ML080280360
| ML080280360 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 01/28/2008 |
| From: | NRC/RGN-II |
| To: | |
| References | |
| 50-348/07-301, 50-364/07-301 | |
| Download: ML080280360 (260) | |
See also: IR 05000348/2007301
Text
Final Submittal
(Blue Paper)
COMBINED RO/SRO WRITTEN EXAM
WITH KAS, ANSWERS, REFERENCES,
QUESTIONS REPORT
for 25 SRO Questions
1. 006 A2.IO 004
Unit 1 is in Mode 1. Chemistry has provided sample results for boron concentration of
1A and 1B Accumulators with the following results:
- 1A A~cumulator boron concentration is 2350 ppm.
- 1B Accumulator boron concentration is 2198 ppm.
Which ONE of the following describes the impact of this condition; and the action
required in accordance with Technical Specifications and SOP-8.0, Safety Injection
System-Accumulators?
A. * Ability to maintain subcriticality after an accident is reduced;
- Drain and fill the 1A Accumulator to lower boron concentration.
B. * Ability to maintain minimum boron precipitation time is reduced;
- Drain and fill the 1A Accumulator to lower boron concentration.
C~ * Ability to maintain subcriticality after an accident is reduced;
- Feed and bleed the 1B Accumulator to raise boron concentration.
D. * Ability to maintain minimum boron precipitation time is reduced;
- Feed and bleed the 1B Accumulator to raise boron concentration.
Monday, January 14, 20082:43:54 PM 1
QUESTIONS REPORT
for 25 SRO Questions
Meets 10 CFR 55.43 (b) 2 requirements for SRO level question.
A. is incorrect; 111 All Accumulator boron concentration is within spec.
SOP-8.0 does provide guidance that fills and/or drains the accumulators in a separate
section, but not to raise or lower the Boron concentration.
Step 4.1 fills the accumulators but does not address TS requirements for level and
pressure while filling. Also does not address boron concentration unless a 12% level
change is made, so sampling is not required due to the fill.
Step 4.2 provides guidance to lower Accumulator level but not to lower boron C.
Appendix 2 is provided expressly to raise boron C. to >2300 PPM
B. incorrect, 111 All Accumulator boron concentration is within spec.
C. Correct.
TS 3.5.4 basis states that boron concentration of the RWST is designed to ensure
subcriticality is maintained with uncontrolled cooldown coincident with most
reactive rod stuck fully out.
For a large break LOCA analysis, the minimum water volume limit of 321 ,000 gallons and the
lower boron concentration limit of 2300 ppm are used to compute the post LOCA sump boron
concentration necessary to assu're subcriticality. The large break LOCA is the limiting case
since the safety analysis assumes that all control rods are out of the core.
Within the same bases the following is found and may cause the applicant to
choose the precipitation idea which is what bounds the upper limit.
A water volume of 506,600 gallons and the upper limit on boron
concentration of 2500 ppm are used to determine the maximum
allowable time to switch to hot leg recirculation following a LOCA. .
The purpose of switching from cold leg to hot leg injection is to avoid
boron precipitation in the core following the accident.
SOP-B.O, APP 2, FEED AND BLEED OF ACCUMULATOR 1A (18, 1C) TO RAISE
BORON CONCENTRA TION >2300 PPM would be used to raise boron concentration
SOP-8.0
CAUTION: Accumulator boron concentration must be maintained between 2200 and
2500 ppm; the intent of this appendix is to raise accumulator boron
concentration> 2300 ppm.
D. Incorrect. Basis is incorrect
Monday, January 14, 2008 2:43:54 PM 2
QUESTIONS REPORT
for 25 SRO Questions
006 A2.1 0 Emergency Core Cooling
Ability to (a) predict the impacts following malfunctions or operations on the and
(b) based on those predictions, use procedures to correct, control, or mitigate the
consequences of those malfunctions or operations:, Low boron concentration in SIS.
Question Number: 86
Tier 2 Group 1
Importance Rating: 3.9
Technical Reference: TS 83.5.1/3.5.2 and basis, SOP-8, SOP-2.3,
Proposed references to be provided to applicants d,uring examination: None
Learning Objective: OPS521 02801
10 CFR Part 55 Content: 43.2
Comments:
fixed per FJE comments
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: CAD C B DCA A D Scramble Range: A - D
Source: MODIFIED Source if Bank:
Cognitive Level: LOWER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:54 PM 3
QUESTIONS REPORT
for 25SRO Questions
2. 009"EA2.01 005
Given the following:
- A small break LOCA has occurred on Unit 1.
- The crew is performing EEP-1.0, Loss of Reactor or Secondary Coolant,
step 7 that checks SI Termination Criteria.
- Containment pressure is 3.6 psig.
- Subcboled Margin Monitor value is 14°F in CETC mode.
- RCS pressure is 1100 psig and stable.
- Pressurizer level is 20% and rising slowly.
Which ONE of the following correctly describes the procedure flow path when
evaluating step 7, Check SI Termination Crite~ia, of EEP-1.0, and the reason?
The crew wilL ..
A':' remain in EEP-1 because RCS subcooling is too low.
B. remain in EEP-1 because RCS pressure is NOT rising.
C. remain in EEP-1 because pressurizer level is too low.
D. go to ESP-1.1, SI Termination, because all SI Termination criteria are met.
Meets 10 CFR 55.43 (b) 5 requirements for SRO level question.
A:Correct. Subcooling does not meet requirements.
Check SUB COOL"ED MARGIN
MONITOR indication - GREATER
THAN 16F{45F} SUBCOOLED IN
CETC MODE.
B: Incorrect. pressure may be stable or rising.
7.3 Check RCS pressure - STABLE OR RISING.
C: Incorrect. level meets requirement.
7.4 Check pressurizer level - GREATER THAN 13% {43 % }.
D: Incorrect. Subcooling must be raised by cooldown or pressure increase
NOTE: For certain break sizes, SI termination criteria may be met due to injection flow
exceeding mass flow out of the break. Step 7.5 is not intended to terminate SI wh~n a known
7.5 IF all SI termination criteria satisfied, THEN go to FNP-1-ESP-1.1, SI TERMINATION.
Monday, January 14, 20082:43:54 PM 4
QUESTIONS REPORT
for 25 SRO Questions
009 EA2.01
009 small break LOCA
Ability to determine or interpret the following as they app~y to a small break LOCA: Actions to
be taken, based on RCS temperature and pressure~ saturated and superheated
Question Number: 76
Tier 1 Group 1
Importance Rating: SRO 4.8
Technical Reference: EEP-1, Step 7
Proposed references to be provided to applicants during examinatio'n: 'None
Learning Objective: OPS52301 B09
10 CFR Part 55 Content: 43.5
Comments:
fixed per FJE comments
009 smaH break LOCA
AbiHty to determine or interpret the following as they apply to a smaU break LOCA: Actions to
be taken, based on ReS temperature and pressure~ saturated and superheated
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: AAAAC B C C CA Scramble Range: A - D
Source: NEW Source if Bank:
Cogniti~e Level: illGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:54 PM 5
QUESTIONS REPORT
for 25 SRO Questions
3.011 A2.12 001
Given the following:
- A reactor trip has occurred and 1C RCP is the only operating RCP.
- Auxiliary Spray has been placed in service lAW ESP-0.1, Reactor Trip
Response, to control and reduce RCS pressure.
- The plant is preparing for a cooldown lAW UOP-2.2, Shutdown of Unit from Hot
Standby to Cold Shutdown, with the following parameters:
- RCS pressure is 2230 psig.
- RCS temperature is 537°F.
- Pressurizer level is 23%.
- The crew is at step 5.2 to begin raising pressurizer level to 55%.
Which ONE of the following correctly describes the limit associated with cooling down
the pressurizer, and while raising pressurizer level, the method used to prevent thermal
stratification in accordance with UOP-2.2?
A. * The temperature difference between the pressurizer steam space and charging
water must not exceed 320°F;
- A pressurizer- insurge must occur during the pressurizer cooldown.
B!' * The temperature difference between the pressurizer steam* space and charging
water must not exceed 320°F;
- A pressurizer outsurge must occur during the pressurizer cooldown.
C. * The pressurizer cooldown rate must not exceed 100°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period;
- A pressurizer insurge must occur during the pressurizer cooldown.
D. * The pressurizer cooldown rate must not exceed 100°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period;
- A pressurizer outsurge must occur during the pressurizer cooldown.
Monday, January 14, 2008 2:43:54 PM 6
QUESTIONS REPORT
for 25 SRO Questions
Meets 10 CFR 55.43 (b) 2 and 5 requirements for SRO level question.
5.4 IF auxiliary spray is in operation, THEN on Data Sheet 1 record the time and the
differential temperature between regenerative heat exchanger outlet charging TI-
123 and pressurizer vapor space TI-454. Ensure that the differential temperature
does not exceed 320 o P.
A incorrect; outsurge is required; correct parameters used to determine differential
temperature.
B correct;
step 5.2 begins raising the level to 55% and there is a caution prior to and a note after
that step to limit delta T and level <63.5% and an outsurge is to be maintained.
5.35.1 Verify Delta T between pressurizer and charging < 320°F.
5.35.1.1 Commence recording delta T in PNP-1-STP-1.0, OPERATIONS DAILY AND SHIFf
SURVEILLAN'CE REQUIREMENTS, misc. section every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during auxiliary spray
operation.
P&L ofUOP-2.2
3.3.3 Do not exceed a 200 0 PIhr pressurizer cooldown rate.
3.3.4 The temperature differential between the pressurizer and the RCS must
not exceed 320 0 P. The pressurizer liquid, surge line and loop B hot leg
temperatures should be monitored to ensure that a pressurizer outsurge is
taking place whenever the pressurizer is being cooled or filled. This will
prevent thermal stratification from taking place. A pressurizer outsurge is
indicated by surge line temperature approximately equal to pressurizer
liquid temperature and greater than "B" Hot Leg temperature.
C incorrect; PRZR cooldown rate is 200°F
CAUTION: PRZR cooldown rate must be limited to < 200°Flhr.
P&L
3.3.2 Do not exceed RCS cooldown rate specified in PTLR section 2.0, Operating
Limits. The maximum cooldown rate is 100 0 P in anyone hour period.
D incorrect; PRZR cooldown rate limit is 200°F
The following flow path could cause entry into UOP-2.2 with aux spray in service. If 1C
RCP is the only running pump after a Rx trip aux spray would be put on service as long
as Letdown is in service.. ESP-O.1 will send you to UOP-2.3 which could send you to
UOP-2.2. In this scenario the P&Ls of UOP-2.2 would apply and the limits of 320°F
and 200°F would be applica,.LJ..bu:;Ie::....------------------------
Monday, January 14, 20082:43:54 PM 7
QUESTIONS REPORT
. for 25 SRO Questions
011 Pressurizer Leve~ control
12 Ability to (a) predict the impacts of the following malfunctions or operations on the
PZR LCS; and (b) bas on those predictions, use procedures to correct} control, or
mitigate the consequences of Operation of auxiliary spray
Question Number: 91
Tier 2 Group 2
Importance Rating: 3.3
Technical Reference: UOP-2.2, TRM B 13.4, TS B 3.4.3, STP-35.0
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS52510E06
10 CFR Part 55 Content: 43.5
Comments:
fixed per FJE comments and added some verbiage to stem to clarify where procedurally you
are at. Otherwise the question does not make sense.
MCS Time: 1 Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: BDC C B B DCAA Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: IDGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:43:54 PM 8
QUESTIONS REPORT
for 25 SRO Questions
4. 016 G2.4.31 001
Unit 1 is at 95% power when the following occurred:
- LT-474, 1A SG NR LVL, was declared INOPERABLE and the channel placed
in trip 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ago lAW Tech Specs 3.3.1, Reactor Trip System (RTS)
Instrumentation and 3.3.2, Engineered Safety Feature Actuation System
(ESFAS) Instrumentation.
Subsequently, the card power supply for LT-475, 1A SG NR LVL, failed.
- The following MCB annunciators are in alarm:
- JC1, 1A SG LO-LO LVL ALERT
- JD1, 1A SG HI-HI LVL Alert
- JF1, 1A SG LVL DEV
Which one of the following is the appropriate procedure(s) and actions to be taken for
this condition?
A~ * Enter EEP-O, Reactor Trip and Safety Injection, and then go to ESP-O.1, Reactor
Trip Response.
- Maintain SG levels 31 %-65°/6 when conditions permit.
B. * Enter EEP-O, Reactor Trip and Safety Injection, and then go to ESP-O.1, Reactor
Trip Response. Implement FRP-H.3, Response to Steam Generator High Level,
in conjunction with ESP-O.1.
SGs.
C. * Enter AOP-1 00, Instrumentation Malfunction.
- Return SGWLC to Automatic when conditions permit.
D. * Enter AOP-1 90, Instrumentation Malfunction.
- Reference T.S. 3.3.1 and 3.3.2, and notify the Shift Manager.
Monday, January 14, 20082:43:54 PM 9
QUESTIONS REPORT
for 25 SRO Questions
Meets 10 CFR 55.43 (b) 5 requirements for SRO level question.
A: Correct. With 2 LT$ on one SG less than 28%, a Rx trip w,ill be generated and an
autostart of MDAFWPs is generated. The appropriate path is EEP-O to
No SI signal is generated.
B: Incorrect. With 2 LTs on one SG less than 28%, a Rx trip will be generated and an
autostart of MDAFWPs is generated. If a candidate thought that a card
failure would cause a high level and the card caused a high level condition which is
plausible in that JD1 is in alarm, then this would be an appropriate action to take.
Since the failures listed cause a low level alarm and condition, a Rx trip occurs and
ESP-0.1 actions taken.
C: Incorrect. LT failure meets entry conditions for AOP-1 00 and subsequent required
actions, however, is the incorrect procedure to enter based upon ERG
entry requirement
D: Incorrect. LT failure meets entry conditions for AOP-1 00 and subsequent required
actions, however, is the incorrect procedure to enter based upon ERG
entry requirement
With 2 LTs on one SG less than 280/0, a Rx trip will be generated and an autostart of
MDAFWPs is generated. The appropriate path is EEP-O to ESP-0.1 .
Monday, January 14, 2008 2:43:54 PM 10
QUESTIONS REPORT
for 25 SRO Questions
016 Non-Nuclear Instrurnentation System (NNIS)~
2.4.31 Emergency Procedures I Plan Knowledge of annunciators alarms and indications, and
use of the response instructions.
Question Number: 92
Tier 2 Group 2
Importance Rating: 3.4
Technical Reference: EEP-O and AOP-100
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: . 43.2
Comments:
changed to a different bank question and modified it due to many technical issues.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: AC AAB AAB B C Scramble Range: A - D
Source: MODIFIED Source if Banle FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:55 PM 11
QUESTIONS REPORT
for 25 SRO Questions
5. 022 G2.1.14 002
Given the following:
- Unit 2 is at 100% power
- The following alarms are received:
- HA1, PRZR LVL HI RX TRIP ALERT
- HA2, PRZR LVL DEV HI B/U HTRS ON
- HB1, PRZR LVL HI
- DE1, REGEN HX LTDN FLOW DISCH TEMP HI
- EA2, CHG HDR FLOW HI-LO
- Actual Pressurizer level is 46% and trending DOWN.
- VCT level is 43% and trending UP.
- RCS temperature and pressure are stable.
Which ONE of the following describes the procedure entry required, and a required
notification for the event in progress?
A. * Enter AOP-1 00, Instrument Malfunction;
- Notify the Shift Manager to initiate a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report lAW EIP-8.0,Non-Emergency
Notifications.
B~ * Enter AOP-1 00, Instrumentation Malfunction;
- Initiate a CR. and notify the Work Week Coordinator.
C. * Enter AOP-1.0, RCS Leakage;
- Notify the Shift Manager to initiate a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report lAW EIP-8.0, Non-Emergency
Notifications..
D. * Enter AOP-1.0, RCS Leakage;
- Initiate a CR and notify the Work Week Coordinator.
Monday, January 14,20082:43:55 PM 12
QUESTIONS REPORT
for 25 8RO Questions
Meets 10 CFR 55.43 (b) 5 requirements for 8RO level question.
A. Incorrect. Credible due to correct procedure for a failed LT. This question gives the
indicatio'ns for a failed LT and lAW AOP-1 00 the WWC would be notified and the crew
would initiate a CR. This is not a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report.
AOP entry is found in EIP-8.0 under 20.0 Additional Corporate Duty Manager
Notifications 20.9 Events requiring entry into the EOPs or AOPs
The time .required to notify the CDM is not defined in this section but would be done as
soon as reasonably possible. The point of the above is that a notification is made lAW
EI P-8 and the candidate would have to know that the notification is not a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
notification.
B. Correct. Due to a failed level instrument, AOP-100 would be entered. The following
people need to be notified, both the 8M and the WWC. The reason for the 8M in
the other distracters is not correct since the E-plan does not need to be
implemented for this condition.
C. Incorrect. Credible due to PZR level trend, and this is not a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report.
AOP entry is found in EIP-8.0 under 20.0 Additional Corporate Duty Manager
Notifications 20.9 Events requiring entry into the EOPs or AOPs
D. Incorrect. Incorrect pro.cedure and incorrect person to notify for an AOP-1 00 entry,
but correct for an AOP-1 00 entry.
LT 459 has failed high, BU heaters will be on, Charging flow will go to a minimum
value, and due to letdown still on service with charging at a minimum, DE1 will be in
alarm.
Due to this failure, Pressurizer level is trending DOWN and VCT level is trending up
due to the charging flow to a minimum.
AOP-100 actions
8 Notify the Shift Manager.
9 Submit a Condition Report for the failed level channel, and notify the Work Week
Coordinator (Maintenance ATL on backshifts) of the Condition Report.
Monday, January 14,2008 2:43:55 PM 13
QUESTIONS REPORT
for 25 SRO Questions
022 G2.1.14
APE 022 Loss of Reactor coolant makeup (this affects charging system)
Conduct of Operations: Knowledge of system status criteria which require the notification of
plant personnel..
Question Number: 77
Tier 1 Group 1
Importance Rating: SRO 3.3
Technical Reference: AOP-100 and above ARPs
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.5
Comments:
changed to meet KA for notification requirements and procedural entry.
MCS Time: 1 Points: 1.00 Version: a12 34 5 6 7 89
Answer: BAB B CADDAA Scramble Range: A - D
Source: NEW Source if Banlc
Cognitive Level: HIGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:55 PM
14
QUESTIONS REPORT
for 25 SRO Questions
6. 025 AA2.04 004
Given the* following:
- Unit 1 is in Mode 5 at 120°F with SG manways open to remove nozzle da-ms
after core reload.
- The following alarms are received:
- EC5 - RCS LVL HI-LO
- BE5 - BOP PANELS ALARM
The operator observes the following indications:
- RCS level 123'1" and falling.
- The leak is estimated to be 25 gpm.
- 1A RHR pump flow, amps, and discharge pressure are stable.
Which ONE of the following correctly describes the*procedure required to be entered,
what the procedure will accomplish, and how to apply Technical Specification 3.4.13
. for the conditions above?
A~ * AOP-12.0, Res'idual Heat Removal System Malfunction, is required to be entered
and will id.entify the location of the leak and WILL isolate the leak.
- LCO 3.4.13, RCS Operational LEAKAGE, is NOT applicable in Mode 5.
B. * AOP-12.0, Residual Heat Removal System. Malfunction, is required to be entered
and will identify the location of the leak but will NOT isolate the leak.
- LCO 3.4.13, RCS Operational LEAKAGE, is NOT applicable in Mode 5.
C. * AOP-1.0, RCS Leakage, is required to be entered and will identify the location
of the leak and WILL isolate the leak.
- Enter LCO 3.4.13, RCS Operational LEAKAGE, for IDENTIFIED leakage.
D. * AOP-1.0, RCS Leakage, is required to be entered and will identify the location
of the leak but will NOT isolate the leak.
- Enter LCO 3.4.13, RCS Operational LEAKAGE, for IDENTIFIED leakage.
Monday, January 14, 2008 2:43:55 PM 15
QUESTIONS REPORT
for 25 SRO Questions
Meets 10 CFR 55.43 (b) 5 requirements for SRO level question.
A. Correct ~ lAW EC5, validation of the low level alarm would send the operator to
AOP-12. This procedure applies to mode 4,5 6. AOP~12 will identify where the leak is
and aid in isolting the leak at step 5.
EC5 setpoint
PROBABLE CAUSE
1. Improper RCS level control
2. Im.proper valve lineup
3. RCS leakage
ACTION
2. IF low level condition exists, THEN monitor RHR pump(s) for evidence of cavitation
and if necessary, THEN refer to FNP-1-AOP-12.0, RHR SYSTEM MALFUNCTION.
lAW LE2, the operator could secure the running -pump and then go to SOP-7.0, but this
is not an option. The leak is outside ctmt due to the running sump pumps which shows
the leak to be in the 1A RHR pump room.
The TS for RCS leakage is NOT applicable in modes 5 or 6 but is applicable in modes
1-4.
B. Incorrect-AOP-12 will isolate the leak and this question says AOP-12 will not isolate
the leak.
C. Incorrect-
wrong procedure Since AOP-1.0 is not applicable in- this mode. AOP-1 is only
applicable in Mode 1 - 3
The TS for ReS leakage is NOT applicable in modes 5 or 6 but is applicable in modes
1-4. If .in mode 1-4 then this would be correct.
D. incorrect. wrong procedure. AOP-1 only applicable in Mode 1 - 3
Monday, January 14, 2008 2:43:55 PM 16
QUESTIONS REPORT
for 25 SRO Questions
025 Loss of the RH RS
2x04 Ability to deterrnine and interpret following as they apply to the Loss Residual
Heat Removal System: location and isolability of leaks
Question Number: 78
Tier 1 Group 1
Importance Rating: SRO 3.6
Technical Reference: AOP-12.0 and AOP-1.0
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.5
Comments:
Fixed per FJE comment to include the location of the leak to meet the KA and then the
procedural guidance to be entered to meet the SRO portion of the question.
MCS Time: 1 Points: '1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: ADAB C B DBAD Scramble Range: A - D
Source: BANK Source if Banle
Cognitive Level: . IDGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:55 PM 17
QUESTIONS REPORT
for 25 SRO Questions
7. 029 G2.1.33 002
Given the following: .
- The plant was at 100% power.
- At 1000, Both Reactor Trip Breakers were declared INOPERABLE.
- SSPS has been determined to be operable.
- The crew immediately initiated a plant shutdown.
- At 1025, a reactor trip signal was generated.
- The Reactor Trip Breakers did NOT open.
At 1030, ALL Reactor Trip and Bypass Breakers were verified open.
UOP-2.3, Shutdown of Unit following Reactor Trip, has been entered.
Which ONE of the following correctly describes the mode the unit is allowed to remain
in or must be placed in lAW Technical Specifications and the reason?
A~ * The plant can remain in Mode 3 indefinitely;
- since the RTBs are now open and rod control is no longer capable of rod
withdrawal.
B. * The plant must proceed to Mode 4, but can remain in Mode 4 indefinitely;
- since the RTBs are now open and rod control is no longer capable of rod
withdrawal..
C. * The plant can remain in Mode 3 for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, but must be in Mode 4 in
13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> and Mode 5 in 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />;
- sfnce BOTH RTBs are inoperable and Technical Specification 3.0.3 is in effect.
D. * The plant can remain in Mode 3 for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> while trying to repair one RTB,
but if one RTB cannot be fixed" the plant must be in Mode 4 in 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> and
Mode 5 in 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />;
- since BOTH RTBs are inoperable and Technical Specification 3.0.3 is in effect.
Monday, January 14, 20082:43:55 PM 18
QUESTIONS REPORT
for 25 SRO Questions
Meets 10 CFR 55.43 (b) 2 requirements for SRO level question due to the application
of 3.0.3 and knowledge that 3.0.3 applies in this case and how it applies, specifically.
A. Correct, 3.0.3 no longer applies since the RTBs are opened. The (a) With RTBs
closed and Rod Control System capable of rod withdrawal for modes 3, 4, 5 show that
when the RTBs are open, the spec no longer applies.
B.incorrect - but plausible because 3.0.3 applies until the RTBs are open and has to
be evaluated.
C. incorrect. The plant can remain in Mode 3 indefinitely so the plant can remain in
mode 3 for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, however the plant does not have .to go to mode 4 and 3.0.3 is no
longer in effect due to the RTBs being open.
D is incorrect - The plant can remain in Mode 3 indefinitely so the plant can remain in
mode 3 for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, however the plant does not have to go to mode 4 and 3.0.3 is no
longer in effect due to the RTBsbeing open.
3.3.1
18. Reactor Trip Breakers (j) 1,2 2 trains *R, V
3 (a) , 4 (a) , 5 (a) 2 trains C,V
(a) WithRTBsclosed and Rod Control System capable of rod withdrawal.
v. Two RTS trains V.1 Enter LCO 3.0.3. mediately
LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an
associated ACTION is not provided, or if directed by the associated
ACTIONS, the unit shall be placed in a MODE or other specified
condition in which the LCO is not applicable. Action shall be initiated
within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in:
a. MODE 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />;
b. MODE 4 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and
c. MODE 5 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.
Monday, January 14, 20082:43:55 PM 19
QUESTIONS REPORT
for 25 SRO Questions
EPE 029 A T
G2x 1.33 Conduct of Operations: AbiUty to recognize indications for system operating
parameters which are entry level conditions for technical specifications
Question Number: 79
Tier 1 Group 1
Importance Rating: SRO 4.0
Technical Reference: TS 3.3.1 and 3.0.3
Proposed references to be provided to applicants during examination: no reference
Learning Objective:
10 CFR Part 55 Content: 43.2
Comments:
This was rewritten to incorporate an ATWT into the stem and in such a way as to make TS
entry a requirement to meet. Just entering Mode 3 on an ATWT event with 3.0.3 in effect
would require the plant to be in mode 5, 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> after the RTBs were found to be inoperable
as long as they can not be opened. Since they are opened in the stem, and as expected per
procedure, then the plant is no longer bound to be in mode 4 or 5 and no time limit applies.
MCS Time: 1 Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: AADC CAAAB A Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: IDGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Mon,day, January 14,2008 2:43:55 PM 20
QUESTIONS REPORT
for 25 SRO Questions
8. 035 A2.06 003
Given the following:
- A small break LOCA has occurred on Unit 2.
- RCS pressure is 1550 psig and lowering slowly.
- SG pressu"res are 1000 psig and stable.
- Total AFW flow is 500gpm.
- SG narrow range levels are 5% and rising slowly.
- PRZR level is off scale low.
- Containment pressure is 3 psig and stable.
- The crew is in EEP-1, Loss of Reactor or Secondary Coolant.
Which ONE of the following describes the correct sequence of actiens the crew must
use to cool down the RCS in order to place RHR in service?
A. * Cooldown in accordance with EEP-1 until RCS pressure is less than SG pressure;
RHR entry conditions.
B~ * NO cooldown will be performed in EEP-1 ;
RHR entry conditions.
C. * Cooldown in accordance with EEP-1 until RCS pressure is less than SG pressure;
Standby;
- Then go to UOP-2.2, Shutdown of Unit from Hot Standby to Cold Shutdown, and
cooldown to RHR entry conditions.
D. * NO cooldown will be performed in EEP-1;
- * Go to ESP-1.2, Post LOCA Cooldown and Depressurization, and cooldown to Hot
Standby;
- Then go to UOP-2.2, Shutdown of Unit from Hot Standby to Cold Shutdown, and
cooldown to RHR entry conditions.
Monday, January 14, 2008 2:43:55 PM 21
QUESTIONS REPORT
for 25 SRO Questions
Meets 10 CFR 55.43 (b) 5 requirements for SRO level question.
A. Incorrect. A cooldown is not done in EEP-1. SGWL is maintained and the
procedure transitions to either ESP-1.2 or ESP-1.3 for a small break LOCA.
B. Correct. Entry to ESP-1.2 is required and cooldown to RHR entry is directed in
C. Incorrect. A cooldown is not done in EEP-1. SGWL is maintained and the
procedure transitions to either ESP-1.2 or ESP-1.3 for a small break LOCA. ESP-1.2
does not senq you to UOP-2.2. see below discussion.
D. Incorrect. ESP-1.2"cools the plant down to RHR entry conditions, not Hot Standby
conditions.
Plausibility-
While UOP-2.2 is not required or directed by EOPs, it could be used in part to recover the plant from this
point. It might be directed by the TSC staff to enter at a step that would consider getting the plant into a
condition in which the appropriate procedure would be used. Since UOP-2.2 is cooldown from Hot Stby
and the unit will be on RHR with temp < 200°F, this UOP would not be appropriate at certain steps and
sections being used and others being N/A ed. However, The TSC would Evaluate long term plant
status lAW ESP-1.2 and then look for an appropriate procedure to use to clean up the plant and get back
to operational status. They could decide many different and/or appropriate procedures depending on the
plant conditions.
What is entirely incorrrect with this statement is that ESP-1.2 would not cooldown to
Hot Standby, it actually goes all the way down to 200°F before the appropriate
procedure would be addressed.
035 SG system
.06 Ability to (a) predict the impacts of the following mal-functions or operations on the
S; and (b) base.d on those predictions, use procedures to correct, control~ or mitigate the
consequences of those malfunctions or operations: Small break LOCA
Question Number: 93
Tier 2 Group 2
Importance Rating: 4.6
Technical Reference: EEP-1 and ESP-~.2
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.5
Comments:
fixed per FJE comments
Monday, January 14, 20082:43:56 PM 22
QUESTIONS REPORT
for 25 SRO Questions
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: BDC C C B B CAA Scramble Range: A - D
Source: NEW Source if Banle
Cognitive Level: HIGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:56 PM 23
QUESTIONS REPORT
for 25 SRO Questions
9.037 G2.4.11 001
The following Unit 1 conditions*exist while at 10% power:
The Shift Radiochemist reports the following:
- 1A SG Primary to Secondary Leakage = 148 gpd
- 1B SG Primary to Secondary Leakage*= 185 gpd
- 1C SG Primary to Secondary Leakage = 134 gpd
The OATC reports the following:
- Pressurizer PORV-445A is leaking to the PRT at 2.2 gpm.
Which ONE of the following correctly describes the procedure that must be entered
and the required action and completion time lAW Technical Specification LCO 3.4.13,
A. * Enter AOP-2.0, Steam Generator Tube Leakage.
- Reduce leakage to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
B~ * Enter AOP-2.0, Steam Generator Tube Leakage.
- Be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
C. * Enter AOP-1.0, RCS Leakage.
- Reduce leakage to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
D. * Enter AOP-1.0, RCS Leakage.
- Be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Monday, January 14,20082:43:56 PM 24
QUESTIONS REPORT
for 25 SRO Questions
Meets 10 CFR 55.43 (b) 5 requirements for SRO level question.
ANSWER / DISTRACTOR ANALYSIS
A. Incorrect. All leakage listed in stem is identified leakage. Plausible because
applicant may think that PORV leakage is unidentified and at 2.2 gpm, this would
exceed the limit.
AOP-2 is correct but the actions are not correct.
B. Correct. 150 gpd is the TS limit.
bases 3.4.13-3
"The ReS operational primary to secondary leakage through any
one SG shall be limited to 150 gallons per day." The limit is based
on operating experience with SG tube degradation mechanisms
that result in tube leakage. The operational leakage rate criterion
in conjunction with the implementation of the Steam Generator
Program is an effective measure for minimizing the frequency of
steam generator tube ruptures.
AOP-2 is the correct procedure to enter for the above conditions and the TS is a
immediately go to mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
C. Incorrect. All leakage listed in stem is identified leakage. Plausible for same
reason as A above and AOP-1 could be entered but not for the reasons given. The
action is not correct for a tube leak.
D. Incorrect. Incorrect procedure. Plausible because it is partially correct in that the
action is correct.
REFERENCES
1. Technical Specification 3.4.13, Operational Leakage.
2. Technical Specification 3.4.13 Basis.
AOP-2 lesson plan
The guidance is based on anticipating a tube rupture and is more restrictive than the
required actions of Technical Specifications. If the steam generator leak rate is
determined to be greater than 150 gpd in any steam generator and the unit is in MODE
1 or 2, then the unit is to be placed in MODE 3 per UOP-3.1 and UOP-2.1 , within 6
hours.
Monday, January 14, 20082:43:56 PM 25
QUESTIONS REPORT
for 25 SRO Questions
037 Steam generator tube leakage
G2.4.11 Knowledge of Abnormal operating procedures.
Question Number: original question # 82
Tier 1 Group 2
Importance Rating: 3.6
Technical Reference: AOP-2 and bases 3.4.13-3
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.5 B 5
wrote new question for new KA approved by FJE. 10-30-2007(was KA 059G2.4.30)
KIA MATCH ANALYSIS
This question tests the AOP selection and the TS involved. Since it is an immediate be in
mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> then it is required knowldge and reference is not provided.
Tech Specs can be considered a procedure that is used by the operators. The question tests
the knowledge of whether or not a 'limit is violated. The applicant must have this knowledge in
order to have the ability to execute the Tech Specs. Testing the procedural entry requirement
makes it SRO-only level.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: Be CADDC CDA Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: HIGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:56 PM 26
QUESTIONS REPORT
for 25 SRO Questions
10. 039 A2.01 002
A Large Break LOCA has occurred on Unit 1 with the following conditions:
- B Train is the on service train.
- An LOSP has occurred and B Train emergency power is not available.
- SG pressures are 680 psig and stable.
- Containment pressure rose to 31 psig and is now 8 psig and slowly lowering.
Recirculation, with *the following conditions:
Containment Spray is aligned to the RWST.
1A RHR pump is running with proper flow and is aligned to the containment
sump.
1A Charging pump has tripped on overcurrent.
Which ONE of the following describes the procedure flow path required and the action
that would be taken to reduce SG pressure?
A. * Transition to ECP-1.1, Loss of Emergency Coolant Recirculation;
- Dump steam from the SGs using the steam dumps and maintain the cooldown
rate less than 1OQoF per hour.
B. * Transition to ECP-1.1, Loss of Emergency Coolant Recirculation;
- Dump steam from the SGs using the Atmospheric Relief Valves and maintain the
cooldown rate less than 100°F per hour.
C. * Continue in ESP-1.3 and align the CS system for recirculation, then transition
back to EEP-1.0, Loss of Reactor or Secondary Coolant;
- Dump steam from the SGs using the steam dumps at the maximum attainable
rate.
D~ * Continue in ESP-1.3 and align the CS system for recirculation, then transition
back to EEP-1.0, Loss of Reactor or Secondary Coolant;
- Dump steam from the SGs using the Atmospheric Relief Valves at the maximum
attainable rate.
Monday, January 14, 2008 2:43:56 PM 27
QUESTIONS REPORT
for 25 SRO Questions
Meets 10 CFR 55.43 (b) 5 requirements for SRO level question.
A. incorrect - The ste*p in ESP-1.3 (7.27) has the crew verify SI flow is stable. If it is,
then continue in ESP-1.3 and place CS on recirculation, then go to procedure and step
in effect. Since a LBLOCA has occurred, EEP-1 would be the procedure used to get to
ESP-1.3. Since the 1A RHR pump is running with proper flow, SI flow will be stable
and the transition to ECP-1.1 not required. The way the step is written, as shown
below, could confuse the candidate in that they could assume all SI flow)s stable when
only one is required to be stable.
7.27 Verify SI flow - STABLE. 7.27 IF at least one train of flow
from the containment sump to
ATRN the RCS* can NOT be
HHSI FLOW established or maintained,
[] FI 943 THEN go to FNP-1-ECP-1.1,
LOSS OF EMERGENCY COOLANT
HHSI RECIRCULATION.
B TRN RECIRC
FLOW
[] FI 940
1A(1 B)
FLOW
[] FI 605A
[] FI 605B
If the applicant decided ECP-1.1 was the proper procedural flow path, then the two
choices of using dumps or ARVs is available and the CDR would be correct for this
procedure. Dumps are not available since CTMT pressure went to 31 psig and the
MSIVs are closed.
B. incorrect - see above
C. incorrect - since the MSIVs would be closed due to the LBLOCA and the LOSP, the
dumps would not and could not be used. Our design of the MSIVs do not allow them to
be re-opened until' a 50 PSID is reached across the valve.
D. Correct - The step in ESP-1.3 (7.27) has the crew verify SI flow is stable. Since it
is, then they continue in ESP-1.3 and place CS on recirculation, then go to procedure
and step in effect. Since a LBLOCA has occurred, EEP-1 would be the procedure
used to get to ESP-1.3 and would be returned to. Then EEP-1 has the crew decide to
release pressure from the SGs to decrease the delta P across the tubes. Since the
dumps are not available, the ARVs would be used. The max attainable rate is
procedurally driven by EEP-1 and makes for a great distracter as well because
someone not familiar with the reason for decreasing pressure at this time would be
confused and would probau.bL'lly----lr.t::-ellthlllinJ1k~et::JnLUt..t:::e_llriUJng~Ed:E~P~--I--1.----------------;;-----------
Monday, January 14,20082:43:56 PM 28
QUESTIONS REPORT
for 25 SRO Questions
039 A2.01 Main and Reheat Steam
Ability t (a) predict the hllpacts of the following mal-functions or operations on the Main
Reheat Stearn System; and (b) based on predictions, use procedures to correct,. control, or
rnitigate consequences of those malfunctions or operations: Flow paths of steam during a
lOCA~
Question Number: 87
Tier 2 Group 1
Importance Rating: 3.2
Technical Reference: E-1 and ESP-1.3
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.5
Comments:
This question was rewritten in entirety. One reason is the original submitted question did not
meet the KA. I had to rewrite it to a LB LOCA since 035 A2.06 on this exam tests the
procedural transition to ESP-1.2. This would have been ideal for this question but was double
jeopardy. I did not find a procedural transition question to ECP-1.1 or back to EEP-1 on this
exam.
This question also had to deal with steam flows during a LOCA. There is no steam flow in
EEP-1 except at step 18 which depressurizes the SGs. Since ESP-1.2 has been taken away
by a previous question, the logical step was to go to step 18. To get the procedural transition
piece, I had to place enough in the stem for the applicant to analyze and to decide which
procedure would be best.
If this is not satisfactory, I will need another suggestion or a replacement KA.
We ran on simulator to get proper pressures and temperatures
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: DAB B A CAD D D I Scramble Range: A-D*
Source: NEW Source if Bank:
Cognitive Level: HIGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January*14, 20082:43:56 PM 29
QUESTIONS REPORT
for 25 SRO Questions
11 . 040 AA2.0 1 002
Given the following on Unit 1:
- Reactor Trip and Safety Injection have 'occurred.
- RCS Pressure is 2010 psig and DECREASING.
- Pressurizer level is 22% and rising.
- Containment Pressure is 16 psig and INCREASING.
- 1A SG Pressure is 520 psig and DECREASING.
- 1Band 1C SG pressures are 840 psig and STABLE.
- Sub Cooled Margin Monitor is reading 130°F.
Which ONE of the following describes the location of the break and the next procedure
the crew will perform after transition from EEP-O.O, Reactor Trip or Safety Injection?
A. Downstream of 1A MSIV; ESP-1.1, SI Termination.
B. Downstream of 1A MSIV; EEP-2.0, Faulted Steam Generator Isolation.
C. Upstream of 1A MSIV; ESP-1.1, SI Termination.
D~ Upstream of 1A MSIV; EEP-2.0, Faulted Steam Generator Isolation.
Meets 10 CFR 55.43 (b) 5 requirements for SRO level question.
A. incorrect because the break is in the wrong place, as indicated by containment
pressure.
B. incorrect because the break is in the wrong place, as indicated by containment
pressure.
C. incorrect due to incorrect procedure, but credible because the break is in the correct
place and the procedure would be correct for a downstream break.
D. correct. Indications of a Faulted SG upstream of MSIV due to containment
pressure. EEP-2 will be addressed because the SG will eventually depressurize.
Monday, January 14,2008 2:43:56 PM 30
QUESTIONS REPORT
for 25 SRO Questions
040 Steam line rupture
AA2.01 Ability to determine and interpret the following as they apply to the Steam Rupture:
Occurrence and location of a steam line rupture from pressure and flow indications
Question Number: 80
Tier 1 Group 1
Importance Rating: SRO 4.7
Technical Reference: EEP-O.O
Proposed references to be provided to applicants*during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.5
Comments:
The parameters were picked based on running this event on the simulator and picking
hypothetical values this could happen depending on the reaction time of the crew and a small .
steam break.
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: D B B D C B C B B B Scramble Range: A - D
Source: NEW Source if Banle
Cognitive Level: HIGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:56 PM 31
QUESTIONS REPORT
for 25 SRO Questions
12. 061 AA2.03 001
Given the following:
- Unit 1 is at 100% power.
- A fuel shuffle is in progress in the SFP room.
- The following alarm is received:
- FH1, RMS HI-RAD
The OATC reports the following:
- R-5, Spent Fuel Pool Area Monitor, Red HIGH alarm light is LIT.
- R-25A and R-25B, SFP VENT, radiation monitor
- Amber ALERT light is LIT.
- Red HIGH alarm light is NOT LIT.
Which ONE of the following describes the current status of Spent Fuel Pool Supply and
Exhaust Fans, and the actions that will be required lAW FH1, RMS HI-RAD?
A. * Spent Fu~1 Pool Supply and Exhaust fans are running;
- Implement EIP-9, Emergency Actions, determine if Automated Rapid Dose
Assessment (ARDA) has actuated, and verify both trains of PRF running.
B. * Spent Fuel Pool Supply and Exhaust fans are tripped;
- Implement EIP-9, Emergency Actions, determine if Automated Rapid Dose
Assessment (ARDA) has actuated, and verify both trains of PRF running.
C. * Spent Fuel Pool Supply and Exhaust fans are tripped;
- Enter AOP-30.0, Refueling Accident, isolate the Control Room and place
Control Room Emergency Filtration system (CREFs) in service.
D~ * Spent Fuel Pool Supply and Exhaust fans are running;
- Enter AOP-30.0, Refueling Accident, isolate the Control Room and place the
Control Room Emergency Filtration system (CREFs) in service.
Monday, January 14, 2008 2:43:56 PM 32
QUESTIONS REPORT
for 25 SRO Questions
Meets 10 CFR 55.43 (b) 4 and 5 requirements for SRO level question.
A. incorrect. because the actions are inco'rrect for the event taking place. Plausi'ble
because it is an action that would be performed for different conditions. ARDA will
activate when R-25A/B go red, and then EIP-9 would be referred to. Both trains of PRF
would actuate for R-25A/B high alarm
SFP exhaust goes to the AB exhaust plenum which feeds the plant vent stack. The
plant vent stack is monitored by R-14, 29 and 22 which would activate ARDA.
EIP-9.1
ARDA will automatically start when any of the following monitors go into
alarm for two consecutive system polls one minute apart on the applicable
unit and use the latest 15 minute average monitor value to perform the
calculations:
Monitor Setpoint
Plant Vent Stack R29 (SPING)
Noble Gas 4.44e-4 IJc/ml
Iodine 1.20e-6 IJc/ml
Particulate 4.00e-5 IJc/ml
Steam Jet air Ejector R15C .027 RIhr
TDAFW Exhaust R60D .038 RIhr
Steam Generator AlBIC R60AIBIC .038 R/hr
ARDA will also automatically start when any of the following monitors
go into alarm for two consecutive system polls one minute apart on the
applicable unit. The ARDA system will use the plant Vent stack SPING
latest 15 minute average monitor value to perform the calculations when
these monitors activate the system:
Monitor Setpoint
Plant Vent stack Monitors
Gas monitor R 14 13000 (VI) 11571 (V2) CPM
Particulate monitor R 21 1800 (VI) 4280 (U2) CPM
Gas monitor R 22 156 (VI) 143 (V2) CPM
B. incorrect. R-25A or B ,RED alarm light realigns FHB ventilation NOT the AMBER
alert light.
C is incorrect. because FHB fans are running. Credible because R-25A/B RED alarm
setpoint has not been reached, so applicant may think FHB ventilation has realigned.
AOP-30 directs Control Room isolation and starting CREFs .
D. Correct. because FHB fans are running. Credible bec,ause R-25A/B RED alarm
setpoint has not been reached, .AOP-30 directed Control Room isolation and starting
Monday, January 14, 2008 2:43:56 PM 33
QUESTIONS REPORT
for 25 SRO Questions
061 Area Rad Monitoring alarms:
AA2.03 Ability to determine and interpret the following as they apply to the Area Radiation
Monitoring (ARM) System Alarms: Setpoints for alert and high alarms .
Question Number: 83
Tier 1 Group 2
Importance Rating: 3.3
Technical Reference: ARP FH5.and FH1 and EIP-9.1; U258400; AOP-30; A181015;
OPS-52106D
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.5/6 and 55.43(b) (4 and 5)
Comments:
Rewrote to FJE comments.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: DCA C D DAD B D Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: HIGHER Difficulty:
Job Position: SRO Plant: ' FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:56 PM 34
QUESTIONS REPORT
for 25 SRO Questions
13. 062 G2.1.33 002
Given the following:
- Unit 1 is in Mode 1.
- WD2, 1B INV FAULT, comes into alarm.
- The ROVER reports the following indications on the 1B Inverter panel:
- The BYPASS SOURCE POWERING LOAD light is LIT.
~ The' INVERTER POWERING LOAD light is NOT LIT.
- The battery input breaker has tripped open.
- The BYPASS SOURCE AVAILABLE light is LIT.
Which ONE of the following statements describes the Technical Specification ACTION
statement(s) that MUST be entered?
Art * Enter the TS LCO action statement for 3.8.7, Inverters - Operating.
- LCO 3.8.9, Distribution Systems - Operating, entry is NOT required.
B. * Enter the TS LCO action statement for 3.8.7, Inverters - Operating.
- Enter the TS LCO action statement for 3.8.9, Distribution Systems - Operating.
C. * LCO 3.8.7, Inverters - Operating, entry is NOT required.
- Enter the TS LCO action statement for LCO 3.8.9, Distribution Systems -
Operating.
D. * LCO 3.8.7, Inverters - Operating, entry is NOT required.
- LCO 3.8.9, Distribution Systems - Operating, entry is NOT required.
Monday, January 14, 2008 2:43:56 PM 35
QUESTIONS REPORT
for 25 SRO Questions
Meets 10 CFR 55.43 (b) 2 requirements for SRO level question. This is SRO only in that
a candidate would have to look at bases to determine whether 3.8.7 and 3.8.*9 apply. The
definitions of operability are found in bases..
A. Correct.
Since the 1B Inverter has lost the DC source, the indications above show that the
inverter swapped to the bypass source. Since this is true, the 1B vital panel is still
energized. T8 3.8.7 applies and Condition A has the SRO look at applicability of 3.8.9.
If the inverter did not swap per design and the vital panel was de-energized, then 3.8.9 .
. would be required to be entered also. The reason the inverter is INOPERABLE is blc
of the DC source is required to be the primary source of power to the inverter. Ac is
just the backup. I *
The vital panel is operable since it is powered from an inverter. 3.8.9 says the vital
panel can be powered from an inverter that is powered from either AC or DC source..
3.8.7
Operable inverters require the associated vital bus to be powered by the inverter with output
voltage and frequency within tolerances, and power input to the inverter from a 125 VDC
station battery.
~.1 ------------~OTE:------------
E:nter applicable Conditions and Required ~ctions of LCO 3.8.9,
IIDistribution System~ - Operating" with any vital bus deenergized.
bases of 3.8.7 .
With a required inverter inoperable, its associated ~C vital bus becomes inoperable until it is
re-energized from its Class 1E: CVT. For this reason a ~ote has been included in Condition ~
requiring the entry into the Conditions and Required ~ctions of LCO 3.8.9, "Distribution
Systems-Operating." This ensures that the vital bus is re-energized within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The'
associated static transfer switch normally provides a bumpless transfer of power to the
alternate ~C source (Class 1E CVT).
3.8.9
OPE:R~BLE vital bus electrical power distribution subsystems require the associated buses to
be energized to their proper voltage from the associated inverter via inverted DC voltage or
Class 1E: constant voltage transformer.
B. Incorrect. 3.8.7 is entered and 3.8.9 is NOT.
C. Incorrect. 3.8.7 is entered and 3.8.9 is NOT.
D. Incorrect. 3.8.7 is entered and 3.8.9 is NOT.
Monday, January 14, 2008 2:43:57 PM 36
QUESTIONS REPORT
for 25 SRO Questions
062 AC electrical distribution -
G221.33 Conduct of operations: Ability to recognize indications for systerTI operating
parameters which are entry conditions for technical specifications
Question Number: 88
Tier 2 Group 1
Importance Rating: 4.0
Technical Reference: TS 3.8.7, 3.8~9 and bases
Proposed references to be provided to applicants during examination: None-
Learning Objective:
10 CFR Part 55 Content: 43.2
Comments:
This KA tests the recognition of entry conditions to TSs and isSRO in that bases knowledge
has to be understood for the TS referenced and also detailed knowledge of the note in LCO
3.8.7 that sends the SRO to 3.8.9 and why.
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: AD C C B D C BCD Scramble Range: A - D
Source: NEW Source if Banle
Cognitive Level: illGHER Difficulty:
Job Position~ SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:57 PM 37
QUESTIONS REPORT
for 25 SRO Questions
14. 073 A2.01 001
Given the following:
- Unit 1 is at 100% power.
- The following alarm is received:
R-23S, SGSD TO DI*LUTION, radiation monitor is indicating downscale with all
indicating lights extinguished.
Which ONE of the following describes the effect of the failure and the associated
ODeM requirement?
A. * FGV-1152, SGSD Heat Exchanger Discharge valve, will close;
- SGSD releases to the environment can NOT continue.
B. * FCV-1152, SGSD Heat Exchanger Discharge valve, will close;
- SGSD releases can continue provided chemistry analyzes grab samples.
C~ * RCV-23S, SGSD Dilution Discharge valve, will close;
- SGSD releases can continue provided chemistry analyzes grab samples.
D. * RCV-23S, SGBD Dilution Discharge valve, will close;
- SGSD releases to the environment can NOT continue.
Meets 10 CFR 55.43 (b) 2 requirements for SRO level question in the realm of the
ODCM requirements when a radiation monitor fails.
A. incorrect. FCV-1152, SGSD Heat Exchanger Discharge valve will not close for this
rad monitor. credible due to R-23A will close 1152. The release can be continued.
Purification Outlet Radiation Monitor (RE-23A)
The purification outlet radiation monitor determines the activity level of the fluid
entering the surge tank. When the demineralizer train is bypassed, this instrument indicates the
activity of the untreated blowdown fluid. If the blowdown is being processed, this instrument
will indicate a radioactive breakthrough across the demineralizers. In any case, a high activity
signal from RE-23A closes FCV-1152, which stops the blowdown. Indication and a high alarm
are on the radiation monitoring system (RMS) panel in the main control room.
B. Incorrect - FC'v'-1152, SGSD Heat Exchanger Discharge valve will not close for this
rad monitor. The action listed is correct.
c. correct. since the rad monitor gives a high signal upon a loss of power, the actions for the
high alarm will occur. This will close RCV-23B.
Monday, January 14,20082:43:57 PM 38
QUESTIONS REPORT
for 25 SRO Questions
FH2 AUTOMATIC ACTION
1. The radiation monitors fail to a "High Radiation" condition on loss of
instrument and/or control power that-will result in actuation of associated
automatic functions. Refer to annunciator FH1 for automatic actions.
ODeM requirements
Instrument Minimum Channels
OPERABLE ACTION
. Gross Radioactivity Monitors Providing Automatic Termination of Release
a. Liquid Radwaste Effluent Line (RE-18) 1 28
b. Steam Generator Slowdown Effluent Line 1 29
(RE-23S)
ODCM page 2-4
ACTION 29 - With the num.ber of channels OPERABLE less than required by the Minimum'
Channels OPERABLE requirement, effluent releases via this pathway may continue,
provided grab samples are analyzed for gross radioactivity (beta or gamma) at a
MINIMUM DETECTABLE CONCENTRATION no higher than 1 x 10 -7 micro Ci/mL.
Discharge Radiation Monitor (RE-23Bl
RE-23B monitors the activity level of the fluid"leaving the SGBD. A high activity signal
from this instrument closes RCV-23B, which prevents the discharge of high activity fluid to the
environment. Indication and a high alarm are located on the RMS panel in the main control
room.
D is incorrect since the release can continue.
FH1 guidance
R-23A SG Slowdown Surge Tank Liquid Scint. (W ) Closes Perform
Step
Inlet (AS 1301) FCV-1152 4.23
R-23B SG Blowdown Surge Tank Liquid Scint. (W ) Closes 'Perform
Step
ODCM Discharge (AB 130 1) RCV-23S 4.23
Monday, January 14,20082:43:57 PM 39
QUESTIONS REPORT
for 25 SRO Questions*
073 Process radiation monitoring
A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the PRM
system; and (b) based on those predictions, use procedures to correct~ control, or mitigate the
consequences of those malfunctions or operations:
Erratic or failed power supply
Question Number: 89
Tier 2 Group 1
Importance Rating: 2.9
Technical Reference: ARP-1.6, FH2, FH1 SGSD lesson plan and aDCM page 2-4
Proposed references to be provided to applicants during examination: None
Learning Objective: .
10 CFR Part 55 Content: 43.4
Comments:
This was rewritten to incorporate comments from FJE and made the ODCM applicable to make
it SRO level
MCS Time: Points: 1.00 Version: 0 12 3 4 5 6 7 8 9
Answer: C C A CAB C C A A Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: HIGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:57 PM 40
SRO QUESTION 076G2.1.2
Distractor "8" is also a correct answer. See examination report 05000348/2007301 and
05000364/2007301 Enclosure 2.
QUESTIONS REPORT
for 25 SRO Questions
15. 076 G2.1.2 001
Given the following:
- Unit 2 is at 100% power withllA II Train on service.
- At 1200 on 11/7/2007, 2E Service Water pump tripped and IIBII Train SW was
declared INOPERABLE.
Which ONE of the following describes the Technical Specification REQUIRED ACTION
lAW 3.7.8, Service Water System, and the action required to make IIB II Train Service
Water OPERABLE?
Art * Immediately declare the DG supported by Train IIB II Service Water INOPERABLE.
pump lAW SOP-24.0, Service Water System.
B. * Immediately declare the DG supported by Train IIB II Service Water INOPERABLE.
Water.
C. * Declare the DG supported by Train IIBII Service Water INOPERABLE no later than
1600 on 11/7/2007 (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> later).
Water.
D. * Declare the DG supported by Train IIB II Service Water INOPERABLE no later than
1600 on 11/7/2007 (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> later).
pump lAW SOP-24.0, Service Water System.
Meets 10 CFR 55.43 (b) 2 and 5 requirements for SRO level question
A. Correct. This TS im.mediately entered from 3.7.8 and the DG is declared INOP.
Then the 2C SW pump is aligned to auto start for 2E. The trains are swapped to do
this. This will allow both trains to be operable. *
CAUTION: Based on plant needs, shifting electrical trains in FNP-1-S0P-24.0,
SERVICE WATER SYSTEMS, may be delayed. Subsequent shifting of
electrical trains is required for train separation.
19 IF affected train NOT leaking, THEN evaluate aligning 1C SW pump to affected train using
FNP-2-S0P-24.0, SERVICE WATER SYSTEM.
Bases 3.7.8
LCO Two SWS trains are required to be OPERABLE to provide the required redundancy to
ensure that the system functions to remove post accident heat loads, assuming that the worst
Monday, January 14, 2008 2:43:57 PM 41
QUESTIONS REPORT
for 25 SRO Questions
case single active failure occurs coincident with the loss of offsite power.
An SWS train is considered OPERABLE during MODES 1, 2, 3, and 4 when:
a. Two pumps are OPERABLE; and
b. The associated piping, valves, and instrumentation and controls
required to perform the safety related function are OPERABLE.
Note from A.1
The first Note indicates that the applicable Conditions and Required Actions of LCO 3.8.1, "AC
Sources-Operating," should be entered if an inoperable SWS train results in an inoperable
FSD 181001
3.1.5.1 The Service Water pumps shall be automatically started by a signal from the LOSP or
ESS sequencer. The Service Water swing pump shall be automatically started by a signal from
the LOSP or ESS sequencer when in service replacing one of the train oriented pumps.
(References 6.7.039 and 6.1.009)
SOP-24 P&L
3.3 Service Water pump Ie may be selected for auto-start from the ESS or the LOSP
sequencers, instead of an A Train or B Train pump, by using key-interlocked
selector switches at the SW local control panels. Normal position of both the A
Train and B Train selector switches will be the lC position and lC SW pump will
not autostart.
B. Incorrect.
This TS is immediately entered from 3.7.8 and the DG is declared INOP.
The second part is in part correct but B Train would be however AOP-1 0 sends the
operator to SOP-24 to select the 2C SWP to autostart and if this was done wlo
swapping trains it would be in an incorrect alignment. This has to be done in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
(3 days later) lAW TS 3.7.8. NOT 7 days. .
c. incorrect.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> would NOT be allowed to declare inop if DG was OOS.
The second part is in NOT correct. See above.
D. incorrect.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> would NOT be allowed to declare inop if DG was OOS.
Second part of this is correct.
Monday, January 14, 2008 2:43:57 PM 42
QUESTIONS REPORT
for 25 SRO Questions
076 Service Water System
G2.1 .2 Conduct of Operations: Knowledge of operator responsibilities during all modes of
plant operation.
Question Number: 90
Tier 2 Group 1
Importance Rating: 4.0
Technical Reference: TS 3.7.8,3.8.1, AOP-10, SOP-24
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.2
Comments:
fixed per FJE comments and added how to restore B train to operable status.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: ADD D A C C BCD Scramble Range: A - D
Source: NEW Source if Banle
Cognitive Level: ffiGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:57 PM
43
QUESTIONS REPORT
for 25 SRO Questions
16. E03 02.4.4 006
A spurious SI has occurred on Unit 1 with the following conditions:
- All systems functioned as required.
- ALL but one charging pump has been secured lAW EEP-O, Reactor Trip or
Safety Injection.
After establishing normal charging in EEP-O, RCS pressure started to decrease and
PRZR level started trending down from 35% and is now 14%.
Which ONE of the following describes the actions and procedural transition the SRO
must direct at this point?
A. Reinitiate a manual Safety Injection and transition back to step 1 of ~EP-O.
B. Return to the diagnostic steps (13 through 15) of EEP-O, and then transition to
EEP-1, Loss of Reactor or Secondary Coolant.
C. Transition to ESP-1.1, SI Termination, step 6, and apply the foldout page of
ESP-1.1 to re-establish HHSI flow.
D~ Re-establish HHSI flow per EEP-O, and then transition to E8P-1.2, Post LOCA
Cooldown and Depressurization.
Meets 10 CFR 55.43 (b) 5 requirements for SRO level question
A - Incorrect; If PZR level can not be maintained, the flow path must be reestablished
and a transition to ESP-1.2 is warranted. There is no need to manually 81 and
transition back to step 1 of EEP-O.
B - Incorrect; returning to diagnostics once they have been completed and additional
actions taken may seem plausible to get to EEP-1, but this is not allowed per sop-O.8
procedural use guidelines procedure. Also the RNO column of EEP-O directs the
correct actions for this condition.
C - Incorrect; the very next step of EEP-O sends the user to ESP-1.1. While this would
eventually lead the crew to the right place, ESP-1.1 foldout page has the operator go to
EEP-1 after the HHSI was re-established. This may seem to be a method to use but is
'not procedurally correct.
D - Correct; From EEP-O, step 19 and 21, RNO, says that if PZR level or pressure
cann'ot be maintained, the procedural requirement is to go to E8P-1.2.
Monday, January 14, 2008 2:43:57 PM 44
QUESTIONS REPORT
for 25 SRO Questions
E03' lOCA cooldown and depressurization
G2.4.4 Emergency Procedures / Plan Ability to recognize abnormal indications for system
operating parameters which are entry-level conditions for emergency and abnormal
operating procedures.
Question Number: B4
Tier 1 Group 2
Importance Rating: 4.3
Technical Reference: EEP-O and SOP-O.B
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.5
Comments:
This question was written to address the double jeopardy transition issue and not giving away
answers by other questions with 035 A2.06. Instead of using EEP-1 to transition to ESP-1.2,
EEP-O is being used to give a similar procedural transition.
MCS Time: 1 Points: 1.00 Version: 0 12 3 4 5 6 7 8 9
Answer: DCAB CDBAAA Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: mGHER Difficulty:
Job Position: SRO
Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:57 PM
45
QUESTIONS REPORT
for 25 SRO Questions
17. E04 G2.2.22 001
Given the following:
- Reactor trip and safety injection have occurred on Unit 1.
- ECP-1 .2, LOCA Outside Containment; has been completed.
- UOP-2.1, Shutdown of Unit from Minimum Load to Hot Standby, is in progress.
- PRZR level is 75% and increasing slowly.
- 1B RCP is operating.
- 1A and 1C RCPs are secured.
- RCS Tavg is 526°F and increasing slowly.
Which ONE of the following describes the Technical Specification LCO action
statement in effect for the given conditions and the basis for the LCO?
A~ * 3.4.9, Pressurizer;
- To maintain pressure control to minimize the consequences of potential
overpressure transients.
B. * 3.4.9, Pressurizer;
- To maintain RCS subcooling during natural circulation conditions.
C. * 3.4.5 RCS Loops - Mode 3;
- To ensure adequate decay ~eat removal from the core in the event of an
inadvertent control rod withdrawal.
D. * 3.4.5 RCS Loops - Mode 3;
- To ensure adequate decay heat removal from the core and proper boron mixing
throughout the RCS.
Monday, January 14, 2008 2:43:57 PM 46
QUESTIONS REPORT
for 25 SRO Questions
Meets 10 CFR 55.43 (b) 2 requirements for SRO level question
A is correct.
Since the PZR must be kept below 63.5% in mode 3. Adequate volume for a steam
bubble for pressure control is the reason for the 63.5% max pzr level in Mode 1-3.
The pressurizer shall be OPERABLE with:
a. Pressurizer water level < or equal to 63.5% indicated;
Bases for LeO 3.4.9 page 83.4.9-2
The LCO requirement for the pressurizer to be OPERABLE with a water volume = 868 cubic
feet, which is equivalent to 63.5% indicated, ensures that a steam bubble exists.' Limiting the
LCO maximum operating water level preserves the steam space for pressure control.
The LCO has been established to ensure the capability to establish and maintain pressure
control for steady state operation and to minimize the consequences of potential
overpressure transients. Requiring the presence of a steam bubble is also consistent with
analytical assumptions.
B. is incorrect. Basis for PZR heaters.
C. is incorrect. 3.4.5 is not in effect since the RTBs are open and there is one RCP
running.
Two RCS loops shall be OPERABLE, and either:
a. Two RCS loops shall be in operation when the Rod Control System is capable of rod
withdrawal; or
b. One RCS loop shall be in operation when the Rod Control System is not capable of rod
withdrawal.
b'ut basis for that spec would be correct in Mode 2 with RTBs closed
D. is incorrect. Incorrect spec (see above), but correct basis for operability
requirements with 1 RCP, RTBs open.
Monday, January 14, 2008 2:43:57 PM 47
QUESTIONS REPORT
for 25 SRO Questions
E04 lOCA outside ctmt
G2.2.22 Equipment Control: Knowledge Limiting Conditions for Operations and safety limits
Question Number: 81
Tier 1 Group 1
Importance Rating: SRO 4.1
Technical Reference: TS 3.4.9 & 3.4.5
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.2
Comments:
Changed per FJE comments note from es-401-9 below:
Exal'Diner Note: Question meets first half of KIA (LOCA outside of containment) because the
event is necessary to provide a credible context for the given plant conditions.
MCS Time: Points: 1.00 Version: 0 12 3 4 5 6 7 8 9
Answer: AD B A A B D BCD Scramble Range: A - D
Source: NEW Source if Banle
Cognitive Level: HIGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:58 PM
48
QUESTIONS REPORT
for 25 SRO Questions
18. E15 EA2.1 005
Given the following:
- A Large Break LOCA has occurred on Unit 1.
- 'RWST level is 14.8 feet and slowly lowering.
The crew is at step 15 in EEP-1.0, Loss of Reactor or Secondary Coolant, to check
LHSI flow in progress, when the following containment indications are reported by the
OATC:
- FI-958A, CS flow, reads a gpm.
- FI-958B, CS flow reads 1850 gpm.
- Ctmt Pressure is 29.5 psig and rising slowly.
- Ctmt Sump Level 8.0 feet and rising slowly.
- Ctmt Radiation Level is 3.6 Rem/Hr on both High Range instruments.
Which ONE of the following 'describes the next action to take for these conditions?
A. Implement FRP-Z.1, Response to High Containment Pressure.
B~ Implement FRP-Z.2, Response to Containment Flooding.
C. Implement FRP-Z.3, Response to High Containment Radiation.
D. Transition to ESP-1 .3, Transfer to Cold Leg Recirculation.
Meets 10 CFR 55.43 (b) 2 requirements for SRO level question
A. Incorrect. 27 psig is ORANGE Path on pressure. Plausible, because if flow for CS
dropped below 1000 gpm this would be the correct procedure on an orange path. As is
it would be a yellow path IF CTMT sump level was less than 7.6 feet.
B. Cor'rect. Since CS flow is > 1000 gpm, and Ctmt sump level> 7.6 feet, this is an
orange path on Z.2.
C. Incorrect. This is a yellow path in the same CSF network and the c,andidate has to
know it is a yellow path and it is below or less critical than Ctmt sump level. This would
be a correct choice if ctmt sump level is < 7.6 feet.
D.lncorrect. Once ESP-1.3 is entered, no CSF applies. In this case, the crew is
holding at step 160f EEP-1.0~aitingforR~ST level todFopbel~~2u.5~f~e~e~ta~r~ld~th~e~n~~~~~
transition to ESP-1.3. Due to the RWST level at 14.8 feet the crew would not wait on
the transition but would enter FRP-z.2 until ESP-1.3 was required.
Monday, January 14, 2008 2:43:58 PM 49
QUESTIONS REPORT
for 25 SRO Questions
E15 CTMT flooding
EA2.1 Ability to determine and interpret the following as they apply to (Containment
Flooding) Facility conditions and selection of appropriate procedures during abnormal and
emergency operations.
Question Number: 85
Tier 1 Group 2
Importance Rating: 3.2
Technical Reference: CSF-O.5, CSFSTs
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS 52533M02
10 CFR Part 55 Content: 43.5
Comments:
Rewrote the question to specifically address the KA. With the conditions given the candidate
has to evaluate the CSFs and determine the appropriate procedure to go to which is tied to the
KA.
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: B A C B C B B B A C Scramble Range: A - D
Source: MODIFIED Source if Bank: FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:58 PM
50
QUESTIONS REPORT
for 25 SRO Questions
19. G2.1.25 004
Unit 1 is in a refueling outage. The following conditions exist:
- Nozzle dams are installed on ALL Steam Generators.
- Both trains of RHR are in service.
1
- RCS level is 123 3 11 and stable.
- RCS temperature is 120°F and stable.
- Secondary side of all SGs is > 85% wide range.
- A charging pump is in service; Band C are tagged out.
- The equipment hatch is closed.
- The time to saturation is 35 minutes.
The OATC reports that both RHR pumps have just tripped due to breaker problems.
Using UOP-4.0 Appendix 1, SHUTDOWN SAFETY ASSESSMENT, which one of the
following is the correct procedure to go to and proper condition based on the events in
progress?
.References Provided
C. Go to AOP-45, SHUTDOWN INVENTORY, under an Orange condition.
D. Go to AOP-45, SHUTDOWN INVENTORY, under a Red condition.
Monday, January 14, 2008 2:43:58 PM
51
QUESTIONS REPORT
for 25 SRO Questions
References: UOP-4.0 Appendix 1, figure 1a page 6 of 16 Version 28
Meets 10 CFR 55.43 (b) requirements for SRO level question in that this is a task only
A. Incorrect- This is a red condition not orange. plausible since time to saturation is a 1
and if the SGs were evaluated improperly, then an orange condition would be selected.
B. Correct- RED
25 min to saturation based on Table B for 100°F. The SG tubes are not filled and
vented which gives them a 0 and #5 is not met as well. With the loss of the RHR
system, AOP-42 would be entered~ Even if a 1 was entered for #5, the condition would
still be an unexpected RED.
CORE COOLING Subtotal Condition AOP
1. ~ 2 SOs Avail with loops filled (Ref step 2.7) _ _0_ 0-1 RED 42
2. Cavity level = 142'1" wI Upper Internals Removed _--",-0_ 2-3 ORANGE 42
3. RHR Subsystems Available (0, 1 or 2) _ _0_ 4 YELLOW
0__ ~5 GREEN
4. RCS level = 126' 6"
1 _ (GREEN if Defueled)
5. Time to saturation> 30 minutes OR RCS press> 325 psig with at least
one RCP available for operation and at least one SO available _ _1_
Core Cooling Subtotal
C. Incorrect- yellow is correct evaluated condition for Inventory, not orange and then if it
was yellow the AOP would not be addressed.
D- Incorrect- yellow is correct. Plausible b/cthe HHSI flow path is not identified and not
in use and could be considered not available, especially in a shutdown mode. The RCS
is intact since the nozzle dams are installed.
INVENTORY Subtotal Condition AOP
1. Refueling Cavity ~ 23 Feet (142' I") Above Fuel °1 'RED
ORANGE
45
45
2. ,LHSI PumplFlowpath Available
2 YELLOW
3. HIlSI PumplFlowpath Available
3-4 GREEN
4. RCS is Intact below the Reactor Vessel Flange (GREEN if Defueled) ,
Inventory Subtotal
Monday, January 14, 20082:43:58 PM 52
QUESTIONS REPORT
for 25 SRO Questions
G2.1.25
Ability to obtain and interpret station reference materials such as graphs, monographs, and
tables which contain perforrnance data.
Question Number: 94
Tier 3 Group 1
Importance Rating: 3.1
Technical Reference: UOP-4.0
Proposed references to be provided to applicants during examination:
UOP-4.0 Appendix 1, figure 1a page 6 of 16 Version 28
Learning Objective:
10 CFR Part 55 Content:
Comments:
This question was changed out to get a better match to the KA. This is an SRO task during an
outage done daily.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: B DB AC B B DAB Scramble Range: A - D
Source: MODIFIED Source if Banle FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:58 PM
53
FARLEY NOVEMBER/DECEMBER 2007
FINAL EXAMINATION
NOS. 05000348/2007301 & 05000364/2007301
THIS PAGE CONTAINING
SRO QUESTION G2.1.6 OMITTED
FROM DISTRIBUTION TO THE PUBLIC
QUESTIONS REPORT
for 25 SRO Questions
21. G2.2.17 008
Given the following:
per minute on a Main Steam Line flange upstream of the MSIVs in the MSVR
that can not be isolated.
leak as TOOLPOUCH MAINTENANCE.
Which ONE* of the following correctly states whether the work ca'n be performed as
TOOLPOUCH MAINTENANCE per FNP-O~ACP-52.1, Guidelines for Scheduling of
On-Line Maintenance, and the reason?
A. May be performed as TOOLPOUCH work because the work will not interrupt the
flow of process fluid.
B. May be performed as TOOLPOUCH work because the flange is not part of a safety
related system.
C~ May NOT be performed as TOOLPOUCH work because the system pressure and
temperature are too high.
D. May NOT be performed as TOOLPOUCH work because the'work will require entry
into a technical specification LCO.
Monday, January 14, 20082:43:58 PM 57
QUESTIONS REPORT
for 25 SRO Questions
Meets 10 CFR 55.43 (b) requirements for SRO level question due to being a
supervisory knowledge of work control procedures. This is also an IR of 2.3 for an RO.
A. Incorrect- This is not acceptable because the flange is part of a high
temperature/high pressure system. Even though the work will not i'nterupt flow, the TPM
can not be perform'ed.
B. Incorrect- This is not acceptable because the flange is part of a high
temperature/high pressure system and because of the high energy of the system.
C. Correct - NOT acceptable because TOOLPOUCH WORK is not allowed on systems
where the pressure is greater than 1000 psig or temperature is greater than 200
degrees F.
ACP-52.1
section 2.0
o Tightening of un-isolatable fittings with process fluids <1000 psig or <200
degrees F can be done as tool pouch work. If the system pressure is >1000 psig
or temperature is >200 degrees F, then a work order is required with Team
Leader or above approval. (AI # 2004202241)
D. Incorrect- it is not acceptable but the reason given is incorrect. TS LCO entry would
not be required to tighten the flange. The stem placed the flange upstream of the
MSIVs to give TS entry credibility.
TOOLPOUCH WORK is defined as work that can be conducted without detailed written
Instructions and without overall plant scheduling.
section 3.0 table
Flanges Tighten to stop leakage (within maximum torque limits)
Monday, January 14,20082:43:58 PM 58
QUESTIONS REPORT
for 25 SRO Questions
G2.2.17
Knowledge of the process for managing m.aintenance activities during power operations.
Question Number: 97
Tier 3 Group 2
Importance Rating: 3.5
Technical Reference: ACP-52.1, Appendix 3; NMP-GM~006
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.5
Comments:
Fixed per FJE comments and removed ctmt from stem. each distracter has a valid reason why
it could or could not be correct for the conditions given.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: C C C C C C C. C C C Items Not Scrambled
Source: MODIFIED Source if Banle
Cognitive Level: HIGHER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 20082:43:58 PM 59
QUESTIONS REPORT
for 25 SRO Questions
22. 02.2.25 005
Technical Specification 3.4.16, RCS Specific Activity, states:
The 'specific activity of the reactor-coolant shall not exceed 100/E bar
microCi/gm of gross activity. If this limit is not satisfie*d, the reactor shall
be shut down and cooled to 500°F or less within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after detection.
Which one of the following is the basis for reducing Tavg to less than 500°F if the
specific activity of the reactor coolant is not within the limits of LCO 3.4.16?
A. Minimize the release of radioactivity in the event of a LOCA outside containment.
B~ Prevent venting a ruptured steam generator to the environment.
C. Ensure that the,1-hour dose at the SITE BOUNDARY will not exceed a small
fraction of the 10 CFR Part 100 dose guideline limits in the event of a SGTR.
D. Ensure that the 1-hour dose at the SITE BOUNDARY will not exceed a small
fraction of the 10 CFR Part 20 dose guideline limits in the event of a LOCA.
Meets 10 CFR 55.43 (b) 2 requirements for SRO level question.
A. Incorrect, LOCA dose is not the bases.
B. Correct- Minimize the release of radioactivity should a steam generator tube
rupture occur.
APPLICABLE: The LCO limits on the specific activity of the reactor coolant ensures SAFETY ANALYSES that
the resulting doses will not exceed an appropriate fraction of the 10 CFR 100 dose guideline limits following
a SGTR accident. The SGTR safety analysis (Ref. 2) assumes the specific activity of the reactor coolant at 0.5
ocCi/gm, a conservatively high letdown flow of 145 gpm, and a bounding reactor coolant steam generator (SG) tube
leakage of 1 gpm total for three SGs. The MSLB analysis assumes a steam generator tube leakage of 500 gpd in
the faulted loop and 470 gpd in each of the intact loops for a total leakage of 1440 gpd.
Condition 8.1
The change within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to MODE 3 and RCS average temperature < 500°F lowers the saturation pressure of the
reactor coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment
in an SGTR event. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach
MODE 3 below 500°F from full power conditions in an orderly manner and without challenging plant systems.
C. Incorrect, this is a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose and the reason for the specific activity limit, not the
temperature.
D. Incorrect,LOCA is not the concern and 10 CFR part 20 is not correct.
TS 3.4.16 'Basis
Monday, January 14, 2008 2:43:58 PM 60
QUESTIONS REPORT
for 25 SRO Questions
G2.2.25
Knowledge of bases in technical specifications for limiting conditions for operations and
safety nmit~.
Question Number: 96
Tier 3 Group 2
Importance Rating: 3.7
Technical Reference: TS 3.4.16 Basis
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.2
Comments:
Changed out the question since the other question did not meet the KA. Per our telephone
discussion this KA can test bases in technical specifications *for limiting conditions for
operations and bases in technical specifications for limiting conditions for safety limits (which
are RO as well as SRO knowledge level questions.)
MCS Time: 1, Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: BDDDC CDDAD Scramble Range: A - D
Source: MODIFIED Source if Banle FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:59 PM 61
QUESTIONS REPORT
for 25 SRO Questions
23. 02.3.8 002
Given the following:
- A Waste Gas release of Waste Gas Decay Tank #5 was started on
November 8, 2007 at 1500.
Which ONE of the following describes a condition that would require termination of the
release once initiated, and the person who is required to be notified lAW with
SOP-51.1, Waste Gas System Gas Decay Tank Release?
A. * R-29A, PLANT VENT STACK, is declared INOPERABLE.
- Shift Supervisor
B. * R-29A, PLANT VENT STACK, is declared INOPERABLE.
- Health Physics Foreman
C. * Waste Gas Decay Tank #4 pressure decreases during the release.
- Health Physics Foreman
D~ * Waste Gas Decay Tank #4 pressure decreases during the release.
- Shift Supervisor
Monday, January 14, 2008 2:43:59 PM 62
QUESTIONS REPORT
for 25 SRO Questions
A is incorrect.
R-29A becoming INOPERABLE will not cause the release to be stopped, but R-14
would (see below). R-29A is a backup to R-29B in the event that R-29B fails and would
be used to comply with ODCM to take grab samples.
3.2 IF R-14 becomes inoperable while discharging gaseous waste to the vent stack,
THEN discharge shall be stopped immediately and the Shift Supervisor notified.
B is incorrect. first not correct (see above).. Second part not correct, see C below.
. C is incorrect.
first part is correct, see below.
second part NOT correct. The HP foreman is in the approval chain for the release, but
is not required to be notified of termination per SOP-51.1
D is correct. If another tank pressure drops, stop the release and Notify the Shift
Supervisor.
SOP-51.1 step 4.1 .15
Monitor all gas decay tank pressures during the release. Ensure that
only the tank which is being released exhibits a pressure decrease and
no other tank pressure increases. Stop the release and notify the Shift
Supervisor if one of the above occurs.
Monday, January 14, 2008 2:43:59 PM 63
QUESTIONS REPORT
for 25 SRO Questions
G2.3.8
Knowledge of the process for performing a planned gaseous radioactive release.
Question Number: 98
Tier 3 Group 3
Importance Rating: 3.2
Technical Reference: FNP-1-80P-51.1, FNP-O-CC*P-213.0
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.4
Comments:
This fits the KA in that it is the 88 job function to know the process for the release and what is
required should a particular P&L not be met or if an instrument should fail such as R-14 or
R-22.
MCS Time: 3 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: D D C C A A A D B D Scramble Range: A - D
Source: MODIFIED Source if Banle
Cognitive Level: LOWER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:59 PM 64
QUESTIONS REPORT
for 25 SRO Questions
24. G2.4.27 002
Given the following:
- Unit 1 and 2 are operating at 100% Power.
- The Outside System Operator reports a fire in the Liquid H2 storage
tank vent stack.
Which ONE of the following describes the INITIAL response the SRO should direct
lAW AOP-29.0, Plant Fire; and contains only notifications REQUIRED, lAW EIP-8.0,
Non-Emergency Notifications?
A. * Assemble the fire brigade and direct the Fire Brigade Leader to extinguish the fire
by spraying water directly on the hydrogen vent stack.
- FNP Duty Manager and the Air Products company.
B. * Assemble the fire brigade and direct the Fire Brigade Leader to extinguish the fire
by spraying water directly on the hydrogen vent stack.
- Corporate Duty Manager and the Nuclear Regulatory Commission Operations
Center (NRCOC).
C~ * Direct the Outside SO to use SOP-34.0 to extinguish the fire by establishing a
. helium purge and isolating the leak.
- FNP Duty Manager and the Air Products company.
D. * Direct the Outside SO to use SOP-34.0 to extinguish the fire by establishing a
helium purge and isolating the leak.
- Corporate Duty Manager and the Nuclear Regulatory Commission Operations
Center (NRCOC).
Meets 10 CFR 55.43 (b) due to EIP notifications are the responsibility of the SRO position.
A. incorrect. The first part is not correct. It is not directed since it would not secure the source
of the hydrogen and may not put the fire out.
In the case of a hydrogen vent stack fire, AOP-29 has the following on the symptoms and entry
page:
I. IF the fire is in the Liquid H2 storage tank vent stack, THEN go to FNP-0-SOP-34.0, section
4.10, HYDROGEN - OXYGEN SYSTEM. IF the actions of FNP-0-SOP-34.0 are
unsuccessful,THEN the Shift Supervisor should enter FNP-O-AOP- 29.0 and at his/her
discretion, assemble the fire brigade to respond.
SOP-34 P&L
3.11 In the event of a fire at exit of vent stack do not spray water on the vent stack or safety
relief valves. Allow fire to continue to burn at top of vent stack until hydrogen source is located,
THEN extinguish per section 4.10.
Monday, January 14, 2008 2:43:59 PM 65
QUESTIONS REPORT
for 25 SRO Questions
The second part is correct lAW EIP-8. SOP-34 and EIP-13.
Notification of the FNP Duty Manager and the Air Products company is required (if the
emergency involves a liquid hydrogen tank, a liquid oxygen tank, or associated use systems.)
by EIP-13.0 Fire Emergencies and EIP-8.0 Non-Emergency Notifications.
B. incorrect. first part and second part is NOT correct.
The Nuclear Regulatory Commission Operations Center (NRCOC) is NOT required to be
notified as delieniated in EIP-8.0 P&Ls unless the fire is an emergency classification per step
'6.2.3 EIP-8.0. Some fires are Emergency classifications, but a vent stack fire is not.
Corporate Duty Manager is correct for this event. There is no requirement to file a
non-emergency report per EIP-8 (which would require notifying the NRCOC per figure 1) nor is
there a requirement to notify the NRCOC in SOP-34.
C. Correct. This is the correct action and notifications
lAW AOP-29, SOP-34 would be used to extinguish the fire per the below steps of SOP-34
4.10.1 Note tank pressure and attempt to quickly determine probable source
of H2 leakage by frost. on lines from relief valves or purge lines. The
most probable source of H2 leakage is PCV-3 which is set at 130 psig
and is isolated by valve NSP14V757 (V-27).
4.10.2 Open isolation valve on installed helium bottle, THEN establish
helium purge of vent stack.
4.10.3 Isolate or attempt to isolate leaking valve.
EIP-8.0
6.0 Notification for EIP-13, "Fire Emergencies"
NOTE: Notifications are required for all plant fires including .small fires and hydrogen
vent stack fires. EXCEPTION: Notifications are not required for intentionally
set fires at th\e Fire Training Facility.
6.1 The Shift Manager shall ensure the following are notified:
6.1.5 The FNP Duty Manager.
6.1.8 Air Products, if the emergency involves a liquid hydrogen tank, a liquid
oxygen tank, or associated use systems.
6.2 The ED / FNP Duty Manager shall notify:
6.2.2 Corporate Duty Manager.
D. incorrect - first part is correct, notifications is incorrect.
Monday, January 14, 2008 2:43:59 PM 66
QUESTIONS REPORT
for 25 SRO Questions
G2.4.27
Knowledge of fire in the plant proCedtH"e~
Question Number: 99
Tier 3 Group 4
Importance Rating: 3.5
Technical Reference: AOP-29.0 and SOP-34 and EIP-8.0
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.5
Comments:
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: C C B D B CAD D D Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: LOWER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:43:59 PM 67
SRO QUESTION G2.4.44
The correct answer for this question is "c" and not "B."
The distractor analysis for "B" should read as follows:
B. Incorrect: First part incorrect (see first part of A,) second part incorrect. For initial
11
notifications the Form "Guideline 1 states: (Unaffected Unit(s) Status Not Required for
Initial Notifications)
The answer analysis for "c" should read as follows:
C: Correct: Notification of Protective Action Recommendations is required to be
completed for the Initial Notification of a General Emergency. (Not required for any other
classification including Site Area Emergency). Announcement with evacuation
instructions required per step II. A. 2. of Guideline 2, EIP-9.0.
QUESTIONS REPORT ..
for 25 SRO Questions
25. G2.4.44 045
A Site Area Emergency was declared 35 minutes ago. Subsequently, conditions have'
degraded and a General Emergency classification needs to be declared.
When upgrading to the General Emergency classification, which one of the following
contains ONLY required actions lAW FNP-O-EIP-9.0, Emergency Actions?
A. * Sounding of the plant emergency alarm.
- Announce needed evacuation instructions to plant personnel.
B~ * Sounding of the plant emergency alarm.
C. * Notify Alabama and Georgia of Protective Action Recommendations.
- Announce needed evacuation instructions to plant personnel.
D. *,Notify Alabama and Georgia of Protective Action Recommendations.
Meets 10 CFR 55.43 (b) requirements for SRO level question since the IR is a 2.1 for
an RO and not required knowledge or an action for an RO.
EIP-9.0
A: Incorrect: This action is required by the General Emergency Guideline Procedure
only when not already previously performed. The SRO must know that it was required,
and was already sounded, for the SAE. Second part correct
II. Emergency Director Actions
NOTE: THE SHIFT MANAGER SHALL PERFORM THE DUTIES OF THE
EMERGENCY DIRECTOR UNTIL HIS ARRIVAL AND ASSUMPTION OF
DUTIES. .
Initials
A. Notify personnel on site
1. If the Plant Emergency alarm has not already been activated, then announce over
the public address system "All Plant Personnel Report to Designated Assembly
Area," activate the PEA [Plant Emergency alarm] for 30 seconds and repeat the
announcement.
2. Announce the classification, and the condition, request setup of the TSC and OSC
and give needed evacuation instructions over plant public address system.
Monday, January 14, 2008 2:43:59 PM 68
QUESTIONS REPORT
for 25 SRO Questions
2(J:@irr&~. N:07tFk:;~,TIctN::: Tf:f..:fE. E
.O!\T...... X-"~n..rrHE?t'7J~~ncti # . . . .
3. $lTE~ ~F~.RLEY NUCLEA1R ~~
m07HER._" .
6. E'~tEftGEWt:c.Y* Rei.E.~E:;
a~~ It::=.s fE*~~OOMe f:$,crma:i
ci!'enSfrMi,fm})m
8t!:T$fCfi~n,g: i :$t:a~bfe' ~Oegnmh'!O;
5, ;}itETEO'Rct..C~t;lc,q o~~;r:~r ;~~~~,:St~~~qn'~,::;***' .. ;::**_~~
- 25 .m:t: e!:e:~:atrcrr'p:~tf:re:d Prec~tat~)).n __
i"ll=&eCL~JRi~(nOM
t:~::):~_g;.*~~I~mtl\~~:
t2.,!Ufiii7 STATtJ!1t .. :B:tti ._~: Po>aer'~ldld::r_i.Jt*Tt:me. Date ._{-:t_:*_
- ~::::;~~:~=:.$tWS lfWt R~@~d f: !lMZ._~: F<7iM~!'r' ~$httstHM' :!d'Tmm,e . O~:te - if_ _}_'_
- L3,lqet.Ai4tR¥~:: t:I~ a;jdl~~ renart:s; D i'feD'3 3d~tHf!Rl)m;~rrem5':~ts QF:{ ~~ .*~~
B. Correct: Notification of Protective Action Recommendations is required to be
completed for the Initial'Notification of a General Emergency. (Not required for any
other classification including Site Area Emergency). Announcement with evacuation
instructions required per step II. A. 2. of Guideline 2, EIP-9.0.
C: Incorrect: First part correct, second part incorrect. For initial notifications the Form
IIGuideline 111 states: (Unaffected Unit(s) Status Not Required for Initial Notifications)
D: Incorrect: First part correct, second part incorrect.
G2.4.44
Knowledge of emergency plan protective action recommendations.
Question Number: 100
Tier 3 Group 4
Importance Rating: 4.0
Technical Reference: EIP-9.0
Proposed references to be provided to applicants during examination: NO
Learning Objective:
10 CFR Part 55 Content: 43.5
Comments: Replaced the question with one that does not require reference material and is not
a direct lookup. It is more closely related to what an SRO duty is in the emergency plan and
required knowledge for an SRO.
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: B B DAB B DDC C Scramble Range: A - D
Monday, January 14, 2008 2:43:59 PM 69
QUESTIONS REPORT
for 25 SRO Questions
Source: BANK Source if Bank: FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: SRO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:43:59 PM 70
QUESTIONS REPORT
for 75 RO Questions
1. 001 AK2.01 001
Initial conditions (Time = 1000) with Rod control in AUTO:
- Tavg - Tref deviation is O°F and stable.
- Pressurizer level is 45% and stable.
- Reactor Power is approximately 75% and stable.
- Control Bank D step counters are at 144 steps.
Current conditions (Time = 1002) with no load change in progress:
- Tavg - Tref deviation is approximately +2°F and rising.
- Pressurizer level 46% and slowly rising.
- Pressurizer spray valves have throttled open.
- Reactor Power is approximately 76% and slowly rising.
- Control Bank D step counters are at 150 steps and rising
at 8 steps per minute.
Which ONE of the following describes the event in progress; and then the
FIRST action that must be performed lAW AOP-19.0, Malfunction of Rod
Control System?
A. * Inadvertent RCS dilution;
- Trip the reactor and enter EEP-O, Reactor Trip or Safety Injection.
B. * Inadvertent RCS dilution;
- Place the rod control mode selector switch to MANUAL and match Tavg with
Tref by inserting rods.
C. * Uncontrolled Continuous Rod Withdrawal;
- Trip the reactor and enter EEP-O, Reactor Trip or Safety Injection.
D~ * Uncontrolled Continuous Rod Withdrawal;
- Place the rod control mode selector switch to MANUAL and verify that
rod motion stops.
Monday, January 14, 20082:42:14 PM 1
QUESTIONS REPORT
for 75 RO Questions
A is incorrect; if an inadvertent dilution were taking place, the rods would go in not OUT
To trip the reactor at this point would be incorrect. t
B is incorrect;
See above for the first part.
Second part is corre'ct lAW AOP-19 for a continuous rod withdrawal.
C is incorrect; is the correct accident, however, the action stated is the RNO' if rods do
not cease moving once they have been placed in manual lAW AOP-19.
D. is correct for the stated situation.
A CRW is taking place due to temperature shows rods should actually be moving in
due to high temperature and the action is to place rods in Manual if they are stepping
while in AUTO
001 AK2.01 Continuous Rod Withdrawal
Knowledge of the interrelations between the Continuous Rod Withdrawal and the following:
Rod bank step counters
Question Number: 57
Tier 1 Group 2
Importance Rating: 2.8
Technical Reference: OPS 52201 E, AOP-19.0
Proposed references to be provided to applicants during examlnation: None
Learning Objective: OPS52520S07
10 CFR Part 55 Content: 41.5
Comments:
Fixed per FJE comments 10/4/2007
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: DAB DAB C B A B Scramble Range: A - D
Source: BANK Source if Banle WBN BANK
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:14 PM 2
QUESTIONS REPORT
for 75 RO Questions
2. 001 K5.36 001
During a power DECREASE, the change in power defect will add (1) reactivity to
the core.
Assuming the operator does NOT borate or dilute, control- rod (2) will initially be
required to maintain Tavg on program.
A. (1) negative
(2) insertion
B. (1) negative
(2) withdrawal
C~ (1) positive
(2) insertion
D. (1) positive
(2) withdrawal
A. incorrect because power defect adds positive reactivity on a power decrease.
B. incorrect because power defect adds positive reactivity on a power decrease.
C. correct. Power defect adds positive reactivity for a negative change in load.
Curve 27 shows for 6000 MWO/MTU power defect will go from -1269 to -658.
Positive reactivity will cause Tavg to rise. With no boration, rods must be inserted.
D. incorrect because rods must be inserted to maintain Tavg on program.
Monday, January 14, 20082:42:14 PM 3
QUESTIONS REPORT
for 75 RO Questions
001 K5.36 Control Rod Drive Systems
Knowledge of the following operational implications as they apply to the CRDS:
Significance of sign .(always minus) of a calculated power defect
Question Number: 29
Tier 2 Group 2
Importance Rating: 3.1
Technical Reference: T&AA, CORE PHYSICS CURVEs pcb-1-voI1-crv27, 34 & 60
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS52510F04
10 CFR Part 55 Content: 41.1
Comments:
Fixed per FJE comments 10/5/2007
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: C DBA A A B D D C Scramble Range: A - D
Source: NEW Source if Banle
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:14 PM 4
QUESTIONS REPORT
for 75 RO Questions
3. 002 K3.02 001 .
Which ONE of the following correctly describes the reason that, in the event of a
design basis Large Break LOCA, the plant is realigned from Cold Leg Recirculation to
Simultaneous Cold and Hot Leg Recirculation?
Art To prevent fuel temperatures from i'ncreasing due to boron precipitation at the
TOP of the core.
B. To prevent fuel temperatures from increasing due to boron precipitation at the
BOTTOM of the core.
C. To prevent a reduction in Shutdown Margin due to boron precipitation at the
TOP of the core.
D. To prevent a reduction in Shutdown Margi'n due to boron precipitation at the
BOTTOM of the core.
l
A. Correct. Hot Leg Recirc is aligned to Ibackflush the core due to boron precipitation
that occurs due to boil-off.
B. Incorrect. Concern is top of the core, not the bottom which will be covered with water
and have continuos flow.
C. Incorrect. Shutdown Margin may be ultimately affected, but core cooling and
blockage of flo~ channels is the concern that Hot Leg Recirc addresses
D. Incorrect. Shutdown Margin may be ultimately affected, but core cooling and
blockage of flow channels is the concern that Hot Leg Recirc addresses, and the
cO'ncern is at the top of the core
Executive volume Rev 2 of ERG guidelines
The operators should continue with the guideline and transfer to cold leg recirculation
(ES-1.3) when the RWST level reaches the switchover setpoint. The plant engineering staff
may also recommend hot leg recirculation (ES-1.4, TRANSFER TO,. HOT LEG
RECIRCULATION ), at a later time, if a boron precipitation concern is possible.
CONCERN(s)
Should the SI system be aligned for hot-leg recirculation in order to prevent boron precipitation
in core? Boron precipitate can plate out on the fuel cladding surface, thereby reducing
heat transfer from the fuel to the coolant. .
This requirement is conservative in that for all cases except the design-basis LOCA, the
actual rate of boron concentration within the core will be less than that assumed in the
FSAR design-basis calculation of the time at which switchover is required from cold- to hot leg
recirculation. This is due to the core boiling rate being less than that assumed in the
calculation. Additionally, for any LOCA smaller than the design-basis LOCA, the saturation
temperature will be higher than that assumed in the calculation, resulting in a boron
precipitation limit that is higher than assumed. These factors substantially lengthen the
Monday, January 14, 20082:42:14 PM 5
QUESTIONS REPORT
for 75 RO Questions
time to the onset of boron precipitation within the core and the time before switchover from
cold- to hot-leg recirculation is required.
Conservative analysis has shown that, following a large cold-leg break in the RCS, the boric
acid concentration limit established by the NRC (the boric acid solubility limit of 27.53%
minus 4% for conservatism) would be exceeded if cold leg recirculation is maintained for an
extended period. The analysis considers the increase in boric acid concentration in the
reactor vessel during the long-term cooling phase of a LOCA assuming a conservatively
small effective vessel volume including only the free volumes of the reactor core and the
upper plenum below the bottom *of the hot leg nozzles. This assumption conservatively
neglects the mixing of boric acid solution with directly connected volumes,such as the
reactor vessel lower plenum.
Effects of Break Location
Cold Leg Break
The calculation of boric acid concentration in the reactor vessel considers a cold I*eg break
of the reactor coolant system in which steam is generated in the core from decay heat while
the boron associated with the boric acid solution is completely separated from the steam
and remains in the effective vessel volume. The cold leg safety injection flow is not
effective in counteracting this boiloff from the core since for larger breaks the downcomer
level is low and the injection flow is primarily refilling the downcomer as opposed to the
core, and no flushing of the core occurs. If the plant is transferred from cold leg to hot leg
recirculation prior to the time the boric acid concentration limit is reached in the reactor
vessel, the hot leg safety injection flow will dilute the vessel boron concentration by passing
relatively dilute boron solution from the hot leg through the vessel to the cold leg break
location and will terminate boiloff from the core. This will prevent boron precipitation in the
core along with any resultant plateout on the fuel cladding which could reduce heat transfer
from .the fuel to the reactor coolant.
. Lesson text ESP-1.3 OPS-52531 G
Approximately 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> following the loss of coolant accident (LOCA),
the cold leg recirculation phase will be terminated and the simultaneous
cold and hot leg recirculation phase is initiated. Switching to a hot leg
recirculation path will wash out the boron that may have plated out on the
fuel rods at the top of the core. Maintaining a ~old leg recirculation path
provides a normal flow path through the core. If the boron were allowed to
build up in the top of the core, it could reduce flow through the core* and
degrade the heat transfer capa~ility of the fuel. This would also result in a
depletion of the boron concentration in the recirculated fluid from the sump.
Monday, January 14, 2008 2:42:14 PM 6
QUESTIONS REPORT
for 75 RO Questions
002 K3.02 Reactor Coolant System
Knowledge of the effect that a loss or malfunction of the ReS will have on the following:
Fuel
Question Number: 30
Tier 2 Group 2
-Importance Rating: 4.2
Technical Reference: OPS-521 028
Lesson text ESP-1.3 OPS-52531G, Executive volume Rev 2 of ERG guidelines
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS521 02803
10 CFR Part 55 Content: 41 .7
Comments:
Fixed per FJE comments 10/5/2007
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: A C B B ABC B C A Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:15 PM 7
QUESTIONS REPORT
for 75 RO Questions
4. 003 A1.04 001
Given the following:
- Unit 1 is in Mode 4.
- 1A RCP has just been started.
- A CCW leak is occurring in the tube section of the upper bearing oil cooler of
the 1A RCP.
Which ONE of the following correctly describes the effect on the 1A RCP Oil Reservoir
level and the MINIMUM motor bearing temperature that requires tripping the RCP?
Oil Reservoir Level MINIMUM Temperature
Art INCREASES
B. INCREASES
c. DECREASES 195 0 F
D. DECREASES
A. correct. CCW would leak into the bearing oil reservoir because it is at a higher
pressure. The correct temperature to trip the RCP is
HG1
2. IF any 1A Rep motor bearing temperature exceeds 195°F, THEN perform
the following actions:
a) Trip the reactor, AND go to FNP-I-EEP-O.O, REACTOR TRIP OR
SAFETY INJECTION.
b) Stop IB RCP.
c) Perform the actions required by FNP-I-AOP-4.0, LOSS OF
REACTOR COOLANT FLOW.
d) Manually close pressurizer spray valve, PK 444C
B. incorrect due to temperature setpoint.
KK5
PHASE 1 alarm setpoint 275°P MFG max safe operating temp. 302°P
c. incorrect because the reservoir level will be high, not low.
D. incorrect because the reservoir level will be high, not low.
195°F correct per ARP, 302°F is Plausible because per ARP-1.10, KK5, Max ~afe
temperature for Rep Motors is 302°F.
On a complete Loss of CCW Flow to RCP Motor Bearing Oil Coolers, the bearing
temperatures will exceed 195°F in approximately 2 minutes.
Monday, January 14, 20082:42:15 PM 8
QUESTIONS REPORT
for 75 RO Questions
003 A1,,04 Reactor Coolant Pump System (RepS)
Ability to predict and/or monitor changes' in parameters (to prevent exceeding design limits)
associated with operating the RepS controls including:
Rep oil reservoir le,vels
Question Number:
Tier 2 Group 1
Importance Rating: 2.6
Technical Reference: HG1 & HH1 ARP-1.8, AOP-4.1
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS52520107
10 CFR Part 55 Content: 41.5
Comments:
Fixed per FJE comments 10/5/2007
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: A B B D DAB B D A Scramble Range: A - D
Source: NEW Source if Banle
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:15 PM 9
QUESTIONS REPORT
for 75 RO Questions
5. 003 K5.03 001
Unit 1 is at 6% reactor power when 1B Rep trips.
Which ONE of the following describes the INITIAL response of Tavg in the 1BLoop
and the reason for that response with no operator action?
A. INCREASE because Thot in the unaffected loops INCREASES.
B. INCREASE due to the reverse flow of primary coolant in the 1Bloop.
C. DECREASE because Tcold in the unaffected loo.ps DECREASES.
D~ DECREASE due to the reverse flow of primary coolant in the 1Bloop.
Monday, January 14,20082:42:15 PM 10
QUESTIONS REPORT
for 75 RO Questions
A. Incorrect; 1Bloop Tave will not increase even though the unaffected loops Tavg will
increase.
B. Incorrect; because BLoop Tavg will not increase.
Plausible because the applicant may misunderstand Thot and Tcold values for
reverse flow in a loop.
C. Incorrect; 1 Bloop Tavg will decrease but the unaffected loops Tc will increase.
o. Correct; Tavg will initially decrease due to reverse flow, which occurs when the 1 B
Loop RCP is tripped.
OPS- 525200
If the reactor is less than 30% power and there is a loss of coolant flow in one loop (two
or more loops if below 10% power), the operator must respond in an efficient manner in
order to minimize the effects on primary and secondary systems. In the loop that has
lost coolant flow, temperatures will stabilize at approximately the cold leg temperature.
(TC) of the unaffected loop(s). This will drop the saturation temperature and pressure
of the affected loop's steam generator (SG), causing SG level to drop (shrink), and will
also reduce the amount of steaming and power output from the affected SG to a
minimum.
The loop flow indications observed by the operators would be as follows: For the affected
loop, flow would slowly decrease to 0 and then return to approximately 10%; for the unaffected
loops, the flow should increase to approximately 105% (each loop). The flow indication in the
idle loop occurs as flow stops and then begins again in the reverse direction. Since flow rates in
the ReS loops are derived from the differential pressure felt in an elbow in each loop, any flow
at all will be indicated, regardless of the direction. The indication observed in the two loops with
the running pumps is due simply to the pumps in those loops picking up' a small portion of the
flow lost in the idle loop.
Monday, January 14, 2008 2:42:15 PM 11
QUESTIONS REPORT
for 75 RO Questions
003 K5.03 Reactor Coolant Pump System (RepS)
Knowledge of the operational implications of the following concepts as they apply to the RepS:
Effects of Rep shutdown on T-ave., including the reason for the unreliability of T-ave. in
the shutdown loop
Question Number: 2
Tier 2 Group 1
Importance Rating: 3.1
Technical Reference: OPS 52520D
Proposed references to be provided to applicants* during examination: None
Learning Objective: OPS40301 AOa
10 CFR Part 55 Content: 41 .7
Comments:
Fixed per FJE comments 10/5/2007
meets the KA in that this addresses the operational implications of the loss of a RCP on Tavg
and the reason the temperature is reading below the other 2 loops (ie., not reliable or different
from the operating loops)
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: DB CDDDDB C C Scramble Range: A - D
Source: BANK Source if Bank: FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:15 PM 12
QUESTIONS REPORT
for 75 RO Questions
6. 004 A4.05 001
Given the following:
- Unit 1 is in Mode 5.
- Solid plant operations are in progress.
- FK-122, CHG FLOW, is in MANUAL.
- PK-145, LTDN PRESS, is in MANUAL.
The OATC lowers the demand on PK-145. Which ONE of the following describes the
effect on PCV-145, Letdown PCV, and RCS pressure?
PCV-145 throttles __(_1)_ _ , RCS pressure (2.-..) _
A. (1) OPEN (2) INCREASES
B. (1)CLOSED (2) INCREASES
C~ (1) OPEN (2) DECREASES
D. (1) CLOSED (2) DECREASES
A. incorrect- PCV-145 WILL open to decrease pressure when the controller is taken to
the the lower position. Due to the location of the valve in the system and with HCV-142
fully open when PCV-145 is opened RCS pressure will drop with no change in charging
flow.
B. incorrect - The valve will open, not close. Distractor is credible because changing
demand does change valve position, and it is easy to associate lowering demand with
valve closure
C. correct. Reducing the demand on PK-145 in manual will cause the valve to open,
reducing backpressure on the letdown line, therefore reducing RCS pressure upstream.
D. incorrect - The valve will open, not close. Distracter is credible because changing
demand does change valve position, and it is easy to associate lowering demand with
valve closure
0% demand on the controller = lower system pressure and the valve will open
100% demand on the controller = higher system pressure and the valve will close
Monday, January 14, 20082:42:15 PM 13
QUESTIONS REPORT
for 75 RO Questions
004 A4.05 Chemical and Volume Control System
Ability to manually operate and/or monitor in the control room:
Letdown pressure and temperature control valves
Question Number: 4
Tier 2 Group 1
Importance Rating: 3.6
Technical Reference: CVCS LP OPS-521 01 F
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS40301 FOa
10 CFR Part 55 Content: 41.5
Comments:
MCS Time: Points: 1.00 Version: a12 3 4 5 6 7 8 9
Answer: CDBCCCBCBA Scramble Range: A - D
Source: . NEW Source if Banle
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42: 15 PM
14
QUESTIONS REPORT
for 75 RO Questions
7. 004 K2.0l 001
Which ONE of the following states the power supply to 1A Boric Acid Transfer Pump?
600 Volt MCC - - - - -
A~1A
B. 1B
C. 1D
D. 1E
A is correct. per the load list for unit 1, page F-92, 1A Boric Acid Transfer pump comes
off FAC4, which comes from ED10 and from DF03.
B is incorrect. Plausible because it supplies power to 1B BAT pump. (FBB4 on MCC
1B which comes from EE10 and DG03)
C is incorrect. Plausible because it supplies power to CVCS components: 1A Charging
pump Aux Lube Oil pump HDL5. This MCC is on the rad side aux bldg and supplies
many rad side AB loads.
D is incorrect. Plausible because this MCC is on the non- rad side aux bldg but
supplies some rad side AB loads such as the Boric acid batching tank cond return unit
and power to CVCS components: 1Band 1C Charging pump Aux Lube Oil pumps from
HEK2 and K3.
Monday, January 14, 20082:42:15 PM 15
QUESTIONS REPORT
for 75 RO Questions
004 K2.01 Chemical and Volume Control System
Knowledge of bus power supplies to the following:
Boric acid makeup pumps
Question Number: 3
Tier 2 Group 1
Importance Rating: 2.9
Technical Reference: OPS 521011, F, & G, FNP-Unit 1 Load List A-506250
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS40301104
10 CFR Part 55 Content: 41.5
Comments:
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: A C CAD B C C C C Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:15 PM 16
QUESTIONS REPORT
for 75 RO Questions
8. 005 K6.03 001
Which one of the following would prevent the 1A(B) RHR heat exchangers from
performing their design function?
A. A loss of air to Heat Exchanger discharge valves HCV-603A and HCV-603B.
Br Closing the component cooling water outlets from the RHR heat exchangers during
Mode 3 operation.
C. Closing the manual valve to HCV-142, RHR Discharge to CVCS Letdown Line,
during Mode 5 solid plant operation.
D. A loss of air to Heat Exchanger bypass valves FCV-605A and FCV-605B.
Distractor analysis:
I
A: Incorrect - Loss of air to the HCV-603 s, HXs discharge valves, will not prevent the
HX from performing their design function, since these valves fail open the HXs are still
available.
B: Correct - CCW system must be able to provide flow through the RHR HXs in order
for them to perform their design function of removing RCS heat to facilitate cooldown
from 350 to 140 within 16 hrs.
C: Incorrect - Manually closing this valve during solid plant operation will result in a loss
of UD and may cause a pressure increase in the RCS, however, this is not the design
D: Incorrect - Loss of air to the FCV-605 I s, HXs bypass valves, will not prevent the HX
from performing their design function, since these valves fail closed the HXs are still
available.
REFERENCES: .
1. FNP-1-S0P-7.0, RESIDUAL HEAT REMOVAL SYSTEM
2. OPS-52102G~40204A COMPONENT COOLING WATER
Also found in bases
bases 3.4.6
In MODE 4, the primary function of the reactor coolant is the removal
of decay heat and the transfer of this heat to either the steam
generator (SG) secondary side coolant or the component cooling
water via the residual heat removal (RHR) heat exchangers. The
seconda ry f1lnption ofihe reactor coola.-nUs-t-O--aCt-as-a-t--;cAa-w-rr~iecl-r f~ou-r- - - - - - - - - - - - - - -
soluble n'eutron poison, boric acid.
bases 3.4.7
In MODE 5 with the ReS loops filled, the primary function of the
Monday, January 14, 20082:42:15 PM 17
QUESTIONS REPORT
for 75 RO Questions
reactor coolant is the removal of decay heat and transfer this heat
either to the steam generator (SG) secondary side coolant via natural
circulation (Ref. 1) or the component cooling water via the residual
heat removal (RHR) heat exchangers. While the principal means for
decay heat removal is via the RHR System, the SGs via natural
circulation (Ref. 1) are specified as a backup means for redundancy
bases 3.7.7
In- MODES 1, 2, 3, and 4, the CCW System is a normally operating
system, which must be prepared to perform its post accident safety
functions, primarily RCS heat removal, which is achieved by cooling
the RHR heat exchanger.
bases 3.9.4
The purpose of the RHR System in MODE 6 is to remove decay heat
and sensible heat from the Reactor Coolant System (RCS), as
required by GDC 34, to provide mixing of borated coolant and to
. prevent boron stratification (Ref. 1). Heat is removed from the RCS
by circulating reactor coolant through the RHR heat exchanger(s),
where the heat is transferred to the Component Cooling Water
System
005 K6.03
Knowledge of the effect of a loss or malfunction on the following will have on the RHRS:
RHR heat exchanger
Question Number: 5
Tier 2 Group 1
Importance Rating: 2.5
Technical Reference:
1. FNP-1-S0P-7.0, RESIDUAL HEAT REMOVAL SYSTEM
2. OPS-52102G-40204A COMPONENT COOLING WATER
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS40301 K06
10 CFR Part 55 Content: 41.7
Comments:
This exact question has been used on 3 NRC exams, 2002 surry exam, 2006 summer exam
and 2006 FNP exam for the same KA.
It tests the knowledge of the loss of the RHR ht exchanger (ie. cooling function) has on the
design basis for the ht exchanger.
MCS Time: 1 Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _----JrlAk+/-l;ns~w¥_\..e4_rr~B~B~C~Al<__FA~AF'lt___Fl.A~C,.__:lD~C-_\0::3Sf_\;r.lcft-aaml-l--ft1-:lbll:\;;r-:e
Ra-I--Hng~e...--i:
l~i.
.. = - D I - : : J - - - - - - - - -
Source: BANK Source if Banlc FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: YES ,
Monday, January 14, 20082:42:15 PM 18
QUESTIONS REPORT
for 75 RO Questions
9. 006 A4.01 011
Given the following:
- Unit 1 was operating at 100% power.
- B Train is on service with 1B charging pump running.
- An SI/LOSP has just occurred.
- At 22 seconds after the SI/LOSP actuation annunciator EB1, CHG PUMP
OVERLOAD TRIP, comes into alarm.
- The operator notices the amber light on the handswitch for the 1C Chg pump.
Which ONE of the following is correct concerning 1B Chg Pump?
1B Chg Pump _
A. must be manually started.
B. will start from the LOSP sequencer.
C!' will start due to 1C Chg Pump tripping on overload.
D. will remain running throughout the event per design.
Monday, January 14, 2008 2:42: 16 PM 19
QUESTIONS REPORT
for 75 RD Questions
A. incorrect; On the LOSP, the 1B chg pump will load shed. 1B charging pump does
not need to be manually started since when the 1C CHG pump trips, the 1B pump
should automatically start due to an overload trip.
B. Incorrect. Plausible because the sequencer will only start the 1B charging pump if
the 1C charging pump breaker is racked out or has tripped on overload. After 22
seconds have passed the sequencer will be at about step 2 of returning equipment to
service. Once a step is complete, the sequencer signal is no longer available to start
any other component on a previous step. Charging pumps come off step 1 and this will
occur about 17 seconds into the event.
C. Correct. The sequencer sequences on the 1C chg pump (unless it is racked out,
then it would sequence on the 1B) 'after about 17 seconds (approx.12 secs for DG to
start and tie on, no more than 5 secs for sequencer to start load.). Then, an overload
trip of 1C will cause 1B chg pump (when aligned to same train) to auto start. In this
case B Train is on service so 1B chg pump is aligned to the B train with 1C Chg pump.
P&L of SDP-2.1
3.30 If the on-service charging pump trips on overload, the off-service charging pump
for the particular train which has two operable charging pumps will automatically
start.
3.31 If lA (IC) Charging Pump trips on overload or is racked out, IB Charging Pump
will automatically start upon safety injection or loss of offsite power.
D. Incorrect. because if there was an SI with no LOSP, the SI Sequencer would leave
1B chg pump running, and would not load shed 1B Chg pump and not start 1C.
006 A4.0.1 Chemical and Volume Control System
Ability to manually operate and/or monitor in the control room:
Pumps
Question Number: 7
Tier 2 Group 1
Importa'nce Rating: 4.1
Technical Reference:
FNP-1-ARP-1.5 EB1
SOP-2.1, CHEMICAL AND VOLUME CONTROL SYSTEM PLANT STARTUP AND
OPERATION version 84
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS521 01 FOe
10 GFR Part 55 Content~:'------------------------
Comments:
Monday, January 14, 2008 2:42: 16 PM 20
QUESTIONS REPORT
for 75 RO Questions
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: C BAD C D B C B A Scramble Range: A - D
Source: BANK Source if Banle FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:16 PM 21
QUESTIONS REPORT
for 75 RO Questions
10. 006 K3.01 001
Given the following:
- A.small break LOCA has occurred on Unit 1 and EEP-O,
Reactor Trip or Safety Injection, is in progress.
- Sub Cooled Margin Monitor is reading 38°F.
- Containment pressure is 9 psig.
ALL RCPs are running.
FI-943, A TRN HHSI FLOW, indicates GPM. °
Which ONE of the following describes how the RCPs must be operated lAW EEP-O
and the reason?
A':' RCPs must remain operating to provide core cooling.
B. RCPs must remain operating to simplify RCS temperature and pressure control
during plant recovery.
C. RCPs must be tripped to prevent damage to the RCPs seals due to the loss of seal
injection flow.
D. RCPs must be tripped to prevent excessive loss of RCS water inventory and to
keep the core covered.
Monday, January 14, 20082:42:16 PM 22
QUESTIONS REPORT
for 75 RO Questions
A. Correct. RCPs may not be tripped because there is no HHSI flow per EEP-O Fold out page.
RCP Trip Criteria
RCP trip criteria have been developed and incorporated into the ERPs
to provide for RCP trip when required for Small Break LOCAs and to
minimize the probability of RCP trip when not required. The RCP trip
criteria consist of two fundamental parts:
. Successful operation of the SI system
AND
- Subcooling less than 16°F {45°F}
In the ERPs, the RCPs are not tripped unless this two-part criterion" is
satisfied.
The following summary is provided from the RCP TRIP/RESTART
document:
If RCPs continue to operate during a small break LOCA, the_ forced
circulation provides core cooling, but also results in greater loss of
coolant inventory due to continued discharge of saturated liquid
(rather than steam) from the break. Continuous operation of the
RCPs during a LOCA cannot be guaranteed since tripping of the
RCPs would occur upon a loss of offsite power or other essential
support conditions which could occur at any time. The reason for
purposely tripping the RCPs during an accident (when the RCP trip
criterion is met) is to prevent excessive loss of RCS water inventory
through a small break which might lead to severe core uncovery if
the RCPs were tripped for some reason later in the accident.
B. incorrect. although RCPs do remain running the "reason is not to make it easier to control
temperature and pressure with no charging flow.
C. Incorrect. Would be correct if HHSI flow was being indicated.
D. incorrect, RCPs would be tripped if HHSI flow was being indicated. A caution in E-O says the
following: .CAUTION: RCP seal degradation may occur if seal injection flow is not maintained
to all RCPs. This could be used to trip the RCPs if Seal injection were lost, however, core
cooling is. more important at this time due to the loss of HHSI flow.
Monday, January 14, 20082:42:16 PM 23
QUESTIONS REPORT
for 75 RO Questions
006 K3.01 Emergency Core Coolant System
Knowledge of the effect that a loss or malfunction of the ECCS will have on the following:
ReS
Question Number: 6
Tier 2 Group 1
Importance Rating: 4.1
Technical Reference: EEP-O.O foldout page
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS52530A03
10 CFR Part 55 Content: 41 .1 0
Comments:
This tests the ability to determine what to do with RCPs running and a loss of subcooling with a
S8 LOCA in EEP-O, and the effects on the RCS of that decision in relation to a loss of HHSI, .
which meets the KA above.
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: A A B B B CAD D A Scramble Range: A - D
Source: NEW Source if Banle
Cognitiv~ Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:16 PM 24
QUESTIONS REPORT
for 75 RO Questions
11. 007 K5.02 001
The crew is forming a pressurizer steam space (drawing a bubble) per UOP-1.1,
Startup of Unit from Cold Shutdown to Hot Standby. The vacuum refill procedure will
NOT be performed.
- Unit 1 is in Mode 5 maintaining 325-375 psig.
- 1B RCP is running.
- A Train RHR is on service with low pressure letdown aligned.
- RCS is in solid plant pressure control with pressurizer temperature at 178°F.
- All PRZR heaters have been energized.
Which ONE of the following correctly describes the condition that will indicate when the
pressurizer is at saturation conditions (ie. a bubble is ready to be formed) lAW
UOP-1.1; and the effect on PRT level during this evolution?
A. * Letdown flow decreases;
- PRT level will remain constant.
B~ * RCS Pressure will increase;
- PRT level will remain constant.
C. * RCS Pressure will increase;
- PRT level will rise.
D. * Letdown flow decreases;
- PRT level will rise.
Monday, January 14, 20082:42:16 PM 25
QUESTIONS REPORT
for 75 RO Questions
A. Incorrect. Plausible, because RCS pressure will start to rise and letdown flow will increase as
pressure starts to rise. The candidate may not know what to expect from the letdown flow as
they may not know the position of PCV-145, LETDOWN PCV, FCV-122, CHG FLOW REG,
and HCV-142, RHR TO LETDOWN LINE.
The PRT parameters will remain constant.
B. Correct. lAW step 5.10, letdown flow will increase as RCS pressure increases.
The PRT parameters will remain constant since the liquid from the pzr is diverted to the RHTs.
Uop-1.1 step 5.10
WHEN pressurizer temperature increases to the saturation temperature for 375 psig
(approximately 442°F) as indicated by increasing RCS pressure or letdown flow, THEN
establish a steam space in the pressurizer as follows:
UOP-1.1 shows that the liquid from the pressurizer will go to the RHTs. There will be no level
increase or liquid that will go to the PRT.
5.10.5 WHEN VCT level increases to 81%, THEN verify VCT HI LVL IDIVERT'VLV
Q1 E21 LCV115A in the fully diverted position.
C Incorrect. First part is correct.
second part is NOT correct. see above.
D. Incorrect. both first and second part are not correct.
007 K5.02 Pressurizer Relief Tank
Knowledge, of the operational implications of the following concepts as the apply to PRTS:
Method of forming a steam bubble in the PZR
Question Number: 8
Tier 2 Group 1
Importance Rating: 3.1
Technical Reference: UOP 1.1, RHR FSD A-181002
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS40301 F06
10 CFR Part 55 Content: 41 .1 0
. Comments: In order to meet the KA, the PRT had to be used to in some fashion. Since the
PZR liquid is directed to the RHTs as the level is being decreased, the PRT level, temp and
pressure will be unaffected. To meet the KA; the method of forming the bubble is addressed
by the indications that will be available when the steam space justbegins to be formed.
Operational 'implications of the PRT are none so level, pressure and temp will remain constant.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: B AAAC B B B B A Scramble Range: A - D
Monday, January 14, 20082:42:16 PM 26
QUESTIONS REPORT
for 75 RO Questions
Source: NEW Source if Banle
Cognitive Level: mGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14,20082:42:16 PM
27
QUESTIONS REPORT
for 75 RO Questions
12. 008 A2.08 001
Given the following:
- Unit 1 is at 100% power.
DISCH TCV, fails low.
is reading at the bottom of the scale (50°F) due to the temperature input failure.
The consequences of the failure is a small RCS 1 ; and the action required
by the OATC would be to use MCB TK-144, LTDN HX OUTLET TEMP, in Manual
Control and 2 CCW flow.
A. 1. dilution
2. increase
B. 1. dilution
2. decrease
C~ 1. boration
2. increase
D. 1. boration
2. decrease
Monday, January 14, 20082:42:16 PM 28
QUESTIONS REPORT
for 75 RO Questions
From SOP-23.0:
CAUTION: CCW temperature should be maintained as stable as possible due to the effects
on reactivity due to changes in letdown temperature. Also, changing CCW temperature
could affect RCP oil levels which could cause level annunciators to come in.
From SOP-2.1 rev 84
CAUTION: Changes in letdown temperature can have a significant effect on
reactor power. Care should be taken to closely coordinate changes
in CCW flow between personnel at LTDN HX CCW TEMP CONT,
QIP17TV3083, and Control Room personnel at LTDN HX OUTLET TEMP TK144.
Letdown Temperature controller, TK-144, failed low. The controller senses a lower
temperature and sends a signal to the CCW valve to close down to provide less cooling
to raise the temperature of Letdown. When Letdown temperature goes up, the
demineralizers have less affinity for boron, and some of the boron in the demineralizers
is released. This is a boration effect.
ARP'S DF5 & DF1 both direct taking manual control of TK-144 when needed to control
temperature.
A. incorrect; a dilution will not occur, increasing is correct.
B. incorrect; a dilution will not occur. Decreasing CCW flow is not correct even though
it would be done using TK-144 IN MANUAL. It is plausible because the TCV failure
does cause a temperature change, just opposite from the change that will cause a
dilution.
C. Correct. a boration will occur and increasing CCW .flow to the Ht exchanger using
TK-144 IN MANUAL is the correct answer. As letdown water heats up, boron will be
released in ion exchangers, resulting in a small boration.
D. incorrect; a boration will occur, Decreasing CCW flow is not correct.
Monday, January 14, 20082:42:16 PM 29
QUESTIONS REPORT
for 75 RD Questions
008 A2.08 Component Cooling Water System
Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and
(b) b.ased on those predictions, use procedures to correct, control, or mitigate the
consequences of those malfunctions or operations:
Effects of shutting (automatically or otherwise) the isolation valves of the letdown
cooler
Question Number: 9
Tier 2 Group 1
Importance Rating: 2.5
Technical Reference: CVCS LP, SOP-2.1, Sec 4.18 & 4.19 cautions
Proposed references to be provided to applicants during examination: ' None
Learning Objective: OPS521 01 F02
10 CFR Part 55 Content: 41 .7
Comments:
This question meets the KA in that it is a failure closed of the CCW isolation valve to a cooler
and has operational procedures and actions to combat the event.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: CBAB CADCAD Scramble Range: A - D
Source: MODIFIED Source if Bank: FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:16 PM 30
QUESTIONS REPORT
for 75 RO Questions
13.008 AK2.01 005
Given the following plant conditions:
- A reactor trip and safety injection have occurred.
- RCS pressure is 1050 psig and lowering.
- Tavg is 550°F and lowering.
- Pressurizer level is 65% and rising rapidly.
- Containment pressure is 2 psig and rising.
Which ONE of the following describes the cause of this event?
A. Letdown line break.
B. Small Break LOCA on an RCS cold leg.
C~ Stuck open pressurizer PORV.
D. Stuck open pressurizer spray valve.
A & B. Incorrect. PZR level would be lowering or off-scale low if either of these
events occurred.
C. Correct. A vapor space LOCA is occurring, due to RCS pressure lowering and
Containment pressure rising with PZR level rising.
D. Incorrect because spray valve failure would not result in containment pressure
rising.
008 AK2.01 Pressurizer Vapor Space Accident-
Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the
following:
Valves
Question Number: 39
Tier 1 Group 1
Importance Rating: R02.7
Technical Reference: AOP-100, HC1 ARP-1.8 HE3 and 4 and 5
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS52201 H12
10 CFR Part 55 Content:
Comments:
This was originally written as a spray valve failure This is not a vapor space accident-and-d~-y-d-----
not meet the KA. Rewritten to meet the KA for vapor space accident and a valve issue.
Subsequent comments from FJE made question unsat, swapped for question on 2004
Robinson NRC exam and also on VC Summer 2007 NRC exam.
Monday, January 14, 2008 2:42: 16 PM 31
QUESTIONS REPORT
for 75 RO Questions
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: CB CADAAAB A Scramble Range: A - D
Source: BANK Source if Bank: FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams:
Monday, January 14, 2008 2:42: 16 PM 32
QUESTIONS REPORT
for 75 RO Questions
14.o09EAl.17001
Given the following:
- A reactor trip and safety injection have occurred.
- RCS pressure is 1450 psig.
- Containment pressure is 7.5 psig.
- SG pressures are 1000 psig.
- All equipment has operated as designed.
Which ONE of the following would have a rising level due to the RCP #1 seal return
flow?
A. VCT
B~ PRT
C. RCDT
D. Containment Sump
A is incorrect. credible because it is the normal #1 seal flowpath.
B is correct. Containment isolation will isolate seal return flow, and the seal return relief
valve will lift and direct the flow to the PRT.
C is incorrect. credible because it is the #2 seal flowpath.
D is incorrect. credible because #3 seal flow path is directed to the Ctmt sump.
Monday, January 14, 20082:42:17 PM 33
QUESTIONS REPORT
for 75 RO Questions
009 EA1.17 Small Break LOCA-
Ability to operate and monitor the following as they apply to a small break LOCA:
Question Number: 40
Tier. 1 Group 1
Importance Rating: R03.4
Technical Reference: OPS-521 01 F EEP-O attachment 3 figure 1
drawing 0-175039 sheet 1 and 0-175037 sheet 2 0-7
Proposed references to be provided to applicants during examination:
Learning Objective:' OPS40301 F05
10 CFR Part 55 Content:
Comments:
I changed the stem from describes where the the RCP seal return flow is being directed to
would have a rising level due to the Rep seal return flow to meet the ability to monitor
the PRT parameters piece of the KA.
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: B C C C B D C B B B Scramble Range: A - D
Source: BANK Source if Bank: SONGS 2005
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:17 PM 34
QUESTIONS REPORT
for 75 RO Questions
15. 010 K3.0l 002
Given the following:
- Unit 1 is at 100% power.
- All control systems are in their normal alignments.
PT-445, Pressurizer Pressure Channel fails HIGH.
Which ONE of the following describes the initial effect on RCS pressure and the
reason for that effect?
RCS pressure _
A. rises due to ONLY Variable heaters energizing.
B. rises due to ALL Backup and Variable heaters energizing.
C~ lowers due to one PRZR PORV opening.
D. "lowers due to ALL Backup and Variable heaters de-energizing,
and both spray valves and one PORV opening.
A incorrect; because pressure will lower initially. Credible because it is consistent with
a controller failure, which could be confused with an input failure from PT-444. Also the
heaters will come on when pressure starts dropping from PK-444A. The pressure will
not initially drop and will not rise until the PORV closes (cycles at 2000#)
B incorrect; see above, only all heaters are involved and could be confused between a
level control failure, PT444 failure and this failure.
C Correct; When PT-445 fails high, PORV 445A opens and will drop pressure to 2000
psig. The valve will close per design at 2000 psig which comes from PT-455, 456 and
457 on 2/3 < 2000 psig. The PORV will cycle at 2000 psig if a rx trip and SI did not
occur.
D incorrect. Would be correct for PT-444 failing high. The lesson plan says that all
heaters will turn on when the RCS pressure drops, sprays will close and PORV 444B
will remain closed.
Monday, January 14, 20082:42:17 PM 35
QUESTIONS REPORT
for 75 RO Questions
010 K3.01 Pressurizer Pressure Control System
Knowledge of the effect that a loss or malfunction of the PZR PC,S will have on the following:
ReS
Question Number: 10
Tier 2 Group 1
Importance Rating: 3.8
Technical Reference: AOP-100, OPS-52201 H
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS52201 H17
10 CFR Part 55 Content: 41.7
Comments:
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: C D B DCA C C C B Scramble Range: A - D
Source: NEW Source if Barne
Cognitive Level: IDGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
'Monday, January 14, 20082:42:17 PM 36
QUESTIONS REPORT
for 75 RO Questions
16.. 011 A3.03 002
Given the following conditions:
- The plant is stable at 90% power.
- Charging, Letdown, and Pressurizer Level Control systems are in
automatic.
- The Pressurizer Level Selector Switch is in the 1/11 Position.
- LT-459, Pressurizer level Transmitter, has failed low.
Which one of the following describes the system response?
No operator action is taken
Charging flow will _--....&.(1~)_ _ and letdown flow will _ _-..,;;;;;;(2;."...) _
A. (1) increase
(2) remain the same
B~ (1) increase
(2) decrease
C. (1) decrease
(2) remain the same
D. (1) decrease
(2) decrease
Monday, January 14, 2008 2:42: 17 PM
37
QUESTIONS REPORT
for 75 ROQuestions
A. Incorrect. charging flow increases through FCV-122, but letdown will isolate and
flow will drop to zero.
B. correct - Charging flow will increase and letdown will isolate and flow will drop to
zero.
C. Incorrect. Charging flow will increase due to indicated PRZR level low and letdown
will isolate and flow will drop to zero.
D. Incorrect. Charging flow will increase due to lower indicated PRZR level and
letdown will isolate and flow will drop to zero.
Reference:
CFR: 41.7 / 45.5
OPS-52201 H (in part
459 (I) Low Low level alarm
LCV-459 closes
Orifice isolation valves close
All pressurizer heaters turn off
Charging flow increases to maximum
Actual pressurizer level increases because of secured letdown
and maximum charging flow
High level alarm from channel 460 (III)
Reactor trip on high pressurizer level if no oper~tor action is
taken
011 A3.03 Pressurizer Level Control System
Ability to monitor automatic operation of the PZR LeS, including:
Charging and letdown
Question Number: 31
Tier 2 Group 2
Importance Rating: 3.2
Technical Reference: PZR level/Press LP OPS-521 01 E & H, UOP-3.1
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS52201 H15
10 CFR Part 55 Content: 41.7
Comments: changed out questior riJtcitwas not correct.
This meets the KA in that the questions has the candidate monitor letdown and charging flows
and the affects on one other system during this failure.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: BCD ABC DAB C Scramble Range: A - D
Monday, January 14, 20082:42:17 PM 38
QUESTIONS REPORT
for 75 RO Questions
Source: MODIFIED Source if Banle FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:17 PM 39
QUESTIONS REPORT
for 75 RO Questions
17. 012 A2.02 001
At 10:00 plant conditions were as follows:
- Unit 1 was at 41 % power, ramping down due to RCS leakage
greater than Tech Spec limit.
- 120V AC vital panel 1A was been de-energized 2 hrs ago due to
damage to the breaker panel.
- DF01, 1A S/U transformer to 1F 4160V bus, tripped open.
At 10:10, a Large Break LOCA occurred.
Which ONE of the following describes the (1) the status of the Reactor Trip Breakers at
10:05; and (2) the action(s) required lAW EEP-O, Reactor Trip or Safety Injection,
concerning ESF components?
At 10:05 the Reactor Trip Breakers will be (1)
After the LBLOCA, the operator is required to manually align (2)
A':' (1) open
(2) "A" Train ESF components ONLY.
B. (1) open
(2) BOTH trains of ESF components.
C. (1) closed
(2) "A" Train ESF components ONLY.
D. (1) closed
(2) BOTH trains of ESF components.
A. Correct, The RTBs will open due to the loss of power from the solas to the RCP
Single loop loss of flow (SLLOF) to 2 RCPs on A Train. At 41 % power, this will result in
a Rx trip.
Then due to the loss of the vital panel, the "A" Train ESF components will not actuate
since the A" train output relay cabinet slave relays will not actuate due to the loss of
II
power.
FSD A-181 007
2. Reactor Coolant Pump Breaker Trip
Opening of one or two reactor coolanLpump-breakers (depe~ndl+li--l---l-ng~-------------
upon power level), which is indicative of an imminent loss of
coolant flow in the loop or loops, will cause a reactor trip. If two
of three pump breakers trip with plant power> P-7 10% RTP, a
reactor trip will occur. Below P-7 (10% RTP), the trip is
Monday, January 14, 20082:42:17 PM 40
QUESTIONS REPORT
for 75 RO Questions
automatically blocked. Also if 1/3 pump breakers are opened with
plant power > P-8 (30% RTP), a reactor trip will occur.
.(References 6.1.003, 6.4.007, 6.7.012)
Low Flow or Rep Breaker Open Trip
There are three low flow protection bistable status lights for each loop (total of nine
lights) on TSLB-2. There is also a protection bistable status light for each reactor coolant pump
(RCP) breaker on TSLB-2. A low flow condition in any loop as detected by the open RCP
breaker or 2/3 low flow signals will energize the respective loop low flow partial reactor trip
alarm, A(B, C) RCS LOOP FLOW LO OR A(B, C) RCP BKR OPEN. If reactor power is greater
than the P-8 setpoint (30 percent), the low flow condition will
cause a reactor trip. This is indicated by the ONE LOOP LO FLOW OR RCP BKR OPEN RX
TRIP alarm. If reactor power is less than the P-8 setpoint but greater than the P-7 setpoint, a low
flow condition in 2/3 loops will cause a reactor trip. This is indicated by the TWO LOOP LO
FLOW OR RCP BKRS OPEN RX TRIP alarm.
B. Incorrect - first part is correct.
second part in not correct in that the master relay is not the relay that causes this issue
and B Train ESF components will actuate from B Train.
c. Incorrect, RTBs will open
second part is correct.
D. Incorrect, both parts incorrect. see above for explanation.
FSD A-181 007
Figure F-l is a block diagram, illustrating the Reactor Protection System FSD boundaries.
The equipment shown depicts the Reactor Protection System and its interfaces as follows:
1. Analog protection system cabinets (W 7300 System Racks) containing the
bistables which input to the Reactor Protection. Although the process instruments
which interface with the analog protection system are considered part of the
reactor protection system as defined by IEEE 279, they are also considered as part
of their respective fluid systems. For completeness, the process sensors are
included in this FSD. The functional requirements associated with the. interfacing
process input components to the RPS are those that are applicable to these type
devices on a generic basis.
2. NIS Racks (the bistables which input to the RPS and the sensors contained
within).
3. Control board switches
4. Field contacts (RCP breakers, turbine stop valves, etc.)
5. Solid State Protection System (SSPS) initiates reactor trip or ESF actuation in
accordance with defined logic that is based on the bistable outputs from the
process racks (730QfNIS-Ra-c~---able T-8 depicts tile system interfaces
associated with the SSPS Output Cabinet providing Reactor Trip and ESF
actuation functions.
6. Reactor trip switchgear, normal and bypass breakers
7. Computer and control board demultiplexers (demux)
Monday, January 14, 2008 2:42:17 PM 41
QUESTIONS REPORT
for 75 RO Questions
8. AMSAC (Anticipated Transient Without Trip (ATWT) Mitigation System
Actuation Circuitry) is not shown nor considered part of the RPS FSD but is being
mentioned here for completeness.
The W 7300 and NIS Racks provide the signal conditioning, setpoint comparison, process
analog signal actuation, control board/control room! miscellaneous indications and
compatible electrical signal output to the protection devices. The bistable outputs
pertaining to these systems which provide this input to the RPS have been included as
part of this FSD and are listed in Section 7. The process instruments which input into the
W 7300 and NIS Racks have been included as part of this FSD and are listed in Table
T-7. (References 6.1.001,6.1.002,6.7.001,6.7.002,6.7.004, 6.7.010, 6.7.057)
Monday, January 14, 2008 2:42: 17 PM 42
QUESTIONS REPORT
for 75 RO Questions
A2.. 02 Ability to (a) predict the impacts of the following malfunctions or operations on the RPS;
and (b) based on those predictions, use procedures to correct, control, or mitigate the
consequences of those malfunctions or operations:
Loss of Instrument Power
Question Number: 25
Tier 2 Group 1
Importance Rating: 3.6
Technical Reference: FSD A-181 007 ,Figure 12 of ops-52201 RPS lesson plan
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS52201 D12
10 CFR Part 55 Content: 41 .1 0
Comments:
This question tests the loss of power to both sides of SSPS, the affects from the loss of power
thru the vital panel and the affects from the loss of power from the solas, and the action
required by the ERGs due to that loss. The second half of the KA is met by knowing from the
question what the effects are to the plant to the failures listed, and then using EEP-O, know
what the operator will have to do based on the failure to correct the malfunctions that are
identified. A loss of power to solid state affects the ESF components as well as RPS
components equally since power is lost to both protection and control as well as RPS.
Logic Cabinet (Figures 4 and 11)
The logic cabinet is to the right of the input relay cabinet. It houses the circuitry to make
logic decisions. The logic circuitry receives signals from the input relay cabinet and if
appropriate signals are received, it will initiate a reactor trip or actuate the ESF systems. In
addition to logic decision making, information is collected, stored, and transmitted (via
multiplexing techniques) to the computer and control board (via demultiplexing techniques).
Both signals come from the logic cabinets.
The logic circuits look for coincidence between protection channels. If the logic
requirements are met for a reactor trip, the circuit sends a signal to the UV driver card. The UV
driver card output drops from 48V DC to zero and de-energizes its associated reactor trip and
bypass breaker UV coils. This action trips open the breakers and de-energizes the control rod
drive mechanisms. This releases the control rod assemblies into the core. The train A UV driver
card sends its trip signal to reactor trip breaker RTA, and to bypass breaker BYB.
If an unsafe condition calls for safeguards actuation, the logic circuits will send a signal
to the safeguards driver card. The card's output will increase from zero to 48V DC and will
energize the required master relays for the specific safeguards actuation. The master relays
energize their slave relays using 120V AC,which supply either AC or DC control power to ESF
loads as appropriate.
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: AC C B AC BDBD Scramble Range: A - D
Monday, January 14, 20082:42:17 PM 43
QUESTIONS REPORT
for 75 RO Questions
Source: MODIFIED Source if Bank: FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams:
Monday, January 14, 20082:42:17 PM 44
QUESTIONS REPORT
for 75 RO Questions
18. 012 A4.01 001
Given the following:
- Unit 1 is operating' at 100% power when a PRZR PORV spuriously opens.
- The control room operators attempt to close the PORV but are unsuccessful.
- The UO closes the block valve for the open PORV.
The following conditions exist:
- Tavg is 575°F.
- PRZR level is 63%.
- PRZR pressure is 1845 psig and rising slowly.
- Reactor power is 98%.
Which ONE of the following actions are required?
A. Restore RCS pressure to >2209 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
B. Commence plant shutdown and be in hot standby per UOP-3.1, Power Operation.
Cr Manually trip the reactor, initiate SI, and enter EEP-O, Reactor Trip or Safety
Injection. .
D. Maintain the PORV block valve closed with power available.
A.lncorrect - TS 3.4.1 requires restoration w/i 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in mode 2 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This
would be true if a reactor trip were not required at this time.
B. 'Incorrect - This would be an action due to rising PRZR level lAW 3.4.9 when pzr
level is greater than 63.5%.
C. Correct - this action is necessary since PRZR pressure is below the reactor trip and
SI setpoint.
D. Incorrect - This is the TS action if the plant would remain at power.
Monday, January 14, 20082:42:17 PM 45
QUESTIONS REPORT
for 75 RO Questions
012 A4.01 Reactor Protection System
Ability to manually operate and/or monitor in the control room:
Manual trip button
Question Number: 11
Tier 2 Group 1
Importance Rating: 4.5
Technical Reference: AOP-100; TS 3.4.11
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41 .1 0
Comments:
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: CBAB B B AADB Scramble Range: A - D
Source: BANK Source if Banle FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42: 18 PM
46
QUESTIONS REPORT
for 75 RO Questions
19. 013 A4.02 002
Given the following:
- A LOCA has occurred.
- RCS pressure is 500 psig and stable.
- Containment pressure is 29 psig and lowering slowly.
- All equipment is operating as designed.
Cooldown and Depressurization, preparing to reset ESF Actuation
signals.
Which ONE of the following describes the conditions required to be met, if any, to reset
Containment Isolation Phase A and B?
A~ * Phase A may be reset without additional conditions.
- Phase B may be reset without additional conditions.
B. * Phase A may be reset without additional conditions.
- Containment Spray must be reset prior to resetting Phase B.
C. ,. Phase A may be reset without additional conditions.
- Phase A must be reset before resetting Phase B.
D. * Safety Injection must be reset before resetting Phase A.
- Containment pressure must be less than the actuation setpoint
before resetting Phase B.
A is correct. Manual resets for Phase A and Phase B may be performed even with
actuating signal present.
B is incorrect. Credible because CTMT spray and Phase B have the same actuating
signal. C8 actuation and Phase B can be reset independently of each other and at any
time.
C is incorrect. Credible because procedure directs Phase B reset after Phase A.
Phase B can be reset at any time.
D is incorrect. Credible because 81 is the automatic initiation signal for Phase A.
Phase A can be reset prior to SI reset.
Monday, January 14,20082:42:18 PM 47
QUESTIONS REPORT
for 75 RO Questions
013 Engineered Safety Features Actuation System
A4.. 02 Ability to manually operate and/or monitor in the control room:
Reset of ESFAS channels
Question Number: 13
Tier 2 Group 1
Importance Rating: 4.3
Technical Reference: EEP-1.0, Reactor Protection System FSD, A181 007, Figure F-2
sheet 8 & Table T-4
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.7 / 45.5 to 45.8
Comments:
meets the KA in that the question asks for the conditions needed to reset ESFAS channels in
the CR
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: ABADBABDDA Scramble Range: A - D
Source: BANK Source if Banle
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:18 PM 48
QUESTIONS REPORT
for 75 RO Questions
20. 013 G2.4.31 008
Plant conditions at 09:00 were as follows:
- Unit 1 was at 100% power.
II
- SSPS train "B surveillance testing was in progress.
- ED4, SSPS B TRN TRBL, was in alarm due to the SSPS testing.
- The "B" Reactor Trip Bypass Breaker were closed and the associated
annunciator alarms have been acknowledged.
At 09:30 the following occurs:
- EC4, SSPS A TRN TRBL, has come into alarm.
- All remaining annunciators are unchanged.
- The plant operator reports that the SSPS train "A" Output Relay
Mode Selector switch was inadvertently placed in the TEST position.
Which ONE of the followi*ng actions are required?
A. Apply Technical Specification 3.0.3 and initiate actions within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to shut down to
Hot Standby.
B. Stop the testing, check the "B" Reactor Trip Breaker closed, then open the "B"
Reactor Trip Bypass Breaker.
C~ Initiate a manual reactor trip; if unsuccessful, enter FRP-*S.1, Response to Nuclear
Power Generation/ATWT.
D. Immediately place the SSPS train "A" Output Relay Mode Selector switch in the
OPERATE position; then verify EC4, SSPS A TRN TRBL, annunciator has cleared.
Monday, January 14, 2008 2:42:18 PM 49
QUESTIONS REPORT
for75 RO Questions
For this condition while testing is on-going in the SSPS B Train cabinets, with the
bypass breaker closed, the general warning light will be lit for B Train. This will cause
ED4 to be in alarm. Then when the SSPS train "A" Output Relay Mode Selector switch
is placed in the TEST position the other trains GW light will be LIT and 2 GW lights
should cause a reactor trip.
some P&Ls of STP-33.0 follow:
4.2 Ensure that the GENERAL WARNING lamps on SOLID-STATE PROTECTION
TRAIN-A and B LOGIC CABINETS are OFF prior to commencing this test.
4.6 IF a failure occurs during testing, THEN hold at the point the failure occurs and
contact Maintenance for troubleshooting and repair.
4.1 The GENERAL WARNING lamps on SOLID-STATE PROTECTION TRAIN-B
and B LOGIC CABINETS will be ON during this test.
A. incorrect. Both trains of Solid State are inoperable, and a common TS which applies
when both trains of safety related equipment is inoperable is 3.0.3, but a reactor trip is .
required for this condition.
B. incorrect. IIBacking ouf' of a test when the other train becomes inoperable is
normally done, but due to the GW on both trains a reactor trip is required.
C. correct. Rea'ctor should have tripped with 2 trains in test due to both trains have a
general warning in which is the input to the automatic trip coincidence of 2/2 GW
alarms in.
ARP EC4
AUTOMATIC ACTION
1. IF both Train A AND B Solid State Protection System Trouble alarms are actuated, THEN a
reactor trip will occur.
D. incorrect. If SSPS train A had not been momentarily inoperable, no further a"ction
would be required. The coincidence for a reactor trip would no longer be met. Quickly
restoring the output switch to it's original position would not change the fact that the
SSPS should have initiated an automatic trip, and met coincidence for one, but it did
not occur. A reactor trip is necessary.
Monday, January 14, 20082:42:18 PM 50
QUESTIONS REPORT
for 75 RO Questions
013 G2.4.. 31 Engineered Safety Features Actuation System
Emergency Procedures I Plan:
Knowledge of annunciators alarms and indications, and use of the response
instructions..
Question Nu.mber: 12
Tier 2 Group 1
Importance Rating: 3.3
Technical Reference: ARP EC4
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.10
Comments:
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 "9
Answer: C C BCD B B A C B . Scramble Range: A - D
Source: BANK Source if Banle
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams:
Monday, January 14, 20082:42:18 PM 51
QUESTIONS REPORT
for 75 RO Questions
21. 015 AK3.01 002
Given the following:
- Unit 2 is in Mode 4 with two RCPs running.
The crew is at a step in UOP-1.1, Startup of Unit from Cold Shutdown to Hot Standby,
to start a third RCP.
Which ONE of the following correctly describes a RCP failure mechanism that will still
allow the remaining RCP to be started, the damage that would occur and the reason?
Art * The anti-reverse rotation device pawls are not engaged in the ratchet plate.
- RCP motor winding damage due to high starting currents.
B. * The anti-reverse rotation device pawls are not engaged in the ratchet plate.
C. * The oil lift pump does not develop 600 psig oil lift pressure.
D. * The oil lift pump does not develop 600 psig oil lift pressure.
A. Correct - the anti rotation device pawls not being engaged in the ratchet plate would
cause the high motor winding temps and possible damage to the windings due to high
starting currents.
A flywheel and an anti-reverse rotation device have been mounted at the top of the motor. Stopping one
or more Reps while other pumps are running will cause a reverse flow through the inactive loops.
Reverse flow will turn the de-energized pump backwards. Although no mechanical damage would
result from reverse rotation, an attempt to start a pump in this condition would cause starting
currents to exist for an exces~ive length of time, resulting in overheating of the motor. To prevent
reverse rotation, each pump has been equipped with an anti-reverse rotation device.
B. Incorrect - first part is correct. second part is not due to the fact that no damage will
result to a RCP due to reverse flow thru the pump.
C. Incorrect - The oil lift pump does not develop 600 psig or greater oil lift pressure
would not allow the RCP to be run. If it were to be started with the pressure not being
>600 psig, then the second part is in fact correct for the oil lift pump but not for the
radial bearing. It would actually potentially damage the windings.
_______3L..L.-'6~D~O~OT attempt to start a Rep unl essitS-oi1 lift plllll.p-has-heelLdeliYering-oiUO'---------------=-----------,-----
the upper thrust shoes for at least two minutes. Observe the oil lift pumps indicating lights to
verify correct oil pump motor operation and oil pressure. The oil lift pumps should run at least 1
minute after the Rep's are started. An interlock will prevent starting a Rep until 600 psig oil
pressure is established.
Monday, January 14, 20082:42:18 PM 52
QUESTIONS REPORT
for 75 RO Questions
system before starting the motor. The oil "lifts" the thrust shoes away from the thrust runner.
D. Incorrect - see above for the first part. RCP would not start in thi-s condition due to
an interlock.
second part deals with the lower bearing. information below. RCP motor winding
damage could result due to the upper thrust shoes.
Rep lesson plan
The lower thrust bearing takes the weight of the rotating parts when the reactor coolant
loop is at low pressure. As the loop pressure increases, the unbalanced force on the number one
seal causes the shaft to lift and transfer the thrust to the upper thrust bearing. By the time loop
pressure is sufficient to allow pump operation (350 psig), all thrust acts on the upper bearing,
which is the normal operating condition.
The lower thrust bearing functions only when the motor runs uncoupled from the pump.
In this condition, the weight of the motor rotating parts acts downward.
In order to reduce starting torque, the thrust bearing shoes receive oil from the oil lift
system before starting the motor. The oil "lifts" the thrust shoes away from the thrust runner.
The lower thrust bearing takes the weight of the rotating parts when the reactor coolant
loop is at low pressure. As the loop pressure increases, the unbalanced force on the number one
seal causes the shaft to lift and transfer the thrust to the upper thrust bearing. By the time loop
pressure is sufficient to allow pump operation (350 psig), all thrust acts on the upper bearing,
which is the normal operating condition.
The lower thrust bearing functions only when the motor runs uncoupled from the pump.
In this condition, the weight of the motor rotating parts acts downward.
015 Reactor Coolant Pump Malfunction -
AK3.01 Knowledge of the reasons for the following responses as they apply to the Reactor
Coolant Pump Malfunctions (Loss of RC Flow):
Potential damage from high winding "and/or bearing temperatures
Question Number: 41
Tier 1 Group 1
Importance Rating: R02.5
Technical Reference: RCP LP OPS-521 01 D, UOP-1.1
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41 .7
Comments: had to replace this question since it did not meet the KA, ie the reason portion of
Hie KA, and it is an exact question from 003 A1.04
meets the KA in that there is a potential failure identified that would cause loss of RC flow form
the RCP and the reason that malfunction would cause that problem, specifically motor
windings. I did not test the bearing side of the KA due to the KA already on the test 003A1.04.
Monday, January 14, 20082:42:18 PM 53
QUESTIONS REPORT
for 75 RO Questions
MCS Time:. Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: ABCBCCAABC Scramble Range: A - D
Source: MODIFIED Source if Bank: FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14,20082:42:18 PM 54
QUESTIONS REPORT
for 75 RO Questions
22. 015 K6.02 002
Given the following:
- Unit 2 was initially at 95% power.
- The reactor has tripped.
- Compensating voltage on N-35, Intermediate Range NI, is set too HIGH.
Which ONE of the following correctly describes the response of Intermediate Range
N-35 following the trip AND the effect of this response on the Source Range (SR) HI
FLUX TRIP?
N-35 will indicate (1) than actual power.
The SR HI FLUX TRIP will reinstate (2)
~~-----
A. (1) LOWER
(2) as soon as N-35 reaches the P-6 s"etpoint.
B~ (1) LOWER
(2) only when N-36 reaches the P-6 setpoint.
C. (1) HIGHER
(2) only when N-35 reaches the P-6 setpoint.
D. (1) HIGHER
(2) as soon as N-36 reaches the P-6 setpoint.
A Incorrect; If one channel indicates low, it will satisfy 1/2 of the required P-6 reset
logic. Therefore, if the logic was 1 of 2, A would be correct.
B Correct; An overcompensated channel means that compensating voltage is too high
for the channel, cancelling out part of the actual signal, resulting in a lower indication.
The P-6 permissive is satisfied when 2 out of 2 IR channels are below the setpoint.
C Incorrect; the channel would indicate lower per above discussion and
Would be correct if the channel was undercompensated with the actual logic (2 out of 2
logic for going below P-6)
D. Incorrect; If the logic was 1 out of 2 versus 2 out of 2, and the IR indicated high
(undercompensated)
Monday, January 14, 20082:42:18 PM 55
QUESTIONS REPORT
for 75 RO Questions
015 Nuclear Instrumentation System
K6.02 Knowledge of the effect of a loss or malfunction on the following will have on the NIS:
Discriminator/compensation circuits
Question Number: 32
Tier 2 Group 2
Importance Rating: 2.6
Technical Reference: UOP-2.3, ESP-O.1, step 11 and NI LP OPS-52201 D
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41 .7
Comments: good match
MCS Time: 4 Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: B C CADAB B BA Scramble Range: A - D
Source: BANK Source if Banle VCS
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:18 PM 56
QUESTIONS REPORT
for 75 RO Questions
23. 017 K1.02 001
The Subcooled Margin Monitor is in the RTD mode. Which ONE of the following
correctly describes the temperature instruments used in the Subcooling calculations for
this mode of operation? .
A~ The highest reading RTD of the 3 Wide Range RCS Hot leg and 3 Wide Range
RCS Cold Legs is used.
B. The highest reading of Core Exit and Upper Head Thermocouples is used.
C. The fifth hottest of all Core Exit .Thermocouples (excluding the Upper Head
thermocouples) is used.
D. The fifth hottest RTD of all Wide Range RCS Hot legs and Cold Legs is used.
A. Correct. This is the method used in the RTD m.ode, but has a greater time delay in
indication of a loss of subcooling due to loop transit time and instrument RTD response
time than the CETC mode.
B. Incorrect. This is correct for the Individual Value display mode.
C. Incorrect. This incorrect for the Individual Value display mode. The upper headTCs
are not excluded from the calculation in the individual value mode, and also the fifth
hottest is not used in the calculation even though the fifth hottest is used for diagnostic
purposes throughout the ERG procedure network.
D. Incorrect. This is incorrect. The fifth hottest of all RTDs is not used
017 K1.02 In-Core Temperature Monitor System
Knowledge of the physical connections and/or cause effect re'lationships between the ITM
system and the following systems:
ReS
Question Number: 33
Tier 2 Group 2
Importance Rating: 3.3
Technical Reference: ICCMS LP OPS-52202E, SOP-68.0
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.5
Comments:
Monday, January 14, 20082:42:18 PM 57
QUESTIONS REPORT
for 75 RO Questions
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: AC B B BAB AB D Scramble Range: A - D
Source: MODIFIED Source if Bank:
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:18 PM 58
QUESTIONS REPORT
for 75 RO Questions
24. 022 AA2.04 002
Given ~the following:
- Unit 1 is at 60% power.
- Pressurizer level i~ on program.
- All charging flow has been lost and NO charging pump is running.
- Letdown has been secured.
- PRZR level is lowering at a rate of 1% every five (5) minutes.
- 100% Tref = 573°F
Approximately how much time will pass before all pressurizer heaters will
automatically secure assuming no operator action?
A. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
B. 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
c~ 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
D. 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
Monday, January 14, 20082:42:19 PM 59
QUESTIONS REPORT
for 75 RO Questions
A. Incorrect; 1/2 the value of C.
B. Incorrect; 60% X 28.8 will result in approx. 17.8%. Level lowering at 1% every 5
minutes yields approx. 89 minutes, about 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
C. Correct; At 60% power, program level is approximately 38.7%. This is calculated by
taking 573 - 547 = 26 which = 2.6 X 10% change in power. 60% power or 6 x 2..6 =
15.6 + 547 = 562.6°F.
Level at this temperature is 38.7% based on 50.2 - 21.4 = 28.8 or 2.88 per 10
6 X 2.88 = 17.28 + 21.4 = 38.68 (level at 50% + 35.8 or 1/2 of 28.8 +21-.4)
Letdown isolates at 15%.
38.7 -15 = 23.68 % level decrease and since 1% per 5 minutes is the level decrease,
the time would be 5 X 23.68 = 118.4 or 2 minutes less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
D. Incorrect; Using the 18.7% and subtracting from 50.2% level yields approx. 34.6% x
5 = 157.5 minutes or 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
calculation:
at 100% power level is 50.2% and at 0% power level is 21 .4%. This gives a level
'change of approx.. 288% .per % power. Letdown isolates at 15%. -
A508617 pzr level setpoint document shows pzr level to be 21.4% at 547°F and 54.9%
at 577.2°F. The lesson plan shows level to be 21.4% at*547°F and 50.2 at 573°F.
figure 6 shows program level to be 21.4 to 50.2 from 547 to 573°F Tavg
Options are plausible b~cause they are symmetrical and not significantly different from
actual time. A math error or misunderstanding of program level at this power could
direct an applicant to any option.
Monday, January 14, 20082:42:19 PM 60
QUESTIONS REPORT
for 75 RO Questions
022 Loss of Reactor Coolant Makeup -
AA2.04 Ability to determine and interpret the following as they apply to the Loss of Reactor
Coolant Pump Makeup:
How long PZR level can be maintained within limits
Question Number: 42
Tier 1 Group 1
Importance Rating: R02.9
Technical Reference: OPS 52201 H FIG 6 and AOP-16
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41 .7
Comments:
Tref is given due to the fact that both units are different and we teach Tref at 573°F and explain
that this value changes most every outage and is a moving target.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: C BAB C C CACD Scramble Range: A - D
Source: BANK Source if Banle WTSI
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams:
Monday, January 14, 20082:42:19 PM 61
QUESTIONS REPORT
for 75 RO Questions
25. 022 Kl.Ol 001
Given the following initial conditions with Unit 1 at 100% power:
There is a smclilleak in the 1A containment cooler. The cooler has been isolated lAW*
SOP-12.1, Containment Cooling System. The following valves are closed for leak
isolation:
- MC)V-3019A, SW TO 1A CTMT CLR AND CTMT FPS
- MC)V-3024A, EMERG SW FROM 1A CTMT CLR
-MC)V-3441A, SW FROM 1A CTMT CLR
A Large Break LOCA occurs at this time. Which ONE of the following correctly
describes the :Service Water flow rate (if any) through the 1A containment cooler with
no operator action?
SW flow will bE3 - - - - -
A. secured
B. approximately 600 gpm
C. approximately 800 gpm
D~ approximately 2000 gpm
Monday, January 14, 20082:42:19 PM 62
QUESTIONS REPORT
for 75 RO Questions
Cooling water normally discharges through a 10-inch line including MOV-3441 A, B, C,
and D, which are located inside containment, then through a 6-inch line and
MOV-3023A, B, C, and D. On an IISII signal, water also discharges through a 10-inch
discharge line through MOV-3024A, B, C, and D, thus increasing the flow through the
coolers.
MOV-3024A, B, C, and D Containment Cooler Emergency Service Water
Discharge Valves (Figure 19)
Each motor-operated valve is controlled by a three-position MCB handswitch
(CLOSE/AUTO/OPEN, spring return to AUTO). In the AUTO position, the valve automatically'
opens upon receiving an S-signal. Valve position indication lights are above each switch.
MOV-3441A, B, C, and D Containment Cooler :
Service Water Discharge
Isolation Valves
The operation of these MCB motor-operated valves (~igure 19) is identical to the
emergency service water discharge valves (3024A, B, C, and D).
MOV-3019A, B, C, and D Containment Cooler Service Water Inlet Isolation
Valves
The operation of these MCB motor-operated valves (Figure 19) is identical to the
emergency service water discharge valves (3024A, B, C, and D).
A. Incorrect - due to the SI signal MOV-3019A, MOV-3441A and MOV-3024A open to
provide emergency SW to the coolers.
MOV-3023A is normally open and is not closed for the leak isolation lAW SOP-12.1.
This would be correct if the candidate did not know what valves rolled open then the
effect on containment temperature due to the failure.
B. Incorrect - This could be confused with the TS bases requirement to have 600 GPM
flow from one Ctmt cooler to meet post LOCA conditio,ns.
C. Incorrect - This could be confused with not knowing the correct valve line u'p and this
is the normal flow rate thru the coolers which if the emergency SW to the valve was not
taken into account for this event, this would be the correct answer.
D. Correct - Since due to the SI signal MOV-3019A, MOV-3441A and MOV-3024A
open to provide emergency SW to the coolers flow rate would increase to 2000 gpm
thru all coolers with two SW pumps running whether the fan is running or not. Cooling
at 2000 gpm will provide the cooling necessary in a LBLOCA to cool containment.
Monday, January 14, 20082:42:19 PM 63
QUESTIONS REPORT
for 75 RO Questions
022 Containment Coo,ling System (CCS)
K1.01 Knowledge of the physical connections and/or cause-effect relationships between the
CCS and the following systems:
SWSI cooling system
Question Number: 14
Tier 2 Group 1
Importance Rating: 3.5
Technical Reference: FSD A181001 Service Water System
OPS-52102F, SW lesson plan and TS bases 3.6.6
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55. Content: 41.5
Comments: This KA was changed to 022K1.01 due to not being able to meet the selected KA.
This was approved by FJE and recorded on ES-401-4 record of rejected KAs.
This meets the KA in that it asks for the cause effect of only having one sw flow path to one
operating ctmt cooler.
MCS Time: 1, Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: DC CDC CDDCA Scramble Range: A - D
Source: MODIFIED Source if Banle FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:19 PM 64
QUESTIONS REPORT
for 75 RO Questions
26. 025 AK2.03 002
The following conditions exist:
- The plant is in Mode 1.
- 1A CCW pump is tagged out to have the motor rebuilt.
- 1Band 1C CCW pumps are both in operation.
- 8 Train is the liOn Service" train.
The 1C CCW pump has just tripped.
Which one of the following correctly describes ONLY components that have lost ALL
CCW flow due to the 1C CCW pump trip?
A. 18 RHR Hx, 18 RHR Pump Seal Cooler, 18 Spent Fuel Pool Hx
8. 18 RHR Hx, 18 RHR Pump Seal Cooler, 1A Spent Fuel Pool Hx
C. 1A RHR Hx, 1A RHR Pump Seal Cooler, 1A Spent Fuel Pool Hx
D¥ 1A RHR Hx, 1A RHR Pump Seal Cooler, 18 Spent Fuel Pool Hx
Monday, January 14, 20082:42:19 PM 65
QUESTIONS REPORT
for 75 RO Questions
C CCW pump is the A train pump-
The B CCW pump is the onservice train and is carrying the misc. header and on B
Train.
A. Incorrect - RHR components are not correct, SFP is correct.
B. Incorrect - All are B Train ESF loads supplied by B CCW pump.
C. Incorrect - RHR components are correct, SFP is not correct.
D. correct - All are A Train ESF loads supplied by C CCW pump
OPS-52102G
The ESS loads consist of the following:
1. Charging pum.ps
2. Spent fuel pool heat exchangers
3. RHR heat exchangers
4. RHR pumps
The secondary heat exchanger loads consist of the following:
1. RCP oil coolers and thermal barrier heat exchangers
2. Reactor coolant drain tank (RCDT) heat exchanger
3. Excess letdown heat exchanger
4. Seal water heat exchanger
5. Letdown heat exchanger
9. Waste gas compressors
10. Sample system heat exchangers
11. Gross failed fuel detector
The CCW system provides cooling for the Train A and Train 8 emergency core cooling
system components. The C CCW pump and the C heat exchanger are designated as Train A. The A
CCW pump and the A CCW heat exchanger are designated as Train B. The 8 CCW pump and the
8 CCW heat exchanger can be aligned to either train.
CCW is normally lined up so that one CCW pump and one CCW heat exchanger is in
operation supplying the on-service train. The on-service train is' the one that supplies the secondary
heat exchangers. The swing pump and heat exchanger (18 CCW pump and 18 HX) is normally
aligned in standby to the on-service train with the heat exchanger outlet valve shut. The remaining
pump and heat exchanger is valved into a closed loop with the redundant safety train. This train is
idle and is designated as the off-service train. The off-service train CCW pump must be running
before starting the off-service train charging pump or RHR pump.
Monday, January 14, 20082:42:19 PM 66
QUESTIONS REPORT
for 75 RO Questions.
25 Loss of Residual Heat Removal System
AK2 . 03 Knowledge of the interrelations between the Loss of Residual Heat Removal System
and the following:
Service water or Closed cooling water pumps
Question Number: 43
Tier 1 Group 1
Importance Rating: R02.7
Technical Reference: ARP-1.1 AD5 and AE4, UOP-1.1, SOP-7.0 and SOP-23
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.7 / 45.7
Comments: meets the KA in that the interrealtionships between RHR and CCW and a CCW
pump trip has to be evaluated to arrive at the correct answer.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: DABDDDB CDC Scramble Range: A - D
Source: MODIFIED Source if Bank: FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14,20082:42:19 PM 67
QUESTIONS REPORT
for 75 RO Questions
27. 026 A3.01 004
Given the following:
- Unit 1 was in Mode 5; Unit 2 was at 100% power.
- A Dual~unit Loss of Offsite Power has occurred.
- Vital load sequencing has been completed.
- ESP-0.1, Reactor Trip Response, has been entered on Unit 2.
- An inadvertent 'A' Train CS actuation signal is received while the crew is
responding to the reactor trip.
Which ONE of the following correctly describes the status of the Unit 2 Train IA 1
Containment Spray (CS) system?
2A CS Pump 2A CS Pump Discharge Valve
A. Stopped Closed
B~ Stopped Open
C. Running Closed
D. Running Open
Monday, January 14, 20082:42:19 PM 68
QUESTIONS REPORT
for 75 RO Questions
A. Incorrect - Neither CS pump will start, due to the LOSP with no SI actuation (pump
controls logic diagram), but the Train A inadvertent actuation for containment spray
would cause the A train pump discharge MOV to open.
B. Correct - See FSD narrative below. ESS loading sequencer requires an SI signal to
be present before sequencing on ESS loads. An SI signal has not occurred for this
transient so the pump will not start but the valve will open.
C. Incorrect - The inadvertent Train A containment spray actuation would not cause
the 2A spray pump to start for reason given in A above. The A train MOV would open
for reasons given in A above. The B Tr?in would be unaffected.
D. Incorrect - See 2A pump will not start per the FSD narrative below.
OPS-52102C
Containment Spray Pumps (Figure 6)
A three-position (STOP/AUTO/START, spring return to AUTO) handswitch controls
each pump. Placing the switch in theSTART position will start the pump. Placing the 'switch in
the STOP position will stop the pump and reset the 86 relay. In the AUTO position, the pump
will automatically start upon receipt of a containment spray actuation signal ("P" signal) if an
LOSP has not occurred. If an LOSP has occurred with the IIp ll signal
present, a safety injection signal to the ESF sequencer must also be
present, or the ESF sequencer must be in test, to start the pump. T
Containment Spray System FSD A 181008
3.1.5.2 With offsite power available, the "P" signal shall start both
CSS pumps. Without offsite power available*, the CSS pumps shall start
by the diesel generator ESS loading sequencer. Starting will occur at
step two of the sequence if the "P" signal is present at that time. If the
"P" signal occurs between the completion of step two and step six of the
ESS sequence, then starting will occur at the completion of step six of
the loading sequence. If the "P" signal occurs after the completion of
step six, starting will take pi_ace immediately. Automatic starting of the
CSS pumps shall not occur unless the pump control switch on the main
control board is in the "AUTO" position (References 6.4.001 , 6.4.006,
6.4.007, 6.4.008).
3.6.1.1
These active valves shall open automatically upon receipt of a containment spray
actuation signal ("P" signal) from the ESFAS and remain open for the containment
spraying function (Reference 6.2.001).
Monday, January 14, 20082:42:19 PM 69
'QUESTIONS REPORT
for 75 RO Questions
026 Containment Spray System'
A3.01 Ability to monitor automatic operation of the CSS, including:
Pump starts and correct MOV positioning
Question Number: 15
Tier 2 Group 1
Importance Rating: 4.3
Technical Reference: CS & Cool OPS-52102C
drawings - 0207195 0207645 0207653 0207646
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41 .7
Comments:
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: BAD C A C DAD C Scramble Range: A - D
Source: BANK Source if Banle FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:19 PM 70
QUESTIONS REPORT
for 75 RO Questions
28. 026 AAl.02 005
The following conditions exist:
- Unit 2 is in Mode 3 preparing for a reactor startup.
- 1B CCW pump is tagged out.
- The on service CCW pump trips due to over-current.
- The other CCW pump wiU not start from the MCB.
Cooling Water.
Which ONE of the following correctly describes operation of the charging pumps while
performing AOP-9.0, Attachment 1, Establishing Firewater Cooling to a Charging
Pump?
A. * Stop all charging pumps until CCW or fire water is established to at least one
charging pump.
- Maximum allowable Charging Pump lube oil temperature is 160°F.
B. * Stop all charging pumps until CCW or fire water is established to at least one
charging pump.
- Maximum allowable Charging Pump lube oil temperature is 140°F.
C~ * Swap operating charging pumps until fire water is established to one charging
pump.
- Maximum allowable Charging Pump lube oil temperature is 160°F.
D. * Swap operating charging pumps until fire water is established to one charging
pump.
- Maximum allowable Charging Pump lube oil temperature is 140°F.
Monday, January 14, 20082:42:19 PM 71
QUESTIONS REPORT
for 75 RO Questions
AOP-9.0, Version 18
A. Incorrect- It is not correct to stop all charging pumps sincethe RCP seals could be
damaged. The temperature is correct.
B. Incorrect - It is not correct to stop all charging pumps since the RCP seals could be
damaged. The temperature is NOT correct.
C. Correct - This is the correct way to operate the chg pumps and the correct
temperature.
D. Incorrect - This is the correct way to operate the chg pumps and NOT the correct
temperature.
AOP-9 attachment 1
Note:
Until alternate cooling is established, swapping the operating CHG
PUMP may lengthen the time that RCP seal injection is maintained.
EA3
Dispatch operator to determine the affected pump and the actual temperature as indicated
on the local temperature indicators. IF local temperature indication is:
1.1 Between 140°F and 155°F, THEN operation may continue during subsequent
troubleshooting.
1.2 Between 155°F and 160°F, THEN consider shutdown of pump.
1.3 > 160°F, THEN immediately shutdown the affected charging pump.
2. IF a loss of CCW has occurred, THEN perform the actions required by FNP-1-AOP-9.0
LOSS OF COMPONENT COOLING WATER.
026 Loss Component Cooling Water
AA1.02 Ability to operate and / or monitor the following as they apply to the Loss of
Component Cooling Water:
Loads on the CCWS in the control room
Question Number: 44
Tier 1 Group 1
Importance Rating: R03.2
Technical Reference: CCW LP OPS-52102G and AOP-9
Proposed references to be provided to applicants during examination: None
Learning Objective:
l11-f--10I-le----t-F.......Rf--IoP~a:l-Flrt_55_6oflt_e_nc'F-"-t:----i4'-f-t11-.5------------------------
Comments:
Monday, January 14, 2008 2:42:20 PM 72
QUESTIONS REPORT
for 75 RO Questions
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
- Answer: CDACAB B B B B Scramble Range: A - D
Source: BANK Source if Bank: FARLEY
Cognitive Level: illGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14,2008 2:42:20 PM 73
QUESTIONS REPORT
for 75 RO Questions
29. 026 G2.1.2 002
Given the following conditions on Unit 2:
- The plant was at 100% power when the 2A SG Main Steam line ruptured inside
containment.
-All systems actuated as per design.
- Containment pressure spiked to 33 psig and is now continuing to decrease
slowly.
- The crew has entered ESP-1.1, SI Termination.
Which one of the following correctly describes the MAXIMUM containment pressure
and MINIMUM recirculation time that will allow the OATC to secure the CS pumps per
Containment Pressure Time Aligned for Recirculation
A. 15 psig 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
B~ 15 psig 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />
C. 18 psig 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
D. 18 psig 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />
Monday, January 14, 20082:42:20 PM 74
QUESTIONS REPORT
for 75 RO Questions
A - Incorrect, 'The 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of operation 'applies to HHSI/LHSI transferring from Cold
Leg recirc to Simultaneous hoVcold leg recirc, but is not long enough to meet the 8
hour minimum for operation of CS on recirc prior to securing CS.
EEP-1
[CAl WHEN 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> have passed since the start of the event, THEN go to FNP-1-ESP-1.4,
TRANSFER TO SIMULTANEOUS COLD AND HOT LEG RECIRCULATION.
B - Correct, CS has been aligned for recirculation flow for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and
containment pressure is15 psig.
ESP-1.3, does provide guidance; Containment pressure is <16# and the time on recirc
is > 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C - Incorrect, containment pressure has to be <16 psig and it is not. The RWST, at 4.5
feet and decreasing would be a time to align the CS pumps for recirc, and if this could
not be done then they would be secured.
D - Incorrect, 18 psig is too high a pressure and CS has not been aligned for recirc for
at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
ESP-1.3. step 10.3
WHEN containment spray recirculation flow has been established for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, AND containment
pressure is less than 16 psig, THEN stop both CS PUMPs.
step 18.3
18.3 [CAl WHEN containment spray recirculation flow has been aligned for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />,
AND containment pressure is less than 16 psig, THEN stop both CS pUMPs.
026 G2.1.2 Containment Spray System
Conduct of Operations:
Knowledge of operator responsibilities during all modes of plant operation.
Question Number: 16
Tier 2 Group 1
Importance Rating: 3.0
Technical Reference: ECP-1.1
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.10
Comments:
replaced to match the KA. original question did not have an operator responsibility. This
question does have an operator responsibility blc this is a Continuing action step and would
_ _ _~_o_c_c_u_ra-----,-Io_ng period of time later in which they would be responsible for identifying and securing
this system properly.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: B B A C B B C B B A Scramble Range: A - D
Monday, January 14, 20082:42:20 PM 75
QUESTIONS REPORT
for 75 RO Questions
Source: MODIFIED Source if Bank: FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:20 PM 76
QUESTIONS REPORT
for 75 RO Questions
30. 027 AA2.16 001
Given the following:
- Unit 2 is at 100% power.
- PT-444, PRZR PRESS, fails LOW.
Which ONE of the following lists an action contained in AOP-1 00, Instrumentation
Malfunction, that will terminate the pressure transient and stabilize ReS pressure?
Ar:' Place all pressurizer heaters to the OFF position and cycle the heaters as required.
B. Fully open both PRZR Spray valves and cycle the heaters as required.
C. Take manual control of PK-444A, PRZR PRESS REFERENCE, and reduce
demand to 0%.
D. Take manual control of PK-444A, PRZR PRESS REFERENCE, and increase
demand to 100% .
Monday, January 14, 2008 2:42:20 PM 77
QUESTIONS REPORT
for 75 RO Questions
A correct. PT-444 inputs to the master controller. If it fails low, heaters will energize in
an attempt to raise pressure. Eventually pressure will rise to a point where 1 PORV
(PORV 445A) will open unless heaters are secured manually.
B incorrect. opening sprays partially would stop the pressure rise and stabilize
pressure, but opening them FULLY will cause pressure to drop and continue dropping.
C incorrect. Lowering demand to 0% will secure heaters, but additionally open spray
valves and one PORV. This would cause pressure to continue to drop.
D incorrect. PT-444 failing low does not cause the controller to fail low. The controller
output demand is actually high at 100% due to the failure.
PT444 Fails Low
Backup heaters tum on
Variable heaters tum on to maximum
Spray valves close (if open)
Actual pressurizer pressure increases to PORV PCV-445A open setpoint causing PORV to open
Actual pressurizer pressure eventually decreases to the PORV
PCV-445A close setp'oint, causing PORV to close
Plant pressure cycles around PORV open/close setpoints
AOP-IOO step 4 actions:
IF an alarm was caused by a CONTROL instrument (PT*444/445) OR component failure,
THEN perform the following as required to restore RCS pressure to desired value.
Take manual control of the following as required:
- Pressurizer Heaters
[] IA PRZR HTR GROUP BACKUP
[] IB PRZR HTR GROUP BACKUP
[ ] IC PRZR HTR GROUP VARIABLE
[] ID PRZR HTR GROUP BACKUP
[] IE PRZR HTR GROUP BACKUP
Monday, January 14, 20082:42:20 PM 78
QUESTIONS REPORT
for 75 RO Questions
027 Pressurizer Pressure Control Malfunction -
AA2.16 Ability to determine and interpret the following as they apply to the Pressurizer
Pressure Control Malfunctions:
Actions to be taken if PZR pres~ure instrument fails low
Question Number: 45
Tier 1 Group 1
Importance Rating: RO 3.6
Technical Reference: AOP-100
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.7
Comments:
This is the action to be taken on a PT-444 failure low lAW AOP-100 and matches the KA.
Revised all choices per FJEs comments.
MCS Time: 1 Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: ADB B DAB DAB Scramble Range: A - D
Source: NEw Source if Bank:
Cognitive Level: illGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:20 PM 79
QUESTIONS REPORT
for 75 RO Questions
31. 028 K2.01 002
Unit 1 has just lost power to 600V Motor Control Center 1B.
Which ONE of the following components will not have power?
A~ 1 B Post LOCA Hydrogen Recombiner.
B. 1 B Containment Cooler Fan - High speed.
C. HHSI TO RCS CL Isolation Valve, MOV-8803B.
D. 1 B Accumulator Discharge Isolation Valve, MOV-8808B.
A. Correct - 1 B Post LOCA Hydrogen Recombiner.
OPS-52102D
B. The recombiners receive power from separate vital electrical power trains.
Recombiners A and B are powered from 600V Mee A and B,
respectively.
B. Incorrect - LCC B is the power supply.
C. Incorrect - M/CC V; valve is outside CTMT.
D. Incorrect - MCC V; valve is inside CTMT.
028K2.01 Hydrogen Recombiner and Purge Control System
Knowledge of bus power supplies to the following:
Hydrogen Recombiners
Question Number: 34
Tier 2 Group 2
Importance Rating: 2.5
Technical Reference: Post LOCA Atm Control, OPS-LP 521 02D
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.5
Comments:
different way to ask the power supply to a component to make it different from 004
K2.01.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: AB CAAB CAB C Scramble Range: A - D
Monday, January 14, 2008 2:42:20 PM 80
QUESTIONS REPORT
for 75 RO Questions
Source: BANK Source if Bank: FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:20 PM 81
QUESTIONS REPORT
for 75 RO Questions
32. 029 EKl.Ol 001
Given the following:
- An ATWT has occurred on Unit 2 during coastdown prior to entering a refueling
outage.
- The crew is performing actions of FRP-S.1, Respon$e to Nuclear Power
Generation/ATWT.
- An operator has been dispatched to trip the reactor locally.
- Attempts to establish Emergency Boration have been unsuccessful.
- Reactor power indicates 6%.
- Intermediate Range Startup rate is slightly positive.
- The RCS temperature is slowly rising.
Which ONE of the following describes the actions required lAW FRP-S.1?
A':' Allow the RCS to heat up, and continue attempts to place the reactor in a subcritical
condition.
B. Allow the RCS to heat up, and open one PORV as necessary to maintain
pressurizer pressure less than 2135 psig to increase charging flow.
C. Stop the RCS heatup by increasing AFW flow to greater than 700 gpm" and verify
dilution paths isolated.
D. Stop the RCS heatup by dumping steam to the main condenser, and continue
attempts to place the reactor in a subcritical condition.
FRP-S.1 version 25
17 Continue emergency boration. 17 Perform the following.
17.1 Determine if moderator temperature coefficient positive or negative.
[] Core Physics Curve 5
17.2 IF moderator temperature coefficient negative, THEN allow RCS to heat up.
A. correct. During coastdown at EOl, MTC is negative under all conditions. Do not
leave FRP-S.1 until power below 5%. IF power was to be >%5 or a positive SUR on the
IR, then in addition to continuing the emergency boration, if the MTC is negative, then
the RCS would be allowed to 'HU to add positive reactivity to the core and help shut it
down.
B. incorrect The RCS is allowed to heatup, but the PORvs are not cycled to maintain
pressure less than 2135 psig unless pressure is > 2335 psig.
~cr.--. nt-r.Cl:-+Qrl-HreG-t--Qe-GaYSe-£-.--:1--Gtoe-s-AQt-have--the AFlAt f-lew to be ~--mmH-GF--tR-i-s-r-easeAI-t-,----
'-t-t-i
but does have an RNO step to increase AFW flow to 700 gpm if SGWls are not >31 %
D. incorrect RCS temperature is not stabilized, it is allow~d to rise:
, Monday, January 14, 2008 2:42:20 PM 82
QUESTIONS REPORT
for 75 RO Questions
029 Anticipated Transient Without Scram (ATWS)
EK1.01 Knowledge of the operational implications of the following concepts as they apply to
the ATWS:
Reactor nucleonics and thermo-hydraulics behavior
Question Number: 46
Tier 1 Group 1
Importance Rating: R02.8
Technical*Reference: FRP-S.1
Proposed references to be provided to applicants during E!xamination: None
Learning Objective:
10 CFR Part 55 Content: 41.10
Comments:
This meets the KAin that this tests the operational implic8ltions during an ATWT
and the effects that we would take if the reactor was still critical after emergency boration and
rods going in what would happen temperature wise, ie. thHrmo-hydralic behavior.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: ADB DCDC B BD Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: IDGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 l~RC exams: NO
Monday, January 14, 20082:42:20 PM 83
QUESTIONS REPORT
for 75 RO Questions
33. 033 AK3.01 057
Given the following:
- Unit 1 is in Mode 2, a reactor startup is in progress.
- Intermediate Range Instrument N-35 indicates 1 X 10-8 Amps.
- FB3, NI 36 LOSS OF COMPENSATING VOLTAGE, has come in to alarm.
- N-36 indicates 1 X 10-11 Amps and has failed LOW.
- Repairs to N-36 will take 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Which one of the following will satisfy the requirements of FB3 and Technical
Specifications, and the reason for those actions?
At:' Shutdown the reactor since two IR Nls are required to remain at the current power
level for the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
B. Remain at 1 X 10-8 Amps since two IR Nls are required to raise power to the POAH.
C. Remain at 1 X 10-8 Amps since positive reactivity additions must be suspended at
this power level with one IR NI failed.
D. Increase power to >5% since the plant can remain at >5% power indefinitely with
one IR NI failed.
Monday, January 14, 2008 2:42:20 PM 84
QUESTIONS REPORT
for 75 RO Questions
A. Correct- Due to TS 3.3.1 below with IR power above P-6 and below P-1 0, tWQ IR Nls
are required or power has to be decreased below P-6 in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or >10% power in 2
hours. Since no answer allows to go to >10°/0 power, this is the only option with one IR
NI broke.
B. incorrect- power can not remain at 10-8 Amps and 2 IR range Nls are not required to
go to the POAH. This is where critical data is taken and power is leveled off during a
startup.
C. Incorrect, with power in the IR and loss of 1 channel, TS 3.3.1 requires power to be
>P-10 where the PR instruments are operable, or <P-6 where SR will be operable to
provide protection against uncontrolled rod withdrawl (TS Basis).
D. incorrect- if the plant could get to >10% power then the plant could remain in this
mode indefinitely. Since the reactor power is in the IR there is not time to get 100/0
power due tot he hold at 85 AND due to placing a SGFP onservice and meeting all
m'ode 1 entry requirements. Also the 8% power is not high enough to stay there and
the UOP- 1.3 has the plant stabilize at 8%.
F THERMAL POWER> P-6 F.1 Reduce THERMAL 2
hours
and < P-10, one POWER to < P-6.
Intermediate Range
Neutron Flux channel OR
F.2 Increase THERMAL 2
hours
POWER to > P-10.
With thermal power >P-6 and <P-IO, one intermediate range neutron flux channel
inoperable requires that thermal power be either reduced to <P-6 or raised to >P-IO within 2
hours.
A failure of both intermediate range detectors, with thermal power between P-6 and P-IO,
requires suspension of operations involving positive reactivity additions immediately and
reduction of thermal power to <P-6 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. If thermal power is <P-6 and either one'or
both of the intermediate range detectors become inoperable, actions must be taken to restore
channel(s) to operable status prior to increasing thermal power to >P-6.
Monday, January 14,20082:42:21 PM 85
QUESTIONS REPORT
for 75 RO Questions
033 AK3.01 Loss of Intermediate Range Nuclear Instrumentation
Knowledge of the reasons forthe following responses as they apply to the Loss of
Intermediate Range Nuclear Instrumentation:
Terminati*on of startup following loss of intermediate- range instrumentation
Question Number: 58
Tier 1 Group 2
Importance Rating: 3.2
Technical Reference: TS 3.3.1 Function 4; A-181007 T5-1'
Proposed references to be provided to applicants during examination: none
Learning Objective:
10 CFR Part 55 Content: 41.10/43.2
Comments: .
replaced question as the original question required SRO knowledge.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: AC CDAABDAB Scramble Range: A - D
Source: MODIFIED Source if Bank: FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:21 PM 86
QUESTIONS REPORT
for 75 RO Questions
34. 035 K4.06 001
Which ONE of the following describes the design requirement of the Steam Generator
Safety Valves?
Limits SG pressure to no greater than 1100/0 of design _
A. assuming a 100% loss of load ,with no credit taken for reactor'trip on High PRZR
pressure.
B~ assuming a 100% loss of load with no credit taken for automatic steam dump or rod
control operation.
C. assuming a limiting ATWT initiated by a loss of feedwater with no credit taken for
operation of primary PORVs or secondary ARVs.
D. assuming a limiting ATWT initiated by a loss of feedwater with no credit taken for
operation of primary PORVs or automatic steam dump operation.
A is incorrect but credible because credit is not taken for the ionitiating event, although
the turbine trip could ultimately give a high pressure trip - It is credited.
B is correct.
C is incorrect because the initiating event is wrong. Credible because turbine trip and
PORV operation are tied to ~afety valves and their capacity, and the events not
credited in these options are similar to actual conditions for the safety analysis.
D is incorrect because the initiating event is wrong and the PORV operations is wrong.
035 Steam Generator System
K4.06 Knowledge of S/GS design feature(s) and/or interlock(s) which provide for the following:
S/G pressure
Question Number: 35
Tier 2 Group 2
Importance Rating: 3.1
Technical Reference: TS basis, 3.7.1 FSAR, 15.2
Proposed references to b~ provided to applicants during examination: None
Learning Objective:
_ _ _--"--"10 CFR Part ~5~_0JJ1ent: 41 ~3--L-2,"--. _
Comments:
this is written at an S'RO level due to the references, but it is a base level of knowledge about
the design of the SG safety valves.
Monday, January 14, 2008 2:42:21 PM
QUESTIONS REPORT
for 75 RO Questions
MCS Time: 1 Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: B C B BDB CADD Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:21 PM 88
QUESTIONS REPORT
for 75 RO Questions
35. 038 EKl.02 001
Given the following:
- A Steam Generator Tube Rupture has occurred on Unit 1.
- RCS cooldown and depressurization are complete.
- The creW is maintaining the plant stable while preparing to transition to
ESP-3.1, Post SGTR Cooldown using Backfill.
- The ruptured SG narrow range level is 73% and slowly decreasing.
- PRZR level is approximately 38% and slowly increasing.
- The OATC turns on PRZR heaters lAW the guidance in EEP-3, Steam
Generator Tube Rupture.
Which ONE of the following describes the reason for this action?
To maintain pressurizer saturation temperature _
A~ corresponding to ruptured SG pressure to minimize SG leakage into the RCS.
B. above the intact SG pressure to maintain adequate secondary heat sink with int~ct
SGs.
C. above the corresponding ruptured SG pressure to ensure RCS Subcooling is
maintained.
D. corresponding to intact SG pressure to ensure RCS Subcooling is maintained.
A is correct. Attempting to maintain an inventory balance between RCS and ruptured
SG prior to ruptured SG cooldown.
B is incorrect because if intact SG pressure was higher than RCS pressure, EEP-3.0
would not be the governing procedure, ECP-3.1 would.
C. is incorrect. Coold'own is to ensure subcooling. RCS and ruptured SG will act like 2
pressurizers. Subcooling is not the issue.
D. is incorrect. Cooldown is to ensure subcooling. RCS and ruptured SG will act like 2
pressurizers. Subcooling is not the issue.
Background document for EEP-3 page 83 of 119
Purpose: To control RCS pressure and charging flow to maintain an indicated pressurizer level while
minimizing primary-to-secondary leakage.
Basis: In order to explain the basis for the guidance provided in this step, consider again equilibrium
conditions between leakage through the failed SG tube and charging flow, as shown in Figure
30. For primary system pressures greater than the ruptured steam generator pressure (PSG),
primary-to-secondary leakage will occur so that excess charging flow, i.e., greater than
letdown and coolant shrinkage, is necessary to maintain pressurizer inventory. Conversely,
for letdown flows greater than charging flow, the equilibrium RCS pressure is less than the
Monday, January 14,20082:42:21 PM 89
QUESTIONS REPORT
for 75 RO Questions
ruptured steam generator pressure and secondary-to-primary leakage will occur. The ideal
conditions, shown by Point B, occur when charging flow exactly compensates for letdown
and coolant shrinkage so that RCS pressure and the ruptured steam generator pressure
equalize. For these conditions both the pressurizer and ruptured steam generator inventories
will remain constant. Obviously fluctuations about these ideal conditions will occur due to
variations in ruptured steam generator pressure, cooldown rates, and letdown flows.
Consequently, the operator must continuously adjust RCS pressure and charging flow to
control pressurizer and ruptured steam generator inventories. This step provides guidance for
performing these actions in the form of a table. Figure 30 can be divided into four different
regions which are characterized by pressurizer and ruptured steam generator level behavior.
For primary pressures greater than the ruptured steam generator pressure, leakage into the
steam generator will increase steam generator water level (LSG). Alternatively, water level
will decrease forRCS pressures less than the ruptured steam generator pressure. Similarly,
pressurizer level (LPRZR) will increase for RCS pressures less than equilibrium. This leads
to the four regions illustrated in Figure 30. The steps one performs to stabilize the plant at
the ideal, equilibrium conditions depend on the pressurizer inventory and ruptured, steam
generator water level behavior. For example, if pressurizer level is low, region II or region ill
must be entered to increase pressurizer level. This requires one to increase charging flow or
decrease RCS pressure, as shown in Figure 30. The further into these regions, the more
rapidly pressurizer level will increase. Of course, if pressurizer level is high, the opposite
response would be necessary. However, the ruptured steam generator water level must also
be considered. STEP DESCRIPTION TABLE FOR E-3Step29 If the steam generator water
level is increasing, RCS pressure must be reduced to stop primary_to_secondary leakage. If
the steam generator water level is decreasing, primary pressure should be increased by
energizing pressurizer heaters to minimize leakage into the ReS. Note that in some cases,
actions which address pressurizer level conflict with those which address steam generator*
level. For example, if steam generator level is increasing one must decrease RCS pressure.
Since this will also increase pressurizer level, the pressurizer could fill with water if level is
initially high. However, by reducing charging flow, pressurizer level will decrease. Since this
will also decrease RCS pressure if heaters are not energized, steam generator level will also
stabilize. Hence, for this situation the preferred action is to reduce charging flow.
038 Steam Generator Tube Rupture
EK1 . 02 Knowledge of the operational implications of the following concepts as they apply to the
SGTR: Leak rate vs. pressure drop
Question Number: 47
Tier 1 Group 1
Importance Rating: R03.2
Technical Reference: EEP-3 and background documents page 83
Proposed references to be provided to applicants during examination: None
Learning Objective:
1OCFR Part 55 Content: 41.5
Comments:
this meets the KA in that the question tests the concept of why PRZR pressure and temp affect
the leak rate into or out of the Przr
MCS Time: 1 Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: ADB B B B DDC C Scramble Range: A - D
Monday, January 14, 20082:42:21 PM 90
QUESTIONS REPORT
for 75 RO Questions
Source: BANK Source if Banle MCGUIRE 2003
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:21 PM 91
QUESTIONS REPORT
for 75 RO Questions
36. 039 Al.lO 001
Given the following:
- Unit 1 is at 100% power.
- A Steam Generator Tube Leak has developed.
Which ONE of the following describes:
(1 ) the R-15A indication that will be observed if SJAE Filtration is placed on service
and (2) the R-15A indication that will be observed if a reactor trip occurs?
R-15A . . . (_1.-.-)
.. when SJAE Filtration is placed on service.
R-15A (2) if the reactor trips.
A. (1) Trends down
(2) Remains stable
B. (1) Trends down
(2) Trends down
C. (1) Remains stable
(2) Remains stable
D~ (1) Remains stable
(2) Trends down
Monday, January 14,20082:42:21 PM 92
QUESTIONS REPORT
for 75 RO' Questions
A. Incorrect
(1) is incorrect, Plausible because it is true for R-158 & R-.15C which are both
downstream of the SJAE Filtration system (which is normally bypassed)
(2) is incorrect. but plausible because the reactor trip does not isolate or stop the SG
Tube leak, and may not consider the dip across the tube is proportional to leak flow.
8. Incorrect
(1) is incorrect, Plausible because it is true for R-158 & R-15C which are both
downstream of the SJAE Filtration system (which is normally bypassed)
(2) is correct.
C. Incorrect
(1) is correct.
(2) is incorrect, but plausible because, the reactor trip does not isolate or stop the SG
Tube leak, and may not consider the dip across the tube is proportional to leak flow.
D. Correct. When the Steam Jet air ejector Filtration system is placed on service there
is no change in the reading since the SJAE is upstream of the Filtration system.
When the reactor trips, steam flow is decreased, steam pressure goes up, & dip across
SG U-tubes goes down. This causes tube leakage rate to decrease which causes R-15
indication to decrease.
039 A1.10 Main and Reheat Steam System (MRSS)
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits)
associated with operating the MRSS controls including:
Air ejector PRM
Question Number: 17
Tier 2 Group 1
Importance Rating: 2.9
Technical Reference: RAD MONITORING LP OPS 52106D
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.5 / 45.5
. Comments:
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: DCA C C B B A B D Scramble Range: A - D
Source: NEW Source if Banle
Cognitive Level: HIGHER Difficulty:
Job Position: RO P l a n t : . FARLEY
1'"reviewed-o--.-----t-G-.rt-Q-l---------J,;1-Prro-'evious 2 NRC exaUlS. NO
Monday, January 14,20082:42:21 PM 93
QUESTIONS REPORT
for 75 RO Questions
37. 041 A1.02 016
Given the following:
- Unit 1 reactor power is steady at 14%.
- Tavg is at 551 0 F.
- Rod control is in Manual.
- Turbine power is steady at 7%.
- Steam Dumps are open in the STM PRESS mode.
at 551 °F~
Which ONE of the following is the correct response of the steam dump system if
PT-464, STM HDR PRESS, fails HIGH under these conditions?
Assume no operator action is taken.
A~ * All steam dumps will open and then close at P-12.
- PK-464 will shift to MANUAL.
B. * All steam dumps will open and then close at P-12.
- PK-464 will remain in AUTO.
c. * All steam dumps will open and then cycle at P-12.
- PK-464 will remain in AUTO.
D. * All steam dumps will open and then cycle at P-12.
- PK-464 will shift to MANUAL.
Monday, January 14, 2008 2:42:21 PM 94
QUESTIONS REPORT*
for 75 RO Questions
A. Correct-
The low-low TAVG (P-12) block actuates when 2/3 TAVG instruments indicate below
543°F.
It should be noted that the AUTO feature of PK-464 can be selected only under certain
conditions. First, if the mode selector switch is in the TAVG mode, PK-464 shifts to'
manual control. Secondly, if the low-low TAVG signal (P-12) exists and th.e BYP
INTLK position on both A and B Train STEAM DUMP INTERLOCK SWITCHES has not
been selected, PK-464 will shift to manual control. By shifting to manual control, the
output of the P+I portion of the controller is set to zero and thus prevents small
pressure errors from being integrated into large controller output signals.
B. Incorrect- they do go closed at 543°F, and the block does reset at 545°F, however,
the controller shifts to minimum and manucil and the dumps do not cycle.
C. Incorrect- the steam dumps will go CIOSE~d and shift to minimum and manual. They
will not cycle.
D. Incorrect- the dumps do open and PK 4164 will go to manual but the dumps will
remain closed and.not cycle to control Tav~~.
041 Main Turbine Generator System
A1.02 Ability to predict and/or monitor changes in parameters (to prevent exceeding design
limits) associated with operating the 80S controls including:
Steam pressure
Question Number: 36
Tier 2 Group 2
Importance Rating: 3.1
Technical Reference: SD LP OPS-52201 G; AOP-100
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.5 / 45.5
Comments:
this meets the KA in that the failure of the stm pressure transmitter affects the dumps and RCS
temp and the operator has to predict and monitor the RCS for these changes.
MCS Time: 1 Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: AC BADBDCB C Scramble Range: A - D
Source: BANK Source if Banle FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:21 PM 95
QUESTIONS REPORT
for 75 RO Questions
38. 054 AA2.05 001
The following plant conditions exist:
- Unit 1 is operating at 28% power.
- A Feed Water control malfunction has caused 1A SG NR level
to reach 84%.
As'suming the plant re'sponds AS DESIGNED, with NO operator action, which ONE of
the following describes the current valve alignment?
Assume all valves were open prior to the event
Reg Valves Discharge Valves Isolation Valves
FCV-478,488, 498 MOV503A, B MOV-3232A,B,C
A. All open All open All shut
B. All shut All open AII'shut
C. All open All shut All open
D~ All shut All shut All shut
A. Incorrect - The SGFP trip will cause the discharge valves to go closed. The FWI
signal causes the FRV and bypasses to close.
B. Incorrect - SGFP discharge valves go shut due to the SGFP trip at 82% level.
C. Incorrect - FRVs close on FWIS and FWI valves close on SGFP trip.
D. Correct. See FSD, Student text, & SOP-21.0.
The FWI signal will cause the SGFPs to trip and the main turbine to trip. The SGFP trip
will cause the discharge valves to go closed and the FW isolation.valves to go closed.
The FWI signal causes the FRV and bypasses to close.
FNP Units 1 & 2 REACTOR PROTECTION SYSTEM A-181 007 [FSD]:
2-26 Rev. 10
2.7.1
4. Main Feedwater Isolation and Turbine Trip
The Main Feed Line Isolation is initiated to prevent excessive cooldown of the reactor or to lessen the
severity of the transient overall. The following signals are utilized to initiate the Main
Feed line Isolation:
a. Safety injection
b. High-high steam generator water level (P-14) set at = 820/0
of narrow range steam generator span on 2/3 coincidence
c. Low Tavg; = 554°F in coincidence with reactor trip P-4
Monday, January 14,20082:42:21 PM 96
QUESTIONS REPORT
for 75 RO Questions
manual reset to clear.
(References 6.1.022,6.4.007,6.4.015,6.7.012,6.4.21)
7. High-High Steam Generator Water Level
If water level in a steam generator increases to 82% on 2/3 narrow range level instruments, the main
turbine trips, the main feedwater pumps trip and main feedwater isolation signals are initiated.
Tripping the turbine is a protective measure to ensure no damage occurs from moisture carry-over. Main
feedwater is isolated so that no further water is added to the steam generator with the high high level to
protect the primary side from excessive cooldown when safety injection is actuated. (References 6.1.022,
6.4.007,6.4.014,6.7.012)
SOP-21.0, CONDENSATE AND FEEDWATER SYSTEM
4.6.3 At approximately 2450 RPM, trip the SGFP.
4.6.4 Verify that the SGFP high and low pressure stop valves are closed.
NOTE: In the following step, annunciator KC3 should clear after approximately three
minutes.
4.6.5 Verify that annunciator KC3, lA OR IB SGFP TRIPPED comes in.
4.6.6 Verify closed the lA(lB) SGFP DISCH VALVE, NIN21V503A(B).
4.6.7 At the lA(lB) SGFP Oil Test Station, depress the LOW LEVEL
TEST push-button until the low level alarm light is illuminated.
CONDENSATE & FEEDWATER STUDENT TEXT, OPS-52104C, OPS-40201B
Main Feedwater Stop Valves (3232A, B, and C) (Figure 14)
A three-position handswitch (CLOSE/AUTO/OPEN, spring return to AUTO) on the MCB
controls each motor-operated isolation valve. In AUTO, the valve automatically closes on a SGFP
trip signal from both pumps. Valve position lights indicate above each switch.
APE 054 Loss of Main Feedwater -
AA2.05 Ability to determine and interpret the following as they apply to the Loss of Main
Status of MFW pumps, regulating and stop valves
Question Number: 48
Tier 1 Group 1
Importance Rating: R03.5
Technical Reference: FSD, Student text, & SOP-21.0.
Proposed references to be provided to applicants during examination: None
1a CFR Part 55 Content: 41 .1 0
Comments:
original question did not meet KA. Replaced. This question meets the KA in that when the
SGWL reached 82%, the SGFPs tripped. The operator has to determine the status of the
Monday, January 14, 20082:42:21 PM 97
QUESTIONS REPORT
for 75 RO Questions
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: DADBDAABDA Scramble Range: A - D
Source: BANK Source if Bank: FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:22 PM 98
QUESTIONS REPORT
for 75 RO Questions
39. 055 EAl.02 001
Given the following:
- A LOSS OF ALL AC POWER has occurred on Unit 1.
- VA2, 1B DG GEN FAULT TRIP, annunciator has come into alarm.
- The crew is at the step in ECP-O.O, Loss of All AC Power, to verify breakers for
'major, loads OPEN.
- A Safety Injection occurs on Unit 1 at this time.
Which ONE of the following describes how the 2C DG will be started and the events
that will take place or need to take place to energize the ESF equipment?
A. * Start 2C DG from EPB in Mode 2 using the start pushbutton.
B. * Start 2C DG from EPB in Mode 2 using the start pushbutton.
- ALL ESF loads will have to be manually aligned.
C. * Start 2C DG from EPB in Mode 1 using the start pushbutton.
D~ * Start 2C DG from EPB in Mode 1 using the start pushbutton.
- ALL ESF loads will have to be manually aligned.
Monday, January 14, 20082:42:22 PM 99
QUESTIONS REPORT
for 75 RO Questions
A. Incorrect. not mode 2 - ECP-O.O step 5 directs starting 2C DG in MODE 1. the
LOSP sequencer will not run due to the SI signal present. Using MODE 2 is plausible
since all other DGs are started in Mode 2 in this condition lAW ECP-O. 2C DG is the
only DG that is started in Mode 1. The sequencer is plausible since it would run if the
SI signal was not present.
B. Incorrect. not mode 2. second part is correct.
C. Incorrect. first part is correct. Second part is NOT correct.
D. Correct. ECP-O.O step 5 directs starting 2C DG in MODE 1. the LOSP sequencer
will not run due to the SI signal present. see below.
The LOSP sequencer will not run per the note below in ECP-O with an SI signal
present.
ECP-O note at step 5.2.1.5 RNO
NOTE: The LOSP sequencer should run when output breaker closes, if no SI
signal is present. If an SI signal is present, neither sequencer
will run and Slloads must be started manually.
5.2 Perform the following:
5.2.1 Perform 2C DG SSO start as follows.
5.2.1.1 Verify 2C DG MODE SELECTOR switch in MODE 1.
5.2.1.3 WHEN load shed verified,THEN depress 2C DG DIESEL START pushbutton.
055 EA1.02 Station Blackout
Ability to operate and monitor the f.ollowing as they apply to a Station Blackout:
Manual ED/G start
Question Num-ber: 49
Tier 1 Group 1
Importance Rating: RO 4.3
Technical Reference: FNP-1-ECP-O.O
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41 ..10
Comments:
replaced this question since the original did not meet the KA. This demonstrates the ability to
start and monitor a manual start of the 2C DG and the subsequent actions to energize
equipment In that train after the start.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: DADCDCAAAC Scramble Range: A - D
Monday, January 14, 20082:42:22 PM 100
QUESTIONS REPORT
for 75 RO Questions
Source: NEW Source if Banle
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14,2008 2:42:22 PM . 101
QUESTIONS REPORT
for 75 RO Questions
40. 056 A2.04 010
Given the following:
- Unit 2 is at 80% power ramping to 100% .
- 2C Condensate Pump has tripped.
- Annunciator KB4, SGFP SUCT PRESS LO, came into alarm 35 seconds ago.
and is decreasing.
Which ONE of the following is the expected result of this condition and action required
per AOP-13.0, Condensate and Feedwater Malfunction?
A. * The standby condensate pump should have AUTO started;
- Trip the reactor and enter EEP-O, Reactor Trip or Safety Injection.
B~ * The standby condensate pump should have AUTO started;
- Verify the standby condensate pump started and if suction pressure is still falling,
then reduce load rapidly lAW AOP-17, Rapid Load Reduction.
C. * The standby condensate pump should NOT have AUTO started;
- Trip the reactor and enter EEP-O, Reactor Trip or Safety Injection.
D. * The standby condensate pump should NOT have AUTO started;
- Verify the standby condensate pump started and if suction pressure is still falling,
then reduce load rapidly lAW AOP-17, Rapid Load Reduction.
Monday, January 14, 2008 2:42:22 PM 102
QUESTIONS REPORT
for 75 RO Questions
The condensate pumps have 2 auto starts, one on the trip of the other pump and one
for suction pressure <275 psig for> 10 sec. In this question the auto start for suction
pressure is not met but the, other one is. This has been a high miss question since
most do not think of the autostart for the tripped pump.
A - Incorrect; The stby condensate pump should start immediately when the other
condensate pump trips (if the stby pump is in AUTO, which is the normal alignment at
80% power). Plausible because the low suction pressure auto start of < 275 psig for>
10 seconds has not yet been met. Tripping the reactor is not required unless
approaching trip criteria or if 80TH SGFPs are tripped and this has not happened.
8 - Correct; The standby condensate pump SHOULD auto start immediately when the
other condensate pump trips. With suction pressure dropping, AOP-13 directs verifying
stby pump started prior to 275 psig decreasing, If pressure continues to drop, rapidly
ramp down lAW AOP-17, Rapid Load Reduction.
C - Incorrect; first part is NOT true: condensate pump should have auto started on the
trip of the other pump.
second part is not correct for this situation. It is plausible in that if the candidate
thought both SGFPs tripped due to the alarm being in for >275 psig, then this would
be correct.
D - Incorrect; first part is NOT true: condensate pump should have auto started on the
trip of the other pump.
Second part is incorrect for this condition also, but plausible because the ramp at :s. 5
MW/MIN is incorrect for a condensate pump but plausible since it is correct for a HDT
pump trip also covered by AOP-13.
At 275 psig falling the standby condensate pump will start after 10 sec. IF suction pressure is
NOT greater than 275 psig within 30 sec, THEN the SGFP's will trip. This could result in a
Monday, January 14, 20082:42:22 PM 103
QUESTIONS REPORT
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056 A2.04 Condensate System
Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate
System; and (b) based on those predictions, use procedures to correct, control, or mitigate the
consequences of those mal-functions or operations:
Loss of condensate pumps
Question Number: 37
Tier 2 Group 2
Importance Rating: 2.6
Technical Reference: AOP-13 & ARP 1.10 KB4
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41 .1 0
Comments:
significantly modified the distracters to make sure this does not answer 059 A1.03 and this is
also a different part of AOP-13.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: B DCADADDB B Scramble Range: A - D
Source: MODIFIED Source if Banle
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:22 PM 104
QUESTIONS REPORT
for 75 RO Questions
41. 057 AK3.01 001
Given the following:
- Unit 1 is at 14% power and ramping up in preparation for rolling the main
turbine.
- The bypass feed regulating valves are in service, and SG level is being'
maintained at 65%.
- 1G. 4160 V Bus has been de-energized due to DG15,.1B. S/U XFMR to 4160V
Bus 1G, tripping open.
- The 1B DG has re-energized the 'BI Train emergency buses.
Which ONE of the following describes the correct operator actions lAW AOP-5.0, Loss
of A or B Train Electrical power? .
A. Trip the reactor and restore the 1G 4160 V bus to the grid.
B. Shut down the reactor and place the unit in Mode 3.
c~ Stabilize the plant and restore the 1G 4160 V' bus to the grid.
D. Verify the reactor tripped and stabilize the unit in Mode 3.
AOP-5.0 Version 24
A. incorrect - AOP-5.0 no longer directs tripping the reactor.
B. Incorrect- Ramping off 'Iine is not required in AOP-5.
C. correct - lAW AOP-5, step 14 the long term status is to maintain the reactor stable
and then restore the grid, then continue with whatever procedure the operatorwas in at
the time of the problem. This was changed a few years ago to prevent an unnecessary
reactor trip for this condition.
D. Incorrect - an automatic reactor trip will not occur.
Monday, January 14, 2008 2:42:22 PM 105
QUESTIONS REPORT
for 75 RO Questions
. APE 057 Loss of Vital AC Instrument Bus -
AK3.01 Loss of Vital AC Instrument Bus, Knowledge of the reasons for the following
responses as they apply to the (Loss of Vital AC Instrument Bus):
Actions contained in the EOP for loss of vital AC electrical instrument bus.
Question Number: 56
Tier 1 Group 1
Importance Rating: 4.1
Technical Reference: AOP-5.0
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41 .10
Comments:
this KA is for a loss ofa vital bus which is an emergency bus and/or panel at FNP. This meets
the KA in that there is a loss of the vital bus and the procedure guidance of AOP-5 is followed.
The reason for those actions is implied and agreed upon as part of the actions that would be
done.. FNP does not have a loss of a vital panel procedure and that issue is vaguely discussed
in an ARP in which it says if a vital panel has been lost, then recover from it when the event is
over by doing the following that apply.
MCS Time: 1 Points: 1.00 Version: o 123 4 5 6 7 8 9
Answer: CCCCBAACBA Scramble Range: A - D
Source: MODIFIED Source if Banle FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:22 PM 106
QUESTIONS REPORT
for 75 RO Questions
42. 058 G2.1.32 002
Given the following on Unit 1:
- A loss of "A" Train Auxiliary Building Battery bus has occurred.
- The crew is performing AOP-29.1, Plant. Stabilization in Hot Standby and
Cooldown Without "A" Train AC or DC Power.
- The RCS temperature is 547°F and pressure is being maintained 2220 psig.
- Seal injection flow has been lost to all three Reactor Coolant Pumps (RCPs).
- LB3, RCP THRM BARR ISO HV-3184 AIR PRESS LO, annunciator has
come into alarm.
- All three RCPs have been tripped.
Which ONE of the following correctly describes one of the actions required by
AOP-29.1 for the conditions given above, and the reason for performing that action?
A':' Isolate seal injection flow to all RCPs to prevent potential RCP seal damage.
B. Isolate seal injection flow to all RCPs to prevent a potential radioactive release to
the Auxilia'ry Building from occurring.
C. Isolate seal return flow to all RCPs to prevent potential RCP seal damage.
D. Isolate seal return flow to all RCPs to prevent potential ~hermal barrier heat
exchanger damage.
A is correct. CAUTION (for step 5): CAUTION: To prevent potential' seal damage,
neither seal injection flow nor CCW flow to the thermal barrier shall be re-established to
an RCP which has lost both seal injection and CCW cooling.
the background documents for a loss of all ac describe what happens on a loss of chg
and CCW to a seal. The seal injection is isolated to prevent potential seal damage.
B is incorrect. This is a correct action but the reason is for the seal return flow.
C. is incorrect. see explanation below.
D is incorrect. seal return is isolated but not to prevent seal damage. It is isolated for
several reasons:
Isolating the seal return line prevents seal leakage from filling the volume control tank (VCT)
(via seal return relief valve outside containment) and subsequent transfer to other auxiliary
building holdup tanks (via VCT relief valve) with the potential for radioactive release within
the auxiliary building. Such a release, without auxiliary building ventilation available, could
.---1-1-'1imit-p-er-so-rm-et-aceess--f-or-toeatoperations.
Isolating the s.eal return line also enables pressure in the number 1 sealleakoff line to increase
up to the relief valve setpoint of 150 psig. Maintaining a backpressure in the sealleakoff line. of
at least 150 psig enables development of high pressure in the number 1 seal leakoff cavity with
a steady-state seal leakage rate established due to the self-limiting leakage characteristic of
Monday, January 14, 2008 2:42:22 PM 107
QUESTIONS REPORT
for 75 RO Questions
the number 1 seal. Under these conditions, with the number 1 seal functioning as expected and
the number 2 seal remaining closed, the expected leakage flow rate is 21.1 gpm/pump.
This is consistent with the steady state pressure distribution and seal leakage determined in the
WCAP-10541 analysis and used in the latest RCP seal leakage PRA model in WCAP-15603
3.0 IF ONE of the following conditions occur at anytime during the event,
AND cannot be readily restored.
[] A total loss of RCP seal cooling as indicated by loss of seal
injection and loss of CCW to the Thermal Barrier Heat Exchanger.
[] A total loss of the operating train of charging without the ability to
quickly restore the redundant train, and RCP sealleakoff before the
loss was less than 2.5 gpm per pump.
bac,kground document for ECP-O
Purpose: To isolate the RCP seals
Basis: This step groups three'actions, with different purposes, aimed at isolating the RCP
seals. The actions are grouped since all require an auxiliary operator, dispatched from the
control room, to locally close containment isolation valves (the reference plant utilizes motor
operated valves for the RCP seal return, RCP thermal barrier CCW return lines and RCP seal
injection lines). This grouping assumes that the subject valves are located in the same
penetration room area and that they are accessible. Concurrent with dispatching the auxiliary
operator, the control room operator should place the valve switches for the motor operated
valves in the closed position so that the valves remain closed when ac power is restored .
. Isolating the RCP seal injection lines prepares the plant for recovery while protecting
the RCPs from seal and shaft damage that may occur when a charging/51 pump is
started as part of the recovery. With the RCP seal STEP DESCRIPTION TABLE FOR
ECA-O.OStep 8 injection lines isolated, a charging/51 pump can be started in the normal
charging mode without concern for cold seal injection flow thermally shocking the
RCPs. Seal injection can subsequently by established to the RCP consistent with
appropriate plant specific procedures. Isolating the RCP thermal barrier CCW return
outside containment isolation valve prepares the plant for recovery while protecting the
CCW system from steam formation due to RCP thermal barrier heating.
Following the loss of all ac power, hot reactor coolant will gradually replace the
normally cool seal injection water in the RCP seal area.
Monday, January 14, 20082:42:22 PM 108
QUESTIONS REPORT
for 75 RO Questions
APE 58 G2.. 1.32 Loss of DC Power
Conduct of Operations: Ability to explain and apply all system limits and precautions.
Question Number: 50
Tier 1 Group 1
Importance Rating: 3.4
Technical Reference: AOP-29.1
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41 .7/1 0
Comments:
This meets the KA in that under the conditions given, a loss of DC power, a system caution to
isolate CCW and Seal injection to the RCP seal to prevent seal damage upon re-initiation of
either Seal injection or CCW flow which could cause a SBLOCA from the RCP seal of 300
gpm. This question asks the operator to explain why these actions are taken.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: AAAAAAAAAA Items Not Scrambled
Source: NEW Source if Banle
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:22 PM 109
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for 75 RO Questions
43. 059 Al.03 001
Given th'e following:
- Unit 1 is at 100% power.
- 1A SGFP has tripped.
- Emergency Boration is in progress.
Which ONE of the following describes the subsequent restrictions on operation of the
unit in accordance with AOP-13.0?
(1) Load must be reduced to _
(2) The reactor must be tripped immediately if _
A. (1) ~ 730 MWe;
(2) SG NR levels cannot be maintained above the minimum value specified.
B. (1) ~ 730 MWe;
(2) FE1, CONT ROD BANK POSITION LO, annunciator comes into alarm
with Tavg at 577°F.
c~ (1) ~ 540 MWe;
(2) SG NR levels cannot be maintained above the minimum value specified.
D. (1) ~ 540 MWe;
(2) FE1, CONT ROD BANK POSITION LO, annunciator comes into alarm
with Tavg at 577°F.
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A. is incorrect. Due to load setpoint, although plausible because this is the value at
which the decrease load button is released when DEH manual load reduction is used
per the RNO column.
B is incorrect. Due to load setpoint, although plausible because this is the value at
which the decrease load button is released when DEH manual load reduction is used
per the RNO column.
TAVG 541°F - 580°F 577 F is still within the band to not trip.
AOP-13 has a step to evaluate the plant per the below:
1.14 IF the Team is NOT confident that a parameter is being restored, THEN trip the reactor and go to
FNP-1-EEP-O, REACTOR TRIP OR SAFETY INJECTION.
one of the parameters checked is FE2, not FE1. Since the team is emergency
borating per the procedure, it is unlikely this alarm would come in, but if it did the
reactor is not required to be immediately tripped. If FE2 came into alarm, then action
here would be to place rods in manual and with the emergency boration in progress,
evaluate the plant, not immediately trip the reactor.
C. Correct. AOP-13.0 directs load to be reduced to 540 MWe, and SG NR levels must
remain above 28%.
AOP-13 step 1.8
IF SG narrow range levels NOT maintained greater than 28%, THEN trip the reactor
and go to FNP-1-EEP-0, REACTOR TRIP OR SAFETY INJECTION.
D is incorrect. see B above for second part.
TAVG 541°F - 580°F 577 F is still within the band to not trip.
OPERATOR ACTION for FE!
1. Check indications and determine that actual control bank rod
position is at low insertion limit.
1.1 Click on Rod Supervision button on Applications Menu.
1.2 Click on Rod Insertion Limits button.
1.3 Determine if low insertion limit exceeded.
2. IF reactor coolant system dilution is in progress,
THEN stop dilution.
3. IF a plant transient is in progress,
THEN place the turbine load on "HOLD".
4. Refer to FNP-1-UOP-3.1, POWER OPERATIONS.
5. Borate the Control Bank "OUT" as necessary using the Boron
Addition Nomographs. {CMT 0008900}
6. Refer to the Technical Specifications section on Reactivity
.-C-'-F\ontror+--l.----------------------------------
Monday, January 14, 2008 2:42:22 PM 111
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for 75 RO Questions
059 A1.03 Main Feedwater System
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits)
associated with operating the MFW controls including:
Power level restrictions for operation of MFW pumps and valves.
Question Number: 18
Tier 2 Group 1
Importance Rating: 2.7
Technical Reference: AOP-13.0, FE2 and KB4
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.7
Comments: good match for KA since it tests power level restrictions with 1 SGFP and
monitoring levels to prevent exceeding design limits and the actions required.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: C C C A C C D A C C Scramble Range: A - D
Source: NEW Source if Barue
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:23 PM 112
QUESTIONS REPORT
for 75 RO Questions
44. 061 K1.07 001
Which ONE of the following describes the Service Water Train normally aligned to the
TDAFW Pump for Emergency Makeup, and if the TDAFW Pump is running and
aligned to the CST, the MINIMUM time available to swap to the emergency supply
when JD4 and JE4, CST LVL La-La A and B TRN, alarms are received?
(1) Servic~ Water Train normally aligned to _
(2) The minimum time available to swap to the emergency supply after receiving the
CST La-La level alarm is _
Atf (1) A Train
(2) 20 minutes
B. (1) A Train
(2) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
C. (1) B Train
(2) 20 minutes
D. (1) B Train
(2) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
A is correct. The CST La-La level alarm received means that at least 20 minutes of
normal supply remains.
1
65,300 Gallons (5 3") is when the La-La comes in.
B is incorrect. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is the time that 150,000 gallons of CST will maintain Hot
Standby..
C is incorrect. Wrong train, plausible because B train can be aligned' and would be if A
train power or SW was unavailable.
D is incorrect. Wrong train and time.
SOP-22, AFW section 4.7
4.7 Aligning Service Water to the AFW System. This shows how to swap SW to B train and
how to align it to A Train. The initial valve line up per the sOP also shows it is aligned to A
Train and D-l 75007 shows the normal line uR is to A Train.
FSAR 6.55 instrumentation after recieving low level alarm setpoint
Monday, January 14, 20082:42:23 PM 113
QUESTIONS REPORT
for 75 RO Questions
061 K1 . 07 Auxiliary Emergency Feedwater (AFW) System
Knowledge of the physical connections and/or cause-effect relationships between the AFW and
the following systems:
Emergency water source.
Question Number: 19
Tier 2 Group 1
Importance Rating: 3.6
Technical Reference: LP 52102H AFW D-175007 and SOP-22 AFW
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.5
Comments:
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: A A A A DCA D B C Scramble Range: A -" D
Source: NEW Source if Banle
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14,2008 2:42:23 PM 114
QUESTIONS REPORT
for 75 RO Questions
45. 062 G2.1.23 001
Given the following:
1
- Unit 1 is operating at 38% power with IA Train on service.
- The 1C Service Water Pump is tagged out for motor replacement.
A fire is reported on the 1K 4160 volt bus and the plant operators de-energize 1K 4160
volt bus. The following conditions exist:
slowly.
Which one of the following actions are required lAW AOP-1 0.0, Loss of Service
Water?
A. Trip the reactor and perform EEP-O, Reactor Trip or Safety Injection.
B. Reduce power to less than 35%, then trip the Main Turbine and refer to AOP-3.0,
Turbine Trip below the P-9 Setpoint.
C~ Start 1A CCW pump and 1C charging pump, secure 1A charging pump, and swap
on service trains of CCW.
D. Start 1C CCW pump and 1A charging pump, secure 1C charging pump, and swap
on service trains of CCW.
Monday, January 14,2008 2:42:23 PM 115
QUESTIONS REPORT
for 75 RO Questions
I*
A. Incorrect - a rx trip is not required at this time. AOP-1 0.0 Step 4.2.2.2 RNO and step
6.3 has the crew trip the reactor if one train of SW does not have at least 60 psig.
Since one train i.s available and operating, a reactor trip is not called for.
B. Incorrect - removing the main turbine from service would be an option if both trains
of SW were less than 60 psig per the RNO step 6 of AOP-10 if power was less than
35%. Being so close to 35% in the stem and temperatures are elevated slightly
makes this plausible since it would make sense to ramp to below 35% and remove
the turbine from service vs trip the rx.
C. Correct - Start a CCW pump and charging pump in the nonaffected train, secure
affected train charging pump, and swap on service trains of CCW.
This is the correct response because enough time is allowed before RCP temps
increase to 195°F to mitigate the loss of SW and prevent the need to trip the reactor.
It will take time to heat up the On service train of CCW, and AOP-1 0.0 takes that
into account.
D. Incorrect - Swapping on-service trains is partially correct, but the unit is not ramped
to 35% power and the main turbine tripped.
APE 062 Loss of Nuclear Service Water -
G2.1.23 Conduct of Operations: Ability to perform specific system and integrated plant
procedures during all modes of plant operation.
Question Number: 51
Tier 1 Group 1
Importance Rating: 3.9
Technical Reference: AOP-10
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.10
Comments:
this meets the KA in that the ability to perform AOP-1 0 is demonstrated during mode 1 at a low
power level.
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: C A CAB BCD A A Scramble Range: A - D
~SO"\-l-'lut'rFlc~e....L-.-----------------H-BANK:-------S~o~u~lce~____EY-----------
Cognitive Level: IDGHER Difficulty:
Job Position: R.O Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:23 PM . 116
QUESTIONS REPORT
for 75 RO Questions
46. 062 K2.01 001
Given the following conditions on Unit 2:
- IAI train is the liOn Service train.
ll
- 2A Charging Pump breaker has been racked out for maintenance.
- 2G 4.160 V Bus has been de-energized due to a fault.
Which one of the following states the ECCS pumps that will have power based on
current conditions?
A':' 2B Charging Pump, 2A RHR Pump.
B. 2B Charging Pump, 2B RHR Pump.
C. 2C Charging Pump, 2A RHR Pump.
D. 2C Charging Pump, 28 RHR Pump.
A. correct, both powered from 2F
B. Incorrect, 28 chg powered from 2F. 28 RHR powered from 2G
C. Incorrect, 2C chg powered from 2G. 2A RHR powered from 2F
D. Incorrect, both powered from 2G
062 K2.01 A.C.. Electrical Distribution
Knowledge of bus power supplies to the following:
Major system loads .
Question Number: 20
Tier 2 Group 1
Importance Rating: 3.3
Technical Reference OPS 521038, Unit 1 Equipment Load List A506250
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.5
Comments:
KA match in that these are major loads on the emergency busses.
MCS 'rIme: POInts: lJJO VerSIon: O-r2-r45 6 7 8 9
Answer: ADDCDB CAB D Scramble Range: A - D
Monday, January 14, 2008 2:42:23 PM 117
QUESTIONS REPORT
for 75 RO Questions
Source: MODIFIED Source if Bank: FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:23 PM 118
QUESTIONS REPORT
for 75 RO Questions
47. 063 A3.0l 001
Unit 1 is at 1000/0 power with the following conditions:
- 1A Battery Charger is on service.
- EM personnel are doing preventative maintenance on the 1A battery..
The following indications are received:
- The UNIT 1 AUX BLDG DC BUS - A TRN GROUND DET white light comes on
momentarily and then goes OFF. .
- Then the following alarms are received:
- WC3, 1A 125V DC BUS BATT BKR 72-LA05 TRIPPED
- THEN WC2 clears.
Which ONE of the following describes the status of the indications on the EPB for the
1A DC BUS and the1A and 1B Inverters?
1A DC BUS VOLTAGE reads approximately (1)
1A and 1B INVERTER AMPERES are reading approximately (2)
A. (1) 0 DC VOLTS.
(2) 25 amps and being powered from the bypass source.
B. (1) 0 DC VOLTS.
(2) 0 amps and being powered from the normal source.
C. (1) 125 DC VOLTS.
(2) 0 amps and being powered from the bypass source.
D~ (1) 125 DC VOLTS.
(2) 25 amps and being powered from the normal source.
Monday, January 14, 20082:42:23 PM 119
QUESTIONS REPORT
for 75 RO Questions
explanation
When the Battery output breaker is opened, LA-05, WC3 will come into alarm due to
the b contact from breaker LA05. WC2 shows either a low voltage condition or a
ground. In this case it would be a ground.
The battery output breaker has opened due to a ground on the battery and when it
opens WC2 clears. The annunciators provide indication that the breaker opened and
the white light provides indication of the ground. For this set of circumstances, the
battery'is no longer aligned to the bus and the battery charger is carrying the load. The
indications will remain normal and the inverters will have normal indications. The
inverters will not swap to the bypass source and will still be powered from the BC.
A. Incorrect. 0 DC volts on the 1A DC bus indicates the bus is de-energized. The bus
still has power from the Batt. chger. The inverters will be powered from the BC or the
normal supply and will indicate 25 amps. If it were to swap to the bypass source, it
would still have amp readings, but if the manual bypass switch were to be placed in the
bypass position, then the amps would be 0 amps.
B. Incorrect. 0 is not correct for both. Normal is correct.
C. Incorrect - 125 is correct. 0 is not correct and bypass is not correct.
D. Correct. 125 is correct and 25 is correct from the normal source.
063 D.C. Electrical Distribution
A3.01 Ability to monitor automatic operation of the de electrical system, including:
Meters, annunciators, dials, recorders, and indicating lights
Question Number: 21
Tier 2 Group 1
Importance Rating: 2.7
Technical Reference: ARP WC2, WC3 and 0177082
Proposed references to be provided to applicants during examination: None
Learning Objective:
lO CFR Part 55 Content: 41.7
Comments: This was replaced to fully meet the KA.
It meets the KA in that it tests the ability to determine the proper readings on the EPB for an
abnormal condition based on the indications and alarms received (white light and
annunciators). The automatic portion of the KA is the breaker opening on an overcurrent
con-rntlnn.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: D C B D D DCA B A Scramble Range: A - D
Monday, January 14, 20082:42:23 PM 120
QUESTIONS REPORT
for 75 RO Questions
Source: NEW Source if Banlc
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:23 PM 121
QUESTIONS REPORT
for 75 RO Questions
48. 063 'K4.04 001
Given the following:
Unit 2 is at 100% power. Reactor Trip Breakers A (RTA) and B (RTB) are closed,
Reactor Trip Bypass Breakers are open.
- 125V DC distribution panel breaker 2B-16, IIAII Reactor Trip switchgear
control power to Bypass breaker and Reactor Trip breaker, has tripped
open.
Which one of the following statements correctly describes how a loss of DC to the
A Train reactor trip switchgear would effect the operation of Reactor Trip Breaker A?
A. RTA will immediately trip open.
B~ RTA will remai'n closed and will still open from either a manual or automatic signal.
C. RTA will remain closed and will not open from either a manual or automatic signal.
D. RTA will remain closed and will not open from a manual reactor trip signal; an
automatic trip will still open the breaker.
Monday, January 14, 2008 2:42:23 PM 122
QUESTIONS REPORT
for 75 RO Questions
A. Incorrect - This will not cause a direct Rx trip due to SSPS still powered up and the
only loss is DC to the STC Rx Trip.
B. Correct- RTA would still open from either a manual or automatic signal.
125 V DC bus A allows the RTA shunt trip coil to energize on a Rx trip. With no DC
available, the Rx trip Brkers will still open on a signal from SSPS A train by
de-energizing the UV coil. With the loss of the DC, no RT brker will open immediately
because SSPS is still energized and the breaker does not have a trip signal.
C. Incorrect - It will open on both.
D. Incorrect - manual Rx trips cause the STC to be energized and the UV coil to be
de-energized. A Rx trip will still operate on a loss of DC due to the UV coil.
according to Table 6 of OPS-521 03C, 125V DC distribution panel feeds to Rx trip swgr
- 1. According to the load list 'page F-51/52 LA-13 feeds 125V DC A Rx trip swgr
control power to Byp brker & Rx trip bker. (2B-16)
FSD A-181007
2.2.18 The RPS shall be designed for fail safe operation. Loss of power to the protection logic
or rod control system shall trip the reactor. The only exception to fail safe criteria shall be
containment spray (H1-3) and shunt trip attachment for reactor trips. (References 6.1.022,
6.7.031) page 3-10
The Shunt Trip Attachment coil- shall operate on 125 Vdc and function as a backup for the
undervoltage trip device.
From the load list. this is the breaker designation and the nomenclature for this panel
II
125V DC distribution panel breaker 2B-16, IIA Reactor Trip switchgear control power to Bypass
breaker and Reactor Trip breaker,
Monday, January 14, 20082:42:23 PM 123
QUESTIONS REPORT
for 75 RO Questions
063 K4.. 04 D"C.. Electrical Distribution
Knowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the
following:
Trips
Question Number: 22
Tier 2 Group 1
Importance Rating: 2.6
Technical Reference: Electrical Dist FSD, DC LP; SEQ LP OPS-52103F reactor
protection lesson plan and FSD A-18100?
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.7
Comments: replaced this question to meet the KA. This is a backward way to meet the KA in
that DC power is provided to theRx trip breakers to cause a rx trip via the shunt trip coil. A
loss of the DC will not allow DC to provide for a rx trip so one will not occur from the STC but
will occur from SSPS via the UV coil. This requires knowledge of the DC bus supply to the RT
- breakers and what the loss means to the operator.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: B CDC CAACDB Scramble Range: A - D
Source: BANK Source if Banle FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:23 PM 124
QUESTIONS REPORT
for 75 RO Questions
49. 064 K6.08 001
Given the following:
- Hurricane force winds have caused damage in the HVSYD and a dual unit
LOSP.
- Long term Emergency Diesel Generator operation is anticipated.
- It is not known how soon the on-site Diesel supplies can be replenished
- The Fuel Oil Day Tank for 1-2A DG is full.
Which ONE of the following describes the maximum time available for 1-2A DG to
cO.ntinue to run at full load?
A. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
B~ 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
c. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
D. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
Monday, January 14, 20082:42:24 'PM 125
QUESTIONS REPORT
for 75 RO Questions
A. Incorrect. The day tank is sized to supply 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at full load, 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> at minimum
TS level.
B.Correct.
C. Incorrect. Plausible because it is close to the amount of time and symmetrical as a
distractor
D. Incorrect. Plausible due to amount of time and also name of tank (day tank)
Lesson plan 52101 i
Diesel Fuel Oil Storage System
Refer to Figure 14. Each diesel generator is connected to a shared fuel oil storage and
transfer system, which consists of five storage tanks, two fuel oil transfer pumps per storage
tank, a day tank for each diesel generator, and interconnecting piping and valves. The diesel fuel
oil storage system has a total of five 40,000-gallon storage tanks, two 1000-gallon day tanks (for
the little diesels), three 1325-gallon day tanks (for the big diesels), *and two redundant capacity
fuel oil transfer pumps per storage tank. The storage tanks are designed with sufficient fuel oil
storage capacity to supply the minimum number of diesels required for seven days of operation
with ten percent excess for testing using the deliverable capacity of four of the five storage tanks.
The electrical distribution system supplies Class 1E, 120V AC power to each storage tank level
transm itter.
Two fuel oil transfer pumps are mounted on each storage tank. The pumps are motordriven,
vertical, submersible, wet pit-type pumps. The capacity of each pump is in excess of the
amount required to simultaneously supply the diesel generator full load fuel requirements and fill
the associated day tank. One pump automatically maintains the required day tank level, and the
other is strictly manual. When full, the day tanks provide sufficient storage for four hours of full
load *operation. Tech Spec required minimum day tank volume ensures 'sufficient fuel oil is
available to allow 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of full load operation of the respective diesel generator.
FSD 3.6.3.2 A-181005
Functional Requirements
Each diesel engine has an individual fuel oil day tank. When
full, the fuel oil day tanks have sufficient fuel oil volume to
supply their respective diesel for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at continuous rated
load. However, the Technical Specifications state that the
minimum amount of fuel oil is 900 gallons each in the day
tanks of the 1-2A, 1Band 2B diesel engines and 700 gallons
each in the day tanks of the 1Cand 2C diesel engines. This
capacity is sufficient to ensure that each day tank can supply
its corresponding diesel with at least three hours of fuel at
continuous rated load (References 6.1.006, 6.1.014, 6.3.012,
6.5.014,6.7.009 and 6.7.030).
Monday, January 14, 20082:42:24 PM 126
QUESTIONS REPORT
for 75 RO Questions
064 KG.OS Emergency Diesel Generators -
Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system:
Fuel oil storage tanks
Question Number: 23
Tier 2 Group 1
Importance Rating: 3.2
Technical Reference: TS 3.8.3.a, f; DG LP OPS-521 021
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.10/43.2
Comments: good question that tests length of time a DG can run with the day tank full and no-
other fuel available. This is the loss of portion of the KA and the affect as it relates to the DGs.
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: B BCD C C C C D D Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:24 PM 127
QUESTIONS REPORT
for 75 RO Questions
50. 065 AK3.03 003
Given the following:
- Unit 2 is at 100% power.
- N2P19HV3825, Instrument Air to Penetration Room valve, has closed and
cannot be opened.
Which ONE of the following will occur with no operator action taken?
A. Pressurizer pressure and level will remain stable.
B. Pressurizer pressure will increase until the PORVs lift.
C!' Pressurizer pressure and level will increase until a reactor trip occurs.
D. Pressurizer level will decrease until letdown isolates and backup heaters turn off,
then increase until a reactor trip occurs.
HV 3825 supplies air to AB loads. FCV-122 will go full open, letdown will secure sprays,
PORVs will go closed and stay closed.
A. Incorrect- Due to the loss of air pre'ssure Przr pressure will be rising and level will be
rising due to FCV-122 and loss of letdown.
B. Incorrect- there is no air to the PORVs
C. Correct- due to no air, charging will be at a max rate and letdown will secure.
Pressure will also be rising and a Rx trip on high pressure will occur.
D. Incorrect- Level will actually rise.
APE 065 Loss of Instrument Air
AK3.03 Knowledge of the reasons for the following responses as they apply to the Loss of
Instrument Air:
Knowing effects on plant operation of isolating certain equipment from instrument air
Question Number: 52
Tier 1 Group 1
Importance Rating: 2.9
Technical Reference: AOP-6.0
____----"-P--=-r-=-toPQ-s-fid---Le1e re ncJ~sj!Lb_e-PmJljded1o_'lppJicants-CiurinQ_6xaminatioJL---None
Learning Objective:
10 CFR Part 55 Content: 41 .7
Comments: changed out the question completely to meet the KA.
Monday, January 14, 20082:42:24 PM 128
QUESTIONS REPORT
for 75 RO Questions
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: CDCBDCBACD Scramble Range: A - D
Source: MODIFIED Source if Bank: FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:24 PM 129
QUESTIONS REPORT
for 75 RO Questions
51. 068 G2.1.20 002
Given the following:
- Both Units are operating at 100% power.
- Toxic gas has made the Control Room inaccessible.
- AOP-28.0, Control Room Inaccessibility, has been implemented.
Which ONE of the following are the minimum and complete actions required lAW
AOP-28.0 before leaving the control room?
A. Trip the reactor an'd trip the main turbine ONLY.
B. Trip the reactor, trip the main turbine, trip both SGFPs, and sound the plant
emergency alarm.
C~ Trip the reactor, trip the main turbine, verify at least one train of 4160 V ESF buses
are energized, and sound the plant emergency alarm.
D. Trip the reactor, trip the main turbine, verify at least one train of 4160 V ESF buses
are energized and a,ctuate a safety injection.
Monday, January 14, 2008 2:42:24 PM 130
QUESTIONS REPORT
for 75 RO Questions
A. Incorrect. These are the Immediate actions of FRP-S.1, not lAW AOP-28. These
actions are not complete lAW AOP-28.
B. Incorrect. Both SGFPs are not required to be tripped at this point in the procedure.
It is done locally at step 15. Sounding the plant emergency alarm is the only place this
can be done and is in the first four steps..
C. Correct. These are the strategies addressed by the steps in the procedure.
AOP-28.0 rev 11 actions:
1.0 Verify reactor tripped.
2.0 Verify the turbine tripped.
3.0 Verify at least o,ne train of 4160 V ESF buses energized.
4.0 Perform the following.
4.1 Direct Operation's personnel to man the Hot Shutdown Panels.
4.2 Actuate the plant emergency alarm.
4.3 Announce "Main control room evacuation. Report to your
designated assembly areas."
4.4 Verify control room and C.A.S. evacuated.
4.5 Notify appropriate support groups to report to the Hot Shutdown ??
Panels.
4.6 Direct Security to station personnel at each co'ntrol room door to
prevent entry.
D. Incorrect. These are almost the first 4 steps of E-O, an SI is actuated if one is
required. however, if no signal is calling for an SI, it would not be conservative to
actuate an SI, and might be considered a good idea, but not lAW AOP-28. Checking
the SI actuated is not required.
Monday, January 14, 2008 2:42:24 PM 131
QUESTIONS REPORT
for 75 RO Questions
APE068 Control Room Evacuation
G2.1 . 20 Conduct of Operations:
Ability to execute procedure steps.
Question Number: 65
Tier 1 Group 2
Importance Rating: 4.3
Technical Reference: AOP-28
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS 52533M02
10 CFR Part 55 Content: 41 .1 0
Comments:
This is a new KA replaced per FJE. This question asks the operator the actions required that
an RO should know to evacuate the control room. These should be committed to memory and
if not properly executed would cause operational concerns. This demonstrates the ability to
execute procedural steps in AOP-28.0.
MCS. Time: 1 Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: CADDDABBDB Scramble Range: A - D
Source: MODIFIED Source if Banle FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO .. Previous 2 NRC exams: YES
Monday, January 14, 2008 2:42:24 PM 132
QUESTIONS REPORT
for 75 RO Questions
52. 069 AA2.01 007
Which ONE of the following conditions represents a loss of containment integrity and
would cause entry into Technical Specification 3.6.1, Containment?
A. . Mode 3 and one of the Personnel Airlock doors will not close.
B. rtI Mode 4 and Integrated Leak Rate test determines that leakage is not within limits.
C. Mode 5 and it is discovered that the Phase IBI isolation valve for CCW to the
RCPs, will not close.
D. Mode 6 and the Equipment Hatch is held in place by 4 bolts.
Containment integrity
A is incorrect. Both doors inop would be a loss of Containment Integrity, this is just an
inop of one of the doors in the Personnel Airlock Plausible becasue one of two series
valves makes containment integrity LCO not met.
B is correct. Surveillance requires ILRT to be within limits for Containment Integrity to
be set.
C is incorrect. because Containment Integrity is not required in Mode 5, plausible
because the valve is part of a containment penetration that would affect integrity in
modes 1-4.
D is incorrect. 4 bolts meets the minimum requirement for Containment Closure in
Mode 6, but not containment integrity in the modes that containment integrity is
required.
Monday, January 14, 20082:42:24 PM 133
QUESTIONS REPORT'
for 75 RO Questions
APE 069 AA2.. 01 Loss of Containment Integrity -
Ability to determine and interpr~t the following as they apply to the Loss of Containment
Integrity:
Loss of containment integrity
Question Number: 59
Tier 1 Group 2
Importance Rating: 3.7
Technical Reference: TS section 3.6
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS-52102A-1
10 CFR Part 55 Content: 43.2/41.10
Comments: meets the KA in that it tests the ability to determine IF Ctmt integrity is met in
different modes lAW Tech Specs.
Mode applicability (1-4) & one hour or less tech specs (one or more air locks with one
door inoperable) are RO Knowledge and this question meets KIA for ROs.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: B B B B B B B B B B Items Not Scrambled
Source: BANK Source if Bank: HARRIS
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:24 PM 134
QUESTIONS REPORT
for 75 RO Questio'ns
53. 072 G2.1.2 001
Which ONE of the following Area Radiation Monitors requires entry to a Technical
Specification Action Statement if it is declared INOPERABLE?
A. R-1 A, Control Room Area Radiation (Unit 1)
B. R-2, Containment Area Radiation
C. R-4, Charging Pump Area Radiation
D~ R-27A, Containment Area Radiation (High Range)
A. Incorrect. Plausible because this is an important radiation monitor indicating the
habitability of the Control Room, but it is not in TS.
B. 'Incorrect. Plausible because this is an important radiation monitor indicating the
abnormally high radiation level in containment. This is used in the emergency
procedures for diagnosis of a LOCA, but it is not in TS.
C. Incorrect. Plausible because this is an important radiation monitor indicating
radiation levels in the Charging Room area. This is used in the emergency procedures
for diagnosis of a LOCA outside Containment, but it is not in TS.
D is correct. This radiation monitor is in TS in 3.3.3 table, and monitored on STP-1.0
every shift to ensure operable.
19. Containment Area Radiation (High Range)
072 G2.1.2 Area Radiation Monitor
Conduct of ,Operations:
Knowledge of operator responsibilities during all modes of plant operation.
Question Number: 38
Tier 2 Group 2
Importance Rating: 3.0
Technical Reference: FNP-1-ARP-1.6 FH1
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.10/43.2
Comments: This meetsttieKAmt11at1f1Sfne operator responsfolltty to know what area rad
monitors are entry conditions to TSs.
Monday, January 14, 2008 2:42:24 PM 135 '
QUESTIONS REPORT
for 75 RO Questions
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: DB C CDADB AA Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:24 PM 136
QUESTIONS REPORT
for 75 RO Questions
54. 073 A4.02 001
Given the following:
- R-19, SGBD SAMPLE, radiation monitor is in alarm and stable above the -alarm
setpoint.
- The Shift Chemist requests to sample the Steam Generators.
Which ONE of the following correctly describes the actions that will allow the shift
chemist to obtain a sample of the SGs lAW SOP-45.0, Radiation Monitoring System?
A. Manually open the sample valves one at a time.
B. Pull the INSTRUMENT power fuses for R-19 to allow opening the sample valves.
C. Pull the DC power fuses to each sample valve solenoid to fail the valve open.
D~ Place R-19 Operations Selector Switch to the RESET position, then open the
sample valves.
A. Incorrect. These valves can not be manually opened. The SGBD sample valves do
not have manual jacks as they have solenoids powered from DC power and fail closed.
B. Incorrect. This is the procedure directed action-for a monitor in saturation, but not to
clear a valid alarm.
c. Incorrect. these solenoid valves are DC powered and fail closed. Even though the
designator is HV3328 and there is no manual operator on the valve.
D. Correct. SOP-45.0, Section 4.4 directs this.
QIPI5HV3328 IA Steam Generator Blowdown sample valve
QIPI5HV3329 IB Steam Generator Blowdown sample valve
QIPI5HV3330 Ie Steam Generator Blowdown sample valve
Monday, January 14, 20082:42:24 PM 137
QUESTIONS REPORT
for 75 RO Questions
073 Process*Radiation Monitoring System
A4.02 Ability to manually operate and/or monitor in the control room:
Radiation monitoring system contr~*1 panel
Question Number: 24
Tier 2 Group 1
Importance Rating: 3.7
Technical Reference: OPS 521060; SOP-45.0 section 4.4
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.11
Comments:
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: DAB B D A A C A A Scramble Range: A - D
Source: NEW Source if Banle
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:24 PM 138
QUESTIONS REPORT
for 75 RO Questions
55. '076 K3.0l 001
Given the following:
- Unit 2 is at 39% power.
- IAI Train is On-Service.
- 1-2A DG is running at full load lAW STP-80.1, Diesel Generator 1-2A
Operability Test, and tied to Unit 2.
I
- A Loss of IA Train Service Water is occurring due to SW Pump failure.
- The crew is performing actions of AOP-1 0.0, Loss of Train A or B Service
Water.
Which ONE of the following describes a potential effect on the unit and the actions
required in accordance with AOP-1 O.O?
A. * RCP Motor Air Coolers will lose cooling water flow;
temperature limit.
B~ * RCP bearing temperatures will rise;
- Trip the reactor and any RCP if its bearing temperature exceeds the
temperature limit.
C. * 1-2A DG will lose cooling water flow;
- Isolate Service Water to the Turbine Building and trip the reactor.
D. * Main Generator bearing and Hydrogen temperatures will rise;
- Isolate Service Water to the Turbine' Building and trip the reactor.
Monday, January 14, 2008 2:42:25 PM 139
QUESTIONS REPORT
for 75 RO Questions
A. Incorrect. plausible because Service Water does supply cooling to the motor air
coolers, but Train B does, not Train A.
B. Correct. CCW temperature will rise, and as it does, RCP bearing temperatures will
rise. This action is done in both AOP-10 and AOP-9 which AOP-10 sends the user to
to accomplish in conjunction with AOP-1 0 at step 11.
C. Incorrect. plausible because Service Water does supply cooling to this DG and is in
fact aligned to both units SW. Therefore a loss of Unit 2 does not cause a loss of
cooling water to the DG. The candidate may believe that for a unit 2 STP, SW would be
secured from unit 1 or flow is affected since STP-80.1 has the following note.
If service water from either unit is secured, a partial surveillance may be
performed for the non affected unit. Full surveillance credit for this STP may be
taken once service water is returned to service and verified operable by
rerunning this STP.
The actions would be performed if the DG was required. It is NOT required. The RNO
of step 4.2 says to isolate SW to the TB for the affected train and trip the reactor if
BOTH trains were isolated. In this case the DG would be secured or left running if SW
flow was sufficient from unit 1.
D. Incorrect.
First part is correct. second part incorrect since at step 6, SW pressure is checked to
be > 60 psig. For this event, the SW pressure is 72 psig. (If it was less than 60 psig,
then actions would be performed to isolate Service water to the Turbine Bldg. If this
were done, then a reactor trip would be required.) Since it is not required to go to the
RNO column, then it is not correct to do these actions.
076 K3.01 Service Water System
Knowledge of the effect that a loss or malfunction of the SWS will have on the following:
Closed cooling water
Question Number: 26
Tier 2 Group 1
Importance Rating: 3.4
Technical Reference: FNP-2-AOP-10.0
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.10
Comments:
This meets the KA in that there is a loss of SW to a train and requires knowledge of hOV\l_th----Cis=----_
affects the equipment that receives CCW and is cooled by the SWS. This also tests the actions
required for this event.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: B ADAB C B CDA Scramble Range: A - D
Monday, January 14, 2008 2:42:25 PM 140
QUESTI"ONS REPORT
for 75 RD Questions
Source: NEW Source if Bank:
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:25 PM 141
QUESTIONS REPORT
for 75 RO Questions
56. 078 G2.1.32 001
Which ONE of the following describes the reason why one Air Compressor should be
aligned with the air compressor panel key switch in 'LOCAL OR in the MCB position
'with the AUTOMATIC OPERATION LED (green light) LIT?
A-I To prevent a complete loss of Instrument Air pressure due to a single failure of the
sequencer panel pressure transducer.
B. To allow one air compressor to be started from the MCB and operate from the
selected sequencer for any complete loss of Instrument Air pressure situation.
C. To allow one air compressor to start after an LOSP and load and unload based on
its Internal Mode" pressure setting.
D. To prevent all three air compressors from running at the same time to prevent a
complete loss of Instrument Air pressure in the event that Service Water is lost to
the Turbine Building.
Monday, January 14, 2008 2:42:25 PM 142
QUESTIONS REPORT
for 75 RO Questions
A. Correct. see P&L below
3.19 Failure of the sequencer panel pressure transducer could unload all air
compressors selected (integrated) and result in complete loss of air pressure. To
prevent loss of air from a single failure, at least one air compressor should be
aligned with the air compressor panel key switch in LOCAL OR in MCB with the
AUTOMATIC OPERATION LED lit (green).
B is incorrect. The air compressor in LOCAL will start from the MCB if OFF is selected first,
then AUTO, but will not operate on the sequencer; but will start by its internal mode pressure
,switch.
3.8 Any air c9mpressor with the panel key switch in the MCB position will (1) stop if
the MCB handswitch is selected to OFF and (2) start and load if the MCB
handswitch is taken to the START/RUN position and returned to AUTO position,
based on the Internal Mode pressure settings on the air compressor. The
AUTOMATIC OPERATION LED on the air compressor panel will be lit (green)
when the MCB handswitch has been taken to the START/RUN position and
returned to AUTO. The lit LED indicates the air compressor will load and unload
based on its Internal Mode pressure settings. IF the MCB handswitch is taken
from START/RUN to OFF, THEN the air compressor panel AUTOMATIC
OPERATION LED will not be lit AND the LED will remain off if the MCB
handswitch is taken from OFF back to AUTO without going to START/RUN
AND the air compressor will not load and unload based on its Internal Mode
pressure settings.
C is incorrect. There is a P&L applicable to 1C air compressor and the compressor will cycle on
the sequencer after the load shed and LOSP is complete. This is not necessarily true for any
air compressor operation after an LOSP. The Air compressor will also sequence b,ack on***and
run on the sequencer, not the Internal Mode pressure setting.
3.11 During an LOSP or SI/LOSP the emergency section of Load Center 1A will
automatically align to Load Center 1D, and 1C air compressor supply breaker
EA-15 will automatically close. If the air compressor was operating prior to the
LOSP, the compressor will resume operation after the LOSP if (1) the MCB
handswitch is in AUTO (returned from START/RUN and not been take.n to OFF)
and the 1C panel key switch is in MCBposition OR (2) the 1C panel key switch
is in the SEQ position and 1C is selected (integrated) to the sequencer OR (3) the
1C panel key switch is in the LOCAL position.
D. incorrect. If air pressure drops with the switches in the above configuration, then all 3 air
compressors will be running. This does not prevent 3 a/cs from running.
The normal system line-up is three air compressors in AUTO on the MCB, two air
compressor selected (integrated) on the sequencer, and one air compressor de-selected (isolated)
from the sequencer. This will allow the sequencer to control two air compressors based on
header pressure and allow the de-selected (isolated) air compressor to auto start based on its
receiver pressure.
Monday, January 14, 2008 2:42:25 PM 143
QUESTIONS REPORT
for 75 RO Questions
078 G2.1.32 Instrument Air System
Conduct of Operations:
Ability to explain and apply all system limits and precautions.
Question Number: 27
Tier 2 Group 1
Importance Rating: 3.4
Technical Reference: FNP 1-S0P-31.0
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.10
Comments:
This meets the KA in that a precaution is required to be known and applied for a failure of a
pressure switch for the instrument air compressors..
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: A CAD A B A A C A Scramble Range: A - D
Source: NEW Source if Banle
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
'---------------------------
Monday, January 14, 20082:42:25 PM 144
QUESTIONS REPORT
for 75 RO Questions
57. 103 K4.06 001
Given the following conditions:
- A LOCA has occurred.
- Containment pressure is currently 19 psig and rising.
- All automatic actions have occurred as required.
- No manual actions have been taken.
Which ONE of the following describes the ESF actuations that have taken place?
A. Safety Injection ONLY.
B. Safety Injection and Containment Isolation Phase A ONLY.
C~ Safety Injection, Containment Isolation Phase A, and Main Steam Line
Isolation ONLY.
D. Safety Injection, Containment Isolation Phase A, Main Steam Line Isolation, "and
Containment Isolation Phase B.
A is incorrect. because if SI actuates, Phase A will also be actuated.
B is incorrect. because Phase A is actuated, but MSLI is also actuated.
C is correct. Containment Isolation Phase A, and Main Steam Line Isolation ONLY,
due to containment pressure.
D is incorrect. because containment pressure is not high enough for phase B.
Monday, January 14, 20082:42:25 PM 145
QUESTIONS REPORT
for 75 RO Questions
103 K4.06 Containment System
Knowledge of containment system design feature(s) and/or interlock(s) which provide for the
following:
Containment isolation system
Question Number: 28
Tier 2 Group 1
Importance Rating: 3.1
Technical Reference: E-O Attachment 3
Proposed references to be provided to applicants during examination: None
Learning Objective:
1 0 CFR Part 55 Content:
Comments:
This tests the KA appropriately in that these are design features that provide for ctmt isolation
at an RO level of knowledge.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: CDCDDAB C B B Scramble Range: A - D
Source: BANK Source if Banle WOLFCREEK
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:25 PM 146
QUESTIONS REPORT
for 75 RO Questions
58. E02 EAl.3 001
Given the following:
- Unit 1 was operating at 10% Reactor power when a Loss of Off-Site Power caused '
a loss of ALL RCP*s and a spurious safety injection.
- The crew has just adjusted the atmospheric relief valves.
determine if adequate natural circulation exists.
Which ONE of the following correctly lists indications that are consistent with adequate
natural circulation lAW ESP-1 .1 ?
1- RCS hot leg temperature --- stable or decreasing
2- RCS hot leg temperature --- increasing
3- SG pressure --- stable or decreasing
4'- SG pressure --- increasing
5 - RCS hot leg temperature --- at saturation for SG pressure
6 - RCS cold leg temperature --- at saturation for SG pressure
A. 2,3, and 5
B. 2,4, and 6
C. 1,4, and 5
D!' 1, 3, and 6
Monday, January 14, 2008 '2:42:25 PM 147
QUESTIONS REPORT
for 75 RO Questions
A. incorrect. RCS HL temps would not be increasing, and not RCS HL at SG saturation
temperature.
B. incorrect. RCS HL temps would not be increasing, SG pressure would not be
increasing
C. incorrect. SG pressure would not be increasing
D. Correct. 1, 3, and 6 -
ESP-1.1 step 21.4 RNO lists NC flow requirements:
Verify adequate natural circulation.
a) Check SG pressure stable or falling. #3
b) Check SUB COOLED MARGIN MO-NITOR indication greater than 16°F subcooled in CETC
mode.
c) Check RCS hot leg temperatures stable or falling. #1
d) Check core exit TICs stable or falling.
e) Check RCS cold leg temperatures at saturation temperature for SG pressure. #6
W/E02 81 Termination -
EA1.3 Ability to operate and I or monitor the following as they apply to the (81 Termination):
Desired operating results during abnormal and emergency situations.
Question Number: 60
Tier 1 Group 2
Importance Rating: 3.8
Technical Reference: ESP-1.1 steps' 2-18
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.10
Comments:
This question tests the ability to monitor for desired operating parameters lAW ESP-1.1, SI
termination, during natural circ flow conditions.
+/-'tMCS~~-;--l--.OO____V_e_FS_ion: 0 1 2 3 4 5 &-1-8-9"-1-------------
Answer: DAAC CADC B B Scramble Range: A - D
Monday, January 14, 2008 2:42:25 PM 148
QUESTIONS REPORT
for 75 RO Questions
Source: MODIFIED Source if Banle FARLEY
Cognitive Level: . LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:25 PM 149
QUESTIONS REPORT
for 75 RO Questions
59. E04 EKl.3 002
Given the following:
- A reactor trip and an SI have occurred.
- Containment pressure is reading 2 psig.
- RCSpressure is reading 1755 psig.
- All systems have operated as required."
At the step in EEP-1.0, Loss of Reactor or Secondary Coolant, the following indications
are observed by the Unit Operator:
- The following BOP annunciator is in alarm:
- NE2, 1B RHR PUMP RM SUMP LVL HI-HI OR TRBL
- 1A and 1B RHR pump discharge pressures are reading 750 psig.
- MK4, LIQ OR GAS PROC PNL ALARM, has just come into alarm.
Which ONE of the following describes 1) the operator actions; AND 2) the operational
implications of those actions performed lAW ECP-1.2, LOCA Outside Containment, in
an attempt to mitigate this leak?
Art 1) Isolate the discharge to ONE train of RHR and check RCS pressure rising;
2) Loss of one train of LHSI for injection and recirculation.
B. 1) Isolate the RWST suction to ONE train of RHR and check RWST level stable;
2) Loss of one train of LHSI for injection 9NLY. '
C. 1) Isolate the discharge to BOTH trains of RHR and check RCS pressure rising;
2) Loss of,BOTH trains of LHSI for inje'ction and recirculation.
D. 1) Isolate the RWST suction to BOTH trains of RHR and check RWST level stable;
2) Loss of BOTH trains of LHSI for injection ONLY.
A. Correct. At FNP, the most credible source of an ISLOCA is from the RCS to the
LHSI pump suction piping which is a low pressure system. The high level action steps
of ECP-1.2 are to verify proper valve alignment, attempt to isolate the break, check if
the break is isolated. The first system to be isolated is the RHR system which in this
case would cause a loss of one train of LHSI due to the isolation of RHR valves. MOVs
8888A and 8887A are closed and the leak checked, then the other train. Only one train
is checked at a time and isolated so BOTH trains are not affected.
At FNP, the most credible mechanism for initiation of RHR suction ISLOCA during power
operation is the catastrophic rupture of the closed MOVs isolating the RHR pump suction from
the RCS.
Following the RCS to RHR pressure boundary failure, the RHR system will be able to withstand
RCS pressure if the hoop stress imposed on the RHR system by exposure to RCS pressures is
below the yield stress. However, the RHR pump seals in both trains are expected to
rupture. This rupture of pump seals 'is assumed to result in failure of both RHR pumps
Monday, January 14, 2008'2:42:25 PM . 150'
QUESTIONS REPORT
for 75 RO Questions
(i.e., motors short due to water spray).
B. incorrect~ correct location, incorrect strategy. incorrect operational implication.
The RHR system is not secured to both trains at the same time. ECP-1.1 isolates one
train first and then restores and isolates the other train if the problem is not corrected.
ECP-1.2 never isolates the RWST to any component. However, ECP-1.1 does. The
loss of the train would be for injection and recirc for one train.
C. Incorrect. Incorrect location, incorrect strategy. incorrect operational implication.
D. incorrect. incorrect location. incorrect strategy. incorrect operational implication.
Location - Components between the RCS and the low pressure LHSI Pump hot leg
injection piping include three check valves. A LOCA through the upstream hot leg
injection piping is less likely than through the cold leg piping due to the addition of an
additional in-series check valve and because the upstream isolation valves are
normally closed. Also the Background states that the piping is able to withstand RCS
pressure if the hoop stress imposed on the RHR system by exposure to RCS pressures is
below the yield stress.
The RWST will not be lost due to the isolation of the sy.stem, it will be saved due to this
action. ECP-1.2 never isolates the RWST to any component. However, ECP-1.1
does.
Further background-
Purpose: To ensure that normally closed valves are closed
Basis: This step instructs the operator, to verify that all normally closed valves in low pressure lines
and other plant specific lines that penetrate containment are closed. The valving connecting
the RHR System to the RCSis of particular interest in this step since the RHR System is a
low pressure system (600 psig) connected to the high pressure reactor coolant system (2500
psig). Therefore, a rupture or break outside containment is most probable to occur in the low
pressure RHR System piping.
ERG StepText: Check If Break Is Isolated
Purpose: To determine if the LOCA outside containment has been isolated from previous actions
Basis: This step instructs the operator to check RCS pressure to determine if the break has been
isolated by previous actions. If the break is isolated in Step 2, a significant RCS pressure
increase .will occur due to the SI flow filling up the RCS with break flow stopped. The
operator transfers to E-1, LOSS OF REACTOR OR SECONDARY COOLANT, if the
break has been isolated, for further recovery actions. If the break has not been isolated, the
operator is sent to ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, for
further recovery actions since there will be no inventory in the sump.
ERP StepText: Identify source of leak.
ERG StepText: Try To Identify And Isolate Break
Purpose: To attempt to identify and isolate a LOCA outside containment
Basis: This step instructs the operator to sequentially close and open all normally opened valves in
paths that penetrate containment to identify and isolate the break outside containment.
Again, as in Step 1, the valving connecting the low pressure (600~) RH~s~te~m~to~t~h~e~~~~~~
high pressure (2500 psig) RCS is of primary interest, since the probability of a break
occurring outside containment is most probable to occur in the low pressure RHR System
piping.
Knowledge: The potential exists for RWST inventory to be lost to the auxiliary building for a LOCA that
occurs outside containment in the RHR system piping. The RWST could be drained to the
auxiliary building if the RCS pressure is reduced to below the static head pressure in the
Monday, January 14, 2008 2:42:26 PM 151
QUESTIONS REPORT
for 75 RO Questions
RWST. If this condition occurs, actions should be taken to isolate this potential leakage path
and loss of inventory from the RWST.
W/E04 LOCA Outside Containment
EK1.3 Knowledge of the operational implications of the following concepts as they apply to
the (LOCA Outside Containment):
Annunciators and conditions indicating signals, and remedial actions associated with
the (LOCA Outside Containment).
Question Number: 53
Tier 1 Group 1
Importance Rating: 3.5
Technical Reference: ECP-1.2,"FNP--O-ECB-1.2 specific background document for
ECP-1.2, lesson plan for ECP-1.2, OPS-52532E, WOG background document
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.10
Comments:
This question gives the indications of an ISLOCA and then the operational implications and
remedial actions or high level actions. The procedure would have the operator isolate one train
at a time so a complete loss of injection is not done. The implications of this action if the
system is isolated lAW ECP-1.2 is to lose one train of LHSI. The RWST is a credible distracter
in that ECP-1.1 which is where this procedure could send the operator to does isolate and turn
off pumps due to low RWST level.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: A C D A C B B B B A Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:26 PM 152
QUESTIONS REPORT
for 75 RO Questions
60. E05 02.1.27003
A Reactor Trip and Safety Injection have occurred on Unit 1:
The crew has entered FRP-H.1, Response to Loss of Secondary Heat Sink, from
EEP-1, Loss of Primary or Secondary Coolant, with the following conditions:
- RCS Pressure is 175 psig and decreasing.
- Intact SG pressures are 475 psig and trending down.
Which ONE of the following describes the status of the Steam Generators and the
associated procedural requirement for the conditions given above?
The Steam Generators are - - - - - - - - - - -
A. * Available to provide secondary heat sink.
- Remain in FRP-H.1.
B~ * NOT Available to provide secondary heat sink.
- Return to EEP-1.
C. * Available to provide secondary heat sink.
- Return to EEP-1.
D. * NOT Available to provide secondary heat sink.
- Remain in FRP-H.1 .
A - Incorrect. Secondary heat sink is not required if SGs are at a higher pressure than
the RCS. They act as a heat source. Plausible because the conditions for FRP-H.1
entry are met except for the first step of FRP-H.1. RNO sends to procedure and step in
effect.
B - Correct. If SGs are NOT required for heat sink, the crew will return to EEP-1.
C - tncorrect. SGs are NOT required, because RCS pressure is below SG pressure.
D - Incorrect. LBLOCA, RCS less than SG pressure, return to EEP-1.
Monday, January 14, 20082:42:26 PM 153*
QUESTIONS REPORT
for 75 RO Questions
W/E05 loss Secondary Heat Sink
G2.1 . 27 Conduct of Operations: Knowledge of system purpose and or function.
Question Number: 54
Tier 1 Group 1
Importance Rating: 2.8
Technical Reference:' FRP-H.1
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: ' 41.10
Comments: meets the KA in that the operator has to know the purpose of the SGs and the
function/role they play in removing/adding heat on a LOCA. .
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: B B DADB ACB B Scramble Range: A - D
Source: BANK Source if Bank: WTSI
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:26 PM 154
QUESTIONS REPORT
for 75 RO Questions
61. E06 EAl.2 064
Given the following:
- A LOCA has occurred.
- Due to ECCS failures, the crew is performing FRP-C.2, Response to
Degraded Core Cooling.
- The crew is depressurizing ALL Steam Generators to 100 psig.
- The Shift Supervisor continues in FRP-C.2.
Which ONE of the following describes the reason that the SS remains in FRP-C.2?
A. Actions cannot be taken for a RED condition on the Integrity CSF because there is
inadequate ECCS equipment available to mitigate the degraded core cooling
condition.
B. The RED condition on the Integrity CSF is not valid because of the dumping of
steam to the condenser at a maximum rate lAW FRP-C.2.
C. FRP-C.2 has a higher priority than any lower level CSF and no other procedural
actions are allowed to be implemented until a transition is directed lAW FRP-C.2.
D~ The RED condition on the Integrity CSF is expected and is based upon the
accumulators injecting.
Monday, January 14, 2008 2:42:26 PM 155
QUESTIONS REPORT
for 75 RO Questions
A. Incorrect. Actions taken would actually be to reduce ECCS flow, so unavailability of
SI for a PTS issue would not be a priority. -
B. Incorrect. The red condition is valid. The dumps are NOT opened to dump steam at
a maximum rate, the limit of 60°F/hour is the limit in C.2. In FRP-C.1, Steam is
dumped at a maximum rate.
C. Incorrect. Core Cooling is high priority, but FRP-C.-2 is entered on an orange
condition, so a red condition on another CSF tree would take priority. However, in the
specific case of the integrity CSF, the dumping of the accumulators is expected and
subsequent entry into this FRP would only cause core temperatures to rise and C.1
entry could be required.
D. Correct. Due to the dumping of the steam above, accumulator injection is the goal
and the dumping of steam is done at a rate of 60°FI hr.
FRP-C.2 Caution prior to step 12 just before depressurizing all intact SGs to 100 psig.
CAUTION: Performance of step 12 will cause accumulator injection which may result in a red path on the
INTEGRITY st,atus tree. This procedure should be completed before transition to FNP-1-FRP-P.1 ,
RESPONSE TO-IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS.
background documentation
ERP StepText: Performance of step 12 will cause accumulator injection which may result in a
red path on the INTEGRITY status tree. This procedure should be completed before transition
to FRP-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK
CONDITIONS.
ERG StepText: The following step will cause accumulator injection which may cause a red
path condition in F-O.4, INTEGRITY Status Tree. This guideline should be completed before
transition to FRP. 1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK.
Purpose: To alert the operator to complete entire guideline FR-C.2 even if a red path occurs in
the Integrity Status Tree, F-0.4.
Basis: Once the RCS is cooled/depressurized in step 10 to the point at which the accumulators
inject, the RCS cold leg temperature could be reduced such that a transition to FR-P.1,
Response to Imminent Pressurized Thermal Shock Condition, is required via the red path of
Status Tree F-0.4. The operator would stop the cooldown after entering FR-P.1.
While the operator is allowing the thermal shock to soak out, the core will continue to boil away
-the injected accumulator water and begin to uncover once again. Eventually, core exit
temperatures and/or RVLIS level values could existwhich would require the operator to
transfer to FR-C.1, Response to Inadequate Core Cooling, via one of the red paths on Status
Tree F-0.2. Thus, by going from FR-C.2 to FR-P.1 and stopping the cooldown and soaking,
a degraded core cooling condition could be allowed to deteriorate -to an inadequate core
cooling condition. Therefore, this caution will require the operator to complete guideline FRC.
~2---tG-e-r+Sl.I+e-GGf-e-GGGAA~-ve-r+-H_a_FeG_f}at-R-00flEl-iti*oo-eec-tJfs-i-A-ttle-l-RteWity--St-a-tlffi~-r-ee,F----~------
0.4. '
Monday, January 14,20082:42:26 PM 156
QUESTIONS REPORT
for 75 RO Questions
W/E06 Degraded Core Cooling
EA1.2 Knowledge of the reasons for the following responses as they apply to the (Degraded
Core Cooling) Operating behavior characteristics of the facility.
Question Number: 61
Tier 1 Group 2
Importance Rating: 3.5
Technical Reference: 1-FRP-C.2 and FNP-O-FRB-C.2 background document
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41 .7
Comments: This question tests the operating behavior characteristic in that when this condition
is entered and the SGs are depressurized, the resulting accumulator injection is expected to
cause FRP-P.1 conditions due to the rapid cooldown. This is a high level action of FRP-C.2 that
an RO is expected to know.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: D D DCA D B D D A Scramble Range: A - D
Source: MODIFIED Source if Bank:
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:26 PM 157
QUESTIONS REPORT
for 75 RO Questions
62. E08 EA2.1 001
Given the following:
- While operating at 100% power, Unit 1 experienced a steam break accident
inside CTMT.
- After the transition from EEP-O, Reactor Trip or Safety Injection, RCS cold
. leg temperatures had dropped to 240°F in 20 minutes.
- Reactor power is currently < 1% and 80th Intermediate Range SUR meters
are reading -.03 dpm.
- AFW flow is reading 300 gpm.
- SG narrow range water levels are reading:
1A SG = 49%
18 SG = 29%
1C SG = 30%
- Sub Cooled.Margin Monitor is reading 34°F in the CETC mode.
- Containment pressure is currently 28 psig and slowly decreasing.
Which ONE of the following is the highest level Functional Restoration Procedure
(FRP) required to be entered under these conditions?
A. FRP-Z..1, Response to High Containment Pressure.
8. FRP-S.1, Response to Nuclear Power Generation- ATWT.
C~ FRP-P.1, Response to Imminent Pressurized Thermal Shock Conditions.
D. FRP-H.1, Response to Loss of Secondary Heat Sink.
reference provided is the RCS pressure - temperature graph CSF-O.4 rev 17 so a
determination can be made to which area of the graph applies.
All distracters are plausible in that evaluation has to be made to determine which FRP is valid
and what condition it is in, orange green yellow. Then a determination of which order these are
referenced in.
Answer A is incorrect: Z.1 is an Orange path but it is lower than P.1 orange path.
Answer 8 is incorrect: FRP S.1 is a green path based on <5% power, IR SUR more
negative than -.02.
Answer C is correct: P.1 entered on an orange path.
Answer.D is incorrect. H.1 entry conditions are < 395 gpm or all sg nr levels < 310/0.
if one sgwl NR is > 31 % no entry is required*.
Monday, January 14, 2008 2:42:26 PM 158
QUESTIONS REPORT
for 75 RO Questions
E08 EA2.1 Pressurized Thermal Shock - EPE
Ability to operate and I or monitor the following as they 'apply to the (Pressurized Thermal
Shock) Facility conditions and selection of appropriate procedures during abnormal and
emergency operations.
Question Number: 62
Tier 1 Group 2
Importance Rating: 3.4
Technical Reference: none
Learning Objective: OPS 52533K-8
10 CFR Part 55 Content: 41.10
Comments: matches KA in that the operator has to monitor the correct parameters and then
based on those parameters select the appropriate procedure to follow during the emergency
event and it deals with a PTS event.
All distracters are plausible in that evaluation has to be made to determine which FRP is valid
and what condition it is in, orange green yellow. Then a determination of which order these are
referenced in.
This does not overlap with G2.4.22
MCS Time: 1 Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: CAB eBB B BCD Scramble Range: A - D
Source: MODIFIED Source if Banle FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:26 PM 159
QUESTIONS REPORT
for 75 RO Questions
63. EIO EKI.I 005
Given the following:
with Allowance for Reactor Vessel Head Steam Voiding (WITH RVLIS).
- The plant is being depressurized using auxiliary spray.
- Charging and Letdown flows are matched.
- As RCS pressure drops through 1300 psig, a rapid rise in pressurizer level is
observed.
- Pressurizer level has increased to 66%.
- Reactor vessel level indication has dropped below the minimum required
UPPER PLENUM level of 44%.
Which ONE of the following correctly describes the required response lAW ESP-0.3?
A. Increase auxiliary spray flow and verify Both CRDM cooling fans running.
B. Reduce charging flow, increase letdown flow and stop the cooidown in progress.
C. Increase the RCS cooldown rate while maintaining charging and letdown flows
matched.
D~ Reduce the auxiliary spray flow and energize additional pressurizer heaters.
A. Incorrect - Plausible because verifying both CRDM cooling fans running would cool
the head and reduce void formation & is required later in procedure. Spray flow should
be reduced, not increased, to establish subcooling.
B. Incorrect - Plausible because ESP-0.3 says to control or reduce charging and
increase letdown or continue the cooldown if PRZR level is >90% . Level is only 660/0.
the method at this point in the procedure.
C. Incorrect - Plausible because cooling down the RCS will cool the head, but with a
time delay. With the head still hot, this will cause the void to increase and PRZR to go
solid.
D. Correct - This is the correct response lAW ESP-0.3 when Reactor vessel level
indication drops to less than 44% upper plenum raise RCS pressure and this would be
done by controlling pressure with sprays and heaters.
Monday, January 14, 2008 2:42:26 PM 160
QUESTIONS REPORT
for 75 RO Questions
W/E10 EK1.1 Natural Circulation with Steam Void in Vessel with/without RVLIS-
Knowledge of the operational implications of the following concepts as they apply to the
(Natural Circulation with Steam Void in Vessel with/without RVLIS) Components, capacity, and
function of ef)lergency systems.
Question Number: 63
Tier 1 Group 2
Importance Rating: 3.3
Technical Reference: ESP-0.3
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS 52531 C06
1,0 CFR Part 55 Content: 41 .1 0
Comments:
matches KA in that it tests the knowledge of ESP-0.3 and the implications of void formation and
what to do about it using the components that the operator os required to control during this
evolution.' .
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: DAAAACDB B D Scramble Range: A - D
Source: BANK Source if Banle FARLEY
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:26 PM 161
QUESTIONS REPORT
for 75 RO Questions
64. Ell EK2.2 004
The following conditions exist on Unit 1:
- 1B RHR pump is tagged out and the oil is drained from the motor.
- The Control room team is responding to a LOCA.
- The reactor was tripped and an SI manually actuated.
- RCS pressure is 1000 psig..
- 1A RHR pump has tripped.
The control room team has transitioned to ECP-1.1, Loss Of Emergency Coolant
Recirculation. Make up has been established to the RWST.
Which one of the following describes the correct actions to take in ECP-1.1 under
these conditions?
A~ * Initiate an RCS cooldown to Cold Shutdown at less than 100°F/hr,
- establish only one charging pump running and
- reduce RCS pressure to reduce break flow.
B. * Initiate an RCS cooldown to Cold Shutdown at the maximum rate possible,
- establish only one charging pump running and
- reduce RCS pressure to dump the accumulators.
C. * Initiate an RCS cooldown to Cold Shutdown at less than 100°F/hr,
- establish two charging pumps running and
- reduce RCS pressure to dump the accumulators.
D. * Initiate an RCS cooldown to Cold Shutdown at the maximum rate possible,
- establish two charging pumps running and .
- reduce RCS pressure to *reduce break flow.
Monday, January 14, 2008 2:42:26 PM 162
QUESTIONS REPORT
for 75 RO Questions
ECP-l.l, LOSS OF EMERGENCY COOLANT RECIRCULATION OPS-52532D
A. Correct - Start makeup to the RWST, initiate an RCS cooldown, minimize ECCS flow and reduce
RCS pressure.
The following criteria are the high level actions needed to be successful in ECP-1.1
Makeup to the RWST is necessary
Inventory in the RWST is a concern for recovery from a loss of ECR capability. Makeup is
added to the RWST to extend the time the SI pumps and containment spray pumps (if operating) can
take suction from the RWST and provide core cooling to the RCS.
Begin Cool Down to Cold Shutdown
The purpose is to begin a controlled RCS cool down to cold shutdown temperature using a
'preferred or alternate method with a specified maximum cool down rate. Shutdown margin should be
monitored during RCS cool down using Curve 61 and/or 61A.
The objective is to reduce the overall temperature of the RCS coolant and metal to reduce the
need for supporting plant systems and equipment required for heat removal. The maximum cool
down rate of 1OO°F/hr will preclude violation of the integrity status tree, thermal shock limits.
Stop SI Pumps
To reduce flow into the RCS, the low-head injection pumps and all but one high-head pump
are stopped. Satisfaction of conditions for SI termination implies that control can be maintained by the
operator without all of the ECCS pumps running. In this step, all but one high-head pump are stopped and
placed in standby for future use.
Reduce RCS Pressure to Reduce Subcooling
This step is performed to decrease RCS pressure to the lowest pressure possible without
losing adequate subcooling. The RCS pressure reduction is done to decrease RCS break flow. The RCS
should be depressurized until RCS subcooling indicates between 16°F (45°F) and 26°F (55°F) on the
Subcooled Margin Monitor in CETC mode. A second criterion for stopping the pressure reduction is PRZR
level greater than 730/0 (50%).
B. Incorrect - See above, first and third part incorrect, second part correct
C. Incorrect - See above, first part correct, second and third part incorrect
D. Incorrect - See above, first and second part incorrect, third part correct
Monday, January 14, 2008 2:42:27 PM 163
QUESTIONS REPORT
for 75 RO Questions
W/E11 Loss of Emergency Coolant Recirculation -
EK2.. 2 Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation)
and the following:
Facility*s heat removal systems, including primary coolant, emergency coolant, the
decay heat removal systems, and relations between the proper operation of these
systems to the operation of the facility.
Question Number: 55
Tier 1 Group 1
Importance Rating: 3.9
Technical Reference: ECP-1.1 and ECP-l.l, LOSS OF EMERGENCY COOLANT
RECIRCULATION OPS-52532D and FNP-O-ECB-l.l
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.10
Comments: This meets the KA in that the operator has to know the strategy of the procedure
and the proper operation of the the various heat removal systems to control the casualty in
progress.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: AAAAAAAAAA Items Not Scrambled
Source: BANK Source if Bank: FARLEY
Cognitive Level: IDGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:27 PM 164
QUESTIONS REPORT
for 75 RO Questions
65. E15 EK3.4 003
Which ONE of the following is a potential source of the flooding that is checked for in
FRP-Z.2; and the concern if the maximum expected post-accident containment water
level (design basis containment flood level) is exceeded?
A. * Condensate Storage Tank.
. * Thermal shock to the reactor vessel lower head due to quenching.
B. * Condensate Storage Tank.
- Damage to vital system or components rendering them inoperable.
C~ * Service Water system.
- Damage to vital system or components rendering them inoperable.
D. * Service Water system.
- Thermal shock to the reactor vessel lower head due to quenching.
Monday, January 14, 2008 2:42:27 PM 165
QUESTIONS REPORT
for 75 RO Questions
A. Incorrect. CST is not one of the potential sources of water FRP-Z.2 addresses and
the background does not mention. However, since AFW goes into ctmt and is in use, it
is plausible that this large source of water would be checked for and isa concern.
The maintenence sump is isolated from the bottom of the reactor vessel by a wall with
an elevation higher than the vital equipment of concern. Plausible, because a high
enough containment level would allow water to potentially thermally shock the hot post
accident reactor vessel.
B. Incorrect. CST is not one of the potential sources of water FRP-Z.2 addresses and
the background does not mention. The reason is correct.
C. Correct. Service water is the most likely source since it does not isolate to the ctmt
on a phase A or B signal and is the largest source of water available to ctmt.
The purpose of the sump is to collect and divert water in areas that will not affect vital
plant equipment. Flooding may jeopardize that function. RWST, CST, & RCS are
expected to provide their full volumes to the CTMT sump in accident analysis.
D. Incorrect. SW .is correct.
The reason is incorrect. Plausible, because a high enough containment level would
allow water to potentially thermal shock the hot post accident reactor vessel.
FNP-O-FRB-Z.2 specific background document for FRP-Z.2 for step 1
Basis: This step instructs the operator to try to identify the unexpected source of the water in
the containment sump. Containment flooding is a concern since critical plant components
necessary for plant recovery may be damaged and rendered inoperable. A water level greater
than the design basis flood level provides an indication that water volumes other than those
represented by the emergency stored water sources (e.g., RWST, accumulators, etc.) have
been introduced into the containment sump. Typical sources which penetrate containment
are service water, component cooling water, primary makeup water and demineralized water.
All possible *plant specific sources which penetrate containment should be included in this
step. These systems provide large water flow rates to components inside the containment
and a major leak or break in one of these lines could introduce large quantities of water into
the sump. Identification and isolation of any broken or leaking water line inside containment
is essential to maintaining the water level below the design basis flood* level.
Monday, January 14, 20082:42:27 PM 166
QUESTIONS REPORT
for 75 RO Questions
E15 EK3.4 Containment Flooding - EPE
Knowledge of the reasons for the following responses as they apply to the (Containment
Flooding)
RO or SRO function as a within the control room team as appropriate to the assigned
position, in such a way that procedures are adhered to and the limitations in the
facilities license and amendments are not violated.
Question Number: 64
Tier 1 Group 2
Importance Rating: 2.9
Technical Reference: FRP-Z.2 and FNP-O-FRB-Z.2 specific background document for FRP-Z.2
Proposed references to be provided to applicants during examination: None
Learning Objective: OPS 52533M01
10 CFR Part 55 Content: 41 .1 0
Comments:
This meets the KA in that it tests the knowledge of the operator on where the most likely source
of water would come from and then the limitations that the bkgrd documents speak of.
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: . C A A A C C"D C C A Scramble Range: A - D
Source: NEW Source if Bank:
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed:" GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:27 PM 167
QUESTIONS REPORT
for 75 RO Questions
66. G2.1.10 001
Unit 2 is in MODE 2 and a reactor startup is in prqgress.
In accordance with Technical Specifications 2.1, Safety Limits (SLs), which ONE of the
following describes the RCS Pressure Safety Limit, and the MAXIMUM time to take
action' if it is exceeded?
Limit MAXIMUM time
A. 2735 psig 5 minutes
B~ 2735 psig 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
C. 2750 psig . 5 minutes
D. 2750 psig 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
A. Incorrect. In Modes 3,4,5, TS allows 5 minutes to restore pressure
B. Correct. In Modes 1 or 2, TS allows 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to Mode 3. In Modes 3,4,5, TS aUows 5
minutes to restore pressure.
C. Incorrect. 2750 would be correct in psia, but not in psig.
D. Incorrect. 2750 would be correct in psia, but not in psig.
G2.1.10 Conduct of Operations
Knowledge of conditions and limitations in the facility license.
Question Number: 67
Tier 3 Group 1
Importance Rating: 2.7
Technical Reference: TS 2.2.2
Proposed references to be pro.vided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.2/41.10
Comments: This meets the KA in that it tests the knowledg,e of a RCS safety limit and the
applicable tech spec requirements for that limit for RO knowledge.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
"Answer: B C C C D B D B B C Scramble Range: A - D
Monday, January 14, 20082:42:27 PM 168
QUESTIONS REPORT
for 75 RO Questions
Source: MODIFIED Source if Banle
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:27 PM 169
QUESTIONS REPORT
for 75 RO Questions
67. G2.l.l2 001
Unit 1 is at 1000/0 power. All three Auxiliary Feedwater Pumps have just been declared
Which ONE of the following actions MUST be taken?
A. Be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in Mode 4 in in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B. Take action to restore at least one AFW pump to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,
and a second AFW pump within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or be in Mode 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
C~ Immediately take action to restore at least one AFW pump to OPERABLE status.
D. Immediately enter LCO 3.0.3 and take actions to initiate a shutdown within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
A. Incorrect. T.S. 3.7.5 with 2 trains INOP or Required Action of A or B not met, then
the action is to enter mode 3 in 6 and mode 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B. Incorrect. T.S. 3.7.5 does not have a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Action for AFW. If the AFW pump was
returned in one hour a case can be made it was RTS immediately, but the second
pump being INOP is not allowed to be INOP for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> wlo action. Therefore, the
distracter is incorrect since when one AFW pump is RTS, Required Action for C. is to
place the unit in mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, not wait an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to fix it. This is a sly
way of using a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> LCO from memory in that the unit has to be placed in mode 3 in
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> but the LCO C is entered immediately once one AFW pump is RTS.
C. Correct. This answer reflects the Note contained in action D as discussed above.
TS 3.7.5 in Condition D has a NOTE that states: LCO 3.0.3 and all other action
statements requiring a Mode change are suspended until one AFW train is restored to
operable status.
This prevents placing the plant in a much higher risk condition than required.
D.1 says: Initiate action to restore one AFW train to OPERABLE status.
D. Incorrect. note in LCO 3.7.5
LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until
one AFW train is restored to OPERABLE status.
Monday, January 14, 20082:42:27 PM 170
QUESTIONS REPORT
for 75 RO Questions
G2.1.12 Conduct of Operations
Ability to apply technical specif-ications for a system.
Question Number: 66
Tier 3 Group 1
Importance Rating: 2.9
Technical Reference: T8 section 3.7.5
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.2/41.10
Comments: this meets the KA at an RO expected level of knowledge for the AFW system.
MCS Time: 1 Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: CDCAAADACB Scramble Range: A - D
Source: MODIFIED Source if Bank: FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:27 PM 171
QUESTIONS REPORT
for 75 RO Questions
68. G2.1.8 004
Given the following:
- An oil spill has occurred from a non-PCB oil source in the Turbine Building.
- The following conditions exist:
- The oil has reached the Turbine Building sump.
- The sump pump is* running and releasing water to the environment.
Which ONE of the following actions is required to be performed by the control room
team lAW AP-60, Oil Spill Prevention Control and Countermeasure Plan, Hazardous
Waste Contingency Plan?
A':' * Dispatch *the SSS to the scene.
discharge valve.
B. * Dispatch the shift chemist to the scene.
can be analyzed.
C. * Dispatch the SSS to the scene.
.can be analyzed.
D. * Dispatch the shift chemist to the scene.
discharge valve ..
A. Correct - Dispatch the SSS to the scene, have the TB System Operator stop the
sump pump, close and tag the discharge valve. AP-60 requires these actions: SSS to
the scene, stop the release and close and tag sump discharge valves.
B. Incorrect - The shift radiochemist is not required to be dispatched, though this may
be a good idea, however chemistry supervision is required to be notified. The shift
radiochemist is not necessarily supervision. The TB System Operator should not
place the sump on recirc. With a release in progress the requirement is to stop the
release.
C. Incorrect- The TB System Operator is required to stop the release immediately, not
evaluate how much more water can be released. The release needs to analyzed prior
_ _ _ _-\o'to~startin-Q-an(lneedss'lpervisor-apptO¥aLto-stactc-l-itb-.- - - - - - - - - - - - - - - - -
D. Incorrect - The shift radiochemist is not required to be dispatched, the release
would not continue with oil going to it due to the potential for release to the
enviro.nment, and the discharge valve would be closed, and also tagged.
Monday, January 14, 2008 2:42:27 PM 172
QUESTIONS REPORT
for 75 RD Questions
G2.1.8 Conduct of Operations
Ability to coordinate personnel activities outside the control room.
Question Number: 68
Tier 3 Group 1
Importance Rating: 3.8
Technical Reference: AP-60 Appendix 1
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content:
Comments:
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: AAAAAAAAAA Items Not Scrambled
Source: BANK Source if Banle .FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: YES
Monday, January 14, 2008 2:42:27 PM 173
QUESTIONS REPORT
for 75 RO Questions
69. G2.2.22 002
Unit 1 is in a Refueling Outage with fuel being loaded into the core.
Which one of the following describes the MINIMUM temperature and the MINIMUM
borated water volume that must be met to maintain an operable Boric Acid Storage
Tank (BAT Ta.nk)?
Solution Temperature Borated Water Volume
A. 35°F 2,000 gal.
B. 35°F 11 ,336 gal.
C~ 65°F 2,000 gal.
D. 65°F 11 ,336 gal.
Reference:
Technical R~quirements *Manual, TRM 13.1.6.4 and 13.1.6.6
A. Incorrect, Mode 5 and 6. TRS 13.1.6.6 Verify the contained borated water volume
in the boric acid storage tank is 2: 2,000 gal., TRS 13.1.6.1 Verify RWST solution
temperature is > or equal to 35°F
B. Incorrect, Mode 5 and 6, TRS 13.1.6.1 Verify RWST solution temperature is >...Q[
equal to 35°F. Mode 1,2,3&4, TRS 13.1.7.4 Verify the contained borated water volume
in the boric acid storage tank is 2: 11 ,336 gal
C. Correct, Plant is in Mode 6. The following TRSs apply. TRS 13.1 .6.4 Verify boric
acid storage tank solution temperature is > or equal to 65°F, TRS 13.1.6.6 Verify the
contained borated water volume in the boric acid storage tank is 2: 2,000 gal
D. Incorrect, Mode 5 and 6, TRS 13.1.6.4 Verify boric acid storage tank solution
temperature is > or equal to 65°F. Mode 1,2,3&4, TRS 13.1.7.4 Verify the contained
borated water volume in the boric acid storage tank is 2: 11 ,336 gal
Monday, January 14, 2008 2:42:27 PM 174
QUESTIONS REPORT
for 75 RO Questions
G2*.2.22 Equipment Control
Knowledge of limiting conditions for operations and safety limits.
Question Number: 69
Tier 3 Group 2 .
Importance Rating: 3.4
Technical Reference: TRM 13.1.6.4 and 13.1.6.6
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.10/43.2
Comments:
replaced this question since the concept is already tested on the SRO portion of the exam, ie.,
IR instrument failed at 10-8 amps and what to do and why.
This is more of an RO question with no overlap on the exam 'and meets the LCO requirements
above.
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: CDADBDCADB Scramble Range: A - D
Source: BANK Source if Banle FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:27 PM 175
QUESTIONS REPORT
for 75 RO Questions
70. 02.2.34002
Given the following:
- Unit 1 is in Mode 3.
- Reactor tripped from 1000/0 RTP.
- ECC has been calculated for a startup 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the trip.
- Estimated critical rod position is Control Bank D at 100 steps.
- Startup is delayed for TWO (2) hours.
Which ONE of the following describes the effect on 11M plot data taken during the
approach to critical?
The 11M plot will predict criticality at a.....
A. LOWER rod height due to Xenon concentration greater than that assumed in ECC
calculation.
B~ LOWER rod height due to Xenon concentration less than that assumed in ECC
calculation.
C. HIGHER rod height due to Xenon concentration less than that assumed in ECC
calculation.
D. HIGHER rod height due to Xenon concentration greater than that assumed in ECC
calculation.
Monday, January 14, 20082:42:28 PM 176
QUESTIONS REPORT
.for 75 RO Questions
A: Incorrect. Lower rod height is correct. Xenon concentration greater is incorrect.
Xenon concentration will be less but will be adding positive reactivity which will result in
a lower rod height for criticality to be obtained.
B: Correct. Delay will affect core reactivity since Xenon is decaying, reducing the
negative reactivity in the core. Rods will not have to be withdrawn as far to make the
reactor critical.
C: Incorrect. Higher rod height is incorrect. Xenon concentration will be less but will be
adding positive reactivity which will result in a lower rod height for criticality to be
obtained.
D: Incorrect. Rods will not have to be withdrawn as far to make the reactor critical.
Delay will affect core reactivity since Xenon is decaying, reducing the negative reactivity
in the core. Candidate needs to demonstrate an understanding of the time that it takes
Xenon to peak from a full power trip, which is typically the square root of the equilibrium
power level. .
G2.2.34 Equipment Control
Knowledge of the process for determining the internal and external effects on core reactivity.
Question Number: 70
Tier 3 Group 2
Importance Rating: 2.8
Technical Reference: Physics curve 60
Proposed references to be provided to applicants during examination: None
Learning ObJective:
10 CFR Part 55 Content: 41.1
Comments: This KA tests the operator to evaluate core reactivity and the effects of xenon after
a rx trip for a startup.
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: B D-cAD C D A C D Scramble Range: A - D
Source: MODIFIED Source if Bank: SEQUOYAH 2004
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:28 PM 177
QUESTIONS REPORT
for 75 RO Questions
71. 02.3.10007
1
What precaution is required to be taken at the 121 Piping Penetration Room (PPR)
prior to lowering RCS level to mid-loop lAW UOP-4.3, Mid Loop Operations?
1
A. The door to the 121 P.PR must be locked.
1
B. Health physics (HP) must survey the 121 PPR.
C~
1
A caution sign must be placed at the entrance of the 121 PPR.
1
D. All vent valves on systems in the 121 PPR penetrating containment
must be verified closed.
UOP-4.3
2.24 Prior to reducing Res level, a caution sign concerning the establishment of
containment closure must be placed at the entrance of the following locations.
NOTE: The signs can be obtained from the Shift Support Supervisor and are normally
stored in the CCW Storage Room on Unit 2.
2.24.1 139' Electrical Penetration Room
2.24.2 121' Piping Penetration Room
2.24.3 100' Piping Penetration Room
2.24.4 Main Steam Valve Room
2.24.5 Personnel Access Hatch
2.24.6 Auxiliary Access Hatch
A. Incorrect - Door not required to be locked SOP 0.0, 15.3.5
The following doors will be locked closed when unattended during unit operation in
Modes 1 through 4:
- .139' Electrical Penetration Room Doors 317A/2317
- 121' Piping Penetration Room Doors 214/2214
B. Incorrect - Surveys are not required prior to reducing level.
C. Correct - per the above initial condition of UOP-4.3
D. Incorrect - Air to air barrier not required for midloop integrity (ctmt closure in 2 hrs)
Monday, January 14, 20082:42:28 PM 178
QUESTIONS REPORT
for 75 RO Questions
G2,,3.. 10 Radiation Control
Ability to perform procedures to reduce excessive levels of radiation and guard against
personnel exposure.
Question Number: 71
Tier 3 Group 3
Importance Rating: 2.9
Technical Reference: Health .Physics manual
Proposed references to be provided to applicants during examination: None
Learning Objective: Describe how the rad monitoring system helps to protect the health
and safety of plant workers and the public. (ESP521 06008)
10 CFR Part 55 Content: 43.4
Comments: This question tests the basic generic applicability of the KA at an RO level
of knowledge.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: C B CACAAADA Scramble Range: A - D
Source: BANK Source if Bank: FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:28 PM 179
QUESTIONS REPORT
for 75 RO Questions
72. G2.3.9 001
The following conditions exist on Unit 2 while at 100% power:
- Containment Main Purge and Mini-Purge are secured.
- CTMT to ATMOS DP is currently 0.3 psid.
- All pre-requisites to perform a batch release of the containment atmosphere
have been met.
Which ONE of the following describes where the containment purge system discharges
to and how the system is operated to reduce containment pressure lAW SOP-12.2,
Containment Purge and Pre-access Filtration System, Appendix 3, Batch Releases of
Containment Atmosphere?
A. * Discharges directly to the plant vent stack;
- Open the Mini-Purge dampers, then start the Mini-Purge supply and exhaust fans
to initiate the release.
B. * Discharges directly to the exhaust plenum;
is <0.25 psid, then start the Mini-Purge s'upply and exhaust fans.
C. * Discharges directly to the plant vent stack;
- Open Purge Filter Outlet Valve, V-294, then open the Mini-Purge dampers and
start the Mini-Purge supply and exhaust fans. Then close V-294.
D~ * Discharges directly to the exhaust plenum;
- Open Purge Filter Outlet Valve, V-294, then open the Mini-Purge dampers. When
CTMT to ATMOS DP is <0.25 psid, then close V-294 and start the Mini-Purge
supply and exhaust fans.
Monday, January 14, 2008 2:42:28 PM 180
QUESTIONS REPORT
for 75 RO Questions
DISTRACTOR ANALYSIS:
A Incorrect. not the correct order, the incorrect release path and v294 is not being
used as required.
B Incorrect. Incorrect order and v294 is not being used as required.
C Incorrect. incorrect release path and the pressure has to be. checked < .25 psid to
start the fans.
D Correct. per the procedure below and the prints
REFERENCES:
1. SOP-12.2 Containment Purge and Pre-access filtration system, Rev. 34/27
3.2 Open N2P13V294, PURGE FILTER COOLING OUTLET VALVE.
3.3 WHEN performing the following valve manipulations, THEN note the
start time for recording purposes:
3.3.1 Place the following CTMT Purge DMPRS hand switches to MINI
to initiate CTMT Batch Release:
HS-3196
HS-3198
3.3.2 Record start date/time data in Part III of the Batch Gaseous Waste
Release Permit.
3.4 WHEN CTMT DIFF PRESSURE decreases to = 0.25 psid, THEN
perform, the following, noting the fan start time and
containment-to-atmosphere delta pressure for recording purposes:
_/- 3.4.1 Close N2P13V294, PURGE FILTER COOLING OUTLET
IY VALVE.
3.4.2 Start MINI PURGE SUPP/EXH FAN.
3.4.3 Record fan start date/time data in Part III of the Batch Gaseous
Waste Release Permit.
Monday, January 14, 20082:42:28 PM 181
QUESTIONS REPORT
for 75 RO Questions
G2.3.9 Radiation Control
Knowledge of the process for performing a containment purge.
Question Number: 72
Tier 3 Group 3
Importance Rating: 2.5
Technical Reference: SOP-12.2, step 4.4 version 33.0
P& lOs:
5.1.1 0-205010, sheets 1 and 2, Containment Cooling and Purge System P & 10
5.1.2 0-207783, Elem. Oiag., Containment Mini-Purge Fans
5.1.3 0-207204, Elem. Oiag., Containment Purge Iso. Oampers Train A
5.1.4 0-207199, Elem. Oiag., Containment Purge Iso. Oampers Train B
5.1.5 0-207236, Elem. Oiag., Containment Purge Air Handling Unit Fan
5.1.6 0-207237, Elem. Oiag., Containment Purge Exhaust Fan
5.1.7 0-204654, Conn. Diag., Containment Purge Starter Panels
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 43.4
Comments:
This is according to SOP-12.2 appendix 3 for the batch release.
This question requires knowledge of process (procedure) for equalizing containment
pressure with atmospheric pressure when initiating a containment purge. One' answer
choice (distractor) may result in system damage.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: DBACADCADA Scramble Range: A - D
Source: BANK Source if Bank: FARLEY
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:28 PM 182
QUESTIONS REPORT
for 75 RO Questions
73. 02.4.22 002
Given. the following:
- FRP-H.1, Response to Loss of Secondary Heat Sink, is in progress in
response to a Red Heat Sink condition.
- The crew is still progressing through FRP-H.1 when critical safety function
status tree conditions are reported as follows:
- Subcriticality: Orange
- Core Cooling: Green
- Heat Sink: Yellow
- Integrity: Green
- Containment: Red
- Inventory: Yellow
Which ONE of the following describes the actions the ,crew should take in response to
the conditions given above?
A. Complete FRP-H.1, then transition to FRP-S.1.
B~ Complete FRP-H.1, then transition to FRP-Z.1.
c. Immediately exit FRP-H.1 and transition to FRP-S.1.
D. Immediately exit FRP-H.l and transition to FRP-Z.1.
A is incorrect. S.1 is higher priority CSF, but lower challenge (Orange).
B is correct. Entered H.1 on red condition, must complete prior to any other lower
procedure.
C is incorrect. Even though heat sink is no longer red, would not immediately leave.
D same as C, except that current conditions indicate a transition to Z.1 is required.
Monday, January 14, 2008 2:42:28 PM 183
QUESTIONS REPORT
for 75 RO Questions
G2.. 4.. 22 Emergency Procedures I Plan
Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations.
Question Number: 75
Tier 3 Group 4
Importance Rating: 3.0
Technical Reference: FRP-H.1 and EOP Users Guide
Proposed references to be provided to applicants during examination: None
Learning Objective: .
10 CFR Part 55 Content: 41 .1 0
Comments:
MCS Time: Points: 1.00 Version: a 1 2 3 4 5 6 7 8 9
Answer: B AAC BADC C B Scramble Range: A - D
Source: BANK Source if Bank: NORTH ANNA
Cognitive Level: HIGHER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GTO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:28 PM 184
QUESTIONS REPORT
for 75 RO Questions
74. G2.4.34 001
Gi'ven the following:
- A fire required evacuation of the control room.
- The crew is performing actions of AOP-28.2, Fire in the Control Room.
- HSO Panel A is manned and functional.
The crew is at the step to adjust HIK-122, CHG FLOW, to maintain pressurizer level
within the required band when the following is reported:
- HIK-122 on the HSO panel'is not controlling FCV-122 properly.
- Pressurizer level is 16% and trending down.
Which ONE of the following contains a correct method for controlling PRZR level and
the correct location of the components to be operated lAW AOP~28.2?
A. Close LCV-459 or 460, LTON LINE ISO, from the HSO panel.
Control charging flow using the bypasses around FCV ~ 122 locally in the 100 * Piping
Penetration Room entrance.
B. Close HV-8149A and 8149B or C, LTON ORIF ISO, at the HSO panel.
Control charging flow using the bypasses around FCV-122 locally in the 100'
hallway BIT area.
C. Close LCV-459 or 460, LTON LINE ISO, from the HSO panel.
Control charging flow using MOV-8803A or B, HHSI TO RCS CL, locally in the 100'
Piping Penetration Room entrance.
O~ Close HV-8149A and 8149B or C, LTON ORIF ISO, at the HSO panel.
Control charging flow using MOV-8803A or B, HHSI TO RCS CL, locally in the 100'
hallway BIT area.
Monday, January 14, 2008 2:42:28 PM 185
QUESTIONS REPORT
for 75 RD Questions
A. incorrect. LCV-459 or 460 can not isolated at the HSDP and would not be isolated
procedurally per the note below. Control of charging using the bypass valves is an
option, actually the first option, (step 14.6) but in this case the first part of the distracter
is not correct and the location is correct.
ADP-28.2
NOTE: Isolation of letdown due to low pressurizer level (15%) will unnecessarily complicate plant recovery
(LeV 459 & 460 cannot be re-opened from the HSDP, Reactor head vents must then be used for
removing mass from the primary system). Therefore, emphasis should be placed on controlling charging
flow to establish a stable or slowly rising pressurizer level that compensates for any effect on
level due to cooldown.
B. incorrect. Placing 2 orifices on service is correct at step 25 in the procedure for
controlling level, and bypassing FCV-122 is correct, but the loca~ion is not correct.,
C. incorrect. LCV-459 or 460 can not isolated at the HSDP and would not be isolated
procedurally per the note below.
control charging flow using MOV8803A or B, HHSI TO RCS CL, is correct but the
location is not correct.
D. Correct. Placing 2 orifices on service is correct step 25 in the procedure for
controlling level, and control charging flow using MOV8803A or B, HHSI TO RCS CL, is
correct and the location is correct.
G2.4.34 Emergency Procedures I Plan
Knowledge of RO tasks performed outside the main control room during emergency operations
including system geography and system implications.
Question Number: 74
Tier 3 Group 4
Importance Rating: 3.8
Technical Reference: FNP-1-AOP-28.2 step 14.6 and the note above as well as step
25.2
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41 .1 0
Comments: This question tests the knowledge of an RO task outside the control room during
an emergency and tests geography, plant location as well as procedural guidance and
operational implications, which include the letdown portion of the question and the note
eles-ertb-ifl-g-why-teteewft-is-not-iselatee.b---------------------
MCS Time: Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: DDCAB AB DB B Scramble Range: A - D
Monday, January 14, 20082:42:28 PM' 186
QUESTIONS REPORT
for 75 RO Questions
Source: NEW Source if Bailie
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 20082:42:28 PM 187
QUESTIONS REPORT
for 75 RO Questions
75. 02.4.49 002
Given the following:
- The crew is performing the actions of FRP-H.1, Loss of Secondary Heat
Sink, following a reactor trip due to a loss of feedwater.
- RCS pressure is 2280 psig.
- Containment pressure is 1 psig.
Which one of the following sets of steam generator wide range level parameters meet
the FRP-H.1 foldout page criteria for feed and bleed for the conditions given?
A. 1A SG - 28%
1'B SG - 29%
1C SG - 0%
B. 1A SG - 0%
1B SG - 13%
C.1ASG-31 %
1B SG - 29%
1C SG - 32%
D~ 1A SG - 11 %
1B SG - 11 %
1C SG - 14%
Monday, January 14, 2008 2:42:28 PM 188
QUESTIONS REPORT
for 75 RO Questions
A. Incorrect. Plausible, 2 of 3 SG WR levels < 31% is the figure for adverse
containment initiation of feed and bleed.
B. Incorrect. plausible since 2 of 3 SG WR levels < 28% was chosen since candidate
could confuse with the adverse containment figure for SG NR level control of 28%. Also
this could be chosen if the candidate did not remember 2 of 3 and thought it was 1 of 3
for non adverse numbers.
C. Incorrect. Plausible, 2 of 3 SG WR levels < or equal to 31 % is the setpoint for a Dry
SG during adverse containment. Possible candidate may confuse the setpoint. Also
this could be chosen if the candidate did not remember 2 of 3 and thought it was 1 of 3
for adverse numbers or did not remember greater than 31 %.
D. Correct. 2 of 3 SG WR levels < 12% requires feed and bleed with normal
containment pressure conditions.
FRP-H.1 Foldout page requirements
1 Monitor bleed and feed criteria. (applicable steps 1 thru 11 only)
1.1 Check at least two SG* wide 1.1 Perform the following.
range levels - GREATER THAN
12%{31%}. 1.1.1 Stop all RCPs.
[]1A
[]1B
[] 1C
1.1 .2 Proceed, to Step 12
G2.. 4.49 Emergency Procedures I Plan
Ability to pertorm without reference to procedures those actions that require immediate
operation of system components and controls.
Question Number: 73
Tier 3 Group 4
Importance Rating: 4.0
Technical Reference: FRP-H.1
Proposed references to be provided to applicants during examination: None
Learning Objective:
10 CFR Part 55 Content: 41.10
Comments: meets the KA in that the question tests the FO page of H.1 which are IOAs of that
procedure when those conditions require entry. This is required RO knowledge.
MCS Time: 1 Points: 1.00 Version: 0 1 2 3 4 5 6 7 8 9
Answer: DDAB B B B DBA Scramble Range: A - D
Source: BANK Source if Banle VOGTLE
Cognitive Level: LOWER Difficulty:
Job Position: RO Plant: FARLEY
reviewed: GO Previous 2 NRC exams: NO
Monday, January 14, 2008 2:42:28 PM 189