ML043270578
| ML043270578 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 10/22/2004 |
| From: | Christman R Entergy Nuclear Indian Point 2 |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| References | |
| Download: ML043270578 (200) | |
Text
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Page 1 of 200 Examination Outline Cross-reference:
Level Tier #
1 Group #
KIA #
1 000008 AK3.03 Importance Rating 4.1 4.6 Knowledge of the reasons for actions contained in EOP for PZR vapor space accident/LOCA Proposed Question:
Common #1 Given the following plant conditions:
The Unit has just tripped from 100% power due to a small break LOCA that was caused by a stuck open Pressurizer PORV and Block Valve.
During implementation of E-1, Loss of Reactor or Secondary Coolant, subcooling lowers to 12°F.
The team has just tripped all the Reactor Coolant Pumps.
Which of the following indicates the reason the RCPs were tripped by the team?
A.
To minimize RCS inventory loss.
- 8.
To minimize the cooldown rate.
C.
To prevent RCP damage from cavitation.
D.
To remove pump heat input to RCS Proposed Answer:
A. To minimize RCS inventory loss.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
RCP Trip Criteria EOP Executive Volume Proposed References to be provided to applicants during examination:
NONE Learning Objective:
EOP-C-010-540 (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
Page 2 of 200 New YES Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (3) 55.43 Comments:
L Page 9 of 200 Examination Outline Cross-reference:
Level Tier #
6 Group #
WA #
1 000022 AK1.02 Importance Rating 2.7 3.6 Knowledge of the operational implications of the relationship of charging flow to pressure differential Proposed Question:
Common #2 The position of HCV-142, Charging Line Flow Control Valve, is changed to vary RCP seal injection flow. IF HCV-142 is closed slightly, THEN:
Charging Pump RCP Seal Charging Flow to Discharge Press Injection Flow Regen Hx A.
Increases Increases Decreases B.
Increases Decreases Increases C.
Decreases Increases Decreases D.
Decreases Decreases Increase Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
I P2-SOD-18 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-O30-30105/30106 (As available)
Question Source:
Bank #
INPO 26073 Modified Bank #
New (Note changes or attach parent)
Question History:
Prairie Island 1 9/1/2003
Page 10 of 200 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (3)(8)(10) 55.43 Comments:
Stem reworded to make plant specific
\\.
Page 11 of 200 Examination Outline Cross-reference:
Level Tier #
7 Group #
WA #
1 000025 AK1.01 Importance Rating 3.9 4.3 Knowledge of the operational implications of a loss of RHRS during all modes of operations Proposed Question:
Common #3 Given the following conditions:
0 The Unit is in Mode 6.
RCS temperature is 125°F.
0 21 RHR Pump and 21 RHR Heat Exchanger are in service.
Reactor Vessel level is 68 with Reactor Head detensioned.
HVC-638, 21 RHR Heat Exchanger Flow Control Valve, drifts CLOSED due to an electrical problem.
Assuming NO action by the operating team, which one of the following describes the effect of this failure on plant operation?
A.
Decrease of NPSH to 21 RHR pump due to increased temperature.
B.
OPS actuation due to over pressurization of the RCS.
C.
Loss of RHR letdown resulting in loss of VCT level and operating charging pump.
D.
21 RHR Pump will supply RCS cooling through 22 RHR Heat Exchanger.
Proposed Answer:
A.
Decrease of NPSH to 21 RHR pump due to increased temperature.
Explanation (Optional):
A. Correct, RCS will heat up due to pumps heat and decay heat B. Incorrect, RCS pressure will not increase with Rx Head detensioned C. Incorrect, RHR letdown will NOT be lost since upstream of HVC-638 D. Incorrect, 22 RHR Heat Exchanger would need to be manually aligned Technical Reference(s):
(Attach if not previously provided)
I P2-SOD-020
\\
Page 12 of 200 Proposed References to be provided to applicants during examination:
None Learning Objective:
SYS-C-042-115/126 (As available)
Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (8) 55.43 Comments:
New Question
Page 13 of 200 Examination Outline Cross-reference:
Level Tier #
8 Group ##
WA #
1 000026 AA1.07 Importance Rating 2.9 3.0 Ability to operate and/or monitor flow rates to the components and systems that are serviced by the CCWS; interactions among the components Proposed Question:
Common #4 Given the following conditions:
0 A loss of all AC power has occurred 0
After 15 minutes, power was restored to Busses 5A and 6A 0
The actions of ECA-0.1, Loss of All AC Power Recovery Without SI Required, are being performed to start a CCW Pump Why is MOV-789, RCP Thermal Barrier Return Isolation Valve verified closed prior to restarting the CCW Pump?
A.
Reduce CCW System heat loads to the minimum based on SW loads.
- 6.
Prevent damage to the RCP bearings due to excessive cooldown rate.
C.
Maximize flow to the CVCS components for reestablishing charging, letdown and seal return.
D.
Protect CCW System availability by precluding steam formation in the CCW piping.
Proposed Answer:
D.
Protect CCW availability by precluding steam formation in the CC piping.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Step 1, Page 11 ECA-0.1 Background Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Page 14 of 200 Question Source:
Bank #
Modified Bank #
New INPO 1941 6 (Note changes or attach parent)
Question History:
Kewaunee 1 1 2/11 /2000 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7)(14) 55.43 Comments:
Page 35 of 200 Examination Outline Cross-reference:
Level Tier #
22 Group #
WA #
2 000032 AK2.01 Importance Rating 2.?
3.1 Knowledge of the interrelations between the Loss of Source Range Nuclear Instrumentation and the power supplies, including proper switch positions Proposed Question:
Common #5 Unit 2 is stable at the POAH with Physics testing in progress.
0 PR channel N-44 has been removed from service due to an instrument failure.
A fault occurs that results in a loss of Instrument Bus 21.
Which one of the following describes the effect on SR indications and the basis for this?
A.
SR channel N-32 will remain de-energized due to the P-10 interlock, N-31 will have no power due to the loss of Instrument Bus 21.
B.
SR channel N-31 will re-energize and N-32 will remain de-energized due to the loss of Instrument Bus 21.
SR channels N-31 and N-32 will remain de-energized since the 2/2 permissive cannot be met due to the loss of power to N-36.
x..
C.
D.
SR channel N-31 will de-energized due to the loss of 21 Instrument Bus and N-32 will re-energize due to the loss of P-10 interlock.
Proposed Answer:
A.
SR channel N-32 will remain de-energized due to the P-10 interlock, N-31 will have no power due to the loss of Instrument Bus 21.
Explanation (Optional):
Technical Ref e rence( s) :
(Attach if not previously provided) 2-AOP-IB-1 Page 53 SD-13 Page 78 Proposed References to be provided to applicants during examination:
NONE L
Learning 0 bjective:
(As available)
Page 36 of 200 Question Source:
Bank #
INPO 25731 Modified Bank #
New (Note changes or attach parent)
Question History:
Surry 1 311 412003 Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (7) 55.43 Comments:
Question, answer and distracters modified to make plant specific
Page 187 of 200 Examination Outline Cross-reference:
Level Tier #
114 Group #
WA #
2 0370EA2.12 Importance Rating 3.3 4.1 Ability to determine and interpret flow rate of leak as it applies to a Steam Generator Tube Leak.
Proposed Question:
Common #6 The plant is evaluating a Steam Generator tube leak with the following plant parameters:
0 0
0 b
b b
0 0
Which A.
B.
C.
D.
Letdown flow is at 75 gpm.
One (1) charging pump is running.
Pressurizer level is STABLE.
Seal injection is 29 gpm total Seal return flows are 5 gpm total Charging flow is 60 gpm.
Preexisting RCS leakage was identified as 1 gpm to RCDT.
LCV-112A is in the AUTO position.
ONE of the following is the approximate amount of primary to secondary leakage?
7 gpm 9 gpm Proposed Answer:
B.
8gpm Explanation (Optional):
A. if identified leakage was subtracted from SG leakage B. correct C. if identified leakage was not accounted for D. if seal return was not accounted for Technical Ref erence(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE
Learning Objective:
AOP025SGl-11312 (As available)
Question Source:
Bank #
Modified Bank #
New NEW (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (5) 55.43 Comments:
Replacement question for Question #6
Page 37 of 200 Examination Outline Cross-reference:
Level RO SRO 24 Rev per TF Tier #
Group #
KIA #
1 2
000059 AK3.01 Importance Rating 3.5 3.9 Knowledge of the reasons for the termination of a release of radioactive liquid as it applies to the Accidental Liquid Radwaste Release Proposed Question:
Common #7 Given the following plant status:
A release of 14 Waste Distillate Storage Tank (WDST) is in progress 0
The release permit was approved for the release of the total contents of 14 WDST 0
After proper recirculation of 14 WDST, chemistry obtained a sample of the tank contents to be used on the release permit 0
At the start of the release, R-54, Unit 1 Liquid Waste Distillate Monitor, was in service With the release still in progress, what are the required actions to be taken in accordance with SOP-5.1.5, Calculation and Recording of Radioactive Liquid Releases, should R-54 become inoperable?
A.
Continue with the release, declare R-54 inoperable, direct chemistry to commence taking periodic samples of 14 WDST for the duration of the release, and submit new release permit if any changes are found to existing permit B.
Terminate the release, declare R-54 inoperable, verify actual volume released prior to termination is consistent for expected flow rate during release of 14 WDST, and then continue with the release.
C.
Terminate the release, have chemistry take two independent samples of 14 WDST, have two persons (RO/CRS) independently verify the discharge rate calculations and have two persons independently verify the discharge valve lineup prior to recommencing the release.
D.
Continue with the release since 14 WDST was sampled prior to release, and the sample results indicated a release activity less than the maximum normally allowable for the entire contents of the tank.
Proposed Answer:
C.
Terminate the release, have chemistry take two independent samples of 14 WDST, have two persons (RO/CRS) independently verify the discharge rate calculations and have two persons independently verify the discharge valve lineup prior to recommencing the release.
Explanation (Optional):
Page 38 of 200 L--
Technical Reference(s):
(Attach if not previously provided)
SOP-5.1.2 Pages 1,2 SOP-5.1.5 Pages 5,7 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-051-137 (As available)
Question Source:
Bank #
Modified Bank ##
(Note changes or attach parent)
New Question History:
N/A Question Cognitive Level:
10 CFR Part 55 Content:
YES Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 (5) (10) 55.43 (4)
Comments:
Page 3 of 200 Examination Outline Cross-reference:
Level Tier #
2 Group #
WA #
1 000009 EA2.34
~~
Importance Rating 3.6 4.2 Ability to determine or interpret conditions for throttling or stopping HPI as they apply to a small break LOCA Proposed Question:
Common #8 Given the following plant conditions:
0 0
0 0
The Unit has tripped from 100% due to a small break LOCA.
Conditions have stabilized and operators are evaluating the criteria for terminating Safety Injection.
Containment pressure has stabilized at 1.5 psig.
Containment Radiation levels peaked at 17 Whr.
Which one of the following conditions would PREVENT SI termination per E-1, "Loss of Reactor or Secondary Coolant"?
A.
RCS subcooling is 25°F and stable.
B.
Pressurizer level is 10% and stable.
C.
Only one Steam Generator level is >lo% narrow range.
D.
RCS pressure is 1700 psig. and stable Proposed Answer:
B.
Pressurizer level is 10% and stable.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
E-1 Step 11 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
EOP-C-011-540 (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
Page 4 of 200 New YES Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 (10) 55.43 (5)
Page 55 of 200 Examination Outline Cross-reference:
2 40 Group ##
1 Rev per TF - check KIA ##
056 A2.04 Importance Rating 2.6 2.8 Ability to (a) predict the impacts of a Loss of condensate pump on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of the loss of condensate pumps Proposed Question:
Common #9 Given the following initial plant conditions:
0 Reactor power, 60%
0 0
0 0
0 Tave, 554°F 21 and 23 Condensate Pumps in service 22 Condensate Pump in standby MBFP suction pressure, 450 psig All 4 SG levels, 48 - 49%
21 condensate pumps subsequently trips due to a motor fault.
How is the Condensate System affected by the loss of the condensate pump and what are the operator actions?
A.
No automatic start of 22 Condensate Pump; manually start 22 Condensate Pump and manually open Flexitest switches for 22 Condensate Pump B.
22 Condensate Pump starts automatically; manually open Flexitest switches for 22 Condensate Pump C.
No automatic start of 22 condensate pump; must manually start 22 condensate pump, Flexitest switches for 22 and 23 Condensate Pumps will automatically open D.
22 Condensate Pump starts automatically, Flexitest switches for 22 and 23 Condensate Pump will automatically open Proposed Answer:
B.
22 Condensate Pump starts automatically; manually open Flexitest switches for 22 Condensate Pump Explanation (Optional):
A.
B.
Correct C.
D.
22 Cond Pumps starts automatically 22 Cond Pumps starts automatically and Flexitest switches are only manually opened for 22 Cond Pump Flexetex switches only manually opened for 22 Cond Pump
Technical Reference(s):
2-POP-1.3 2-AOP-FW-1 2-SOP-20.2 (Attach if not previously provided)
Page 34 Pages 25,28 Pages 6,7,13 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
AOP-C-FW1-11503 (As available)
SYS-C-200-364 Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 (5) 55.43 Comments:
New Question
Page 39 of 200 Examination Outline Cross-reference:
25 Rev per TF Tier #
1 Group #
2 Importance Rating 4.1 4.3 KIA #
000068 AAl.03 Ability to operate and / or monitor the S/G levels as they apply to the Control Room Evacuation Proposed Question:
Common #10 The Team has evacuated the Control Room in accordance with 2-AOP-SSD-1, CONTROL ROOM INACCESSABILITY SAFE SHUTDOWN CONTROL.
Which ONE of the following describes the preferred method of maintaining SG inventory during the cooldown per 2-AOP-SSD-l ?
A.
Feed all 4 SGs to maintain corrected Wide Range level at approximately 63 - 68% to ensure symmetric heat removal B.
Feed 21 and 22 SGs unless explicitly directed to steam from the intact SGs by the EOPs or FRPs since they are the only SGs with reliable backup level indication at the Safe shutdown Panel
\\
C.
Feed 22 and 23 SGs unless explicitly directed to steam from the intact SGs by the EOPs or FRPs since they supply steam to 22 Auxiliary Boiler Feed Pump D.
Feed any combination of SGs that ensures greater than 400 gpm total auxiliary feedwater flow to maintain adequate heat sink Proposed Answer:
B.
Feed 21 and 22 SGs unless explicitly directed to steam from the intact SGs by the EOPs or FRPs since they are the only SGs with reliable backup level indication at the Safe shutdown Panel Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided) 2-AOP-SSD-1 Page 5 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Page 40 of 200 Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (10) 55.43 (5)
Comments:
Page 49 of 200 Examination Outline Cross-reference:
37 Knowledge of the CSS design feature@)
via the CSS Level RO SRO Tier #
Group #
WA #
2 1
026 K4.06 Importance Rating 2.8 3.2 and/or interlock(s) which provide for Iodine scavenging Proposed Question:
Common #11 Tri-Sodium Phosphate (TSP) Baskets are located inside containment on the 46' El. for introduction into the Containment environment during the recirculation phase in order to maintain the Containment Sump pH basic.
What is the reason for controlling the pH?
A.
Maintains Iodine in solution.
B.
Maintains Hydrogen in solution.
C.
Reduces Iodine concentration in solution.
D.
Reduces Hydrogen concentration in solution.
Proposed Answer:
A.
Maintains Iodine in solution.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
ES-1.3 Background Page 62 ES-1.2 Background Page 79 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-102 - KR 158b (As available)
Question Source:
Bank ##
INPO 20630 Modified Bank #
New (Note changes or attach parent)
Question History:
Point Beach 1 2/2/02
L' Page 50 of 200 Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (7) 55.43 Comments: Question modified to make plant specific for Na3P04 baskets
i, Page 7 of 200 Exam i nation 0 u tl in e Cross-reference:
Level Tier #
5 Group #
WA #
1 00001 5 AK3.02
~
Importance Rating 3.0 3.1 Knowledge of the reasons for responses of CCW lineup and flow paths to RCP oil coolers during RCP malfunctions Proposed Question:
Common #12 Given the following:
Unit 2 is operating at 100% power.
MOV-784, RCP Bearing Discharge Isolation Valve Phase B closed 2 minutes ago due to an electrical short and cannot be opened.
MOV-786, RCP Bearing Discharge Isolation Valve Phase B has remained open.
Highest Upper Radial Bearing temperature is currently reading 190°F and rising.
RCP Bearings Cooling Water Return High Temperature Alarm NOT illuminated.
In accordance with 2-AOP-CCW-1, Loss of Component Cooling, which one of the following actions, if any, is required and why ?
A.
Pumps can remain in service since CCW flow to oil coolers will be maintained through return valve MOV-786.
B.
Pumps can remain in service until RCP Bearings Cooling Water Return High Temperature alarm is received.
C.
Reactor must be tripped and ALL RCPs stopped due to loss of CCW flow to ALL RCP oil coolers.
D.
Reactor must be tripped and 21 and 22 RCPs stopped due to loss of CCW flow to their respective oil coolers.
Proposed Answer:
C.
Reactor must be tripped and ALL RCPs stopped due to loss of CCW flow to ALL RCP oil coolers.
Explanation (Optional):
MOV-784 and 786 are in series. Either one closes secures CCW flow to all RCP oil coolers Technical Reference(s):
(Attach if not previously provided)
IP2-SOD-014 2-ARP-SG F Window 1 -3
~
~-
Page 8 of 200 ARP SCF Window 1-6 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-Ol3-40/41 (As available)
AOP-C-RCP1-1601039 Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New YES Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (3)(5)(10) 55.43 (5)
Comments:
Page 57 of 200 Examination Outline Cross-reference:
Level Tier #
41 Group ##
KIA #
1 061 K5.01 importance Rating 3.6 Knowledge of the operational implications of the relationship between AFW flow and RCS heat transfer Proposed Question:
Common #13 A reactor trip occurs from 100% power due to a loss of main feedwater.
The following conditions exist:
0 All RCPs are running.
0 0
The turbine driven AFW pump is in service feeding all 4 SGs.
Both motor driven AFW pumps tripped upon startup and remain unavailable.
The turbine driven A M pump speed has begun to slowly lower due to a malfunctioning governor.
Which one of the following describes the impact on Pressurizer level if the turbine driven AFW pump speed CONTINUES to lower?
Pressurizer level:
A.
rises due to increased primary to secondary heat transfer.
B.
rises due to decreased primary to secondary heat transfer.
C.
lowers due to increased primary to secondary heat transfer.
D.
lowers due to decreased primary to secondary heat transfer.
Proposed Answer:
B.
rises due to decreased primary to secondary heat transfer.
Explanation (Optional):
Technical Reference( s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE
Page 58 of 200 Learning Objective:
(As available)
Question Source:
Bank #
INPO 2502 1 Modified Bank #
New (Note changes or attach parent)
Question History:
Beaver Valley 1 12/01/02 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (5) 55.43 Comments:
Question modified to make plant specific
Page 61 of 200 Examination Outline Cross-reference:
Level Tier #
WA #
Importance Rating 43 Group #
2 W/E07 G2.4.18 2.7 3.6 Knowledge of the specific bases for EOPs Proposed Question:
Common #14 Step 1 of FR-C.3, "Response to Saturated Core Conditions", checks if the RHR system has been placed in service in the shutdown cooling mode.
Which of the following describes the basis for this step?
A.
To ensure an ORANGE or RED condition in Core Cooling will not arise while performing this procedure.
B.
If RHR is in service in the shutdown cooling mode, the saturated core cooling condition is a problem with RHR and this procedure will not address this condition.
C.
To verify RHR is aligned for long term cooling if the appropriate conditions are met D.
If RHR is in service in the shutdown cooling mode, the saturated core cooling condition is a problem with RHR and this procedure will identify and isolate the affected train.
Proposed Answer:
B.
If RHR is in service in the shutdown cooling mode, the saturated core cooling condition is a problem with RHR and this procedure will not address this condition.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
FR-C.3 Background Page 8 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 22526 Modified Bank #
New (Note changes or attach parent)
Question History:
Diablo Canyon 1 10/1/2002
Page 62 of 200 Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (10) 55.43 Comments:
Page 63 of 200 Examination Outline Cross-reference:
Level Tier #
44 Group #
WA ##
2 W/E15 EK1.2 Importance Rating 2.7 2.9 Knowledge of the operational implications of normal, abnormal and emergency operating procedures associated with Containment Flooding Proposed Question:
Common #15 The following conditions exist:
0 A large break LOCA has occurred.
0 The plant is tripped and ECCS is operating as expected.
Accumulators have discharged and are isolated.
0 The SM directs performance of FR-Z.2, Containment Flooding.
Which one of the following describes the required actions per FR-Z.2 and their purpose?
A.
Secure all water sources from outside of containment to prevent damaging vital electrical equipment and diluting the containment water inventory.
B.
Secure all water sources from outside of containment to prevent overloading concrete containment structures and diluting the containment water inventory.
C.
Locate source of flooding in an attempt to prevent damaging vital electrical equipment and diluting the containment water inventory.
D.
Locate source of flooding in an attempt to prevent overloading concrete containment structures and diluting the containment water inventory.
Proposed Answer:
C.
Locate source of flooding in an attempt to prevent damaging vital electrical equipment and diluting the containment water inventory.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
FR-Z.2 Page 3 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Page 64 of 200 Question Source:
Bank #
Modified Bank #
New INPO 24609 (Note changes or attach parent)
Question History:
Seabrook 1 5/30/2003 Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (8,lO) 55.43 Comments:
L-'
'\\-
Page 65 of 200 Examination 0 ut I i n e Cross-reference:
Level Tier #
45 Group #
WA #
2 W/E16 EA2.2 Importance Rating 3.0 3.3 Ability to determine and interpret adherence to appropriate procedures and operation withing the limitations in the facility's license and amendments as they apply to High Containment Radiation.
Proposed Question:
Common #16 Given the following plant conditions:
0 0
0 0
A loss of coolant accident has just occurred.
During the initial phases of the accident, containment pressure peaked at 32 psig and containment radiation dose rate peaked at 2.5E6 Whr.
The CRS has directed that adverse containment numbers be used during EOP implementation.
Approximately 30 minutes later, containment pressure has lowered to 6 psig and containment radiation dose rate has lowered to 1 E3 Whr.
The CRS must direct that adverse containment numbers:
A.
still be used until containment pressure is less than 4 psig.
B.
still be used until relaxed by Technical Support Center personnel.
C.
not be used since the containment radiation level is no longer indicative of adverse containment conditions.
D.
not be used since the containment pressure is no longer indicative of adverse containment conditions.
Proposed Answer:
B.
still be used until relaxed by Technical Support Center personnel.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
OAP-012 Page 10 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Page 66 of 200 Question Source:
Bank #
Modified Bank #
New Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
(Note changes or attach parent)
YES Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 (10) 55.43 (5)
Comments:
Page 5 of 200 Examination Outline Cross-reference:
1 3
Group #
1 L
WA #
000011 EK2.02 Importance Rating 2.6 2.7 Knowledge of the interrelations between pumps and a Large Break LOCA Proposed Question:
Common #17 Unit 2 is shutdown following a Loss of Offsite Power with a LOCA event. The operators are performing actions in. E-1, Loss of Reactor or Secondary Coolant.
Which ONE of the following statements is the basis for placing ALL non-running CCW pump control switches in PULL-TO-LOCK prior to resetting the SI signal?
A.
Prevent Thermal shock to the RCP Thermal Barriers.
B.
Prevent an overload condition on the Emergency Diesel Generators.
C.
Prevent CCW System overpressure with all 3 CCW pumps starting simultaneously.
D.
Prevent steam formation in the RHR heat exchanger CCW side.
Proposed Answer:
B.
Prevent an overload condition on the Emergency Diesel Generators.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Step 5, Page 47 E-1 Background Proposed References to be provided to applicants during examination:
NONE Learning Objective:
EOP-C-011-540 (As available)
Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge X
L-Page 6 of 200 Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (7, 10) 55.43 Comments:
\\..
Page 15 of 200 Examination Outline Cross-refer en ce:
Level Tier #
9 Group #
KIA ##
1 000027 AA2.15 Importance Rating 3.7 4.0 Ability to determine and interpret the actions to be taken if PZR pressure instrument fails high Proposed Question:
Common #18 Given the following conditions:
The plant is at 100% power.
0 ALL control systems are in their normal automatic alignments.
Pressurizer pressure channel PT-455 indicates 2275 psig and slowly rising.
0 All other narrow range pressurizer pressure indications are 2220 psig and slowly dropping Which of the following actions is required in accordance with 2AOP-INST-l ?
A.
Place the affected PORV control switch in CLOSE.
B.
Place the pressurizer pressure master controller in MANUAL and control RCS pressure.
C.
RESET and reenergize pressurizer heaters D.
TRIP the reactor, enter E-0, Reactor Trip or Safety Injection.
Proposed Answer:
B.
Place the pressurizer pressure master controller in MANUAL and control RCS pressure.
Explanation (Optional):
A. - Incorrect. Only for channels directly impacting PORVs B. - Correct. Controlling channel is failing C. - Incorrect. Heaters will not energize until pressure control in manual D. - Incorrect. Rx trip criteria not yet met Technical Reference(s):
(Attach if not previously provided) 2-AOP-I NST-1 Page 7 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-014-50 (As available)
Question Source:
Bank #
INPO 23382
-./
Modified Bank #
(Note changes or attach parent)
New Question History:
Indian Point Unit 3 3/10/2003 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7) 55.43 (5)
Comments:
Stem modified to make procedure plant specific
Page 189 of 200 Examination Outline Cross-reference:
1 Group #
1 KIA #
000029 EK2.06 Importance Rating 2.9 3.1 Knowledge of the interrelations between the breakers, relays, and disconnects following an ATWS Proposed Question:
Common #19 Unit 2 is operating in Mode 1. An initiating event occurred such that an automatic Reactor Trip signal was generated, however the Reactor Trip Breakers did not open.
Which one of the following is the initiating event?
A.
With reactor power at 12%, an electrical fault caused the Main Generator F7-9 Disconnect to open while synchronized to the grid.
B.
With reactor power at 12%, 21 MSlV inadvertently closed during turbine roll-up.
C.
With reactor power at 28%, 22 RCP shaft seized causing an overload trip of its supply breaker.
D.
With reactor power at 28%, the RO inadvertently unblocked and energized the Source Range Nls.
Proposed Answer:
C.
With reactor power at 28%, 22 RCP shaft seized causing an overload trip of its supply breaker.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
Question Source:
(As available)
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Page 190 of 200 Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7) 55.43 Comments:
New Question to replace Question 19
Page 17 of 200 Examination Outline Cross-reference:
Level Tier #
11 Group #
WA #
1 000038 EA?. 1 1 Importance Rating 3.8 3.9 Ability to operate and monitor SG level indicators as they apply to a SGTR Proposed Question:
Common #20 A SGTR on 24 SG caused an automatic SI on Unit 2.
0 0
24 SG is isolated 0
TAVE is stable 0
21, 22 and 23 SG NR levels were stable at 30%
0 23 SG NR level has started to rise with no change in ARN flow or steaming rate The operating team is in E-3, Steam Generator Tube Rupture, performing SI RESET actions Which one of the following describes the required action(@?
A.
Continue performing the steps of E-3 B.
Return to E-3, Step 1 C.
Reduce ARN flow to 23 SG and monitor level while continuing in E-3
- 0. Re-initiate SI and return to E-0, Step 1 Proposed Answer:
B.
Return to E-3, Step 1 Explanation (Optional):
(A) E-3 only isolates a ruptured SG if started from the beginning of the procedure, (B) Correct answer when a tube rupture in a second SG is identified, (C) This could mask another SG tube leak or rupture, (0)
E-0 would only identify a SG tube rupture and initiate E-3 at step 1 Technical Reference(@:
(Attach if not previously provided)
E-3 foldout page Proposed References to be provided to applicants during examination:
NONE
Page 18 of 200 Learning Objective:
(As available)
Question Source:
Bank #
INPO 231 22 Modified Bank #
New (Note changes or attach parent)
Question History:
Salem Unit 1 1 1 /04/2002 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7) 55.43 (5)
Comments:
i Page 21 of 200 Examination Outline Cross-reference:
Level Tier #
14 Group #
WA #
1 000056 G2.1.20 Importance Rating 4.3 4.2 Ability to execute procedure steps.
Proposed Question:
Common #21 A Reactor Trip with Loss of Offsite power occurred from 100% power 30 minutes ago.
The current plant conditions are:
ALL SGs are available.
RCS subcooling is 26°F.
0 0
Average of Qualified CETs is 530°F and rising slowly:
Loop T-hots are 520°F and rising slowly.
Loop T-colds are 490°F and rising slowly.
Which one of the following actions are to be taken in accordance with ES-0.1, Reactor Trip Response, to enhance natural circulation?
L.
A.
Turn on available pressurizer heaters B.
Initiate Auxiliary Spray C.
Throttle open the Auxiliary Feed Water control valves D.
Throttle open the Atmospheric Dump Valves Proposed Answer:
D.
Throttle open the Atmospheric Dump Valves Explanation (Optional):
Technical Reference(@:
(Attach if not previously provided)
Step 10 & Attachment 3 ES-0.1 Proposed References to be provided to applicants during examination:
NONE L-Learning Objective:
EOP-C-004-509 (As available)
Page 22 of 200 Question Source:
Bank ##
Modified Bank ##
(Note changes or attach parent)
New YES Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (10) 55.43 (5)
Comments:
Page 25 of 200 Examination Outline Cross-reference:
Level 16 Tier #
Group #
WA #
1 W/E04 EK2.1 Importance Rating 3.5 3.9 Knowledge of the interrelations between the (LOCA Outside Containment and the components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
Proposed Question:
Common #22 The following plant conditions exist:
e e
a e
e 0
e e
e The unit is in MODE 4 RHR cooldown is in progress.
RCS Temperature - 340°F slowly decreasing.
RCS pressure - 300 psig decreasing.
PZR level - 42% decreasing.
Containment pressure - 0.2 psig and steady.
PAB 98 ft. elevation radiation monitor R-5987 is alarming Plant vent monitors R-43 and R-44 are alarming All S/G Narrow Range levels are steady at approximately 42%
S/G pressures are steady at approximately 125 psig Based on the above conditions, what has occurred?
A.
B.
C.
D.
A steam leak has occurred inside containment.
The Low Temperature Overpressure Protection (LTOP) system has actuated.
Letdown line pressure control valve, PCV-135 has failed open.
A LOCA has occurred on the suction of the RHR pump.
Proposed Answer:
D.
A LOCA has occurred on the suction of the RHR pump.
Explanation (Optional):
A, B - inside containment C - Incorrect (wrong lineup)
D.- Correct (radiation levels indicate LOCA in PAB)
Technical Reference(s):
(Attach if not previously provided)
~
~
~-
~~
Page 26 of 200 ARP SAF-1 window 2-8 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 24608 Modified Bank #
New (Note changes or attach parent)
Question History:
Seabrook 1 5/30/2003 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7) 55.43 (5)
Comments:
Question modified to make plant specific
Page 29 of 200 Examination Outline Cross-reference:
Level RO SRO 18 Rev per TF Tier #
1 Group #
1 KIA #
000065 G2.1.2 Importance Rating 3.0 4.0 Knowledge of operator responsibilities during all modes of plant operation.
Proposed Question:
Common #23 Given the following conditions:
The plant is at 100% power.
A Loss of Instrument Air pressure has occurred The CRS has directed entry to 2-AOP-AIR-1, Loss of Instrument Air Which of the following plant conditions will require a reactor trip in accordance with 2-AOP-AIR-l ?
A.
Steam Generator Levels 45% and decreasing slowly B.
VCT Level 4% and decreasing slowly C.
Pressurizer Level is 55% and increasing slowly D.
Instrument Air header pressure is 85 psig and decreasing slowly Proposed Answer:
B.
VCT Level 4% and decreasing slowly Explanation (Optional):
A - Rx trip required for >lo% level change from program B - Correct ( ~ 5 %
Trip Rx)
C - >5% above program requires plant S/D D - No trip required for low air pressure only required to check other parameter if <95 psig Technical Reference(s):
(Attach if not previously provided) 2-AOP-AIR-1 Pages 7,8,9 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-292-474 (As available)
Page 30 of 200 Question Source:
Question History:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
New Question Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (1 0) 55.43 (5)
Comments:
Page 31 of 200 Examination Outline Cross-reference:
Level Tier #
19 Group ##
WA #
1 W/E05 EK1.2 Importance Rating 3.9 4.5 Knowledge of the operational implications of normal, abnormal and emergency operating procedures associated with the (Loss of Secondary Heat Sink)
Proposed Question:
Common #24 The following plant conditions exist:
a The team is responding to a LOCA in accordance with E-1, Loss of Reactor or Secondary Coolant.
A RED path occurs on the "Heat Sink Critical Safety Function Status Tree".
The team transitions to FR-H.1, "Response to Loss of Secondary Heat Sink".
Total available AFW flow is 200 gpm.
RCS pressure is 525 psig and STABLE.
Containment pressure is 10 psig and INCREASING.
SG pressures are all 950 psig and STABLE.
SG wide range levels are 45% and DECREASING.
Which of the following actions are required?
A.
Transition back to E-1, "Loss of Reactor or Secondary Coolant" B.
Attempt to establish feed to the SG using the Main Boiler Feed Pumps C.
Attempt to establish feed to the SG using the Condensate Pumps D.
Steps 9 through 15 are to be petformed immediately to establish bleed and feed Proposed Answer:
A.
Transition back to E-1, "Loss of Reactor or Secondary Coolant" Explanation (Optional):
RCS pressure is less than all non-faulted SGs. Secondary heat sink is not required. Note, "Adverse Conditions " DO NOT change pressure setpoints Technical Reference(s):
(Attach if not previously provided)
FR-H.l Page 2 Proposed References to be provided to applicants during examination:
NONE
Page 32 of 200 Learning Objective:
EOP-C-044-3561 (As available)
Question Source:
Bank #
INPO 2471 5 Modified Bank #
YES (Note changes or attach parent)
New Question History:
Seabrook 1 5/30/2003 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7) 55.43 (5)
Comments:
Question and distracters Modified
Page 33 of 200 Examination Outline Cross-reference:
Level Tier #
21 Group #
WA #
2 000028 AK2.02 Importance Rating 2.6 2.7 Knowledge of the interrelations between the Pressurizer Level Control Malfunctions and sensors and detectors Proposed Question:
Common #25 The plant is operating at 100% power with all control systems operating normally. The reference leg of LT-460 has just developed a leak where the reference leg connects to the D/P cell. LT-459/460 is selected for PZR level control and alarm (Level 3 in DEFEAT).
Which one of the following best describes the plant response from this leak?
A.
LT-459 - indication will decrease, LT-460 indication will increase, LT-461 - indication will increase, charging flow will increase.
B.
LT-459 - indication will decrease, LT-460 indication will increase, LT-461 - indication will decrease, charging flow will decrease C.
LT-459 - indication will increase, LT-460 indication will decrease, LT-461 indication will decrease, backup heaters will deenergize.
D.
LT-459 - indication will decrease, LT-460 indication will decrease, LT-461 indication will increase, backup heaters will energize.
Proposed Answer:
B.
LT-459 - indication will decrease, LT-460 indication will increase, LT-461 - indication will decrease, charging flow will decrease Explanation (Optional):
LT-460 will indicate high, 460 is in control and 459 is in alarm with defeat switch to defeat channel 3 (461), charging pump speed will decrease and actual level will decrease Technical Reference(s):
(Attach if not previously provided) 2-AOP-INST-1 IP2 SOD 007 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Page 34 of 200 Question Source:
Bank #
INPO Modified Bank #
New 2461 2 (Note changes or attach parent)
~~
Question History:
Seabrook 1 5/30/2003 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7) 55.43 Comments:
Question, answer and distracters modified to make plant specific
Page 41 of 200 Examination Outline Cross-reference:
Level i-30 Tier #
Group #
KIA #
1 003 K5.05 Importance Rating 2.8 3.0 Knowledge of the operational implications of the dependency of RCS flow rates upon the number of operating RCPs Proposed Question:
Common #26 The Plant is in MODE 3 with all Steam Generators available. Which statement describes the effect on the Reactor Coolant System (RCS) of the number of operating Reactor Coolant Pumps (RCPs)?
A.
Fifteen minutes after shutting off ALL RCPs there will be NO flow in the RCS, and margin to DNB will be reduced.
B.
Operating ALL RCPs raises RCS flow rate, but results in a reduction in DNB margin due to pump heat input.
C.
Reducing the number of operating RCPs lowers the RCS flow rate causing a REDUCTION in DNB margin.
Reducing the number of operating RCPs lowers the RCS flow rate causing a RISE in DNB margin.
D.
Proposed Answer:
C.
Reducing the number of operating RCPs lowers the RCS flow rate which causes a REDUCTION in DNB margin.
Explanation (Optional):
Technical Ref e rence( s) :
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning 0 bj ective:
(As available)
Question Source:
Bank #
INPO 26298 Modified Bank #
(Note changes or attach parent)
Page 42 of 200 New Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (5) 55.43 Comments:
Page 43 of 200 Examination Outline Cross-reference:
Level Tier #
L.
31 Group #
WA#
1 004 A4.15 Importance Rating 3.6 3.7 Ability to manually operate and/or monitor in the control room Boron concentration Proposed Question:
Common #27 Given the following conditions:
Unit 2 is at 100% power.
Core Burnup is 11,000 MWD/MTU.
Rod Control is in MANUAL.
All other plant controls are in their normal configuration.
AUTO makeup initiated to the VCT.
The boron addition rate is set 5 gpm higher than required for present RCS conditions.
Assuming NO operator action, what will be the effect on the following parameters 15 minutes after the auto makeup is complete?
Reactor Power RCS Tave Main Generator Electrical Output
\\
A.
Lower Lower Higher B.
Higher Higher Lower C.
Higher Higher Higher D.
Lower Lower Lower Proposed Answer:
D.
Lower Lower Lower Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Page 44 of 200 Question Source:
Bank #
Modified Bank #
New INPO 24074 (Note changes or attach parent)
Question History:
Salem Unit 1 5/05/2003 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7) 55.43 Comments:
Page 75 of 200 Examination Outline Cross-reference:
Level Tier #
50 Group #
WA #
1 01 0 A3.02 Importance Rating 3.6 3.5 Ability to monitor automatic operation of PZR PCS, including: PZR pressure.
Proposed Question:
Given the following:
Common #28 0
0 0
PRZR Pressure Control System in position "DEFEAT 3 & 4" Both PORV Block Valves CLOSED with control switches in AUTO Both PORVs CLOSED with control switches in AUTO All PRZR Heater control switches in AUTO PRZR Spray Valve controllers in AUTO PRZR Pressure Master Controller in AUTO Which ONE of the following statements describes the RCS/Pressurizer system response should PRZR pressure transmitter PT-455, Channel 1, fail LOW? {ASSUME: NO operator action.}
A.
ALL pressurizer heaters turn ON and Both PORVs available to automatically cycle for control of pressurizer pressure.
- 9.
ALL pressurizer heaters turn ON and Only 1 PORV available to automatically cycle for control of pressurizer pressure C.
ALL pressurizer heaters turn ON and Spray valves cycle to control pressurizer pressure.
D.
ALL pressurizer heaters turn ON and Reactor trips on high pressurizer pressure condition.
Proposed Answer:
B.
All pressurizer heaters turn ON and Both PORVs available to cycle to control pressure Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
IP2-SOD-007 Proposed References to be provided to applicants during examination:
NONE
Page 76 of 200 Learning Objective:
SYS-C-014-54 (As available)
Question Source:
Bank #
INPO 201 20 L-Modified Bank #
New (Note changes or attach parent)
Question History:
Cook 1 9/10/2001 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (357) 55.43 Comments:
Questions, answer and distracters modified to make plant specific
Page 77 of 200
-\\.---
Examination Outline Cross-reference:
Level Tier #
52 Group #
KIA #
1 012 K1.02 Importance Rating 3.4 3.7 Knowledge of the physical connections and/or cause effect relationships between the RPS and the 125VDC System.
Proposed Question:
Common #29 The plant is operating at 100% when a loss of 125VDC control power occurs.
Which one of the following describes the effect, if any, on the Reactor TRIP and BYPASS breakers?
A.
All TRIP and BYPASS breakers will open automatically B.
The BYPASS breakers will open automatically but the TRIP breakers will NOT automatically open C.
The TRIP breakers will automatically open but the BYPASS breakers will NOT automatically open D.
All TRIP and BYPASS breakers will NOT open Proposed Answer:
D.
All TRIP and BYPASS breakers will NOT open Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided) 2-AOP-DC-1 Pages 99,125 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
Question Source:
SYS-C-161-277 SYS-C-271 B-6454 (As available)
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Page 78 of 200 Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 (6) 55.43 Comments:
Page 101 of 200 Examination Outline Cross-reference:
Level Tier #
66 Group #
WA #
2 011 K5.15 Importance Rating 3.6 4.0 Knowledge of the operational implications of the PZR level indication when RCS is saturated Proposed Question:
Common #30 During a natural circulation cooldown, which of the following pressurizer level responses would indicate the presence of a void in the reactor vessel upper head?
A.
A Pressurizer level increase when charging flow is directed through the auxiliary sprays.
B.
A Pressurizer level decrease when charging flow is directed through the auxiliary sprays.
C.
A Pressurizer level increase when charging flow is directed into the cold legs.
D.
A Pressurizer level decrease when there is an increase in the cooldown rate.
Proposed Answer:
A.
A Pressurizer level increase when charging flow is directed through the auxiliary sprays.
Explanation (Optional):
Auxiliary spray flow would decrease the pressure in the pressurizer and cause a steam void to form in the upper head resulting in water being displaced to the pressurizer Technical Reference(s):
(Attach if not previously provided)
ES-0.3 Background Pages 18-22 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
EOP-C-008-3531 (As available)
Question Source:
Bank #
Modified Bank #
New Yes (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
Page 102 of 200 10 CFR Part 55 Content:
55.41 (3,8,14) 55.43 Comments:
Page 103 of 200
.\\._.--
Examination Outline Cross-reference:
Level Tier #
67 Group #
WA #
2 014 Al.03 Importance Rating 3.6 3.8 Ability to predict and/or monitor changes in parameters associated with operating the RPlS controls, including PDIL, PPDIL Proposed Question:
Common #31 Which one of the following instrument failures would directly cause a change in the computer calculated rod insertion limits?
A.
B.
An impulse pressure channel failing HIGH.
A THOT RTD channel failing HIGH.
C.
A Power Range NIS channel failing HIGH.
D.
A Pressurizer pressure channel failing LOW.
Proposed Answer:
B.
A THOT RTD channel failing HIGH.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SY S-C-1 63-292/293/295 (As available)
Question Source:
Bank #
INPO 20629 Modified Bank #
New (Note changes or attach parent)
Question History:
Point Beach 1 2/02/2002 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
Page 104 of 200 10 CFR Part 55 Content:
55.41 (6) 55.43 Comments:
Page 107 of 200 i-
\\-
Examination Outline Cross-reference:
Level Tier #
69 Group #
WA #
RO 2
SRO 2
017 A2.02 Importance Rating 3.6 4.1 Ability to (a) predict the impacts of core damage on the ITM system; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of core damage Proposed Question:
Common #32 A severe accident has occurred and the operating team is currently implementing the Emergency Operating Procedures. The following conditions exist:
All RCPs are off.
0 0
0 0
PZR level is off scale low.
RVLIS, Natural Circ Range, is 68%
RCS pressure is 400 psig.
Core exit thermocouples are reading 750°F.
What course of action will the operating team take and why?
A.
Transition to FR-C.1, Response to Inadequate Core Cooling, because core damage is occurring.
B.
Transition to FR-C.l, Response to Inadequate Core Cooling, because core uncovery is likely occurring.
C.
Transition to FR-C.2, Response to Degraded Core Cooling, because core damage is occurring.
D.
Transition to FR-C.2, Response to Degraded Core Cooling, because core uncovery is likely occurring.
Proposed Answer:
D.
Transition to FR-C.2, Response to Degraded Core Cooling because core uncovery is likely occurring.
Explanation (Optional):
A. FR-C.1 requires >120O0Ffor core damage B. FR-C.1 requires >700"F with RVLIS <41% for core damage imminent C. Core damage occurs >1200"F D. Correct Technical Reference(s):
(Attach if not previously provided)
F-0.2
~
_ _ _ ~
~
FR-C. 1
Background
Page 108 of 200 FR-C.2
Background
Proposed References to be provided to applicants during examination:
NONE L.'
Learning Objective:
EOP-C-001-500 EOP-C-O18-576,3539 EOP-C-019-3540 (As available)
Question Source:
Bank #
INPO 21 574 Modified Bank #
New (Note changes or attach parent)
Question History:
Kewaunee 1 9/0 6/2 002 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7, 14) 55.43 Comments:
Minor modification to Question given parameters
Page 11 9 of 200 Examination Outline Cross-reference:
Level Tier #
75 Group #
WA #
2 086 A3.02 Importance Rating 2.9 3.3 Ability to monitor automatic operation of the Fire Protection System including actuation of the FPS Proposed Question:
Common #33 Fire Protection System pressure has lowered to 120 psig, and is continuing to lower slowly.
Which one of the following best describes the sequence in which the alternate pumps will start if system pressure were to continue to slowly lower?
A.
11 Fire Main Booster Pump, 12 Fire Main Booster Pump, Diesel Fire Pump.
B.
12 Fire Main Booster Pump, 11 Fire Main Booster Pump, Diesel Fire Pump.
C.
Diesel Fire Pump, 11 Fire Main Booster Pump, 12 Fire Main Booster Pump.
D.
Diesel Fire Pump, 12 Fire Main Booster Pump, 11 Fire Main Booster Pump.
\\
Proposed Answer:
B.
12 Fire Main Booster Pump, 11 Fire Main Booster Pump, Diesel Fire Pump.
Explanation (Optional):
12 Fire Main Booster Pump starts 105 psig, 1 1 Fire Main Booster Pump starts at 90 psisg, Diesel Fire Pump starts at 65 psig Technical Reference(s):
(Attach if not previously provided)
SOP-29.6 Pages 16 & 28 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-(3-296-486 (As available)
Question Source:
Bank #
IP2 TM SYSC296-1 Modified Bank #
New (Note changes or attach parent)
Question History:
u Question Cognitive Level:
Memory or Fundamental Knowledge
~-
-~
~
Page 120 of 200 Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (7)
X 55.43 Comments:
Page 191 of 200 Examination Outline Cross-reference:
1 Group #
1 WA #
000054 AA1.02 Importance Rating 4.4 4.4 Ability to operate and / or monitor the manual startup of electric and steam-driven AFW pumps as they apply to the Loss of Main Feedwater Proposed Question:
Common #34 Given the following conditions:
0 15% Reactor Power 0
0 0
22 MBFP secured 0
Main Generator Breakers 7 and 9 CLOSED 21 MBFP in service supplying all required feedwater 21,22 and 23 ABFPs aligned for normal at power operations A loss of 21 MBFP occurs and the Team trips the Reactor as directed in 2-AOP-FW-1, Loss of Feedwater.
Which one of the below indicates the automatic response of the Aux Feedwater Pumps following the Reactor Trip?
A.
21, 22 and 23 ABFPs secured B.
21,22, and 23 ABFPs running C.
21 and 23 ABFPs running, 22 ABFP secured D.
21 and 23 ABFPs secured, 22 ABFP running Proposed Answer:
C.
21 and 23 ABFPs running, 22 ABFP secured Explanation (Optional):
Power level to low to cause SG levels to decrease to start AFW pumps from AMSAC or normal low water level. Only start signal is MBFP trip which starts only the motor drive pumps only.
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-210-378 (As available)
Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 Comments:
Page 19 of 200 Examination Outline Cross-reference:
-J 13 Tier #
1 Group ##
1 WA #
000055 EK3.02 Importance Rating 4.3 4.6 Knowledge of the reasons for the actions contained in EOP for loss of offsite and onsite power Proposed Question:
Common #35 The team is implementing ECA-0.0, Loss of All AC Power.
Under which one of the following conditions must the team manually initiate SI?
A.
B.
C.
If no Emergency Diesel Generator is running and SI will be reset to facilitate restoring equipment upon power restoration.
Under all conditions in preparation for RCP seal failure and SI will not be reset until transition from ECA-0.0.
Only if the automatic SI signal failed to actuate and SI will be reset after power is restored to at least one (1) 480V bus.
Upon transition from ECA-0.0 to ECA-0.1 and SI will not be reset until directed in E-1 after transition from ECA-0.1, Loss of All AC Power Recovery Without SI Required.
Proposed Answer:
A.
if no Emergency Diesel Generator is running and SI will be reset to facilitate restoring equipment upon power restoration.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
ECA-0.0 Step 5, 6 Caution ECA-0.0 Background Page 110 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
u Question Source:
EOP-C-036-3554 (As available)
Bank #
INPO 24038 Modified Bank #
YES (Note changes or attach parent)
New
Page 20 of 200 Question History:
Salem Unit 1 5/05/2003 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41
@)(lo) 55.43 (5)
Comments:
Answer and distracters completely changed
'L.
Page 193 of 200 Examination Outline Cross-reference:
Level Tier #
WA #
117 Group #
2 W/E03 EK1.l Importance Rating 3.4 4.0 Knowledge of the operational implications of the components, capacity, and function of emergency systems.
Proposed Question:
Common #36 A post-LOCA cooldown and depressurization is in progress following a small-break LOCA, and the following conditions exist:
0 Safety Injection pumps 21 and 22 are running.
0 RCS subcooling is 56°F Tc is550"F.
PZR level is below the indicating range.
Reactor coolant pumps are operating.
The operating crew is on step 9 of ES-1.2, Post-LOCA Cooldown and Depressurization, which directs them to refill the PZR by depressurizing the RCS using normal spray.
Why will depressurizing the RCS refill the PZR?
A.
The water in the RCS will expand at lower pressure, forcing water into the PZR.
B.
The lower RCS pressure will increase SI flow, refilling the PZR.
C.
Voiding throughout the RCS will displace water into the PZR D.
Accumulator injection will force water into the PZR.
Proposed Answer:
B.
The lower RCS pressure will increase SI flow, refilling the PZR.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
ES-1.2 Background Proposed References to be provided to applicants during examination:
NONE Learning Objective:
EOP-C-012-3347 (As available)
~
~
Page 194 of 200 Question Source:
Bank #
Modified Bank #
New Yes (Note changes or attach parent)
Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Memory or Fundamental Knowledge Comprehension or Analysis 55.41 (8) 55.43 Comments:
New Question for Question 3#6
Page 45 of 200 Examination Outline Cross-reference:
Level Tier #
32 Group #
WA #
1 004 K3.08 Importance Rating 3.6 3.8 Knowledge of the effect that a loss or malfunction of the CVCS will have on RCP Seal Injection Proposed Question:
Common #37 The following plant conditions exist on Unit 2:
0 0
Unit 2 is at 90% power.
All systems are in normal at power condition.
The instrument air line to operating charging pump has just broken loose.
How will charging flow and seal injection flow respond?
Charging Flow RCP Seal Inj. Flow A
Increases Decreases B
Decreases Increases C
Increases Increases D
Decreases Decreases Proposed Answer:
C Increases Increases Explanation (Optional):
Charging pumps fails to maximum speed on loss of Instrument Air Tech n ical Ref e rence( s) :
(Attach if not previously provided)
Page 27 2-A0 P-AI R-1 SOD-18 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 22965 Modified Bank #
New (Note changes or attach parent)
Page 46 of 200 Question History:
Prairie Island 2 8/16/2002
'v-Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7) 55.43 Comments:
Question and Answer modified to make plant specific
Page 47 of 200 Examination Outline Cross-reference:
Level RO SRO 33 OK per TF Tier #
2 Group #
1 WA #
005 K2.03 Importance Rating 2.7 2.8 Knowledge of the bus power supplies to the RCS pressure boundary motor-operated valves Proposed Question:
Common #38 A loss of 480V Bus 6A has just occurred.
What is the effect on the power supplies of normally de-energized RHR Heat Loop Inlet Stop valves, MOV-730 and MOV-731 A.
MOV-730 and MOV-731 power unavailable B.
MOV-730 and MOV-731 power available C.
MOV-730 power unavailable and MOV-731 power available D.
MOV-730 power available and MOV-731 power unavailable Proposed Answer:
D.
MOV-730 power available and MOV-731 power unavailable Explanation (Optional):
MOV-730 powered from MCC-26A - Bus 5A and MOV-731 powered from MCC-26B - Bus 6A Technical Reference(s):
(Attach if not previously provided)
SOP-4.2.1 Pages 4,5 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
Question Source:
SYS-C-042 - 118 (As available)
Bank #
Modified Bank #
(Note changes or attach parent)
New YES Question History:
Page 48 of 200 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7) 55.43 (1)
Comments:
730/731 interlock only prevents opening if pressure is >450 psig c
C v
Page 195 of 200 Examination Outline Cross-reference:
Level Tier #
118 Group #
WA #
1 006000 K4.14 Importance Rating 3.9 4.2 Knowledge of ECCS design features(s) and/or interlock(s) which provide for Cross-Connection of HPI/LPI/SIP Proposed Question:
Common #39 Which valve(s) must the operator OPEN when aligning the SI System for Cold Leg Recirculation due to inadequate Low Head Flow?
A.
SI Pump Suction from RWST Isolation Valve MOV-1810
- 6.
RHR Heat Exchanger Outlet to SI Pumps Valves MOV-888A,/B C.
SI Cold Leg Injection Valves MOV-856A/C/D/E D.
Mini Flow Isolation Valves MOV-842 and 843 Proposed Answer:
- 8.
RHR Heat Exchanger Outlet to SI Pumps Valves MOV-888A and 888B Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
ES-1.3 Step 26 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYSD-C-101-147 (As available)
Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
Page 196 of 200 10 CFR Part 55 Content:
55.41 (7) 55.43 Comments:
New Question to replace #39 c
Page 51 of 200 Examination Outline Cross-reference:
Level 38 Tier #
Group #
K/A #
1 039 A2.05 Importance Rating 3.3 Ability to predict the impacts of Increasing steam demand, its relationship to increases in reactor power operation on the MRSS Proposed Question:
Common #40 The Unit is operating at 80% power EOL with all systems in automatic. One Group of condenser steam dump valves fail full OPEN.
Assuming that NO operator action occurs, what will be the approximate Rx power level 5 minutes after the valves fail open?
A.
0%
B.
70%
C.
80%
D.
90%
Proposed Answer:
D.
90%
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Page 52 of 200 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (5) 55.43 (5)
Comments:
Page 67 of 200 Examination 0 utli n e Cross-reference:
2 Group #
1 WA #
003 Al.10 Importance Rating 2.5 2.7 Ability to predict and/or monitor changes in parameters associated with operating the controls including RCP standpipe levels Proposed Question:
Common #41 The Unit is operating at 100% power.
Which one of the following will cause RCP Standpipe level(@ to rise?
A.
Failure of #1 Seal.
B.
Failure of #3 Seal.
C.
MOV-222, RCP Seal Return Isolation Stop Valve, fails closed.
D.
HCV-142, Charging Line Flow Control Valve, fails closed.
Proposed Answer:
A.
Failure of #1 Seal.
Explanation (Optional):
Technical Reference(@:
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-S-013-31/41 (As available)
Question Source:
Bank #
INPO 23064 Modified Bank #
New (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis
Page 68 of 200 10 CFR Part 55 Content:
55.41 (3)(5) 55.43 Comments: Question Distracters modified to make plant specific
Page 71 of 200 Examination Outline Cross-reference:
Level Tier #
48 Group #
WA #
1 007 Al.01
~~
Importance Rating 2.9 3.1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with maintaining quench tank water level within limits Proposed Question:
Common #42 Which ONE of the following describes the adverse affects, if any, of NO operator action with a leaking pressurizer PORV?
A.
There are NO adverse affects. The PRT is designed to handle continuous in-leakage.
B.
The cyclic temperature stresses in combination with inner wall erosion on the PORV tailpipe may lead to premature piping failure C.
Mechanical breakdown of the PORV seating surface may cause the PORV to fail when needed for overpressure protection.
D.
The PRT rupture disc may break with subsequent elevated radiation, temperature and pressure indications in containment.
Proposed Answer:
D.
The PRT rupture disc may break with subsequent elevated radiation, temperature and pressure indications in containment.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-O14-49/62 (As available)
Question Source:
Bank #
INPO 22895 Modified Bank #
New (Note changes or attach parent)
Question History:
Cook 1 12/9/2002
1-Page 72 of 200 Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (3, 14) 55.43 Comments:
Page 79 of 200 Examination Outline Cross-reference:
Level Tier #
53 Group ##
WA #
1 013 K2.01 Importance Rating 3.6 3.8 Knowledge of the bus power supplies to the ESFAS/Safeguards equipment control.
Proposed Question:
Given the following:
Common #43 0
0 0
A Steam Generator Tube Rupture occurred and the team has completed all steps through Step 16 of E-3, Steam Generator Tube Rupture.
Offsite power is lost during preparations to depressurize the RCS to minimize break flow per Step 17 of E-3.
The EDGs automatically supply power to all 480V buses.
Fifteen seconds later, the Watch Engineer observes that the SI pumps are no longer running. They were running before the loss of offsite power.
Should the SI pumps have restarted automatically by this point in time and why?
A.
Yes. The Safety Injection timers should have started the SI pumps.
B.
Yes. The Blackout timers should have started the SI pumps.
C.
No. The Safety Injection timers did not actuate because the SI signal has been reset.
D.
No. The Safety Injection timers did not actuate because they did not have sufficient time to start the pumps.
Proposed Answer:
C.
No. The Safety Injection timers did not actuate because the SI signal has been reset.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Caution prior to step 9 E-3 E-3 Backaround Paae 80 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 22853
~ _ _ _
Page 80 of 200 Modified Bank #
YES (Note changes or attach parent)
New Question History:
Cook 1 12/09/2002 Question Coanitive Level:
Memory or Fundamental Knowledae Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (3, 5) 55.43 (5)
Comments:
Question, answer, and distracters modified to make plant specific
Page 81 of 200 Examination Outline Cross-reference:
Level 54 Tier #
Group #
KIA #
1 022 A4.04 Importance Rating 3.1 3.2 Ability to manually operate and/or monitor in the control room valves in the CCS.
Proposed Question:
Common #44 Given the following conditions:
e 0
The plant is operating at 100% power.
21, 23, and 25 FCUs are in service to provide Containment Cooling.
Subsequently, reactor trip and Loss of Off-Site power occur. All equipment functions as designed.
Which one of the following describes the resulting Containment Cooling lineup?
A.
FCUs must be started manually. Cooling water flow is raised by providing a Service Water flow path parallel to TCV-1103, CNTMT BLDG Air Temperature controller..
B.
FCUs must be started manually. Cooling water flow is maintained by TCV-1103.
1--
C.
Only 21, 23, and 25 FCUs will be in service. Cooling water flow is raised by providing a Service Water flow path parallel to TCV-1103.
D.
All FCUs will be in service. Cooling water flow is maintained by TCV-1103 Proposed Answer:
B.
FCUs must be started manually. Cooling water flow is maintained by TCV-1103.
Explanation (Optional):
A. Incorrect. Parallel flow path only provided in safeguards mode B. Correct. All previously running fans will restart C. Incorrect. All fans start only in safeguards mode and no parallel flow path is provided D. Incorrect. All fans start only in safeguards mode.
Technical Reference(s):
(Attach if not previously provided)
~,
\\-.--
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-102-161 (As available)
Page 82 of 200 Question Source:
Bank #
INPO 23215 IP2
\\--
Modified Bank #
New (Note changes or attach parent)
Question History:
Indian Point 2 (Unit) 3/10/2003 Question Coanitive Level:
Memorv or Fundamental Knowledae Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (4) 55.43 Comments:
Page 23 of 200 Examination Outline Cross-reference:
Level Tier #
15 Group #
WA #
1 000057 AK3.01 Importance Rating 4.1 4.1 Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus Actions contained in EOP for loss of vital AC electrical instrument bus Proposed Question:
Common #45 Given the following conditions:
0 The plant is at 80% power.
0 A loss of Instrument Bus 21 has occurred.
Which one of the following statements describes why the HI-HI Containment Pressure relays are blocked when performing the appropriate attachment in accordance with 2AOP-IB-1, Loss of Instrument Bus?
A.
Blocks inadvertent actuation of Containment Spray in the case of a redundant channel failure B.
Provides a channel trip of Containment Spray to change the coincidence to 1 out of 3 for Spray actuation C.
Makes up part of the coincidence circuitry for Spray initiation, since Containment Spray relays are energized to actuate D.
Blocks the actuation signal from the channel supplied from the de-energized instrument bus from causing an inadvertent Phase B containment isolation signal Proposed Answer:
C.
Makes up part of the coincidence circuitry for Spray initiation, since Containment Spray relays are energized to actuate Explanation (Optional):
A-Incorrect. Makes up part of trip coincidence B-Incorrect. Logic does not change C-Correct D-Incorrect. Channel is energize to actuate Technical Reference(s):
(Attach if not previously provided) 2AOP-l B-1
Page 24 of 200 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 23258 Modified Bank #
New (Note changes or attach parent)
Question History:
Indian Point Unit 2 3/10/2003 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (5)(10) 55.43 Comments:
Page 53 of 200 Examination Outline Cross-reference:
2 39 Group #
1 OK per TF WA #
059 K3.04 Importance Rating 3.6 3.8 Knowledge of the effect that a loss or malfunction of the MFW will have of the RCS Proposed Question:
Common #46 Given the following conditions:
83% Reactor power.
0 0
Both Main Feedwater pumps are operating in AUTOMATIC.
Steam Generator Water Level Controls are in AUTOMATIC.
Which ONE of the following failures will cause RCS TAVE to INITIALLY INCREASE?
A.
21 SG Level Channel, 417B, fails HIGH B.
21 Steam Flow Channel, 419B, (CONTROLLING) Fails HIGH C.
21 Feed Flow Channel, 41 8B, (CONTROLLING) Fails LOW D.
21 Steam Pressure Channel, 419C, Fails LOW Proposed Answer:
A.
21 SG Level Channel, 4178, fails HIGH Explanation (Optional):
A.
Feedflow decreases due to high SG level, Tcold 21 loop increases, Tave increases B. Feedflow increases to match SF, Tcold decreases, Tave decreases C. Feedflow increases to match SF, Tcold decreases, Tave decreases D. Atmos 21 SG opens, SF increases, Tcold decreases, Tave decreases Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-012-27 SYS-C-2 1 1 -3983 SGOPSAOP8-1601066 (As available)
Question Source:
Bank #
~
-~ __
~~
~
~~
Page 54 of 200 Modified Bank #
New YES (Note changes or attach parent)
\\
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7, 14) 55.43 Comments:
New Question
Page 73 of 200 Examination Outline Cross-reference:
Level Tier #
.V' 49 Group #
WA #
1 006 K6.18 Importance Rating 3.6 3.9 Knowledge of the effect that a loss or malfunction of the ECCS will have on Subcooling Margin Indicators Proposed Question:
Common #47 A small break LOCA has occurred on Unit 2.
ES-1.2, Post LOCA Cooldown and Depressurization, is in progress.
Safety Injection Pump 21 has just been stopped in accordance with Step 11.
Safety Injection Pumps 22 and 23 are running.
The following conditions are noted:
Subcooling Monitor, Channel A 75°F and stable Subcooling Monitor, Channel B 10°F and stable Core Exit Thermocouple avg 570°F Containment pressure 5 psig Containment rad levels 4 Whr RCS Wide Range Pressure 1335 psig After comparing the subcooling readings with RCS pressure and CETs, the team will determine that:
A.
B.
C.
D.
Subcooling Monitor Channel A is reading inaccurately, Safety Injection Pump 21 will be started to restore subcooling.
Subcooling Monitor Channel B is reading inaccurately, Safety Injection Pump 22 or 23 will be stopped since adequate subcooling exists.
Subcooling Monitor Channel A is reading accurately, Safety Injection Pump 21 will NOT be started, the team will continue in ES-1.2.
Subcooling Monitor Channel B is reading accurately, Safety Injection Pump 22 or 23 will be stopped since adequate subcooling exists.
Proposed Answer:
A.
Subcooling Monitor Channel A is reading inaccurately, Safety Injection Pump 21 will be started to restore subcooling.
Explanation (Optional):
Page 74 of 200 Tech n ica I Ref e rence( s) :
(Attach if not previously provided)
ES-1.2 Proposed References to be provided to applicants during examination:
Steam Tables Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (1 0,14) 55.43 (5)
Comments:
Page 85 of 200 Examination Outline Cross-reference:
Level RO SRO 56 Rev per TF Tier #
2 Group #
1 WA #
076 A4.01 Importance Rating 2.9 2.9 Ability to manually operate andor monitor in the control room SWS Pumps Proposed Question:
Common #48 Given the following plant conditions:
0 Unit 2 at 100%
0 0
0 24/25/26 Essential Service Water Header Three Header Service Water System Operation Operating Service Water Pumps prior to Rx Trip
- a. 21
- b. 22 powered from 2A
- c. 25 powered from 3A
- d. 26 From the list below, determine the service water pump combination for 22 and 25 Service Water Pumps that would result following a Reactor Trip with a Station Blackout:
SWP A.
B.
C.
D.
22 25 Running - 2A Running - 2A Stopped Running -2A Running -2A Running -3A Stopped Running -3A Proposed Answer:
B.
Stopped Running -2A Explanation (Optional):
A.
B.
C.
D.
No Non-essential SWP receives a start signal Correct - 25 SWP receives start signal first for bus 2A 25 SWP will only start on bus 3A if 2A start failed No Non-essential SWP receives a start signal and 25 SWP will only start on bus 3A if 2A start failed
Page 86 of 200 Technical Reference(s):
(Attach if not previously provided)
SD-10.0 Pages 25,26 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-240-393 SYS-C-271 A-2832 (As available)
Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (4, 7, 8) 55.43 Comments:
Page 87 of 200 Examination Outline Cross-reference:
Level Tier #
57 Group #
WA #
1 076 K1.05 Importance Rating 3.8 4.0 Knowledge of the physical connections and/or cause-effect relationships between the SWS and the D/G Proposed Question:
Common #49 What are three signals that will cause the EDG service water valves FCV-1176 and FCV-1176A to open fully?
A.
B.
C.
D.
High Lube Oil Temp, High Casing Temp and Safety Injection High Lube Oil Temp, High Jacket Water Temp and Phase A High Jacket Temp, High Casing Temp and Phase A High Lube Oil Temp, High Jacket Water Temp and Safety injection Proposed Answer:
D.
High Lube Oil Temp, High Jacket Water Temp and Safety Injection Explanation (Optional):
Technical Ref e rence( s):
(Attach if not previously provided)
Windows 1-3, 1-8 E-0 Page 25 2-A R P-003 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-240-399d (As available)
Question Source:
Bank #
I P2 SYSC240-2 Modified Bank #
New (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
Page 88 of 200 10 CFR Part 55 Content:
55.41 (4,,7)
Comments:
Page 95 of 200 Examination Outline Cross-reference:
Level Tier #
61 Group #
WA #
1 063 K3.02 Importance Rating 3.5 3.7 Knowledge of the effect that a loss or malfunction of the DC Electrical System will have on the following: Components using dc control power.
Proposed Question:
Common #50 A loss of which ONE of the following DC panels will result in the majority of the Control Room Supervisory Alarm Panels losing power?
A.
21 DC Power Panel B.
22 DC Power Panel C.
23 DC Power Panel D.
24 DC Power Panel Proposed Answer:
D.
24 DC Power Panel Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Pages 25,33 2-A0 P-AN N UN-1 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
A 0 P-AN N U N-27 1 4 1 (As available)
Question Source:
Bank #
INPO 2091 2 Modified Bank #
YES (Note changes or attach parent)
New Question History:
Palisades 1 12/21 /2001 Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis
i..
Page 96 of 200 10 CFR Part 55 Content:
55.41 (7) 55.43 Comments:
Question, Answer and distracters changed to make plant specific
Page 181 of 200 Examination Outline Cross-reference:
Level Tier #
111 Group #
WA #
1 078 K3.02 Importance Rating 3.4 3.6 Knowledge of the effect that a loss or malfunction of the IAS will have on systems having pneumatic valves an controls Proposed Question:
Common #51 Given the following conditions:
0 Unit 2 is in Mode 3 0
21 and 23 ABFP maintaining SG levels 0
ABFP suction aligned to the CST 0
Nitrogen Backup in the Aux Feed Building is isolated A complete loss of instrument air to the City Water supply valves to the ABFPs occurs Which one of the following describes the suction valve alignment for 21 and 23 ABFPs?
A.
PCV-1187/1188, CW Supply to 21/23 ABFPs CLOSE and FVC-l205A, CW ABFPs OPEN 1
B.
PCV-1187/1188, CW Supply to 21/23 ABFPs OPEN and FVC-l205A, CW ABFPs CLOSE C.
PCV-1187/1188, CW Supply to 21/23 ABFPs CLOSE and FVC-l205A, CW ABFPs CLOSE D.
PCV-1187/1188, CW Supply to 21/23 ABFPs OPEN and FVC-1205A1 CW ABFPs OPEN Proposed Answer:
A.
PCV-l187/1188, CW Supply to 21/23 ABFPs CLOSE and FVC-l205A, CW ABFPs OPEN Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided) 2-AOP-AIR-1 Page 31 Proposed References to be provided to applicants during examination:
NONE
Page 182 of 200 Learning Objective :
SY S-C-2 92-474 AOP-C-AI R-29301 (As available)
Question Source:
Bank ##
Modified Bank ##
(Note changes or attach parent)
New YES Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (4, 5, 7) 55.43 Comments:
Page 27 of 200 Examination Outline Cross-reference:
Level 17 Tier #
Group #
WA #
1 000058 AA2.03 Importance Rating 3.5 3.9 Ability to determine and interpret DC loads lost; impact on ability to operate and monitor plant systems as they apply to the loss of DC Power Proposed Question:
Common #52 Which one of the following describes how a loss of 125VDC to the Reactor Trip Relays Train A affects the Reactor Trip Breaker, RTA?
A.
B.
The loss of voltage de-energizes the UV coil and the breaker OPENS.
The loss of 125VDC will prevent ALL trip signals to RTA.
C.
D.
The breaker is NOT capable of opening on a signal to the UV trip coil of RTA The loss of voltage de-energizes the shunt coil and the breaker OPENS.
Proposed Answer:
A.
The loss of 125VDC Train A relay power will cause a Reactor Trip.
Explanation (Optional):
125VDC power is used to power the shut trip coil for Reactor Trip Breakers. This is in addition to the Undervoltage Trip which is also on each reactor trip breaker. Power is not required to trip the UV portion of breaker, but power is supplied to energize the shunt trip coil as a backup to the UV trip.
Technical Reference(s):
(Attach if not previously provided) 2-AOP-DC-1 Page 125 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-161-277 SYS-C-2718-6454 Question Source:
Bank #
INPO 24061 Modified Bank #
New (Note changes or attach parent)
Page 28 of 200 Question History:
Salem Unit 1 5/5/2003 Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 55.43 (5)
Comments:
i-Page 153 of 200 Examination Outline Cross-reference:
Level Tier #
WA #
95 Group #
RO 3
2 G2.2.22 SRO Importance Rating 3.4 4.1 Knowledge of limiting conditions for operations and safety limits.
Proposed Question:
Common #53 Given the following plant conditions:
0 0
0 Unit 2 is at 100% power.
22 ABFP is tagged out for maintenance.
Engineering has just notified the Shift Manager that a common electrical problem has been discovered in the controllers for the motor driven ABFP flow control valves, FCV-406A, through FCV-406D The Shift Manager has declared 21 and 23 ABFPs inoperable.
0 In addition to initiating action to restore the flow control valves to service, which of the following statements describes an action required for Unit 2?
A.
A reactor shutdown to Mode 2 is required.
B.
A reactor shutdown to Mode 3 is required.
C.
A reactor shutdown to Mode 4 is required.
D.
Maintain stable plant conditions Proposed Answer:
D.
Maintain stable plant conditions Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
T.S. 3.7.5, Condition D Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-210-383 (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New YES Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 (2)
Comments:
New Question
Page 185 of 200 Examination Outline Cross-reference:
Level L
Tier #
KIA #
113 Group #
L-RO SRO 3
3 G2.3.11 Importance Rating 2.7 3.2 Ability to control radiation releases.
Proposed Question:
Common #54 Given the following information:
0 Plant is operating at 100% power R-46, FAN CLR UNIT SERVICE WTR HI RADTTROUBLE, is in alarm Increased activity from 22 FCU service water has been confirmed What action is required to be taken?
A. Isolate Service Water flow for 22 FCU and raise the R-46 High setpoint above existing reading to clear the alarm B. Stop all FCUs and isolate service water flow to 22 FCU to prevent the spread of contamination in containment C. Initiate Containment Ventilation Isolation to prevent an unmonitored release to the environment.
D. Initiate a Containment Purge and monitor a release of the Containment environment to the plant vent.
Proposed Answer:
A.
Isolate Service Water flow for 22 FCU and raise the R-46 High setpoint above existing reading to clear the alarm Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided) 2-ARP-SAF-1 Window 1-9 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-120-120120 (As available)
Question Source:
Bank #
c Modified Bank #
(Note changes or attach parent)
~.
Page 186 of 200 New YES L
Question History:
Question Cognitive Level:
Memory or Fundamental (nowlec Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (1 2) 55.43 Comments:
New Question to replace existing question #96 in this file (Test Question #54)
Page 155 of 200 Examination Outline Cross-reference:
3 Group #
3 WA #
G2.3.1 Importance Rating 2.6 3.0 Knowledge of 10 CFR: 20 and related facility radiation control requirements.
Proposed Question:
Common #55 During a plant emergency an operator receives a radiation exposure of 10 REM to the lenses of both eyes.
Regarding 10 CFR 20, "Standards for Protection Against Radiation" and Entergy's administrative radiation control limits, which, if any, of these limits have been exceeded?
A.
NEITHER of the exposure limits listed have been exceeded.
B.
BOTH 1 OCFR20 AND plant admin. limits have been exceeded.
C.
Plant quarterly administrative limits only, have been exceeded, but NOT 10CFR20 limits.
D.
Plant annual limits have been exceeded, but not 1 OCFR20 limits..
Proposed Answer:
A.
NEITHER of the exposure limits listed have been exceeded.
Explanation (Optional):
Entergy limits are 80% of 1 OCFR20 resulting in limits of 12 REM to the lens of the eyes Technical Reference(s):
(Attach if not previously provided) 1 OCFR20.1201 IP-SMM-RP-201 Pages 8 &9 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 26338 Modified Bank #
YES (Note changes or attach parent)
New
~.\\.'
Question History:
Palisades 1 8/01 12003 Question Cognitive Level:
Memory or Fundamental Knowledge X
Page 156 of 200 Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (12) 55.43 (4)
Comments:
Question and Answer modified to make plant specific
I Page 179 of 200 Exami nation Out I i ne Cross-reference:
Level Tier #
110 Group #
WA ##
1 008 K1.03 Importance Rating 2.8 3.3 Knowledge of the physical connections and / or cause-effect relationships between the CCWS PRMS Proposed Question:
Common #56 A Reactor Coolant Pump thermal barrier heat exchanger rupture occurred approximately 10 minutes ago. Given the following indications:
0 High Radiation alarm on CCW Radiation Monitor R-47 0
CCW Surge Tank level indicates OFFSCALE HIGH NPO reports the CCW Surge Tank Relief Valve is lifting periodically 0
FCV-625, RCP Thermal Barrier CCW Return Valve indicates OPEN MOV-789, RCP Thermal Barrier CCW Return Isolation Valve indicates OPEN RCV-017, CCW Surge Tank Vent Valve indicates CLOSED Determine which one of the following explains the given status of the CCW System? (Assume NO operator actions have been taken)
A.
RCV-017 has failed to automatically OPEN B.
MOV-789 has failed to automatically CLOSE C.
FCV-625 has failed to automatically CLOSE D.
CCW Surge Tank Level Transmitter has failed high Proposed Answer:
C.
FCV-625 has failed to automatically CLOSE Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
ARP SAF-1 Window 3-8 2-AOP-LICCW-1 Pages 7-1 7 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-041-102 (As available)
Page 180 of 200 Question Source:
Bank #
I P2 SYSCO41-15 Modified Bank #
New (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (3, 7) 55.43 Comments:
i--
Page 83 of 200 Examination Outline Cross-reference:
Level Tier #
55 Group #
WA #
1 062 K2.01 Knowledge of bus power supplies to the major system loads.
Proposed Question:
Common #57 Which one of the choices lists only equipment powered from 480V 3A?
3.3 3.4 A.
B, C.
D.
21 RHR Pump 21 Recirc Pump 22 Charging Pump 23 Fan Cooler Unit 21 Aux Feedwater Pump 21 RHR Pump 22 Charging Pump 24 Fan Cooler Unit 21 RHRPump 22 CCW Pump 24 Fan Cooler Unit MCC-211 25 SW Pump 22 Charging Pump 21 Aux Feedwater Pump MCC-24A Proposed Answer:
B, 21 Aux Feedwater Pump 21 RHRPump 22 Coolant Charging Pump 24 Fan Cooler Unit Explanation (Optional):
Technical Reference( s) :
(Attach if not previously provided)
Page 21 E-1, Attachment 1 Proposed References to be provided to applicants during examination:
NONE
Page 84 of 200 Learning Objective:
(As available)
Question Source:
Bank #
INPO 231 83 Modified Bank #
YES (Note changes or attach parent)
New Question History:
Salem Unit 1 1 1 /04/2002 Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (7) 55.43 Comments:
Question, answer and distracters modified to make plant specific
Page 99 of 200 Examination Outline Cross-reference:
Level L-64 Tier #
Group #
KIA #
1 103 A3.01 Importance Rating 3.9 4.2 Ability to monitor automatic operation of the containment systems including containment isolation Proposed Question:
Common #58 An inadvertent Safety Injection Actuation and automatic Reactor Trip has occurred. It was noted during the performance of E-0, Reactor Trip or Safety Injection, that Letdown Isolation Valve, 201 failed to automatically close as required, and had to be manually closed. SI has subsequently been placed in DEFEAT and has been RESET in ES-1.1, SI Termination.
However, Containment Isolation (CIA) Phase A could NOT be reset when attempted.
Which one of the following could be a cause for the failure of CIA, Phase A, to reset?
A.
Letdown Isolation Valve, 201 failed to AUTOMATICALLY close as required.
B.
Control switch for the Weld Channel & Penetration Pressurization System (WCPPS) is OPEN.
C.
Equipment Hatch Solenoid control switch is in NORM.
\\ -
D.
Isolation Valve Seal Water System valves (1 41 0, 141 3, SOV-3518, and SOV-3519) control switches are OPEN.
Proposed Answer:
C.
Equipment Hatch Solenoid control switch is in NORM.
Explanation (Optional):
A. Daisy Chain made up from switch position not valve position B. Control switch for WCPPS not Phase A valves C. Correct, switch needs to be in INCIDENT to make up Daisy Chain D. IVSWS valves switches should be open for Daisy Chain Technical Reference(@:
(Attach if not previously provided)
Page 29 E-0, Attachment 1, step 12 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-107-107108 (As available)
Page 100 of 200 Question Source:
Bank #
I P2 SYSC107-29 Modified Bank #
New (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7,9) 55.43 Comments:
Page 105 of 200 Examination Outline Cross-reference:
..-/
Importance Rating Common #59 Unit 2 is at 20% p Control rods are in A.
Power Range Nuclear exceeded B.
Intermediate Range Nuclear setpoint (1/4) exceeded
/4) exceeded, TAVE - Avg TAVE (+5"F)
Proposed Answer:
ARP SAF Windows 1-8, 2-8 ARP FCF Windows 1-2, 1-3, 1-4 1
Proposed References to be provided to applicants during examination:
NONE
Page 106 of 200 Learning Objective:
SYS-(2-161-289 (As available)
Question Source:
Bank #
INPO 22962 Modified Bank #
YES (Note changes or attach parent)
New Question History:
Prairie Island 2 8/16/2002 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (6,7) 55.43 Comments:
Question, answer and distracters modified
Page 109 of ZOO Examination Outline Cross-reference:
Level Tier #
70 Group #
WA #
2 033 K4.03 Importance Rating 2.6 2.9 Knowledge of design features(@ and/or interlock(s) which provide for anti-siphon devices Proposed Question:
Common #60 Which one of the following statements describes a design feature that prevents excessive loss of level in of the spent fuel pool through the spent fuel pool cooling (SPFC) System?
A.
SPFC pumps will automatically trip when the low SFP level alarm is annunciated.
B.
SFPC discharge piping has a siphon breaker slightly below the normal water level.
C.
Deepest SFPC piping extends only 6 feet down into the SFP.
D.
Primary makeup valve to the SFP automatically opens on a low level in the SFP.
Proposed Answer:
B.
Explanation (Optional):
SFPC discharge piping has a siphon breaker slightly below the normal water level.
gw e% [ deLaT\\.
A. There is no auto trip from low level for the SFP Pumps B. Correct C. Discharge piping extends to within 54 of top of the spent fuel racks D. Primary makeup to SFP is manual d/ C. ~\\s;&?&~EK C &$o COFfQ&,
p a - +&&(;cy CCWmO Technical Reference(s):
(Attach if not previously provided)
SOD - 004 SD-4.3 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SY S-C-043-2630 (As available)
Question Source:
Bank #
INPO 251 37 Modified Bank #
YES (Note changes or attach parent)
New Question History:
Millstone 2 12/04/2002 Question Cognitive Level:
Memory or Fundamental Knowledge
~
~
Page 11 0 of 200 Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (7) 55.43 X
Comments:
Question, Answer and distracters modified to make plant specific
Page 11 1 of 200 Examination 0 u tl i n e Cross-reference:
2 Group #
2 Importance Rating 3.4 3.9 WA #
035 K5.01 Knowledge of operational implications of the effect of secondary parameters, pressure, and temperature on reactivity Proposed Question:
Common #61 Given the following plant conditions:
0 0
Turbine load is at 100%.
CD-19, "23/24/25, Feedwater Heater Bypass Valve" is inadvertently opened.
How and why will reactor power respond to this condition?
A.
Reactor power will decrease for a very short time due to less steam leaving the turbine extraction lines.
B.
Reactor power will decrease due to the decrease in steam generator pressure caused by the colder feedwater entering the feed ring.
C.
Reactor power will increase due to the colder water entering the steam generators.
D.
Reactor power will increase for a very short time due to MTC adding negative reactivity, but then decrease as the steam generator pressure increases.
Proposed Answer:
C.
Reactor power will increase due to the colder water entering the steam generators causing TC~LD to drop; MTC will add positive reactivity.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 20633 Modified Bank #
(Note changes or attach parent)
New Question History:
Point Beach 1 2/02/2002 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (5, 14) 55.43 Comments:
Minor modifications to question and distracters to make plant specific
L Page 105 of 243 Examination Outline Cross-reference:
Level Tier #
51 Group #
WA #
1 010 K6.01 Importance Rating 2.7 3.1 Knowledge of the effect that a loss or malfunction of the Pressure Detection systems will have on the PZR PCS Proposed Question:
Common #62 The unit is at 100% power, steady state, normal operating temperature and pressure. The Pressurizer Pressure Master Controller setpoint fails to 21 85 psig. Assume a step change in the setpoint and assume that pressurizer pressure control remains in automatic.
Which of the following is the immediate automatic response of the system?
A.
PORV 455C opens, Spray valves open, Variable Heaters energize.
B.
Spray valves open, Variable Heaters energize.
C.
Spray valves close, Variable Heaters de-energize.
D.
Spray valves open, Variable Heaters de-energize Proposed Answer:
D.
Spray valves open, Variable Heaters de-energize Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-(2-014-55 SGOPSAOP8-1601066 (As available)
Question Source:
Bank #
INPO 191 89 Modified Bank #
New (Note changes or attach parent)
Question History:
Braidwood 1 10/20/2000
Page 106 of 243 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (3,7) 55.43 Comments:
Question modified to change correct answer
Page 11 3 of 200 i.:
Examination Outline Cross-reference:
Level Tier #
WA #
72 Group #
2 041 G2.1.10 Importance Rating 2.7 3.9 Knowledge of conditions and limitations in the facility license.
Proposed Question:
Common #63 Which one of the following is considered to be the most limiting event (time critical) concerning operation of the Atmospheric Dump Valves?
A.
B.
C.
D.
Inadequate Core Cooling accident with off site power available.
Main Steam Line Break accident inside containment with a loss of off site power.
Small Break Loss of Coolant accident with off site power available.
Steam Generator Tube Rupture accident with a loss of off site power.
Proposed Answer:
D.
Steam generator Tube Rupture accident with a loss of off site power.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
TAA-C-009-T0325, E02623 (As available)
Question Source:
Bank #
INPO 261 86 Modified Bank #
New (Note changes or attach parent)
Question History:
Point Beach 1 9/29/2003 Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analvsis 10 CFR Part 55 Content:
55.41 (43)
~
~
Page 11 4 of 200 55.43 (1)
Comments:
Minor modification to disctracter
Page 115 of 200 Examination Outline Cross-reference:
Level Tier #
73 Group #
WA #
2 045 Al.06 Importance Rating 3.3 3.7 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MT/G system controls including expected response of secondary plant parameters following T/G trip Proposed Question:
Common #64 The following plant conditions exist:
0 0
The plant is operating at 100%.
All systems are lined up in their normal lineups.
All control systems are in automatic.
A fault signal occurs from MO Disconnect Switch F7-9.
Which one of the following describes the immediate plant response?
A.
S/G pressure initially increases as main turbine is lost, S/G levels initially decrease due to shrink, feed flow initially increases.
B.
S/G pressure initially increases as main turbine is lost, S/G levels initially decrease due to shrink, feed flow initially decreases.
C.
S/G pressure initially decreases as main turbine is lost, S/G levels initially increase due to lower steam pressure, feed flow initially decreases.
D.
S/G pressure initially decreases as main turbine is lost, S/G levels initially decrease due to shrink, feed flow initially increases.
Proposed Answer:
B.
S/G pressure initially increases as main turbine is lost, S/G levels initially decrease due to shrink, feed flow initially decreases.
Explanation (Optional):
D - correct - initially (prior to steam dumps opening), steam header pressure increases due to the loss of steam demand. The increased back-pressure in the S/Gs partially suppresses boiling which causes shrink to occur in the S/Gs. The reduced steam demand inputs to the feed controller to reduce feed thus feed flow decreases.
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE
Page 116 of 200 Learning Objective:
TAA-C-005-2503 (As available)
Question Source:
Bank #
INPO 24696 Modified Bank #
New (Note changes or attach parent)
Question History:
Seabrook 1 5/30/2003 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (5,14) 55.43 Comments:
Page 11 7 of 200 Examination Outline Cross-reference:
Level Tier #
WA #
74 Group #
2 075 G2.1.8 Importance Rating 3.8 3.6 Ability to coordinate personnel activities outside the control room Proposed Question:
Common #65 Switching CW Pump speed from HIGH to LOW or LOW to High requires which one of the following items?
A.
An operator imposed 20 second delay when switching from high to LOW B.
C.
The pump to be stopped prior to switching speeds No delay when switching from LOW to HIGH D.
The pump recirculation valve to be opened Proposed Answer:
- 6.
The pump to be stopped prior to switching speeds Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
SOP-23.1 Page 1 ProDosed References to be provided to awlicants durina examination:
NONE Learning Objective:
SY S-C-240-230 1 09 (As available)
Question Source:
Bank #
I P2 SYSC230-6 Modified Bank #
New (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
Page 1 18 of 200 10 CFR Part 55 Content:
55.41 (7, 10) 55.43 Comments:
Page 97 of 200 Examination Outline Cross-reference:
2 Group #
1 WA #
064 G2.1.28 Importance Rating 3.4 3.3 Knowledge of the purpose and function of the major system components and controls Proposed Question:
Common #66 Select the ONE answer that describes EDG governor and voltage control with the EDG paralleled to the BUS and the UNIT-PARALLEL switch in PARALLEL.
A.
Increasing the governor will INCREASE bus frequency and Increasing the voltage rheostat will INCREASE bus voltage.
B.
Increasing the governor will INCREASE EDG load and Increasing the voltage rheostat will INCREASE bus voltage.
C.
Increasing the governor will INCREASE EDG load and Increasing the voltage rheostat will INCREASE lagging VARs.
D.
Increasing the governor will INCREASE bus frequency and Increasing the voltage rheostat will INCREASE leading VARs.
Proposed Answer:
C.
Increasing the governor will INCREASE EDG load and Increasing the voltage rheostat will INCREASE lagging VARs.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-(2-273-2790 (As available)
Question Source:
Bank #
Taskmaster SYSC273-3 Modified Bank #
New (Note changes or attach parent)
Question History:
Page 98 of 200 Question Cognitive Level:
10 CFR Part 55 Content:
Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 (8) 55.43 Comments:
Page 125 of 200 Examination Outline Cross-reference:
Group #
WA #
2 1
073 K1.01
~
Importance Rating 3.6 3.9 Knowledge of the physical connections and/or cause-eff ect relationships between the PRM system and those systems served by PRMs Proposed Question:
Common #67 What conditions must be met to reset a Containment Ventilation Isolation after a high containment air particulate or radiogas alarm has isolated the Containment Purge and Containment Pressure Relief lines in accordance with 2-SOP-5.4.3, Vapor Containment Purge?
A.
The containment air particulate and radiogas alarms, R41/42, are the only signals that must be clear prior to resetting the Containment Ventilation Isolation.
B.
Containment Phase A and R41/42 and R44 must be below the isolation setpoint prior to resetting the Containment Isolation.
C.
Safety Injection must be reset and R41/42 and R44 must be below the isolation setpoint prior to resetting the Containment Isolation.
D.
No conditions need to be met, the Containment Purge and Pressure Relief Lines do not close on a high containment air particulate or radiogas alarm.
Proposed Answer:
C.
Safety Injection must be reset and R41/42 and R44 must be below the isolation setpoint 6 Explanation (Optional): & T\\IRc @50(u$-)ifl Technical Reference(s):
(Attach if not previously provided) 2-SOP-5.4.3 Page 23 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-120-120112 (As available)
Question Source:
Bank #
I P2 sYsc120-20 Modified Bank #
New (Note changes or attach parent)
Question History:
Page 126 of 200 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7,8,9,
- 11) 55.43 Comments:
Page 147 of 200 Examination Outline Cross-reference:
3 Group #
1 Importance Rating 3.4 3.3 KIA #
G2.1.29 Knowledge of how to conduct and verify valve lineups.
Proposed Question:
Common #68 During an independent verification a valve is found out of position. How is the verifier to handle the component out of position situation in accordance with OAP-19, Component Verification and System Status Control?
A.
Do NOT change valve position. Notify the Shift Manager of the discrepancy.
B.
Do NOT change valve position. Notify the initial valve positioner of the discrepancy.
C.
Correct the valve position. Have Shift Manager obtain new verifier for independent verification for that valve only.
D.
Correct the valve position. Have the initial valve positioner perform the independent verification for that valve only.
Proposed Answer:
A.
Do NOT change valve position. Notify the Shift Manager of the discrepancy.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
OAP-019 Paae 6
\\-
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 261 35 Modified Bank #
New (Note changes or attach parent)
Question History:
Prairie Island 1 10/3/2003 Question Cognitive Level:
Memory or Fundamental Knowledge X
~
~
~
Page 148 of 200 Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (1 0) 55.43 Comments:
Page 123 of 200 Examination Outline Cross-reference:
Level Tier #
77 Group #
WA #
1 064 A4.01 Importance Rating 4.0 4.3 Ability to manually operate and/or monitor in the control room local and remote operation of the ED/G Proposed Question:
Common #69 Determine the affect on Emergency Diesel Generator operation if the Jacket Water Pump on an Emergency Diesel Generator had a broken shaft and the Emergency Diesel Generator received an AUTO start signal?
A.
The Emergency Diesel Generator would start and continue to run, but the field would not
'flash' so there would be no generator output B.
Without jacket water pressure the Emergency Diesel Generator would start, run for 2 minutes and shut down normally C.
The Emergency Diesel Generator would run until it overheated, then low oil pressure would trip the 86 device D.
The Emergency Diesel Generator would start but only run for about 37 seconds, then the 86 would trip Proposed Answer:
D.
The Emergency Diesel Generator would start but only run for about 37 seconds, then the 86 would trip Explanation (Optional):
A. Jacket water pressure is used to allow field flash but the EDG would S/D B. Normal shutdown would not occur with 86 trip C. 37 seconds is not long enough for EDG to overheat D. Correct - Overcrank is sensed from Jacket Water Pressure Technical Reference(s):
(Attach if not previously provided) 2-ARP-003 Window 1 -2 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SY S-C-273-2797/2798 (As available)
Question Source:
Bank #
I P2 27300301
Modified Bank #
New (Note changes or attach parent)
L-Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7,8) 55.43 Comments:
Minor modification to question
Page 149 of 200 Examination Outline Cross-reference:
Level Tier #
93 Group #
KIA #
1 G2.1.7 Importance Rating 3.7 4.4 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
Proposed Question:
Common #70 The following plant parameters exist:
0 RCS pressure is 1600 psig and lowering.
0 Pressurizer level is slowly lowering.
0 PORVs and spray valves are closed.
All steam generator water levels are normal 0
Plant ventilation radiation monitors are rising.
Containment pressure and sump levels are normal.
Which one of the following is the correct plant condition?
A.
Faulted Steam Generator B.
Ruptured Steam Generator C.
LOCA Inside Containment D.
LOCA Outside Containment Proposed Answer:
D.
LOCA Outside Containment Explanation (Optional):
Technical Ref e rence( s) :
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 20585 Modified Bank #
(Note changes or attach parent)
Page 150 of 200 New Question Question History:
Point Beach 1 2/2/2002 Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (5) 55.43 (5)
Comments:
Page 151 of 200 Examination Outline Cross-reference:
Level Tier ##
94 Group #
WA ##
2 G2.2.33 Importance Rating 2.5 2.9 Knowledge of control rod programming.
Proposed Question:
Common #71 Given the following conditions:
Reactor power is 90%
0 Control Bank D is at 200 steps Automatic rod control is selected Which ONE of the following statements describes the response o1 the rod control system i becomes 45°F more than TREF?
A.
- 6.
C.
D.
(Assume no power mismatch effects)
The control rods step in at 32 steps per minute.
The control rods step in at 48 steps per minute.
The control rods step in at 56 steps per minute.
TAVE deviation of more than 4°F will inhibit rod insertion.
TAVE Proposed Answer:
C.
The control rods step in at 56 steps per minute.
Explanation (Optional):
A. 32 Steps per minute is rods speed / O F error B. 48 steps per minute is rod speed for 15°F above dead band minus 8 stepdmin initial C. Correct D. TAVE to Average TAVE deviation blocks rod motion Technical Reference(+:
(Attach if not previously provided)
SD-I 6.1 Figure 27 I P2-SOD-022 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
SYS-C-161-280 (As available)
Question Source:
Bank #
I P2 SYSCl61-11
Page 152 of 200 Modified Bank #
(Note changes or attach parent)
New Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (6) 55.43 (6)
Comments:
Page 157 of 200 Examination Outline Cross-reference:
Level Tier #
98 Group #
WA #
4 G2.4.16 Importance Rating 3.0 4.0 Knowledge of EOP implementation hierarchy and coordination with other support procedures.
Proposed Question:
Common #72 A reactor trip has occurred. During the CRS read-through of E-0, Reactor Trip Or Safety Injection, Step 3, an Orange Path condition is observed by the Watch Engineer to exist on a Critical Safety Function (CSF) Status Tree.
Transition to the Orange Path procedure is to take place:
A.
immediately after confirming the Orange Path condition NO Red Path condition is verified to exist on remaining status trees.
B.
immediately after the CRS completes reading step 4 and NO Red Path condition is verified to exist on remaining status trees.
C.
when transitioning to another E-series procedure and NO Red Path condition is verified to exist on remaining status trees.
D.
as soon as NO Red Path condition is verified to exist on remaining status trees.
Proposed Answer:
C.
when transitioning to another E-series procedure.
Explanation (Optional):
Technical Reference(@:
(Attach if not previously provided)
OAP-012 Page 11 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 261 42 Modified Bank #
New (Note changes or attach parent)
L Page 158 of 200 Question History:
Prairie Island 1 10/03/2003 Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (10) 55.43 (5)
Comments:
Page 159 of 200 Examination Outline Cross-reference:
Level Tier #
WA #
99 Group #
4 G2.4.6 Importance Rating 3.1 4.0 Knowledge symptom based EOP mitigation strategies Proposed Question:
Common #73 Given the following plant conditions:
0 0
Following a series of plant malfunctions, operators are currently implementing ECA 0.0, Loss of All AC Power.
The operators have reached the point in the procedure where they are to begin depressurization of the Steam Generators.
Which of the following statements indicates the reason that a secondary depressurization is performed?
A.
To ensure the reactor remains subcritical and does not result in a restart accident.
B.
To remove stored energy in the Steam Generators to prevent a secondary side Safety Valve from lifting.
C.
To minimize RCS inventory loss through the RCP seals, which maximizes time to core u n cove ry.
D.
To depressurize the RCS in order to prevent a challenge to the "Integrity" Critical Safety Function Status Tree which is being monitored for implementation.
Proposed Answer:
C.
To minimize RCS inventory loss through the RCP seals, which maximizes time to core uncovery.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
ECA-0.0 Background Page 104 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Page 160 of 200 Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New YES Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (1 0) 55.43 (5)
Comments:
Page 161 of 200 Examination Outline Cross-reference:
Level Tier #
L, 100 Group #
WA #
4 G2.4.1 Importance Rating 4.3 Knowledge of EOP entry conditions and immediate action steps.
Proposed Question:
Common #74 The following conditions exist:
0 0
The plant has sustained an ATWS.
The team has entered FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS, from E-0, Reactor Trip or Safety Injection, step 1.
4.6 0
The OTC operator was unable to trip the turbine by pressing the Manual Tur,,,ie Tr push button.
What is the OTC's next action?
P A.
Open the Generator breaker.
B.
Manually run back the turbine.
C.
Close MSIVs.
D.
Commence Emergency Boration of RCS Proposed Answer:
B.
Manually run back the turbine.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
FR-S.l Step 2 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
EOP-C-002-503 (As available) i-Question Source:
Bank #
INPO 24659 Modified Bank #
New (Note changes or attach parent)
~ _ _ _
Page 162 of 200 Question History:
Seabrook 1 5/3 0/2 003 Question Cognitive Level:
Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content:
55.41 (1 0) 55.43 (5)
Comments:
Page 175 of 200 Examination Outline Cross-reference:
Level Tier #
108 Group #
KIA ##
1 G2.1.20 Importance Rating 4.3 4.2 Ability to execute procedure steps.
Proposed Question:
Common #75 Unit 2 is operating at 100% power when a sequence of annunciators actuate, indicating a loss of feedwater and a reactor trip, but NO reactor trip occurs. The following plant status is noted:
All attempts to perform a manual reactor trip fail.
An urgent failure prevents all rod motion.
All auxiliary feedwater pumps are operating.
0 The turbine remains on-line (AUTO turbine trip did not occur).
0 Reactor power remains near 100%.
Reactor coolant system temperature and pressure slowly increase from 100% power values.
Which one of the following correctly states the action that the operator is to take to mitigate the transient?
A.
Reduce turbine load slowly to avoid a rapid reactor coolant system temperature and pressure increase, leading to opening of a pressurizer safety valve.
B.
Trip the turbine to conserve the secondary coolant inventory to allow future RCS cooldown and depressurization.
C.
Open the PORVs immediately because the increasing pressure will take the pressurizer solid, resulting in insufficient water relief.
D.
Align maximum auxiliary feedwater flow to one steam generator to maintain it as a heat sink for cooldown of the reactor coolant system.
Proposed Answer:
B.
Trip the turbine to conserve the secondary coolant inventory to allow future RCS cooldown and depressurization.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
FR-S.l Step 2 FR-S.l Background Page 64
Page 176 of 200 Proposed References to be provided to applicants during examination:
NONE
-v.
Learning Objective:
Question Source:
EOP-C-042-3559/4633 (As available)
Bank #
~
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (1 0) 55.43 (5)
Comments:
Page 173 of 200 Examination Outline Cross-reference:
Level Tier #
107 Group #
WA #
4 G2.4.6 Importance Rating 3.1 4.0 Knowledge of symptom based EOP mitigation strategies Proposed Question:
Common #76 SRO-only Given the following sequence of events:
The plant was initially operating at 35% power.
High containment pressure resulted in Safety Injection, Steamline Isolation and Containment Spray Actuation.
RCS Pressure is 0 psig Containment Pressure is 25 psig No ABFPs can be started from the control room All SG Narrow Range Levels indicate 15%.
All SG pressures have stabilized at 700 psig Which one of the following describes the correct sequence of EOP implementation?
The team will initially enter E-0, Reactor Trip or Safety Injection and then transition to:
A.
E-1, Loss of Reactor or Secondary Coolant, FR-H.1, Loss of Heat Sink, from Status Trees, after SG levels lower to less that 10% Narrow Range B.
FR-H.l, Loss of Heat Sink, when no Aux Feedwater Flow can be established, then back to E-0, with RCS pressure less than SG pressure C.
FR-H.1, Loss of Heat Sink, when no Aux Feedwater Flow can be established, then to E-1, with RCS pressure less than SG pressure D.
E-1, Loss of Reactor or Secondary Coolant, then to ES-1.3 when RWST level drops below 9.24 feet Proposed Answer:
- 6.
FR-H.1, Loss of Heat Sink, when no Aux Feedwater Flow can be established, then back to E-0, with RCS pressure less than SG pressure Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
.\\..
E-0 Step 7 FR-H.l Ster, 1
Page 174 of 200 Proposed References to be provided to applicants during examination:
NONE Learning 0 bject ive:
(As available)
L.
Question Source:
Bank #
INPO IP2 23309 Modified Bank #
YES (Note changes or attach parent)
New Question History:
Indian Point 2 311 012003 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (1 0) 55.43 (5)
Comments:
Question, Answer and Distracters modified
Page 59 of 200 Examination Outline Cross-reference:
Level Tier #
42 Group #
Rev per TF WA #
2 002 K5.07 Importance Rating 3.3 3.6 Knowledge of the operational implications of reactivity effects of RCS boron, pressure and temperature as they apply to the RCS Proposed Question:
Common #77 SRO-only The unit is in MODE 6 for a refueling outage. Fuel movements are in progress to reload the reactor core.
Chemistrys sample of the RCS indicates that the boron concentration of the reactor coolant is 1960 ppm. What TS ACTION is required?
A.
Continue with CORE ALTERATIONS, suspend any positive reactivity additions and initiate actions to restore boron concentration limit
- 6.
Immediately lower any suspended fuel assemblies back into reactor vessel and initiate action to restore boron concentration limit C.
Immediately suspend all core reload operations, core offload operations may continue provided actions to restore boron concentrations limit is commenced D.
Immediately suspend CORE ALTERATIONS, suspend positive reactivity additions and initiate action to restore boron concentration limit Proposed Answer:
D.
Immediately suspend CORE ALTERATIONS, positive reactivity additions and initiate action to restore boron concentration limit Explanation (Optional):
A.
- 6.
C.
D.
Correct Must suspend all core alterations Lowering suspended fuel into core is required for decreasing cavity level Must suspend all core alterations even offload Technical Reference(s):
(Attach if not previously provided)
Page 3.9.1-1 Tech Spec 3.9.1, Cond A COLR Page 5 Proposed References to be provided to applicants during examination: COLR Page 5 Learning Objective:
ITS-C-012-0393, 0394 (As available)
Page 60 of 200 Question Source:
Bank #
~~
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comorehension or Analvsis X
10 CFR Part 55 Content:
55.41 (5) 55.43 (2,671 Comments:
Page 89 of 200 Examination Outline Cross-reference:
Level Tier #
58 Group #
WA #
1 000009 EK2.03 Importance Rating 3.0 3.3 Knowledge of the interrelations between the small break LOCA and the Steam Generators.
Proposed Question:
Common #78 SRO-only The following plant conditions exist:
0 A reactor trip with SI has occurred.
The team transitioned from E-0, Reactor Trip or Safety Injection, to FR-H.l, Loss of Secondary Heat Sink, from step 7, based on AFW flow c400 gpm and all SG levels
<lo%
RCS pressure is 700 psig and slowly decreasing.
0 All S/G pressures are approximately 950 psig and stable.
Which of the following summarizes plant conditions and what procedure is to be implemented?
A.
Remain in FR-H.1 until feed is restored then transition to E-1 where a depressurization of the secondary is prescribed to increase the heat transfer between the RCS and S/Gs.
- 6.
Heat transfer in the RCS during this event is such that the S/Gs are currently not functioning as a heat sink. Remain in FR-H.l to restore S/G levels to normal band.
C.
Heat transfer in the RCS during this casualty is such that the S/Gs are currently not functioning as a heat sink and therefore not required. Transition back to E-0, Reactor Trip or Safety Injection, step 7.
D.
Heat transfer in the RCS during this casualty is such that the S/Gs are currently not functioning as a heat sink and therefore not required. Transition to E-1, Loss of Reactor or Secondary Coolant, step 1.
Proposed Answer:
C.
Heat transfer in the RCS during this casualty is such that the S/Gs are currently not functioning as a heat sink and therefore not required. Transition back to E-0, Reactor Trip or Safety Injection, step 7.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
E-0 Step 7 FR-H.l Step 1
Page 90 of 200 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
EOP-C-044-3561 (As available)
Question Source:
Bank #
INPO Modified Bank #
YES New 2471 7 (Note changes or attach parent)
Question History:
Seabrook 1 5/30/2003 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 (5)
Comments:
Page 91 of 200 Examination Outline Cross-reference:
Level Tier #
59 Group #
WA #
1 00001 5/01 7 A2.02 Importance Rating 2.8 3.0 Ability to determine and interpret abnormalities in RCP air vent flow paths and/or oil cooling system as they apply to the Reactor Coolant Pump Malfunction Proposed Question:
Common #79 SRO-only A Component Cooling water leak inside containment has caused reduced flow to the RCPs.
The following conditions exist:
0 Unit 2 is at 100% power.
Temperatures / RCP#
21 22 23 24 Motor Bearing 189°F 205°F 177°F 181 "F Stator Winding 220°F 225°F 21 5°F 229°F Seal Inlet 195°F 185°F 205°F 200°F Annunicator RCP Hi Not Lit Not Lit Not Lit Not Lit Vi brat ion Which ONE of the following set of actions must be taken?
A.
Trip Reactor, secure all RCPs, Initiate E-0 B.
Trip Reactor, secure 22 RCP, Initiate E-0 C.
Trip Reactor, secure 23 RCP, Initiate E-0 D.
Perform a rapid plant shutdown, secure RCPs as necessary to isolate CCW leak Proposed Answer:
B.
Trip Reactor, secure 22 RCP, Initiate E-0 Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided) 2-AOP-RCP-1, Step 1 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
AOP-C-RCP1-1601040 (As available)
- ~ - _ _ _ -
Page 92 of 200 Question Source:
Bank #
INPO 22799 Modified Bank #
YES (Note changes or attach parent)
New Question History:
Cook 1 12/9/2002 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 (5)
Comments:
Question, answer and distracters modified
Page 93 of 200 Examination Outline Cross-reference:
Level Tier #
WA #
60 Group #
1 000040 G2.4.6 Importance Rating 3.1 4.0 Knowledge of symptom based EOP mitigation strategies Proposed Question:
Common #80 SRO-only You have entered ECA-2.1, "Uncontrolled Depressurization of All Steam Generators" and are performing SI Termination. Steam generator #24 pressure suddenly begins to rise.
Which one of the following actions is correct?
A.
Once the SI termination is complete, ECA-2.1 is complete and you are returned to procedure step in effect.
B.
Continue performing SI Termination and complete remaining steps of ECA-2.1, the RCS is now cooled to a point that the steam generators are beginning to fill.
C.
Stop performing SI Termination and go to E-2 because the pressure boundary has been established in steam generator #24.
D.
Complete performing SI Termination and then go to E-2, because the pressure boundary has been established in steam generator #24.
Proposed Answer:
D.
Complete performing SI Termination and then go to E-2, because the pressure boundary has been established in steam generator #24.
Explanation (Optional):
Technical Reference(@:
(Attach if not previously provided)
ECA-2.1 - Foldout Page Proposed References to be provided to applicants during examination:
NONE Learning Objective:
EOP-(3-026-3546 (As available)
Question Source:
Bank #
INPO 19545 Modified Bank #
New (Note changes or attach parent)
Page 94 of 200 Question History:
Cook 1 512 1 I200 1 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (5) 55.43 Comments:
Minor modifications to answer and distracters
Page 69 of 200 Examination Outline Cross-reference:
3 47 Group ##
2 Replaced per TF KIA #
G2.2.25 Importance Rating 2.5 3.7 Equipment Control; Knowledge of bases in technical specifications for limiting conditions for operations and safety limits Proposed Question:
Common #81 SRO-only It is determined that the Containment Spray System and FCU TS LCO (TS 3.6.6) is not met due to 480VAC Bus 6A being declared inoperable.
Which one of the following describes when a safety function determination must be performed?
A.
When the support system's Required Actions direct entry into Conditions and Required Actions for the supported system.
B.
When the support system's Required Actions direct the supported system to be declared inoperable.
C.
When the Conditions and Required Actions associated with the supported system LCO are NOT entered.
D.
When the Conditions and Required Actions associated with the support system LCO are NOT entered.
Proposed Answer:
C.
When the Conditions and Required Actions associated with the supported system LCO are NOT entered.
Explanation (Optional):
When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. In this event, an evaluation shall be performed in accordance with TS 5.5.1 3.
Technical Reference(s):
(Attach if not previously provided)
Pages 5.5-1 3 & 14 Page 3.0-2 Pages 3.0-8 & 9 TS 5.5.13 TS Section 3.0, LCO 3.0.6 Section 3.0.6 Bases Proposed References to be provided to applicants during examination:
NONE Learning Objective:
ITS-C-001-0103 (As available)
Page 70 of 200 Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
New Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 (2)
Comments:
Page 121 of 200 Examination 0 ut1 ine Cross-ref e rence:
Level Tier #
76 Group #
WA #
1 W/E11 EA2.2 Importance Rating 3.4 4.2 Ability to determine and interpret adherence to appropriate procedures and operation within the limitations in the facility's license and amendments as they apply to Loss of Emergency Coolant Recirculation Proposed Question:
Common #82 SRO-only The following sequence of events occurs:
0 0
0 0
0 Unit 2 was operating at 100% power Small break LOCA occurred 25 minutes ago.
The Team is currently implementing ECA 1.1, Loss of Emergency Coolant Recirculation due to loss of recirculation capability.
RCS Pressure is 450 psig Containment pressure is 4 psig.
Given the attached reference from ECA 1.l, Loss Of Containment Sump Recirculation, which of the following indicates the REQUIRED correct combination of Containment Fan Cooler Units and Containment Spray Pumps that are required to be operating under these conditions?
A.
4 FCUs, 0 Spray Pumps B.
1 FCU, 1 Spray Pump C.
- OFCU, 1 SprayPump D.
0 FCU, 2 Spray Pumps Proposed Answer:
D.
0 FCU, 2 spray Pumps Explanation (Optional):
25 minutes from start of small break LOCA with no RHR pumps injecting, the RWST level will be >9.24 ft. With VC pressure 4 psig, Table Step 4 requirements only met with 0 FCUs and 2 Spray Pumps Technical Reference(s):
(Attach if not previously provided)
ECA-I.1 Step 4 Table Proposed References to be provided to applicants during examination:
ECA-1.1, Step 4 Table
Page 122 of 200 Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank ##
(Note changes or attach parent)
New YES Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 (5)
Comments:
Page 127 of 200 Examination Outline Cross-reference:
Level Tier #
79 Group #
KIA #
1 000038 EK2.02 Importance Rating 2.4 2.5 Knowledge of the interrelations between sensors and detectors and a SGTR Proposed Question:
Common #83 SRO-only After a Reactor Trip the following parameter values and trends are noted:
Pressurizer Level is 0%.
RCS Pressure is 1500 psig and lowering.
RCS TcoL0 is 530°F and slowly trending down.
Containment pressure is 0.2 psiv and steady.
Containment Average Temperature is 105°F and lowering.
Main Steam Line Radiation Monitor R29 is in alarm and trending up.
0 SG 21 22 23 24 Wide Range 46%
31 Yo 47%
45%
Level trending up trending down trending up trending up AFW Flow 205 stable 200 stable 200 stable 21 0 stable Assuming that all other equipment responded as designed, Which one of the following statements describes the events in progress and what is the correct procedural flowpath after transitioning from E-0, Reactor Trip Or Safety Injection?
A.
Main Steam Line Break outside Containment and Steam Generator Tube Rupture; E-2, Faulted Steam Generator Isolation, to ECA-3.1, SGTR With Loss of Reactor Coolant -
Subcooled Recovered Desired.
B.
Main Steam Line Break outside Containment and Steam Generator Tube Rupture; E-2 Faulted Steam Generator Isolation, to E-3 Steam Generator Tube Rupture.
C.
Steam Generator Tube Rupture and isolated LOCA in Letdown; E-3, Steam Generator Tube Rupture, to E-1, Loss of Reactor Coolant or Secondary Coolant.
D.
Steam Generator Tube Rupture and Loss of Coolant Accident outside Containment; E-3, Steam Generator Tube Rupture, to ECA-1.1, Loss of Emergency Coolant Recirculation.
Proposed Answer:
B.
Main Steam Line Break outside Containment and Steam Generator Tube Rupture.
L-Page 128 of 200 Explanation (Optional):
Technical Reference(s1:
(Attach if not previously provided)
I
~
~
~
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
EOP-C-032-635 (As available)
Question Source:
Bank #
INPO 25582 Modified Bank #
YES (Note changes or attach parent)
New Question History:
ANO, Unit 2 711 1 /2003 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7) 55.43 (5)
Comments:
Question modified to make plant specific and add procedure decisions per TF
Page 129 of 200 Examination Outline Cross-reference:
Level Tier #
80 Group #
WA #
1 000054 AK1.01 Importance Rating 4.1 4.3 Knowledge of the operational implications of a MFW line break depressurizes the S/G concepts as it applies to a Loss of Main Feedwater Proposed Question:
Common #84 SRO-only The following conditions exist immediately following a reactor trip from 50% power:
e Pressurizer Level 38% (lowering)
Pressurizer Pressure 21 50 psia (lowering)
Containment sump levels rising Containment temperature1 45°F (rising)
Containment pressure 2.6 psig (rising)
Containment area radiation monitors R41/42 - 2.1 E-1 1 pCi/cc Containment wide range area radiation monitors I32926 - 1.1 R/hr LOOP 21 22 23 24 TCOLD 539°F 538°F 539°F 533°F lowering lowering lowering lowering Stm Gen 21 22 23 24 NR Level 14%
12%
15%
0%
lowering lowering lowering stable WR Level 58%
57%
59%
34%
increasing increasing increasing decreasing Pressure 910 psig 915 psig 912 psig 800 psig Which of the following is indicated by the given plant conditions and what procedure must be implemented?
A.
Pressurizer Steam Space Leak; 2-AOP-LEAK-1, Sudden Increase in Reactor Coolant System Leakage B.
Loop 24 Main Feedwater Break Inside Containment; E-2, Faulted Steam Generator Isolation C.
Loop 24 Main Feedwater Break Outside Containment; 2-AOP-FW-1, Loss of Main Feedwater D.
Loop 24 Cold Leg Small Break LOCA; E-1, Loss of Reactor or Secondary Coolant
Page 130 of 200 Proposed Answer:
B.
Loop 24 Main Feedwater Break Inside Containment; E-2, Faulted Steam Generator Isolation Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 25583 Modified Bank #
YES (Note changes or attach parent)
New Question History:
ANO, Unit 2 711 1 I2003 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (8, IO) 55.43 (5)
Comments:
Question, Answer and Distracters modified
Page 197 of 200 Examination Outline Cross-reference:
Level Tier #
119 Group #
WA #
1 000056 G2.4.6 Importance Rating 3.1 4.0 Knowledge symptom based EOP mitigation strategies Proposed Question:
Common #85 SRO-only A loss of off-site power caused a Unit 2 reactor trip. NO emergency diesel generators energized their respective busses and the operating team entered ECA-0.0, Loss of All AC Power.
The following conditions now exist:
All S/G NR levels are between 1525%
S/G Pressure is 940 psig 22 AFW Pump is running The team was depressurizing intact S/Gs when 480V Bus 5A was energized from 21 EDG 24 SW Pump has been started 21 and 22 FCUs are running RCP seals have been isolated PZR Level is 17%
Low Pressure SI was blocked during the Cooldown RCS Pressure lowered to 1750 psig during the cooldown Hottest In-core Thermocouple - 530°F Which one of the following identifies the next procedure to be implemented?
A.
E-0, Reactor Trip or Safety Injection B.
ES-0.1, Reactor Trip Response C.
ECA-0.1, Loss of All AC Recovery without SI required D.
ECA-0.2, Loss of All AC Recovery with SI Required Proposed Answer:
C.
ECA-0.1, Loss of All AC Recovery without SI required Explanation (Optional):
L Technical Reference(s):
(Attach if not previously provided)
Page 198 of 200 ECA-0.0 Step 27 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
~
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (1 0) 55.43 (5)
Comments:
New Question to replace #85
Page 199 of 200 Examination Outline Cross-reference:
Level Tier #
u 120 Group #
WA #
RO SRO 000001 AA2.03 Importance Rating 4.5 4.8 Ability to determine and interpret the proper actions to be taken as they apply to Continuous Rod Withdrawal.
Proposed Question:
Common #86 SRO-only Unit 2 power escalation was in progress. The OTC withdrew control rods to increase TAVE. When the OTC released the Rod Control IN-OUT switch the control rods continued to withdraw. The OTC placed Rod Control Bank Selector Switch to AUTO.
The following conditions were noted:
Nuclear Instrumentation Power Range Channels 15% and rising Power Range Low Power Trip NOT Blocked Intermediate Range Trip NOT Blocked Control rods withdrawing at 66 steps per minute Which ONE of the following actions are required in accordance with 2-AOP-ROD-1, Rod Control and Indication System Failure?
A.
Place Rod Control in Manual, if rod motion does not stop, Trip the reactor and go to E-0, Reactor Trip or Safety Injection.
B.
Place Rod Control in Manual, if rod motion does not stop, place rod control in Individual Bank Select, if rod motion does not stop, Trip the reactor and go to E-0.
C.
Place Rod Control in Manual, if rod motion does not stop, Open MOV-333, Emergency Boration Valve, Trip the reactor, and go to E-0.
D.
Place Rod Control in Manual, if rod motion does not stop, place shim switch to the IN direction and drive rods in to rod height prior to unwarranted rod motion.
Proposed Answer:
A.
Place Rod Control in Manual, if rod motion does not stop, Trip the reactor and go to E-0, Reactor Trip or Safety Injection.
Explanation (Optional):
L Technical Reference(s):
(Attach if not previously provided) 2-A0 P-ROD-1 Step 4.2
Page 200 of 200 Proposed References to be provided to applicants during examination:
NONE u
Learning Objective:
(As available)
Question Source:
Bank #
INPO 22797 Modified Bank #
New (Note changes or attach parent)
Question History:
Cook 1 1 12/9/2002 Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 (5)
Comments:
New Question to replace #86
Page 133 of 200 Examination Outline Cross-reference:
Level Tier #
84 Group #
WA #
2 WIE 01 EK1.2 Importance Rating 3.4 4.0 Knowledge of the operational implications of normal abnormal and emergency operating procedures associated with Reactor Trip or Safety Injection / Rediagnosis Proposed Question:
Common #87 SRO-only The plant has undergone a Reactor Trip and Safety Injection. The Team transitioned to E-3, Steam Generator Tube Rupture, from E-0, Reactor Trip or Safety Injection] due to elevated Steam Generator Radiation levels from a previous 0.5 gpm Steam Generator tube leak.
The CRS has transitioned to ES-0.0, Rediagnosis, to evaluate if the Team is in the correct guideline.
Given the following:
All SI pumps and 21 RHR pump are running RCS press is 1900 psig and rising slowly PZR level is 30 % and rising slowly TAVE is 544°F and rising slowly SG radiations levels same as from SG tube leak All SG Pressures are 1000 psig and stable All SG Levels are 5% NR and rising slowly AFW Flow is 200 gpm to each SG Based on the given plant conditions, the CRS is required to transition to which of the following procedures?
A.
ES-1.1, SI Termination] Step 1 B.
E-0, Reactor Trip or Safety Injection, Step 1 C.
E-1 Loss of Reactor or Secondary Coolant, Step 1 D.
E-3, Steam Generator Tube rupture, Step 1 Proposed Answer:
C.
E-1, Loss of Reactor or Secondary Coolant, Step 1 Explanation (Optional):
A. SI termination criteria met for given plant conditions but must first transition to E-1 B. E-0 was already completed prior to transition to E-3 C. Correct, there is no faulted SG or SG with uncontrolled increase in level or abnormal radiation levels
Page 134 of 200 D. E-3 not required since radiation levels from SG tube leak and no SG level increasing abnormally Technical Reference(s):
(Attach if not previously provided)
ES-0.0 Pages 2,3 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
EOP-C-003-507 (As available)
Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (8, 10) 55.43 (5)
Comments:
New Question
Page 177 of 200 Examination Outline Cross-reference:
Level Tier #
109 Group #
WA #
2 000069 G2.2.14 Importance Rating 2.1 3.0 Knowledge of the process for making configuration changes Proposed Question:
Common #88 SRO-only The Maintenance Supervisor has requested to override the 80 ft. VC airlock mechanical interlocks to facilitate placing required maintenance equipment into containment.
Per Containment Entry and Egress procedure, the VC airlock mechanical interlocks may be overridden....
A.
B.
ANYTIME with Shift Manager approval.
ONLY with the Unit in Mode 5 or Mode 6 C.
During Mode 4 with a dedicated watch assigned to set containment integrity.
D.
After a Temp Alt has been written in accordance with ENN-DC-136, Temporary Alterations.
Proposed Answer:
B.
ONLY with the Unit in Mode 5 or Mode 6 Explanation (Optional):
Technical Reference( s) :
(Attach if not previously provided)
OAP-007 Page 47 TS 3.6.2 Page 3.6.2-1 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Page 178 of 200 Question Cognitive Level:
10 CFR Part 55 Content:
Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 55.43 (3)
Comments:
New Question c
Page 137 of 200 Examination Outline Cross-reference:
Level Tier #
a7 Group #
WA #
1 004 K5.19 Importance Rating 3.5 3.9 Knowledge of the operational implications of the concept of SDM as it applies to the CVCS Proposed Question:
Common #89 SRO-only Given the following:
The Unit is in Mode 3 0
Plant cooldown is in progress RCS temperature is 520°F 0
RCS Boron Concentration is 600 ppm 0
Reactor Engineering informs the control room that the SDM is 1.2%
Using the attached COLR, determine what action, if any, is required to be performed for the given conditions.
A.
- 6.
Initiate boration to restore SDM to within limits within 15 minutes Initiate boration to restore SDM to within limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C.
Cooldown may continue but do not enter Mode 4 until SDM within limits D.
No action required, SDM within limits of COLR for given conditions Proposed Answer:
A.
Initiate boration to restore SDM to within limits within 15 minutes Explanation (Optional):
A. Correct, TS 3.1.1, Condition A, required SDM from COLR Figure 2 is 1.5 B. Completion time is 15 minutes to restore SDM C. TS 1.1.1 is applicable in Modes 2 with Keff 4.O and 3,4 & 5 D. SDM required to be above curve of COLR Figure 2 Technical Reference(@:
(Attach if not previously provided)
TS 3.1.1 Page 3.1.1 -1 Graph RPC-5 COLR Figure 2 L
Proposed References to be provided to applicants during examination: Graph RPC-6 COLR Figure 2
Page 138 of 200 Learning Objective:
Question Source:
SYS-C-O30-30105/30114 (As available)
Bank #
Modified Bank #
(Note changes or attach parent)
New YES Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (5) 55.43 (1, 2)
Comments:
New Question
Page 139 of 200 Examination Outline Cross-reference:
Level Tier #
88 Group ##
WA #
1 005 A2.03 Importance Rating 2.9 3.1 Ability to predict the impacts of RHR pump/motor malfunctions or operations on the RHRS and based on those predictions use procedures to correct, control or mitigate the consequences of those malfunctions or operations Proposed Question:
Common #90 SRO-only The Unit is in Mode 6 for Vessel Head O-Ring repair with 21 RHR pump in service providing core cooling. A loss of 21 RHR occurs and 2-AOP-RHR-1, Loss of RHR, is implemented.
Given the following plant conditions:
0 Reactor Vessel level is 68 ft.
0 Reactor Vessel Head has been de-tensioned 0
Reactor has been shutdown for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 0
The plant is at 600 EFPDs 0
RCS temperature is 130°F and stable Determine the amount of time remaining to reach 200°F and the required actions the team must take.
A.
8 minutes; evacuate ALL personnel from containment B.
8 minutes; evacuate ALL NON-Essential personnel from containment C.
28 minutes; evacuate ALL personnel from containment D.
28 minutes; evacuate ALL NON-Essential personnel from containment Proposed Answer:
B.
8 minutes; evacuate ALL NON-Essential personnel from containment Explanation (Optional):
A. Correct time; incorrect evacuation B. Correct from Graph ACSPC, RCS level at 66; correct evacuation C. Time to 200°F using Graph ACS-2DI Reactor Cavity level at 92; incorrect evacuation D. Incorrect time; correct evacuation Technical Reference(s):
(Attach if not previously provided) 2-AOP-RHR-1 Graph ACS 2C Rev 9
Page 140 of 200 Graph ACS 2D Rev 8 Proposed References to be provided to applicants during examination: Graph ACS-PC, 2D Y
Learning Objective:
SYS-(2-042-114 (As available)
Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (51 55.43 (5)
Comments:
New Question
'v
'Le-Page 141 of 200 Examination Outline Cross-reference:
Level Tier #
89 Group #
WA #
1 008 A2.02 Importance Rating 3.2 3.5 Ability to predict the impacts of High/low surge tank level malfunction or operations on the CCWS and based on those predictions use procedures to correct, control or mitigate the consequences of those malfunctions or operations Proposed Question:
Common #91 SRO-only Given the following plant conditions:
0 Unit 2 was operating at 100% power when a loss of coolant event occurred.
0 The operating Team has just entered ES-1.3, Transfer To Cold Leg Recirculation.
0 Shortly after initiating ES-1.3, a CCW leak on the bottom of the CCW Surge Tank has caused the surge tank to empty.
The CRS directs that all 3 CCW pumps be placed in pullout.
All other equipment is operating per design.
Based on these conditions, which of the following correctly describes the impact of these events during subsequent actions to establish cold leg recirculation?
A.
Alignment for cold leg Recirc will continue without CCW. Attempt to start 2 CCW pumps while performing ES-1.3.
- 3.
Alignment for cold leg recirculation will NOT continue without CCW. A transition to ECA-1.l, Loss of Emergency Coolant Recirculation, must be made immediately.
C.
Immediately enter 2-AOP-CCW-I, Loss of Component Cooling Water, to address the CCW leak. ES-1.3 will be utilized as a secondary priority until CCW is restored.
D.
Alignment for cold leg recirculation will continue without CCW. Auxiliary Component Cooling Water Pumps will be verified running.
Proposed Answer:
D.
Alignment for cold leg recirculation will continue without CCW. Auxiliary Component Cooling Water Pumps will be verified running.
Explanation (Optional):
Technical Reference(@:
(Attach if not previously provided)
ES-1.3, step 7 Page 5
Page 142 of 200 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
EOP-(2-013-3535 (As available)
Question Source:
Bank ##
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (5) 55.43 (5)
Comments:
Page 135 of 200 Examination Outline Cross-reference:
Level Tier #
86 Group #
WA #
2 W/E09 EA2.2 Importance Rating 3.4 3.8 Ability to determine and interpret adherence to appropriate procedures and operation within the limitations in the facility's license and amendments as they apply to Natural Circulation Operations Proposed Question:
Common #92 SRO-only Given the following plant conditions:
Loss of Offsite power has occurred.
A plant cooldown is in progress per ES-0.2, "Natural Circulation Cooldown" Adequate RCS subcooling is being maintained The Control room operators suspect void formation in the reactor vessel head due to large variations in Pressurizer level Cooldown and Depressurization with vessel void formation is NOT desirable Which of the following is used to collapse the void?
A.
Exit ES-0.2 and transition to ES-0.3, Natural Circulation Cooldown with Steam Void In Vessel, and increase RCS pressure by starting additional charging pumps to collapse the void.
B.
Stay in ES-0.2 and increase Reactor Coolant System (RCS) pressure using Pressurizer Heaters to collapse the void C.
Exit ES-0.2 and transition to ES-0.3 and increase RCS pressure by starting Safety Injection Pumps to collapse the void.
D.
Stay in ES-0.2 and decrease RCS temperature while maintaining RCS pressure constant to collapse the void Proposed Answer:
B.
Increase Reactor Coolant System (RCS) pressure using pressurizer heaters.
Explanation (Optional):
A. ES-0.3 only used if C/D and Depressurization required to be done at rate that will produce a void, Note prior to step 12.
B. Correct C. Cooldown is stopped only if subcooling is lost, step 13 D. Safety Injection pumps are not used in ES-0.2 Technical Reference(s):
(Attach if not previously provided)
Page 136 of 200
_ _ _ _ _ _ _ ~
ES-0.2 & Background Step 14 & Background Page 41 L-Proposed References to be provided to applicants during examination:
NONE Learning Objective:
EOP-C-007-3530 (As available)
Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 (5)
Comments:
Page 143 of 200 Examination Outline Cross-reference:
Level Tier #
90 Group #
WA #
1 062 K4.01 Importance Rating 2.6 3.2 Knowledge of AC distribution system design features andor interlocks which provide for Bus lockouts Proposed Question:
Common #93 SRO-only Unit 2 was operating at 100% power when a winding short developed on 21 RCP. The overcurrent conditions caused a loss of 6.9KV Bus section 1.
Determine the required operator action, if any, and the electrical plant configuration, 480V Bus, 60 seconds after the fault on 6.9KV Bus section.
A.
Manually transfer 6.9KV Bus 2 to Bus 5 and energize Bus 2A from Station Service Transformer (SST) 2; 480V Buses 2A, 3A, 5A & 6A energized from SST 2,3,5 & 6 B.
Manually energize bus 2A from EDG; 480V Buses 3A, 5A & 6A energized from SST 3,5
& 6 and 480V Bus 2A energized for EDG.
C.
No operator action required; 480V Buses 3A, 5A & 6A energized from SST 3, 5 & 6 and 480V Buses 2A de-energized.
D.
No operator action required; 480V Buses 2A, 3A, 5A & 6A energized from EDGs 21,22
& 23.
Proposed Answer:
B.
Manually energize bus 2A from EDG; 480V Buses 3A, 5A & 6A energized from SST 3,5
& 6 and 480V Bus 2A energized for EDG.
Explanation (Optional):
Loss of Bus Section 1 will cause single loop loss of flow Rx trip. Loss of bus section 1 will cause loss of auto transfer of Bus section 2 due to loss of sync check. Bus sections 5 & 6 will remain powered from Station Aux Trans. And Buses 3 &4 will auto transfer to Bus section 6.
480V buses 3A, 5A & 6a remain powered from Bus Sections 3,5 & 6 due to no Blackout signal present. Buses 2A will be de-energized. All 3 EDGs start but do not auto close onto the buses.
ES-0.1 directs operator to energize all 480V buses from EDG if not powered from offsite.
Technical Reference(s):
(Attach if not previously provided)
Proposed References to be provided to applicants during examination:
NONE
L.
Page 144 of 200 Learning Objective:
SYS-C-271-426/427 (As available)
Question Source:
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7) 55.43 (5)
Comments:
Page 145 of 200 Examination Outline Cross-reference:
Level Tier #
WA ##
91 Group #
2 034 A2.01 Importance Rating 3.6 4.4 Ability to predict the impacts of a dropped fuel element on the Fuel Handling System and based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations Proposed Question:
Common #94 SRO-only The following conditions exist:
0 There is a core off-load in progress.
The fuel handler was moving irradiated fuel to a location in the spent fuel pool.
0 You are notified that the spent fuel bundle was accidentally dropped in the spent fuel pool.
The fuel handler reports the fuel bundle fell into the correct pool location.
R44, Plant Vent radiation monitor reads 4E-3 pCi/cc and is steady.
What actions, if any, are required?
A.
Enter procedure 2-AOP-FH-1, Fuel Damage or Loss of SFP/Refueling Cavity Level.
Evacuate non-essential personnel from the FSB, place FSB ventilation in service and monitor R44.
B.
Enter procedure 2-AOP-FH-1, Fuel Damage or Loss of SFP/Refueling Cavity Level.
Suspend ALL fuel handling operations in FSB, Evacuate non-essential personnel from the FSB, secure FSB ventilation and monitor R44.
C.
Enter procedure 2-AOP-FH-1, Fuel Damage or Loss of SFP/Refueling Cavity Level.
Suspend ALL fuel handling operations in FSB, Evacuate ALL personnel from the FSB, dispatch an operator to close ALL FSB doors and monitor R44.
D.
Enter procedure 2-AOP-FH-1, Fuel Damage or Loss of SFP/Refueling Cavity Level.
Suspend ALL fuel handling operations in FSB, Evacuate ALL personnel from the FSB and dispatch an operator to independently verify proper fuel bundle location and monitor R44.
Proposed Answer:
C.
Enter procedure 2-AOP-FH-1, Fuel Damage or Loss of SFP/Refueling Cavity Level.
Suspend ALL fuel handling operations in FSB, Evacuate ALL personnel from the FSB, dispatch an operator to close ALL FSB doors and monitor R44.
Explanation (Optional):
All personnel must be evacuated, ventilation is not secured and independent verification of location not required
Page 146 of 200 Technical Reference(s):
(Attach if not previously provided) 2-A0 P-FH-1 Pages 7,9, 11
\\--
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
Question Source:
AOP032FHl-17304 (As available)
Bank ##
INPO 2471 2 Modified Bank ##
New (Note changes or attach parent)
Question History:
Seabrook 1 5/30/2003 Question Cognitive Level:
Memory or Fundamental Knowledge Comwehension or Analvsis X
10 CFR Part 55 Content:
55.41 (5) 55.43 (57)
Comments:
Minor modifications to make plant specific
Page 163 of 200 Examination Outline Cross-reference:
1 Group #
2 WA #
000076 G2.1.28 Importance Rating 3.2 3.3 Knowledge of the purpose and function of major system components and controls Proposed Question:
Common #95 SRO-only Given the following:
Unit 2 is at 100% power.
2-AOP-HIACT-I, High RCS Activity, has been implemented due to an increase in RCS activity.
WHICH ONE of the following is the required action and reason for the adjustment?
A.
Increase letdown flow to maximum so that more water can flow through the cation bed ion exchangers.
B.
Increase letdown flow to maximum so that more RCS water can be diverted and processed by the waste management system.
C.
Increase letdown flow to maximum so that more water can flow through the mixed bed ion exchangers.
D.
Increase letdown flow so that a maximum amount of water can flow into the VCT to increase sparging of the radioactive gases from the top of the VCT.
Proposed Answer:
C.
Increase letdown flow to maximum so that more water can flow through the letdown ion exchangers.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided) 2-HIACT-1 Page 9 2-H I ACT-1 Background Page 2 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
AOP031 HlAl-28981/28982 (As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New YES Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 (7)
Comments:
Page 165 of 200 Examination Outline Cross-reference:
3 Group #
1 WA #
G2.1.33 Importance Rating 3.4 4.0 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications Proposed Question:
Common #96 SRO-only Given the following plant conditions:
0 0
0 0
0 0
Unit 2 is at 80% power.
Bank D step counter indicates 200 steps.
AFD verified outside target band for 45 minutes in last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> CBD Rod H-8 position is verified at 185 steps 21 SG Activity verified at 0.1 uci/gm 1-131 equivalent 21 SG Tube leakage verified at 0.2 gpm Based on the above conditions, which one of the following statements is correct?
A.
AFD outside target band excess amount of time; reduce power to 40%
B.
Rod H-8 out of position; realign rod H-8 C.
SG activity in excess of allowable limits; commence plant shutdown and cooldown to
<5OO0F D.
SG tube leakage in excess of allowable limits; commence plant shutdown Proposed Answer:
D.
21 SG Tube leakage verified at 0.2 gpm Explanation (Optional):
A. TS 3.2.3 5 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is limit B. TS 3.1.4 524 steps is limit C. TS 3.7.14 50.1 5 vci/gm 1-1 31 equivalent is limit D. TS 3.4.13 150 gpd is limit (0.2 gpm = 288 gpd)
Technical Reference(s):
(Attach if not previously provided) 2-AOP-H IACT-1 TS 3.1.4 TS 3.2.3
Page 166 of 200 TS 3.4.13 TS 3.7.14 L-Proposed References to be provided to applicants during examination:
NONE Learning Objective:
Question Source:
(As available)
Bank #
Modified Bank #
New YES (Note changes or attach parent)
Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 (2,3)
Comments:
New Question
Page 183 of 200 Examination Outline Cross-reference:
Level Tier #
112 Group #
KIA #
1 G2.1.6 Importance Rating 2.1 4.3 Ability to supervise and assume a management role during plant transient and upset conditions Proposed Question:
Common #97 SROOnly As the Shift Manager, you have been informed by the National Weather Service, that a Hurricane Warning for the Mid Atlantic Coast of the United States has just been issued.
In accordance with OAP-008, Severe Weather Preparations, which of the following are you required to perform ?
A.
Declare a Notification of Unusual Event (NUE), commence a plant shutdown, notify NRC of forced plant shutdown B.
Call in Staffing Level I I personnel, coordinate with Consolidated Edison to remove the unit from the system C.
Notify the Unit 2 OM, designate a Storm Coordinator, review the Emergency Plan to determine if emergency classification exists for conditions.
D.
Monitor storm progress and take no actions until the National Weather Service issues a Hurricane Warning for the Peekskill area.
Proposed Answer:
C.
Notify the Unit 2 OM, designate a Storm Coordinator, review the Emergency Plan to determine if emergency classification exists for conditions.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
OAP-008 Page 4, 9 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
Modified Bank #
(Note changes or attach parent)
New
~
YES Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 (6)
Comments:
Page 169 of 200 Examination Outline Cross-reference:
3 Group #
3 KIA #
G2.3.4 Importance Rating 2.5 3.1 Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized Proposed Question:
Common #98 SRO-only Given the following conditions on Unit 2:
0 0
0 A LOCA outside containment occurred A Site Area Emergency was declared The broken line was manually isolated locally, but the operator performing the task was injured and cannot leave the area on his own Initial dose estimates are 1 10 Whr gamma The rescue time for a 2-man team is estimated to be 5 minutes with a maximum of 10 minutes In accordance with IP-EP-630, On Site Medical Emergency, the injured person:
A.
is to be moved immediately, by risk-informed volunteers, unless moving the individual is life threatening.
B.
is NOT to be moved immediately, because moving the individual could cause serious medical complications.
C.
is to be moved, by risk-informed volunteers, as soon as Emergency Medical Technician /
First Aid Responder authorizes.
D.
is NOT to be moved until after dose levels are reduced, by any means, to less than 10 Whr gamma.
Proposed Answer:
A.
should be moved immediately, by risk-informed volunteers, unless moving the individual is life threatening.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
IP-EP-630 Page 6
Page 170 of 200 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 261 67 Modified Bank #
YES (Note changes or attach parent)
New Question History:
Question Cognitive Level:
Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content:
55.41 55.43 (4)
Comments:
Modified Question
Page 171 of 200 Examination Outline Cross-reference:
Level 106 Tier #
Group #
WA #
4 G2.4.27 Importance Rating 3.0 3.5 Knowledge of fire in the plant procedures Proposed Question:
Common #99 SRO-only There is a fire in the plant that cannot be brought under control in a reasonable amount of time.
In accordance with SMM-DC-901, IPEC Fire Protection Program Plan, the Fire Brigade Leader is to:
A.
B.
C.
D.
request additional fire fighting qualified plant personnel to assist in fighting the fire.
call the Verplanck Fire Department directly for assistance.
call the Buchanan Fire Department directly for assistance.
request the CRS notify the Verplanck Fire Department for assistance.
Proposed Answer:
D.
request the CRS notify the Verplanck Fire Department for assistance.
Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
SMM-DC-901 Page 37 Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
INPO 261 68 Modified Bank #
YES (Note changes or attach parent)
New Question History:
Prairie Island 1 9/01 /2003
~
Page 172 of 200 Question Cognitive Level:
b 10 CFR Part 55 Content:
55.41 (1 0)
Memory or Fundamental Knowledge X Comprehension or Analysis 55.43 (5)
Comments:
Question, Answer and Distracters modified
Page 167 of 200 Examination Outline Cross-reference:
Level L-104 Tier #
Group #
WA #
2 G2.2.26 importance Rating 2.5 3.7 Knowledge of refueling administrative requirements Proposed Question:
Common #IO0 SRO-only The RCS is being drained to 63 feet for vacuum refill following refueling. The current RCS level is 66 feet. The operable level indicators are the CCR Foxboro and the Mansell Level Monitoring System. During the draindown the hard drive on the Mansell fails resulting in a loss of indication from this instrument.
What actions, if any, are required?
A.
No actions required, continue with the draindown using the CCR Foxboro.
B.
Stop the draindown; ensure the CCR Foxboro is still tracking accurately, then continue with the draindown.
C.
Stop the draindown and evaluate conditions. Shift Manager approval required and is sufficient before resuming.
Stop the draindown and evaluate conditions. Operations Manager approval required and is sufficient before resuming.
D.
Proposed Answer:
D.
Stop the draindown and evaluate conditions. Operations Manger approval required before proceeding Explanation (Optional):
Technical Reference(s):
(Attach if not previously provided)
SOP-4.2.2 Pages 2 & 3
~~
Proposed References to be provided to applicants during examination:
NONE Learning Objective:
(As available)
Question Source:
Bank #
I P2 SYS171-3 Modified Bank #
YES (Note changes or attach parent)
Page 168 of 200 New Question History:
Question Cognitive Level:
10 CFR Part 55 Content:
Comments:
Memory or Fundamental Knowledge Comprehension or Analysis X
55.41 55.43 (5)