:

knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

knowledge of diagnostic steps and decision points in the emergency operating procedures (EOPs) that involve transitions to event-specific sub-procedures or emergency contingency procedures

knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures The example for this question is knowing what specific step addresses the condition, and what procedure has that step.

Additionally, it is SRO ONLY Level due to job function. It is the SROs job to determine OPERABILITY /

INOPERABILTY.

HIGH COG - requires analysis of data for two control systems to diagnose failure, and then apply knowledge of operability requirements to make operability determination.

Objective:

Given off normal plant conditions, select the applicable Abnormal Operating Procedure to mitigate the event without error. (K/A G2.4.47)

Question 83 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 4 Difficulty: 3.00 System ID: 847672 User-Defined ID: 028AA2.08 Cross Reference Number: IOTF_CK03.0-37 SRO ONLY The Plant is at 50% power with the following Topic:

conditions: LIC-0101B, ?B? Channel PZR Leve Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 304 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 305 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 306 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 84 ID: 851530 Points: 1.00 SRO ONLY:

The Plant is in Mode 5 for a maintenance outage.

An UNPLANNED ENTRY into a Higher Risk Plant Operating States (HRPOS) is required.

Given these conditions, complete the following statements:

Any Hot Work in progress ___(1)___.

LCO 3.0.9, which addresses situations where required barriers are unable to perform their related support functions and provides instructions and conditions for meeting the supported system LCOs ___(2)___ applicable to Fire Barriers.

(1) (2)

A. must be stopped is B. must be stopped is NOT C. can continue is D. can continue is NOT Answer: B Answer Explanation Answer: B B. must be stopped, is NOT; is correct.

Explanation:

Per the reference, Admin 4.49, Section 7.1 requires an announcement that any hot work must be stopped when an unplanned entry into a HRPOS occurs. Thus, part one is correct. (allowing hot work to continue is an acceptable distractor)

Per the Tech spec Bases for LCO 3.0.9, the provisions provided by this LCO do NOT apply to Fire Barriers. Thus, part two is correct. (implying the opposite is an acceptable distractor)

Distractors:

All are plausible for the examinee that is not familiar with the requirements of Admin 4.49, and / or LCO 3.0.9, but incorrect because they don't have both parts correct.

KA: 067 Plant Fire on Site 2.2.38 Knowledge of conditions and limitations in the facility license.

(CFR: 41.7 / 41.10 / 43.1 / 45.13)

IMPORTANCE RO 3.6 SRO 4.5 PLP - 2018 NRC EXAM Page: 307 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Test item meets KA. Tech Specs refer to many other programs that are required to meet various regulations. The Fire Protection Implementing Procedures are one of them. One specific SRO Responsibility stated in FPIP-4 Fire Protection Systems and Fire Protection Equipment, is that the SRO has the responsibility to ensure equipment status and compensatory measures are maintained as required by Palisades Admin Procedure 4.49, 'Non Power Operation Fire Risk Management'. This is the procedure that directs the stoppage of hot work, and LCO 3.09.

MEMORY level SRO ONLY:

A. Conditions and Limitations in the Facility License [10 CFR 55.43(b)(1)] Examples of SRO exam items for this topic include the following:

reporting requirements when the maximum licensed thermal power output is exceeded

administration of fire protection program requirements, such as compensatory actions associated with inoperable sprinkler systems and fire doors

required actions necessary when a facility does not meet the administrative controls listed in Technical Specifications (TS), Section 5 or 6, depending on the facility (e.g., shift staffing requirements)

National Pollutant Discharge Elimination System requirements, if applicable

processes for TS and final safety analysis report changes Objective: Given the need to block open a Fire Door or Barrier, implement the requirements of AP 4.02 and FPIP-04.

Question 84 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 851530 User-Defined ID: 067G2.2.38 Cross Reference Number: APTS_CK21.0 SRO ONLY: The Plant is in Mode 5 for a maintenance outage.

Topic:

An UNPLANNED ENTRY into a Higher Risk Num Field 1:

Num Field 2:

Text Field:

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EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 309 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 85 ID: 850970 Points: 1.00 SRO ONLY An event has occurred requiring a plant trip from 100% power Conditions are such that EOP-9.0, Functional Recovery Procedure, has been entered

ALL PCPS are OFF

Corrected PZR Level is 10%

MSIVs are closed

E-50A, 'A' S/G is completely depressurized

ALL feedwater to 'A' S/G has been stopped Following the current check of Safety Functions 15 minutes ago, it has been determined that the PCS cooldown rate has been exceeded, and the PCS is over subcooled.

PZR Pressure is 1800 psia

PCS Temperature is 400°F What actions will the Control Room Supervisor direct the operators to perform to restore PCS to within the limits?

A. Operate PZR PORVs to reduce PCS Pressure, and stop PCS cooldown for at least 45 minutes B. Operate PZR PORVs to reduce PCS Pressure, and continue PCS cooldown at less than 100°F per hour C. Control Charging and Letdown to reduce PCS Pressure, and stop PCS cooldown for at least 45 minutes D. Control Charging and Letdown to reduce PCS Pressure, and continue cooldown at less than 100°F per hour Answer: A Answer Explanation Answer: A.

A. Operate PZR Porvs to reduce PCS Pressure, and stop PCS cooldown for at least 45 minutes is correct.

Explanation:

EOP-9.0 Functional Recovery Procedure, PC-1, Step 6 (a continuous action Step)

Directs the use of the PZR PORVs to lower PCS pressure, when the limits of EOP Supplement 1 have been exceeded, and it also directs the use of EOP Supplement 33 when the Heatup or Cooldown rates have been exceeded. The specific action of the Supplement for exceeding PCS cooldown rates is to Stop PCS cooldown until the temperature change rate is within acceptable limits. The acceptable limit for PCS Cooldown is 100°F in any 60 minute period. 15 minutes ago the limit was exceeded. That means at least 45 minutes must pass before further cooldown is allowed.

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EXAMINATION ANSWER KEY Daft N 1 Distractors:

Distractors with 'control charging and letdown' are plausible because they are listed as options in EOP-9.0 step 6, but don't apply until the SI throttling criteria are met, which can't be met because PZR Level is too low at 10%.

Distractors with 'continue cooldown' are plausible because that is the overall strategy for the procedure to get to SDC conditions, but it is incorrect because there is no direction to violate TS limits.

Reference:

EOP-9.0, step 6, and EOP Supplement 33 Step 4.

KA: A11 RCS Overcooling AA2. Ability to determine and interpret the following as they apply to the (RCS Overcooling)

(CFR: 43.5 / 45.13)

AA2.2 Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.

IMPORTANCE RO 3.0 SRO 3.4 Test item meets KA. Question asks what the specific procedure actions are for an RCS Overcooling event, with some equipment not available.

HIGH COG - requires assessment of conditions and selection of specific actions, including if SI throttle criteria are met.

SRO ONLY:

Assessment of Facility Conditions and Selection of Appropriate Procedures during Normal, Abnormal, and Emergency Situations [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both (1) assessing plant conditions (normal, abnormal, or emergency) and then (2) selecting a procedure or section of a procedure to mitigate or recover, or with which to proceed. One area of SRO-level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item. The following are examples:

knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

knowledge of diagnostic steps and decision points in the emergency operating procedures (EOPs) that involve transitions to event-specific sub-procedures or emergency contingency procedures This question is SRO ONLY because it tests knowledge of how to use EOP Supplements 1 and 33 which is essentially an attachment or appendix to the EOP in use.

Objective: Given plant conditions involving Emergency Operating Procedure, describe the mitigating strategy of the in use Emergency Operating Procedure in accordance with the Emergency Operating Procedure Bases Document.

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EXAMINATION ANSWER KEY Daft N 1 Question 85 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 850970 User-Defined ID: A11AA2.2 Cross Reference Number: TBCORE_CK01.0 SRO ONLY An event has occurred requiring a plant trip from Topic:

100% power Conditions are such that EO Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 312 of 369 15 January 2020

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EXAMINATION ANSWER KEY Daft N 1 86 ID: 851561 Points: 1.00 SRO ONLY:

The plant is at 100% power with P-52C, Component Cooling Pump, in service.

At time 1000:00, the following annunciator alarms:

EK-1167, Component Clg Pumps P-52A, P-52B, P-52C Trip

Operators immediately determine P-52C tripped

ONE minute later P-52A auto-started At time 1005:00, the following annunciator alarms:

EK-1172, Component Clg Surge Tank T-3 Hi-Lo Level At time 1008:00 The CCW Surge Tank level issue results in isolating the CCW side of E-54B, CCW Heat Exchanger.

Given these conditions, complete the following statements:

Just prior to 1001:00, there ___(1)___ 2 trains of CCW OPERABLE.

At time 1008:00, LCO 3.0.3 ___(2)___ required to be entered.

(1) (2)

A. are is B. are is NOT C. are NOT is D. are NOT is NOT Answer: A Answer Explanation Answer: A.

A, are, is; is correct.

Explanation:

The tech spec bases for LCO 3.7.7 CCW, specifies P-52A and P-52B as the pumps that must be operable, to make their respective Trains operable. There is NO mention of P-52C. The fact that P-52A was not operating does NOT make it inoperable. The EK-1167 annunciator is a multiple input alarm and will annunciate if any of the CCW pumps trip. The trip of P-52C does NOT impact the operability of the two required CCW trains. Thus, part one is correct.

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EXAMINATION ANSWER KEY Daft N 1 The tech spec bases also states that isolating the CCW side of either CCW HX prevents the CCW system from meeting the 100% of the required CCW post accident cooling capability, making Condition C.1 applicable and requiring immediate entry into LCO 3.0.3. Thus, part two is correct.

Distractors All plausible for the examinee that doesn't comprehend the LCO or the bases.

The distractors with 'are NOT' in part one are plausible for the examinees that do NOT know the TS bases for operability requirements of CCW. These distractors are incorrect, because P-52C is not required for either train's operability.

The distractors with 'is NOT' in part two are plausible for the examinees that are not familiar with the actions of TS 3.7.7 for an inoperable CCW HX. These distractors are incorrect as explained in the answer.

KA: 008 CCW A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.02 High/low surge tank level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2 3.5 Test item meets KA. Question poses a low Surge tank level that results in isolating required components, per the procedure. The impact on continued operations is provided by tech specs.

HIGH COG - apply requirements of tech specs, multiple mental steps.

SRO ONLY:

Facility Operating Limitations in the Technical Specifications and Their Bases [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic the following:

application of required actions (TS Section 3) and surveillance requirements (SR) (TS Section 4) in accordance with rules of application requirements (TS, Section 1)

application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).

knowledge of TS bases that are required to analyze TS-required actions and terminology Objective: Given plant conditions and Technical Specifications 3.7.7 determine the following for the Component Cooling Water system in accordance with Technical Specification 3.7.7 BASES for the Component Cooling Water system, LCO Section 1.0, and LCO Section 3.0. (G2.2.40) (SRO ONLY)

a. Status of associated LCO Condition(s) and applicable Required Actions and Completion Times

b. Time period during which any associated Surveillance Requirements must be performed PLP - 2018 NRC EXAM Page: 315 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 86 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 851561 User-Defined ID: 008A2.02 Cross Reference Number: CCW_CK21.0 SRO ONLY: The plant is at 100% power with P-52C, Topic:

Component Cooling Pump, in service. At time 100 Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 316 of 369 15 January 2020

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EXAMINATION ANSWER KEY Daft N 1 87 ID: 851111 Points: 1.00 SRO ONLY:

The plant is in MODE 3 after a trip from 100% power.

The PCS is at normal operating temperature and pressure.

2400 Volt AC Bus 1D develops a fault and is de-energized Given these conditions, complete the following statements:

LCO 3.6.6, Containment Cooling Systems ACTION C for Less than 100% of the required post accident Containment Cooling Capability ___(1)___ required to be entered.

The required support function of Containment Spray for the HPSI pumps ___(2)___ met.

(1) (2)

A. is NOT is NOT B. is NOT is C. is is NOT D. is is Answer: B Answer Explanation Answer: B B is NOT, is; is correct.

Explanation:

The question has little to nothing to do with TS 3.5.2, or 3.5.3. The question is about containment cooling systems (3.6.6), and plant systems (3.7.7 CCW, and 3.7.8 SWS), NOT emergency cooling systems (3.5).

While the loss of Bus 1D would make one LPSI train inoperable, it does not reduce the available ECCS flow to less than 100% of the required ECCS flow, and 3.0.3 is not entered from 3.5.2.

Each of the following Tech Specs has a condition for having less than 100% of the required post accident (containment cooling 3.6.6, CCW 3.7.7, SWS 3.7.8) flow with an action to enter 3.0.3 Immediately. This question is asking about the version in 3.6.6.

The Action in the LCO for the condition of less than 100% of the post accident containment cooling capability available is to enter 3.0.3 immediately. Per the TS Bases, even though the loss of Bus 1D causes the loss of 2 SWS pumps, 1 CS pump and 3 CACs, the 2400 volt Bus 1C carries enough Containment Cooling and SW equipment to meet the requirements for 100% of the cooling capabilities in a post accident situation. Thus, the condition requiring entry into 3.0.3 does not yet exist. (but it would if an MSIV Bypass valve or two were open.)

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EXAMINATION ANSWER KEY Daft N 1 The CS pumps provide a required support function for the HPSI pumps as described in Bases 3.5.2.

That bases only applies to OPERABLE CS pumps. With Bus 1D faulted, P-54A is NOT operable. The remaining two CS pumps can fulfill their support function.

Distractors The distractors are plausible for the examinee that doesn't comprehend the TS bases for LCOs 3.6.6 CCS, and 3.7.6 SWS. (and 3.5.2)

The distractors with 'is' in part one are plausible for the examinee that doesn't understand the requirements for operable equipment are met because of which equipment is on which bus. They are incorrect as explained in the answer.

The distractors with 'is NOT' in part two are plausible for the examinee the doesn't understand the requirements are met for the support function CS provides for HPSI. They are incorrect as explained in the answer.

KA: 022 Containment Cooling 2.2.22 Knowledge of limiting conditions for operations and safety limits.

(CFR: 41.5 / 43.2 / 45.2)

IMPORTANCE RO 4.0 SRO 4.7 Test item meets KA. Question poses a loss of a bus that impacts the status of available CCS and components. Answer requires LCO and TS Bases knowledge to determine if a specific requirement (action) applies. (similar to a procedure step that requires action to correct control or mitigate the consequence of the loss of CCS equipment.)

HIGH COG - assess loss, apply knowledge of bases.

SRO ONLY:

Requires knowledge of TS Bases.

Facility Operating Limitations in the Technical Specifications and Their Bases [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic the following:

application of required actions (TS Section 3) and surveillance requirements (SR) (TS Section 4) in accordance with rules of application requirements (TS, Section 1)

application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).

knowledge of TS bases that are required to analyze TS-required actions and terminology Objective: From memory, describe the Technical Specification bases for the Containment System in accordance with Technical Specification 3.3.3, 3.3.4, 3.3.6, 3.3.7, 3.6.1, 3.6.2, 3.6.3, 3.6.4, 3.6.5 and 3.9.3 Bases. (G2.2.25) (SRO ONLY)

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EXAMINATION ANSWER KEY Daft N 1 Question 87 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 851111 User-Defined ID: 022G2.2.22 Cross Reference Number: CTMT_CK22.0 SRO ONLY: The plant is in MODE 3 after a trip from 100%

Topic:

power. The PCS is at normal operating tem Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 320 of 369 15 January 2020

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EXAMINATION ANSWER KEY Daft N 1 88 ID: 850971 Points: 1.00 SRO ONLY:

A LOCA has occurred that has resulted in the actuation of the Containment Spray System.

The appropriate EOPs are being performed.

Given these conditions complete the following statements:

If the pH of the containment sump was NOT buffered by Sodium Tetraborate after RAS, then the radiological contribution of iodine to a containment atmosphere leak will be ___(1)___.

Allowing the containment sump pH to exceed 8.0 ___(2)___ the conclusions of the Maximum Hypothetical Accident (MHA) for hydrogen generation analysis.

(1) (2)

A. higher could invalidate B. higher would have NO effect on C. unaffected could invalidate D. unaffected would have NO effect on Answer: A Answer Explanation Answer: A.

A. (1) higher, (2) could invalidate Explanation:

The bases for the Tech Spec for the Containment Sump Buffering Agent and Weight Requirements states the potential for more airborne iodine in the containment atmosphere is greater if the sump pH is allowed to become acidic. The same bases states the MHA analysis could be invalidated with regard to hydrogen generation if pH is allowed to exceed 8.0.

Distractors:

Each distractor is plausible for the examinee that doesn't know the TS bases for TS 3.5.5.

The distractors with 'unaffected' in part one are plausible for the examinee that doesn't understand the chemistry and reactions between the leaking PCS, the sodium tetraborate, and the volatility of the resultant iodine. They are incorrect as explained in the answer.

The distractors with 'would have no effect on' are plausible for the examinee that is not familiar with the conclusions of the MHA for H2 generation analysis. They are incorrect as explained in the answer.

KA: 026 Containment Spray PLP - 2018 NRC EXAM Page: 322 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.05 Failure of chemical addition tanks to inject . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7 4.1 Test item meets part (a) of the KA. Question requires TS Bases knowledge which describes the impact of the failure to use the sodium tetraborate after RAS with CS.

Due to the design of the chemical addition portion of the CS system at Palisades, there are no credible failures that would result in the sodium tetraborate NOT being mixed into the RAS stream. (The chemical resides in open baskets in the Containment sump. There are no valves and no piping in the flowpath from these baskets to the sump.)

MEMORY level.

SRO ONLY:

The knowledge being tested is found in the Tech Spec Bases. Facility Operating Limitations in the Technical Specifications and Their Bases [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic the following:

application of required actions (TS Section 3) and surveillance requirements (SR) (TS Section 4) in accordance with rules of application requirements (TS, Section 1)

application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).

knowledge of TS bases that are required to analyze TS-required actions and terminology Objective:

Given plant conditions and Technical Specification 3.5.5 and 3.6.6 determine the following for the Containment Spray system in accordance with Technical Specification 3.5.5 and 3.6.6 BASES for the Containment Spray system, LCO Section 1.0, and LCO Section 3.0. (G2.2.40) (SRO ONLY)

a. Status of associated LCO Condition(s) and applicable Required Actions and Completion Times

b. Time period during which any associated Surveillance Requirements must be performed PLP - 2018 NRC EXAM Page: 323 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 88 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 850971 User-Defined ID: 026A2.05 Cross Reference Number: CSS_CK21.0 SRO ONLY: A LOCA has occurred that has resulted in the Topic:

actuation of the Containment Spray System.

Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 324 of 369 15 January 2020

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EXAMINATION ANSWER KEY Daft N 1 89 ID: 851480 Points: 1.00 SRO ONLY:

The plant is in MODE 1, but considering shutting down due to instrumentation issues listed below:

Instrument Issue LI-0702, 'A' SG Narrow Range Level power supply LI-0757B, 'A' SG Wide Range Level power supply FI-0727A Aux Feedwater Flow to 'B' SG failed high LI-0752D, 'B' SG Safety Channel failed low Given these conditions, complete the following statements:

The number of Steam Generators that have at least one INOPERABLE Post Accident Monitoring (PAM) Instrument is ___(1)___.

The basis behind the required OPERABILITY of the Steam Generator PAMs is that they are used to identify the affected Steam Generator during a(n) ___(2)___.

(1) (2)

A. one ESDE B. one SGTR C. two SGTR D. two ESDE Answer: B Answer Explanation Answer: B B. one, SGTR; is correct.

Explanation:

The Tech Spec Bases for LCO 3.3.7 and FSAR Appendix 7C identify the PAMS that cover SG Level as LI-0757A and B, for the 'A' SG, and LI-0758A and B for the 'B' SG. These two documents also list the bases for their operability as being required as they are used to identify the SG with a tube rupture that is required to be isolated.

Distractors:

All plausible as they are each a real SG Level or flow Instrument, but they each have different purposes and are all affected by the operation of AFW.

LI-0702 is a narrow range level indicator on panel C-01 that is part of the high level override circuit and not tied to any tech spec.

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EXAMINATION ANSWER KEY Daft N 1 FI-0727A is a flow indication in the control room on C-11 and is part of the circuit that determines if there is flow from AFW pumps P-8A and /or P-8B for proper sequencing of AFW pump starts.

LI-0752D are level instruments that are part of the RPS and AFAS instrumentation and tied to tech spec LCO 3.3.1, 3.3.3, and 3.3.8.

The distractors with 'TWO' in part one are plausible for the examinee that believes either the AFW flow indicator or the Safety Channel are PAMs. (They are NOT).

ESDE is plausible in part two because an ESDE affects the SG levels as the Affected SG blows down.

These distractors are incorrect because Steam Flow and Steam Pressure instruments are used to identify the affected SG in an ESDE.

KA:

061 AFW 2.4.3 Ability to identify post-accident instrumentation.

(CFR: 41.6 / 45.4)

IMPORTANCE RO 3.7 SRO 3.9 Test item meets KA. Answer requires the ability to identify from a list, the PAMs associated with a Steam Generator most likely after AFW initiation. Per the TS bases, the only PAM instrument that mentions AFW is the Condensate Storage Tank Level Indication. (A suitable test item could not be developed around just a single PAM for AFW). AFW feeds the SG and level is monitored via these indicators. They are the only PAMS for this system.

MEMORY SRO ONLY:

SRO ONLY, because the information found to answer the bases part of the question is in the TS bases.

Facility Operating Limitations in the Technical Specifications and Their Bases [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic the following:

application of required actions (TS Section 3) and surveillance requirements (SR) (TS Section 4) in accordance with rules of application requirements (TS, Section 1)

application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).

knowledge of TS bases that are required to analyze TS-required actions and terminology Objective: Given plant conditions and Technical Specifications 3.3.3, 3.3.4, 3.3.7, 3.3.8, and 3.7.5 determine the following for the Auxiliary Feedwater System in accordance with Technical Specification 3.3.3, 3.3.4, 3.3.7, 3.3.8, and 3.7.5 BASES for the Auxiliary Feedwater System, LCO Section 1.0, and LCO Section 3.0. (G2.2.40) (SRO ONLY)

a. Status of associated LCO Condition(s) and applicable Required Actions and Completion Times

b. Time period during which any associated Surveillance Requirements must be performed PLP - 2018 NRC EXAM Page: 327 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 89 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 851480 User-Defined ID: 061G2.4.3 Cross Reference Number: AFW_CK21.0 SRO ONLY: The plant is in MODE 1, but considering shutting Topic:

down due to instrumentation issues lis Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 328 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 329 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 90 ID: 851810 Points: 1.00 SRO ONLY:

The plant is in MODE 3 after a trip from 100% power.

The PCS is at normal operating temperature and pressure.

2400 Volt AC Bus 1C develops a fault and is de-energized Given these conditions complete the following statements:

The Tech Spec ACTION for the Condition where Less than 100% of the required post accident SWS Cooling Capability ___(1)___ required to be entered.

Emergency Diesel Generator (DG) 1-1 ___(2)___ INOPERABLE.

(1) (2)

A. is is B. is is NOT C. is NOT is D. is NOT is NOT Answer: C Answer Explanation Answer: C C. is NOT, is: is correct.

Explanation:

The Action in the SWS LCO for the condition of less than 100% of the post accident cooling capability available is to enter 3.0.3 immediately. Per the TS Bases, even though the loss of Bus 1C causes the loss of 1 SWS pump, 2 CS pumps, the 2400 Volt AC Bus 1D carries enough SWS equipment to meet the requirements for 100% of the cooling capabilities in a post accident situation.

Thus, the conditions for immediate entry into 3.0.3 are not met yet.

With Bus 1C faulted, EDG 1-1 can NOT automatically (or manually) re-energize it during a loss of offsite power because the EDG output breaker won't close onto a bus that is faulted. Thus even though the EDG would start and come up to speed and voltage, it can NOT perform its safety function, and should be declared INOPERABLE.

From Tech Spec Bases for AC Sources Operating: Following a loss of offsite power, each DG must be capable of starting and connecting to its respective 2400 volt bus. Proper sequencing of loads...are required functions for DG OPERABILITY.

Distractors:

The distractors are plausible for the examinee that do not comprehend the TS bases, or the OPERABILITY requirements for the EDGs.

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EXAMINATION ANSWER KEY Daft N 1 The distractors with 'is' in part one are plausible for the examinee that doesn't understand the TS bases for the SWS and 100% post accident cooling capability. They are incorrect as explained in the answer.

The distractors with 'is NOT' in part two are plausible for the examinee that doesn't understand how operability of the EDG is maintained. They are incorrect as explained in the answer.

KA: 076 SWS A2 Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 / 43.5 / 45/3 / 45/13)

A2.01 Loss of SWS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5* 3.7*

Test item meets KA. Question tests on predicting impact on components and systems with TS Bases knowledge for a loss of SWS. Procedure Actions to mitigate are the correct implementation of the TS requirements.

HIGH COG - assess loss and apply knowledge of bases.

SRO ONLY:

Facility Operating Limitations in the Technical Specifications and Their Bases [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic the following:

application of required actions (TS Section 3) and surveillance requirements (SR) (TS Section 4) in accordance with rules of application requirements (TS, Section 1)

application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).

knowledge of TS bases that are required to analyze TS-required actions and terminology Objective:

From memory, describe the Technical Specification bases for the Service Water System and Ultimate Heat Sink in accordance with Technical Specification 3.7.8 and 3.7.9 Bases. (G2.2.25) (SRO ONLY)

Given plant conditions and Technical Specifications 3.7.8 and 3.7.9 determine the following for the Service Water System and Ultimate Heat Sink in accordance with Technical Specification 3.7.8 and 3.7.9 BASES for the Service Water System and Ultimate Heat Sink , LCO Section 1.0, and LCO Section 3.0. (G2.2.40) (SRO ONLY)

a. Status of associated LCO Condition(s) and applicable Required Actions and Completion Times

b. Time period during which any associated Surveillance Requirements must be performed PLP - 2018 NRC EXAM Page: 331 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 90 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 851810 User-Defined ID: 076A2.01-1 Cross Reference Number: SWS_CK22.0 SRO ONLY: The plant is in MODE 3 after a trip from 100%

Topic:

power. The PCS is at normal operating tem Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 332 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 333 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 91 ID: 851072 Points: 1.00 SRO ONLY:

With the plant at 100% power, the Rod Mode Select Switch is taken to 'EM OFF".

This action ___(1)___ result in EK-0917, Rod Withdrawal Prohibit, alarming.

While the switch is in the 'EM OFF' position, Tech Spec 3.1.4, Control Rod Alignment, Required Action E.1, to be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ___(2)___ entered.

(1) (2)

A. does is B. does is NOT C. does NOT is D. does NOT is NOT Answer: A Answer Explanation Answer: A A. does, is; is correct.

Explanation:

Placing the mode select switch to the EM OFF position, de-energizes all the CRDM motors and brakes, and the rods can no longer be moved. The Tech Spec bases for LCO 3.1.4 requires the rods to be able to be moved by their CRDMs for operability. It also activates the Rod Withdrawal Prohibit alarm.

Distractors:

ALL plausible for the examinee that doesn't know the inputs to the alarms, or the impact of being in EM OFF with regard to tech spec requirements for rod alignment and operability.

Distractors with 'does NOT' are plausible for the examinee that does not know the inputs to the rod motion inhibit circuit.

Distractors with 'is NOT' are plausible for the examinee that is not aware of the explanation in the tech spec bases.

KA: 001 Control Rod Drive A2 Ability to (a) predict the impacts of the following malfunction or operations on the CRDS- and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5/43.5/45.3/45.13)

A2.14 Urgent failure alarm, including rod-out-of-sequence and motion-inhibit alarms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7 3.9 PLP - 2018 NRC EXAM Page: 334 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Test item meets KA. While Palisades does not have a Rod Control Urgent Failure alarm, it does have a rod motion inhibit alarm. The question tests knowledge of the inputs to that alarm, which meets the 'a' part of the KA - predict the impact of manipulating a rod control mode select switch, and then depends on knowledge of tech specs and bases (like procedures) to correct, control or mitigate the consequence of having all rods INOPERABLE.

SRO ONLY:

Knowledge of what is required for operability is discussed in Tech Spec Bases for LCO 3.1.4. Operability calls, are SRO ONLY job function.

Facility Operating Limitations in the Technical Specifications and Their Bases [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic the following:

application of required actions (TS Section 3) and surveillance requirements (SR) (TS Section 4) in accordance with rules of application requirements (TS, Section 1)

application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).

knowledge of TS bases that are required to analyze TS-required actions and terminology MEMORY Objective: From memory, describe the Technical Specification bases for the Control Rod Drive System in accordance with Technical Specification 3.1.4, 3.1.5, and 3.1.6 Bases. (G2.2.25) (SRO ONLY)

Question 91 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 3.00 System ID: 851072 User-Defined ID: 001A2.14 Cross Reference Number: CRD_CK22.0 SRO ONLY: With the plant at 100% power, the Rod Mode Topic:

Select Switch is taken to 'EM OFF". This ac Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 335 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 336 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 92 ID: 851070 Points: 1.00 SRO ONLY:

Irradiated Sources are being transferred in accordance with SOP-28, Fuel Handling System.

After lowering the source lifting tool directly over the source to be removed, as the SRO you will direct ___(1)___ SFP Overhead Crane prior to closing the jaws of the lifting tool.

The reason for this action is to prevent ___(2)___.

A. (1) de-energizing (2) re-opening of the source lifting tool jaws B. (1) energizing (2) unintended lifting of neutron sources C. (1) energizing (2) re-opening of the source lifting tool jaws D. (1) de-energizing (2) unintended lifting of neutron sources Answer: D Answer Explanation Answer: D.

D. (1) de-energizing, (2) unintended lifting of neutron sources; is correct.

Explanation:

The reference procedure has a Warning in the steps being performed during the transfer of neutron sources evolution. The Warning explains how the unintended lifting can occur, and how it can be prevented.

The following Warning from SOP-28, Fuel Handling System is present in the procedure after energizing the SFP crane:

WARNING Unintended lifting of neutron sources may occur due to spurious operation of powered lifting devices (eg, SFP overhead crane) when source lifting tool is attached to sources. This can be prevented by de-energizing any powered lifting devices prior to closing the jaws of the source lifting tool.

The entire operation is under the supervision of an SRO, which requires license level knowledge.

Knowledge of how the equipment works when it is energized, or de-energized is part of the SROs job. It is entirely possible to confuse these steps, as it is an infrequently performed evolution.

Just five steps before the warning, there is a step to ENSURE the New Fuel Elevator is energized, followed by a step that PLACES the fuel bundle that is to have its source removed in the New Fuel Elevator (while it is energized).

An excerpt from the procedure:

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EXAMINATION ANSWER KEY Daft N 1

b. ENSURE New Fuel Elevator is energized. Refer to Section 7.7.2, "To Energize New Fuel Elevator."

c. PLACE fuel bundle to have its source removed in New Fuel Elevator east box. Refer to Section 7.2, "Spent Fuel Handling Machine (SFHM)."

d. PLACE fuel bundle to receive source in New Fuel Elevator west box. Refer to Section 7.2, "Spent Fuel Handling Machine (SFHM)."

1. MOVE SFHM clear of New Fuel Elevator.

e. RAISE New Fuel Elevator to comfortable working height.

f. IF desired, THEN SUSPEND source handling tool only from the SFP overhead crane (or other suitable support).

g. OPEN jaws of source handling tool.

1. LOWER tool directly over source to be removed.

Distractors:

Plausible for examinees not familiar with the procedure requirements.

Energizing is plausible if it is thought that is necessary to complete the alignment of the crane at this time to perform the lift, or to close the jaws. Or it is confused with the New Fuel Elevator which is also energized and has two boxes associated with it.

Re-opening of the source lifting tool jaws is plausible if it is thought an interlock between the jaws and the crane exists.

KA: 034 Fuel Handling Equipment K6 Knowledge of the effect of a loss or malfunction on the following will have on the Fuel Handling System:

(CFR: 41.7 / 45.7)

K6.01 Fuel handling equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1 3.0 Test item meets KA. Answer requires knowledge of what happens to the fuel handling system when FH equipment malfunctions, or procedures aren't understood or followed precisely.

MEMORY level.

Reference:

SOP-28, Fuel Handling System SRO ONLY:

Question is SRO ONLY because it involves a procedure for replacing neutron sources, an evolution controlled by an SRO.

Procedures and Limitations Involved in Initial Core Loading, Alterations in Core Configuration, Control Rod Programming, and Determination of Various Internal and External Effects on Core Reactivity [10 CFR 55.43(b)(6)]. Some examples of SRO exam items for this topic include the following:

evaluation of core conditions and emergency classifications based on core conditions PLP - 2018 NRC EXAM Page: 338 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1

administrative requirements associated with low-power physics testing processes

administrative requirements associated with refueling activities, such as approvals required to amend core loading sheets or administrative controls of potential dilution paths and/or activities

administrative controls associated with the installation of neutron sources

knowledge of TS bases for reactivity controls Objective: Given fuel handling procedures and fuel movement conditions, identify the operating limitations designed to prevent damage to equipment or the fuel assemblies in accordance with SOP-28 and GOP-11 and/or Refueling Procedures.

Question 92 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 851070 User-Defined ID: 034K6.01 Cross Reference Number: IOTD_E01.02 SRO ONLY: Irradiated Sources are being transferred in Topic:

accordance with SOP-28, Fuel Handling Syste Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 339 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 340 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 93 ID: 851052 Points: 1.00 SRO ONLY:

At time 1000:00, With the plant at 100% power, it is noted that LCO 3.7.6, Condensate Storage and Supply, is NOT met because of volume.

Below is an excerpt from the ACTIONS portion of this Technical Specification:

Given these conditions, complete the following statements:

The LATEST the OPERABILITY of the backup water supplies must be verified is ___(1)___;

For subsequent performances of the Operability Verification of Backup Water Supplies, a 25%

time extension ___(2)___ allowed.

(1) (2)

A. 1400 is NOT B. 1400 is C. 1500 is NOT D. 1500 is Answer: B Answer Explanation Answer: B B. 1400, is; is correct Explanation:

Per the Tech Spec Bases, and SR 3.0.2, the first performance of the ACTION must be performed within the stated time, with NO EXTENSION allowed.

Subsequent performances can employ the 25% extension.

Distractors:

Plausible for those not familiar with the Bases of TS 3.7.6 (or SR 3.0.2) where this is explicitly stated.

1500 is plausible if the 25% extension is erroneously applied to the first performance of the required verification action.

'is NOT' is plausible if the allowed 25% extension is NOT applied when the question implies LATEST.

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EXAMINATION ANSWER KEY Daft N 1

Reference:

TS Bases 3.7.6.

KA: 056 Condensate 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12 / 45.13)

IMPORTANCE RO 4.4 SRO 4.7 Test item meets KA. Question requires knowledge of TS Bases to make an operational decision (when is an action required) based on instrument interpretation (volume of the Condensate Storage Supplies).

HIGH COG - comprehension and application of a TS rule of usage.

Objective: Given plant conditions and Technical Specifications 3.7.3 and 3.7.6 determine the following for the Main Condenser, Condensate and Feedwater system in accordance with Technical Specification 3.7.3 and 3.7.6 BASES for the Main Condenser, Condensate and Feedwater system , LCO Section 1.0, and LCO Section 3.0. (G2.2.40) (SRO ONLY)

a. Status of associated LCO Condition(s) and applicable Required Actions and Completion Times

b. Time period during which any associated Surveillance Requirements must be performed SRO ONLY:

Facility Operating Limitations in the Technical Specifications and Their Bases [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic the following:

application of required actions (TS Section 3) and surveillance requirements (SR) (TS Section 4) in accordance with rules of application requirements (TS, Section 1)

application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).

knowledge of TS bases that are required to analyze TS-required actions and terminology.

Question 93 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 851052 User-Defined ID: 056G2.1.7 Cross Reference Number: CDFW_CK21.0 SRO ONLY: At time 1000:00, With the plant at 100% power, it Topic:

is noted that LCO 3.7.6, Condensate S Num Field 1:

Num Field 2:

Text Field:

Comments: new PLP - 2018 NRC EXAM Page: 342 of 369 15 January 2020

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EXAMINATION ANSWER KEY Daft N 1 94 ID: 847872 Points: 1.00 SRO ONLY Reference Provided The MAXIMUM allowable Linear Heat Rate (LHR) at the top of the peak powered fuel rod is ______

kW/ft.

A. 6.11 B. 9.16 C. 14.21 D. 15.28 Answer: C Answer Explanation Answer: C C. 14.21 kW/ft is CORRECT PROVIDE to the EXAMINEE:

COLR page 4 of 10 - entire page

2.3 Linear Heat Rate (LHR)

Explanation /

Reference:

As LHR is evaluated at higher core elevations, the limit on LHR in the peak powered rod lowers as shown of figure 2.3-1, such that at the top of the fuel, the limit is 93% of 15.28 kW/ft.

Each distractor is a calculated value that is reached if the use of the figure is NOT understood.

6.11 is 40% up the height (reciprocal of 60% - a break point on the graph) 9.16 is 60% up the height 15.28 is 100% of the limit with no penalty for height.

KA:

2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.

(CFR: 41.10 / 43.5 / 45.12)

IMPORTANCE RO 3.9 SRO 4.2 Objective: Describe factors that affect peaking and hot channel factors.

Test item meets KA. Answer requires interpreting the Table from the COLR to determine appropriate limits.

HIGH COG - interpret requirements of graph and instructions in the COLR.

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EXAMINATION ANSWER KEY Daft N 1 SRO ONLY:

SRO Job Function Facility Operating Limitations in the Technical Specifications and Their Bases

[10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic the following:

application of required actions (TS Section 3) and surveillance requirements (SR)

(TS Section 4) in accordance with rules of application requirements (TS, Section 1)

application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).

knowledge of TS bases that are required to analyze TS-required actions and terminology

same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM)

Question 94 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847872 User-Defined ID: G2.1.25 SRO Cross Reference Number: GFAI_E01.3 SRO ONLY Reference Provided The MAXIMUM allowable Topic:

Linear Heat Rate (LHR) at the top of the peak Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 346 of 369 15 January 2020

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EXAMINATION ANSWER KEY Daft N 1 95 ID: 847882 Points: 1.00 SRO ONLY It is day 15 of a refueling outage

CORE ALTERATIONS are about to begin

Equipment Hatch is open The Fuel Handling Ventilation system is aligned as follows:

V-7, Fuel Handling Area Supply Fan is OFF

V-8A, Fuel Handling Exhaust Fan is ON

V-8B, Fuel Handling Exhaust Fan is ON

V-69, Fuel Handling Area Supply Fan is OFF

V-70A, Fuel Handling Area Exhaust Fan is OFF

V-70B, Fuel Handling Area Exhaust Fan is ON In order to meet Technical Specification requirements before CORE ALTERATIONS can begin,

___(1)___ fans have to be started or stopped.

The basis for the required fan alignment is to ___(2)___.

(1) (2)

A. 2 allow Fuel Handling Building atmosphere flow to bypass the HEPA filter B. 2 ensure all fuel handling areas exhaust to the HEPA filter C. 3 ensure all fuel handling areas exhaust to the HEPA filter D. 3 allow Fuel Handling Building atmosphere to bypass the HEPA filter Answer: B Answer Explanation Answer: B.

B. 2, ensure all fuel handling areas exhaust to the HEPA filter: is CORRECT Explanation /Reference GOP-11 rev 51, step 5.7.1.

Tech Spec 3.7.12 Bases States NO more than one V-8A/B Fuel Handling Exhaust Fan operating. The others must all be OFF.

Stopping either 8A or 8B, AND 70B is 2 fans started or stopped.

Bases: During plant evolutions when the possibility for a FH accident exists, the FH Area Ventilation system is configured such that all fans are stopped except one exhaust fan in the original plant subsystem aligned to the emergency filter bank.

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EXAMINATION ANSWER KEY Daft N 1 Distractors:

A. 2, allow FHB atmosphere to bypass the HEPA filter is plausible. Part one is correct. Part two is plausible for the examinee that doesn't understand the TS bases requirement and believes the FHB is not affected by the containment atmosphere.

C. 3, ensure all fuel handling areas exhaust to the HEPA filter: is plausible. Part one is plausible to think starting V-70A would balance the Area Exhaust flow, Starting V-7 because there is no supply fan running, and stopping V-8A or 8B to keep FH Exhaust are pressure from becoming too low totals 3 fans started or stopped. Part two is correct as explained in B.

D. 3, allow FHB atmosphere to bypass the HEPA filter is plausible. Part one is plausible as explained in C. Part two is plausible as explained in A.

KA:

2.1.36 Knowledge of procedures and limitations involved in core alterations.

(CFR: 41.10 / 43.6 / 45.7)

IMPORTANCE RO 3.0 SRO 4.1 Test item meets KA. Question asks about limitations in ventilation alignments specifically for CORE ALTERATIONS.

SRO ONLY:

ensuring the correct system alignments for CORE ALTERATIONS is an SRO job function, as explained in the Tech Spec Bases for TS 3.7.12.

F. Procedures and Limitations Involved in Initial Core Loading, Alterations in Core Configuration, Control Rod Programming, and Determination of Various Internal and External Effects on Core Reactivity [10 CFR 55.43(b)(6)]. Some examples of SRO exam items for this topic include the following:

evaluation of core conditions and emergency classifications based on core conditions

administrative requirements associated with low-power physics testing processes

administrative requirements associated with refueling activities, such as approvals required to amend core loading sheets or administrative controls of potential dilution paths and/or activities

administrative controls associated with the installation of neutron sources

knowledge of TS bases for reactivity controls G. Fuel-Handling Facilities and Procedures [10 CFR 55.43(b)(7)]

Some examples of SRO exam items for this topic include the following:

refuel floor SRO responsibilities

assessment of fuel-handling equipment SR acceptance criteria

prerequisites for vessel disassembly and reassembly

decay heat assessment

assessment of SRs for the refueling mode

reporting requirements

emergency classifications Objective:

Given fuel handling procedures and fuel movement conditions in the Spent Fuel Pool or Reactor Core, determine ventilation equipment alignment requirements for fuel movement in accordance with GOP-11.

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EXAMINATION ANSWER KEY Daft N 1 HIGH COG - application of procedure requirements for allowing the commencement of CORE ALTERATIONS Question 95 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847882 User-Defined ID: 2.1.36 SRO Cross Reference Number: IOTD_E02.03 SRO ONLY It is day 15 of a refueling outage CORE Topic:

ALTERATIONS are about to begin Equipment Hatch i Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 350 of 369 15 January 2020

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EXAMINATION ANSWER KEY Daft N 1 96 ID: 847891 Points: 1.00 SRO ONLY The Plant is in MODE 3

The PCS Safety Limit has been violated

PCS Pressure peaked at 3000 psia The ACTIONS for this Safety Limit violation...

A. require a MODE reduction to MODE 4 in order to perform the necessary engineering evaluation of PCS Integrity.

B. require a MODE reduction to MODE 6, and when the first Reactor Vessel head bolt is less than fully tensioned, this safety limit no longer applies.

C. do NOT require a MODE reduction because the PCS was hydro tested to 125% of design pressure (3125 psia) and that pressure was NOT reached.

D. do NOT require a MODE reduction because reducing MODES would add further thermal gradient stress due to the cooldown to the existing pressure stress.

Answer: D Answer Explanation Answer: D D. do NOT require a MODE reduction in order to limit adding further thermal gradient stresses due to the cooldown to the existing pressure stress.

Explanation /

Reference:

Tech Spec Bases for Safety Limits 2.2.2.2:

In modes 3-6 the vessel temperature may be lower than modes 1 or 2, and the vessel material less ductile. Further reduction in Mode, means further reduction in temperature and adding more thermal gradient stress.

Distractors:

A. require a mode reduction to perform engineering evaluation is plausible, because that is a reason stated in the bases for violating the pressure temperature heatup cooldown curves. But it is incorrect as explained in the answer because a mode reduction is not required for exceeding the safety limit for pressure.

B. require a mode reduction to mode 6... is plausible for the examinee that is not familiar with the Safety Limits Bases that states this SL is applicable until all the fuel is out of the vessel. That is beyond entry into mode 6 whose definition includes the 'less than fully tensioned head bolt' clause.

C. do not require a mode reduction because the PCS was hydro ... is plausible for the examinee that believes that because the PCS did not exceed the initial one time hydrostatic test of the PCS to 3125 psia as described in the tech spec bases for safety limits, that there was not a significant challenge to PCS integrity.

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EXAMINATION ANSWER KEY Daft N 1 KA:

2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2)

IMPORTANCE RO 3.2 SRO 4.2 Test item meets KA. Answer requires TS bases knowledge for reasons for actions on Safety Limit Violation.

MEMORY SRO ONLY: Knowledge found in TS bases for why a mode reduction is not desirable or required for TS SL violation on PCS Pressure, and is required for a violation of Reactor Core SL.

Facility Operating Limitations in the Technical Specifications and Their Bases [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic the following:

application of required actions (TS Section 3) and surveillance requirements (SR) (TS Section 4) in accordance with rules of application requirements (TS, Section 1)

application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).

knowledge of TS bases that are required to analyze TS-required actions and terminology

same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM)

Objective:

Given Technical Specifications, the Operating Requirements Manual (ORM), plant conditions, and a proposed plant Mode change, determine whether or not the Mode change can be made in accordance with Technical Specifications and the ORM. (SRO Only) (G2.2.40)

Question 96 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 847891 User-Defined ID: 2.2.25 Cross Reference Number: APTS_E01.16 SRO ONLY The Plant is in MODE 3 The PCS Safety Limit has Topic:

been violated PCS Pressure peaked at 300 Num Field 1:

Num Field 2:

Text Field:

Comments: New PLP - 2018 NRC EXAM Page: 353 of 369 15 January 2020

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EXAMINATION ANSWER KEY Daft N 1 97 ID: 847926 Points: 1.00 SRO ONLY For 100% power conditions, which of the following conditions affecting the ability to makeup to the Primary Coolant System requires notification to the Nuclear Regulatory Commission?

A. P-67A, LPSI pump, is INOPERABLE and will be restored in 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months .

B. P-55B, Charging Pump, is INOPERABLE and will be restored in 4 days.

C. T-82A, Safety Injection Tank, pressure is 180 psig and lowering. It will be restored to normal in 4 days.

D. P-56A, Boric Acid Pump, spuriously started and was manually stopped. Repairs will require 68 hours7.87037e-4 days 0.0189 hours 1.124339e-4 weeks 2.5874e-5 months .

Answer: C Answer Explanation Answer: C.

C. Safety Injection Tank T-82 pressure is 180 psig and lowering. It will be restored to normal in 4 days.

Explanation /

Reference:

TS 3.5.1, Action A, C TS 3.5.2, Action A

a. Candidate incorrectly assesses the Tech Spec implications of this condition and believes that a plant shutdown is required, or that NRC notification is required. The associated LCO (3.5.2) is 7 days.

b. There are no Tech Specs for the CVCS equipment.

c. CORRECT - The associated LCO has been exceeded and since a plant shutdown is now required, notification to the NRC is required (4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months reportable).

d. Candidate incorrectly believes this was a safeguards equipment unplanned actuation, which requires a NRC notification.

KA:

2.2.44 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.

(CFR: 41.5 / 43.5 / 45.12)

IMPORTANCE RO 4.2 SRO 4.4 Test item meets KA. Question provides a control room indication (SIT low pressure), and a time frame for restoration that exceeds the allowable completion time for the TS action. The answer requires the candidate to recognize the failure to restore the pressure in the allowable time has an effect on continued plant operations (requires a shutdown to Mode 3) because of this required system's condition.

HIGH COG - multiple mental steps. Determine if Tech Specs requires a shutdown, based on a given condition that results in a TS Required Action and Completion time being exceeded.

SRO ONLY:

PLP - 2018 NRC EXAM Page: 356 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Test item is SRO ONLY, because the failing to meet the TS action in the time allowed result in entry into a second CONDITON of the applicable Tech Spec and the requirement to Shutdown (an SRO decision) and notify the NRC (another SRO responsibility). This is supported by:

Facility Operating Limitations in the Technical Specifications and Their Bases [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic the following:

application of required actions (TS Section 3) and surveillance requirements (SR) (TS Section 4) in accordance with rules of application requirements (TS, Section 1)

application of generic limiting condition for operation (LCO) requirements (LCO 3.0.1 through 3.0.7; SR 4.0.1 through 4.0.4).

knowledge of TS bases that are required to analyze TS-required actions and terminology

same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM)

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action statements and the safety limits since reactor operators (ROs) are typically required to know these items.

Objective: From memory and given plant conditions involving a non-emergency event,

a. determine if the event requires notifications.

b. describe the steps required to perform the notifications

c. identify the individuals or agencies that require notification in accordance with EN-LI-102, AP-4.00, 10CFR50.72, 10CFR72.75, and 10CFR73.71. (G2.1.14)

(G2.4.30) (SRO only)

Question 97 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 3 Difficulty: 3.00 System ID: 847926 User-Defined ID: G2.2.44 SRO Cross Reference Number: APOR_E01.02-5 SRO ONLY For 100% power conditions, which of the following Topic:

conditions affecting the ability to ma Num Field 1:

Num Field 2:

Text Field:

Comments: Bank PAL-LOI-6095 Not found on 2014 or 2017 NRC exams at PAL.

PLP - 2018 NRC EXAM Page: 357 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 358 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 98 ID: 851010 Points: 1.00 SRO ONLY A waste gas release is being planned for Waste Gas Decay Tanks, T-101A, T-101B and T-68C.

Samples have been completed and analyzed per CH 6.22, Sampling Waste Gas Decay Tank, and CH 6.23, Evaluation and Release of Waste Gas Decay Tank."

There are NO Operational or Maintenance needs driving the release of any Waste Gas Decay Tank The following information is contained on the batch release forms for T-101A, T-101B and T-68C:

T-101A T-101B T-68C Isolation date 52 days ago 50 days ago 18 days ago Isolation pressure 90 psig 91 psig 93 psig Current pressure 93 psig 97 psig 95 psig Which tank(s), if any, should be approved for release?

A. ONLY T-101A B. ONLY T-68C C. ONLY T-101A and T-101B D. T-101A, T-101B, and T-68C Answer: A Answer Explanation Answer: A A. ONLY T-101A Explanation:

Procedure states 'If tank pressure at time of release is more than 5 psig greater than tank pressure after sampling, then do not release tank. Reference CH 6.23 Attachment 3 (page 2 of 2)

T-101B went from 91 psig to 97 psig.

T-68C meets the pressure rise criteria, but does not have enough hold up time per the following Precaution and Limitation from CH 6.23:

4.3 Notwithstanding the minimum tank hold-up time described in Section 4.1, ALARA principles (dose to the public) shall be considered prior to releasing a tank even after the 15-day criteria is met. Unless driven by Operational or Maintenance needs, for example preparation for an upcoming outage, a WGDT should not be released prior to a 50-day hold up time.

Distractors are plausible for the examinee that doesn't know the restrictions.

DISTRACTOR ANALYSIS

a. Correct PLP - 2018 NRC EXAM Page: 359 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1

b. Plausible if the student believes that a pressure rise of 3 or more psig rules out tank T-101A and T-101B, and 18 days is enough hold up time for tank T-68C.

c. Plausible if the student believes that the 50+ days of hold up time is necessary but neglects limit on the rise in pressure.

d. Plausible if the student believes that there are no requirements (or all the requirements are met) for tank pressure or holdup time. The procedure does have explicit instructions for shorter hold up times, but the conditions given in the stem negate them.

K/A: G2.3.11-Ability to control radiation releases.

10CFR: 41.11 / 43.4 / 45.10 SRO Imp: 4.3 Test item meets KA. Question requires the examinee to know what attributes must be met before approving release.

HIGH COG - apply limits of a procedure SRO ONLY Applicable 10CFR55 Section: 43.4 - Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various containment conditions. This exam questions meets the criteria for an SRO-only question because the candidate must analyze the conditions in the waste gas tanks and apply knowledge of the chemistry requirements to determine the controls that apply.

Also, authorizing Waste discharge permits is an SRO-only duty. (task PL-341 012 03 03)

Palisades Learning Objective: RMS_E02.01 002/, From memory, describe the information expected to be completed on Form CH 6.23-3 (WGDT Release Authorization) prior to authorizing the WGDT release to be initiated in accordance with Form CH 6.23-3

References:

CH 6.23 PLP - 2018 NRC EXAM Page: 360 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 98 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 851010 User-Defined ID: 2.3.11 Cross Reference Number: RMS_E02.01-2 SRO ONLY A waste gas release is being planned for Waste Topic:

Gas Decay Tanks, T-101A, T-101B and T-6 Num Field 1:

Num Field 2:

Text Field:

PLP - 2018 NRC EXAM Page: 361 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Comments: Modified from PAL-LOI-5889 which was on the 2014 PLP NRC exam:

A waste gas release is being planned for Waste Gas Decay Tanks, T-101C and T-68A.

Samples have been completed and analyzed per CH 6.22, Sampling Waste Gas Decay Tank, and CH 6.23, Evaluation and Release of Waste Gas Decay Tank."

The following information is contained on the batch release forms for T-101C and T-68A:

NOTE: Todays date/time is 10/18/2020@1800 T-101C T-68A Sample date/time - 10/15/2020@1400 Sample date/time - 10/17/2020@1500 Isolation pressure - 93 psig Isolation Pressure - 92 psig Current pressure - 89 psig Current Pressure - 99 psig Which of the following describes which tank(s), if any, should be approved for release and why?

A. (1) None.

(2) Pressure at the start of release exceeds procedure restrictions for both tanks.

B. (1) T-68A only.

(2) Time since the sample analysis exceeds procedure restrictions for T-101C C. (1) T-101C only.

(2) Pressure at the start of the release exceeds procedure restrictions for T-68A.

D. (1) Both T-68A and T-101C.

(2) All procedural restrictions for sample times and pressures are met for both tanks answer: C PLP - 2018 NRC EXAM Page: 362 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 363 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 99 ID: 850994 Points: 1.00 SRO ONLY:

The plant is tripped, after an accident in which the General Emergency Thresholds are exceeded.

A General Emergency has just been declared.

The State Emergency Operations Center is NOT yet activated.

The INITIAL Protective Action Recommendation (PAR) to be made to the State and Local authorities must be approved by the ___(1)___, and personnally communicated to Van Buren county by ___(2)___.

(1) (2)

A. Shift Manager the Shift Manager B. Shift Manager a qualified ENS Communicator C. Operations Manager the Shift Manager D. Operations Manager a qualified ENS Communicator Answer: A Answer Explanation Answer: A A. Shift Manager; the Shift Manager Explanation:

The governing procedure (EI-6.13), Protective Recommendations for Offsite Populations states that the Shift Manager is responsible to do these two actions.

Step 6.3.1:

6.3.1 Initial Recommendation The declaration of a General Emergency requires that an INITIAL PROTECTIVE ACTION RECOMMENDATION be formulated using Attachment 1, "INITIAL Protective Action Recommendations (PARs)," and communicated to offsite authorities.

a. IF the State Emergency Operations Center is not activated, THEN the Shift Manager/Emergency Director shall approve the PAR and personally communicate (via telephone) the General Emergency and the initial protective action recommendation to Van Buren County.

KA:

2.4.38 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

(CFR: 41.10 / 43.5 / 45.11) l IMPORTANCE RO 2.4 SRO 4.4 PLP - 2018 NRC EXAM Page: 364 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Test item meets KA. Question requires knowledge of an SRO ONLY responsibility / action to take during the implementation of the emergency plan.

MEMORY level.

Objective:

PL-NI00113_E01.05:

Given EI-1, list the ACTIONS to be taken by the Palisades Emergency Response Organization for a given emergency level, per EI-1 SRO ONLY:

Assessment of Facility Conditions and Selection of Appropriate Procedures during Normal, Abnormal, and Emergency Situations [10 CFR 55.43(b)(5)] This 10 CFR 55.43 topic involves both (1) assessing plant conditions (normal, abnormal, or emergency) and then (2) selecting a procedure or section of a procedure to mitigate or recover, or with which to proceed. One area of SRO-level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item. The following are examples:

knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

knowledge of diagnostic steps and decision points in the emergency operating procedures (EOPs) that involve transitions to event-specific sub-procedures or emergency contingency procedures

knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures Question is SRO ONLY because the tasked being performed is a responsibility of an SRO.

PLP - 2018 NRC EXAM Page: 365 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Question 99 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 850994 User-Defined ID: 2.4.38 Cross Reference Number: PL-N00113_E01.05 SRO ONLY: The plant is tripped, after an accident in which the Topic:

General Emergency Thresholds are e Num Field 1:

Num Field 2:

Text Field:

Comments: Modified from PAL-LOI-7861 (96101) which has different distractors and answer.

Not used on 2014 or 2017 NRC exams Due to the ongoing fire in the Turbine building, Covert Fire Department and South Haven Area Emergency Services have arrived on site. With limited number of Security Officers and large number of firefighters is it necessary to suspend the Security Safeguard measure to allow all the firefighters into the Protected Area.

Who has the authority to authorize the suspension of the Security Safeguard measure?

A. On-watch Senior Reactor Operator (SRO)

B. Any licensed Reactor Operator (RO) not in the Control Room.

C. Technical Support Center (TSC) Emergency Plant Manager (EPM).

D. Emergency Operations Facility (EOF) Emergency Director (ED).

Answer: A PLP - 2018 NRC EXAM Page: 366 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 PLP - 2018 NRC EXAM Page: 367 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 100 ID: 850993 Points: 1.00 SRO ONLY The Plant was operating at full power when an event occurs.

The Shift Manager determines that the Primary Coolant System barrier is lost and the Containment barrier is potentially lost.

Which of the following describes the Emergency Classification that would apply for these conditions?

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Answer: C Answer Explanation Answer: C C. Site Area Emergency Explanation:

Per the EP wall charts, the Loss or Potential Loss of ANY two barriers from Table F-1 is Classified as FS1.1.

Distractors are plausible as they are other levels of classification.

KA:

2.4.41 Knowledge of the emergency action level thresholds and classifications.

2.9/4.6 Test item meets KA. Answer requires basic knowledge of the thresholds that give the appropriate level of classification.

MEMORY Level.

Reference:

EI-1 Attachment 1 Table F-1.

Objective:

Given EI-1, EAL Basis and SEP Supplement 1, along with plant emergency condition(s), classify the emergency, per given procedures.

PLP - 2018 NRC EXAM Page: 368 of 369 15 January 2020

EXAMINATION ANSWER KEY Daft N 1 Given EI-1, define each of the four emergency levels to include references to potential releases, per EI-1.

SRO ONLY: Job Function to classify events at Palisades Assessment of Facility Conditions and Selection of Appropriate Procedures during Normal, Abnormal, and Emergency Situations [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both (1) assessing plant conditions (normal, abnormal, or emergency) and then (2) selecting a procedure or section of a procedure to mitigate or recover, or with which to proceed. One area of SRO-level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

Question 100 Info Question Type: Multiple Choice Status: Active Always select on test? No Authorized for practice? No Points: 1.00 Time to Complete: 2 Difficulty: 2.00 System ID: 850993 User-Defined ID: 2.4.41 Cross Reference Number: PL-N00109_E01.09-2 SRO ONLY The Plant was operating at full power when an Topic:

event occurs. The Shift Manager determine Num Field 1:

Num Field 2:

Text Field:

Comments: Bank NOT used on 2014 or 2017 NRC exams.

PAL-LOR-4004 PLP - 2018 NRC EXAM Page: 369 of 369 15 January 2020

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