ML13073A452
| ML13073A452 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/01/2013 |
| From: | Vincent Gaddy Operations Branch IV |
| To: | Entergy Operations |
| laura hurley | |
| References | |
| 50-313/13-002 | |
| Download: ML13073A452 (201) | |
Text
Question 1 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # _____007EK1.04___
Importance Rating __3.6_ _____
EK1. Knowledge of the operational implications of the following concepts as they apply to the reactor trip: Decrease in reactor power following reactor trip (prompt drop and subsequent decay)
Proposed Question:
The plant has been operating at full power for 60 days. On day 61, a disturbance on the electrical grid causes an automatic reactor trip to occur.
Reactor thermal power, due to decay heat, should be approximately ________ per cent of rated power 10 seconds after the reactor trip, and approximately ________ per cent of rated power 10 minutes after the reactor trip (choose the closest values).
A. 7.0; 0.2 B. 12.0; 0.2 C. 7.0; 2.0 D. 12.0; 2.0 Proposed Answer: ___C___
Explanation (Optional):
Answer C is the correct answer with the other distracters being plausible if the drop in reactor power due to the insertion of negative reactivity combined with the positive reactivity addition of decay heat is not fully understood.
Technical Reference(s): _ASLP-RO-RXT14, GP Reactor Theory Chapter 8 Reactor Operational Physics, Revision 2 Page 115 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis ___X_
10 CFR Part 55 Content: 55.41 __1__
55.43 _____
Comments:
Question: 2 Examination Outline Cross-
Reference:
Level RO Tier # __1__ _____
Group # __1__ _____
K/A # __008AK2.02______
Importance Rating _2.7 _ _____
AK2. Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: AK2.02 Sensors and detectors Proposed Question:
A leak exists on the upper tap of a Pressurizer level transmitter sensing line, causing a PZR steam space leak.
Indicated PZR level will ___________ and actual PZR level will _________.
A. Drop, Drop B. Drop, Rise C. Rise, Drop D. Rise, Rise Proposed Answer: __D____
Explanation (Optional):
Answer [D] is correct since a leak on the upper tap will cause the differential pressure to decrease on the affected transmitter, thus causing indicated level to rise. Likewise a steam space leak will cause actual level to increase. Answers [A] thru [C] are combinations of the correct answer, and could be correct if the leak was elsewhere.
Technical Reference(s): _ASLP-AO-I&C___
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _0371__
Modified Bank # _______ (Note changes or attach parent)
New _______
Question History: Last NRC Exam _2004___
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __5__
55.43 _____
Comments:
Question 3 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # __009 EA1.13_____
Importance Rating __4.4 _____
- CHANGE IN K/A from EA1.12 to EA1.13**
009 - Small Break LOCA EA1.13 - Ability to operate and monitor the following as they apply to a small break LOCA:
ESFAS Proposed Question:
Which of the following ESAS Digital Channel Trip Conditions are designed to preclude a release path from containment to the environment, and to protect fuel cladding, during the worst case Small Break LOCA event?
A. ONLY High reactor building pressure at 4 psig B. ONLY High reactor building pressure at 30 psig C. ONLY Low reactor coolant system pressure at 1590 psig D. BOTH high reactor building pressure at 4 psig AND low reactor coolant system pressure at 1590 psig Proposed Answer: __D___
Explanation (Optional):
D. Per STM 1-65, ESAS, post-TMI changes were made at ANO1 to preclude potential release paths from containment to the environment. Some components actuated by digital channels 5 or 6 (actuated on high building pressure only) were moved to digital channels 1 through 4 (actuated by either low reactor coolant system pressure or high reactor building pressure).
Technical Reference(s): ____STM 1-65, ESAS (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _________________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New __X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 _____
55.43 _____
Comments:
Question: 4 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # 011EK2.02________
Importance Rating _2.6__ _____
K/A: Knowledge of the interrelations between the Large Break LOCA and the following:
Pumps Proposed Question:
Given:
Large break LOCA has occurred ESAS has actuated.
A3 bus is locked out and cannot be re-energized.
LPI/HPI flow rates are as follows:
- "B" LPI flow 3150 gpm
- "B" HPI pump total flow is 275 gpm Which of the following actions is required per the ESAS EOP for these conditions?
A. Restore full HPI flow on "B" HPI pump.
B. Secure the "B" HPI pump.
C. Energize bus B-5 from bus B-6.
D. Swap to RB sump recirculation.
Answer: B Answer "B" is correct since LPI flow has been greater than 3050 gpm.
Answer "A" is incorrect, this action would only be taken if LPI flow was inadequate.
Answer "C" is incorrect, this would help to restore electrical alignment but is incorrect for the given conditions.
Answer "D" is incorrect for BWST is not less than 6 ft. but is not a specific required action for the given conditions.
Technical Reference(s): 1202.001, 1202.013, 1202.002 _____________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___N/A______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __10_
55.43 _____
Comments:
Question: 5 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # __015 AA1.05______
Importance Rating __3.8_ _____
K/A Statement: [015/017 Reactor Coolant Pump (RCP) Malfunctions] Ability to operate and / or monitor the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): RCS Flow Proposed Question:
The plant is steady state at 70% power per Dispatcher request.
Subsequently, you observe the following indications:
- "A" MFW flow 2.3 e 6 lbm/hr
- "B" MFW flow 4.7 e 6 lbm/hr What event would cause this MFW flow discrepancy?
A. "B" MFW pump trip B. "A" T cold instrument failed high C. "A" RCP trip D. "D" RCP trip Answer: D Notes:
Answer [D] is correct, a RCP trip caused FW to re-ratio with the highest flow in the opposite loop.
Answer [a] is incorrect, a MFW trip would cause FW flow to decrease to both SGs.
Answer [b] is incorrect, this would cause FW to decrease to the "A" SG and increase to the "B" SG.
Answer [C] is incorrect, this would cause FW to decrease to the "A" SG and increase to the "B" SG.
Reference:
STM 1-64, Rev. 14, Integrated Control System, pages 38 - 39 History:
Original question created for 2001 RO/SRO Exam.
Proposed references to be provided to applicants during examination: _None________
Learning Objective: _________________________ (As available)
Question Source: Bank # ___X____
Modified Bank # _______ (Note changes or attach parent)
New _______
Question History: Last NRC Exam _____2001___
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X___
10 CFR Part 55 Content: 55.41 __7___
55.43 _____
Comments:
Question # _6__
Examination Outline Cross-
Reference:
Level RO SRO Tier # __1___ _____
Group # __1___ _____
K/A # __022 AA2.04 _____
Importance Rating __2.9__ _____
K/A Statement: (022 Loss of Reactor Coolant Makeup) Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: How long PZR level can be maintained within limits.
Proposed Question:
Given the following conditions:
- Loss of all AC power has occurred
- RCS system leakage of 30 gpm is occurring
- Atmospheric dump system is operating as expected
- Turbine driven Emergency Feedwater Pump, P-7A, is running
- Makeup flow to the reactor coolant system is not occurring
- PZR level was 145 inches at the time AC power was lost Which of the following is the MAXIMUM amount of time that can elapse before the pressurizer empties?
A. 20 minutes B. 30 minutes C. 45 minutes D. 60 minutes Proposed Answer: __D___
Explanation: Per TRM B 3.4.5, Reactor Coolant System (RCS) Operational Leakage, Given the above conditions it will take more than 60 minutes to empty the pressurizer from the combined effects of system leakage and cooldown. The equal sign in conjunction with answers A, B and C make these answers incorrect.
Technical Reference(s): TRM B 3.4.5 Reactor Coolant System (RCS) Operational Leakage, Rev. 5 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: None
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X___
10 CFR Part 55 Content: 55.41 _(5)__
55.43 _____
Comments:
Question 7 Examination Outline Cross-
Reference:
Level RO SRO Tier # _1___ _____
Group # _1 __ _____
K/A # 025 AA2.03 Importance Rating _3.6__ _____
Proposed Question:
025 Loss of Residual Heat Removal System (RHRS)
Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: Increasing reactor building sump level Proposed Question:
Which of the following Decay Heat Removal System valves relieve to the Reactor Building Sump?
- 2) PSV-1404/-1405, BWST/RBS SUCTION LINE RELIEF VALVES
- 3) PSV-1406/-1407, DHR PUMPS P-34B/P-34A DISCHARGE RELIEF VALVES A. 1 only B. 1 and 2 only C. 2 and 3 only D. 3 only Proposed Answer: _A_
Explanation (Optional):
A) Correct. PSV-1403 relieves to the RBS.
B) Incorrect. PSV-1404/-1405 relieve to the Aux Bldg Sump.
C) Incorrect. PSV-1404/-1405 and PSV-1406/-1407 relieve to the Aux Bldg Sump.
D) Incorrect. PSV-1406/-1407 relieve to the Aux Bldg Sump.
Technical Reference(s): STM 1-05, DHR System Training Manual, R-18.
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: None Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New __X_____
Question History: Last NRC Exam ___N_______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _X____
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 .8___
55.43 _____
Comments:
Question 8 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # _____026AA1.01___
Importance Rating __3.1_ _____
AA1. Ability to operate and/or monitor in the control room as they apply to the Loss of Component Cooling Water: CCW temperature indications Proposed Question:
Given:
- The plant is at 330 MWe following a reactor plant startup.
- A momentary loss of intermediate cooling water (ICW) occurred.
- Group 6 rod 1 is noted to be 185 degrees F five minutes after ICW had been lost.
- Group 7 rod 2 stator temperature is noted to be 205 degrees F on the plant computer.
- All other control rod stator temperatures are between 155 degrees F and 165 degrees F and stable.
According to Procedure 1203.003, Control Rod Drive Malfunction, which of the following actions is required to be taken FIRST?
A. Manually trip the reactor and enter Procedure 1202.001, Reactor Trip B. Reduce power to less than 300 MWe and remain in Procedure 1203.003 C. Remove the six stator fuses associated with Group 7, rod 2 D. Transfer Group 6, rod1 to the auxiliary power supply Proposed Answer: ___A___
Explanation (Optional):
A is correct since greater than one control rod is greater than 180 degrees F B would be correct if the direction was to reduce power to less than 360 MWe C is performed AFTER actions for the group 7 rods are performed D would be the correct answer if only one rod was overheating Technical Reference(s): _1203.003, Control Rod Drive Malfunction Action, Revision 024 pages 35-38 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis ___X_
10 CFR Part 55 Content: 55.41 __4__
55.43 _____
Comments:
Question: 9 Examination Outline Cross-
Reference:
Level RO Tier # __1__ _____
Group # __1__ _____
K/A # __027AK2.02______
Importance Rating _2.6 _ _____
AK2. Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: AK2.02 Controllers and positioners Proposed Question:
The plant is operating at rated thermal power when the loop A RCS narrow range pressure instrument, PT-1021, INSTANTANEOUSLY fails high.
The effect of this failure is:
A. The Electro-Magnetic Relief Valve will open.
B. Reactor Protection System channel A will trip.
C. The pressurizer heaters will turn off.
D. The pressurizer spray valve will open.
Proposed Answer: ___B__
Explanation (Optional):
All of the distractors would be valid if it were not for the automatic transfer to the other pressure channel by the SASS. This action eliminates all the choices except B since this signal is processed up stream of SASS. Therefore, only B is correct.
Technical Reference(s): __ ________
(Attach if not previously provided) __STM 1-63, Reactor Protection System, Rev 9 ________
(including version/revision number) __
Proposed references to be provided to applicants during examination: ___None__________
Learning Objective: _________________________ (As available)
Question Source: Bank # _0155__
Modified Bank # _______ (Note changes or attach parent)
New _______
Question History: Last NRC Exam __1998______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __XX_
10 CFR Part 55 Content: 55.41 __7__
55.43 _____
Comments:
Question 10 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # __038 EA2.10_____
Importance Rating __3.1_ _____
EA2.10 - Ability to determine or interpret the following as they apply to a SGTR: Flowpath for charging and letdown flows Proposed Question:
Given:
- Leak rate of 15 gpm has been identified on A OTSG
- P36A is OP, P36C is ES
- PZR level is 195 and slowly lowering Per 1202.006, Tube Rupture, which of the following identifies the FIRST operator action that should be taken to raise PZR level?
B. Open CV-1220, HPI Block valve C. Start P36C and open CV-1285, HPI Block valve D. Place Pressurizer Level Control (CV-1235) in HAND and open Proposed Answer: __B___
Explanation (Optional):
A. Incorrect: RT-3 is only initiated if SCM is inadequate, however similar sounding RT-2 is initiated if opening the HPI block valve is unsuccessful in raising PZR level.
B. Correct: Per 1202.006, Step 4 RNO (PZR level not > 200), B - Operate HPI Block associated with OP HPI pump to restore PZR level > or = 200 C. Incorrect: RNO actions do not direct starting STBY pump D. Incorrect: RNO actions do not direct taking manual control of CV-1235.
Technical Reference(s): _____1202.006, Tube Rupture, Step 4_______________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ____None________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 _____
55.43 _____
Comments:
Question: 11 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # 054AK3.04________
Importance Rating _4.4__ _____
K/A: Knowledge of the reasons for the following responses as they apply to the Loss of Main Feedwater (MFW): Actions contained in EOPs for loss of MFW Proposed Question:
While at 100 percent power, Unit 1 experienced a momentary grid disturbance and a loss of both Main Feedwater pumps.
EFW is not available.
The crew transitions to 1202.004, Overheating, and the following conditions exist:
- Vital and non-vital buses remain energized and voltage is normal
- Both Main Feedwater pumps are NOT available
- All Reactor Coolant Pumps are off
- Subcooling margin is adequate
- Both EFIC OTSG levels are 20 inches and lowering Which of the following actions is procedurally required?
A. Perform RT-16, Feeding Intact SG, to refill OTSGs to between 20 and 40 inches B. Perform RT-16, Feeding Intact SG, to refill OTSGs to between 300 and 340 inches C. Perform RT-16, Feeding Intact SG, to refill OTSGs to between 370 and 410 inches D. Initiate HPI Cooling per RT-4 Proposed Answer: __B__
Explanation (Optional):
Implicit in each answer is the knowledge of the reason for performing the action. (OTSG is not dry and establish natural recirculation with the 300-340 band.)
A. (Incorrect) While 20 to 40 is the normal SG level with RCPs running, RCPs are off so this level is too low.
B. (Correct) OTSGs should be refilled to between 300 and 340.
C. (Incorrect) 370 to 410 is the reflux boiling OTSG level for inadequate SCM, but SCM is given as adequate. As long as SCM is adequate then this higher level should not be attempted due to problems with possible overfill and tube to shell DT concerns.
D. (Incorrect) Initiate HPI cooling per RT-4 is required by the contingency action for step 5 if RCS pressure became excessive or any form of secondary feed was not expected to be available, but the conditions state that buses are energized so AFW should be available.
Technical Reference(s): 1202.004, 1202.012 ______________________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___N/A______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __10_
55.43 _____
Comments:
Question: 12 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # __055 G 2.2.37______
Importance Rating __3.6_ _____
K/A Statement: [055 Loss of Offsite and Onsite Power (Station Blackout)] Ability to determine operability and/or availability of safety related equipment.
Proposed Question:
When recovering from a Station Blackout condition with no offsite power available and only one EDG operable, which of the following is true?
A. The EDG output breaker should automatically close once the EDG has started and ONE other feeder breaker to its respective bus is closed B. The EDG output breaker hand switch must be momentarily held in the "close" position to energize its respective bus C. ONE A3-A4 crosstie breaker must be open prior to closing the EDG output breaker D. The EDG should be used to back feed A2 to restore essential loads.
Answer : C Notes:
(a.) is incorrect. The EDG should start upon either A3/A4 or B5/B6 bus under voltage but the output breaker will not automatically close if the normal feeder breaker to the other feeder breakers is open.
(b.) is incorrect. The EDG output breaker should automatically close.
(c.) is correct. At least one A3/A4 crosstie breaker must be open for the EDG output breaker to close.
(d.) is incorrect. With only one EDG available, only the ES bus loads should be placed on the diesel.
Reference:
STM 1-32, Electrical Distribution, Rev. 20, page 47, step 3.3.6 History:
Developed for use in 98 RO Re-exam Selected for use in 2002 RO/SRO exam.
Selected for use in 2013 RO/SRO exam.
Technical Reference(s):
(Attach if not previously provided) _
(including version/revision number) _STM 1-32, Electrical Distribution, Rev. 20, Pg. 47___
Proposed references to be provided to applicants during examination: _________________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # ___X____ (Note changes or attach parent)
New ___2002____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis ___X__
10 CFR Part 55 Content: 55.41 __7___
55.43 _____
Comments:
Original Bank Question: 174 (Direct)
When recovering from a Station Blackout condition with no offsite power available and only one EDG operable, which of the following is true?
- a. The EDG output breaker will automatically close once the EDG has started and all other feeder breakers to it's respective bus are open.
- b. The EDG output breaker hand switch must be momentarily held in the "close" position to override the ES bus under voltage relay.
- c. The A3-A4 crosstie breakers are closed prior to closing the EDG output breaker.
- d. The EDG should be used to back feed A2 to restore essential loads.
Answer : A Notes:
(a.) is correct. The EDG is automatically started upon either A3/A4 or B5/B6 bus undervoltage and the output breaker will automatically close if the normal feeder breaker to the 4160V ES bus has tripped and one of the A3/A4 tie breakers is open.
(b.) is incorrect. The undervoltage relay will not prevent the EDG output breaker from closing.
(c.) is incorrect. At least one A3/A4 crosstie breaker must be open for the EDG output breaker to close.
(d.) is incorrect. With only one EDG available, only the ES bus loads should be placed on the diesel.
Reference:
STM 1-32, Electrical Distribution, Rev. 20, page 47, step 3.3.6 History:
Developed for use in 98 RO Re-exam Selected for use in 2002 RO/SRO exam.
Selected for use in 2013 RO/SRO exam.
Question # _13__
Examination Outline Cross-
Reference:
Level RO SRO Tier # __1___ _____
Group # __1___ _____
K/A # __056 AK1.03 _____
Importance Rating __3.1__ _____
K/A Statement: (056 Loss of Offsite Power) Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite Power: Definition of subcooling: use of steam tables to determine it.
Proposed Question:
Given the following conditions:
- Degraded power event has occurred
- EDGs are powering the 4160 volt buses
- Core exit temperature is 600F
A. 1543 psia B. 1574 psia C. 1588 psia D. 1606 psia Proposed Answer: __A__
Explanation: Per the Steam Tables at 600F and 1650 psia the condition of the RCS will be subcooled. At 600F the RCS will not reach saturation until 1543 psia. This renders B, C and D incorrect.
Technical Reference(s): Steam Tables (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __ ___
Comprehension or Analysis __X___
10 CFR Part 55 Content: 55.41 _(8)__
55.43 _____
Comments:
Question 14 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # __057 AG 2.4.6_____
Importance Rating _3.7__ _____
057 Loss of Vital AC Electrical Instrument Bus: Knowledge of EOP mitigation strategies.
Proposed Question:
Given the following:
- The unit is in Mode 1
- Annunciator K01-A6, RS2 INVERTER TROUBLE, is in alarm.
- RS-2 is aligned to its normal inverter, Y22.
- A local operator reports that annunciator K1622-05, INVERTER FAILURE is in alarm locally.
Per Technical Specifications and plant procedures, the Inverter Y22 is ___________ and the action taken should be to ________________________ .
A. Inoperable, transfer RS2 to Swing Inverter Y25.
B. Operable, manually transfer Y22 static switch to ALT SOURCE.
C. Inoperable, manually transfer Y22 static switch to ALT SOURCE..
D. Operable, transfer RS2 to Swing Inverter Y25.
Proposed Answer: __A___
Explanation (Optional):
A) Correct. The inverter is inoperable even though RS2 is operable and the action to align RS2 to the swing inverter is correct.
B) Incorrect. RS2 is operable, but the static switch will automatically transfer.
C) Incorrect. Although RS2 is inoperable, the static switch will automatically transfer. This action is similar to the action to transfer the manual selector switch to alternate source.
D) Incorrect. RS2 is operable but the inverter Y22 is inoperable however, the action is correct.
Technical Reference(s): STM 1-32-3 Rev 1, 120 Volt Vital AC Distribution; (Attach if not previously provided) 1107.003, Rev 22, Inverter and 120V Vital AC Distribution; 1203.012A Rev 42, Annunciator K01 Corrective Action.
(including version/revision number)
Proposed references to be provided to applicants during examination: _None___
Learning Objective: ___A1LP-RO-TS, objective 4___________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____NO________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X___
10 CFR Part 55 Content: 55.41 .10__
55.43 _____
Comments:
Question 15 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # _____058AK1.01___
Importance Rating __2.8_ _____
AK1. Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation Proposed Question:
Given the following indications at 100% power:
- Annunciator D01 UNDERVOLTAGE (K01-A7) in alarm
- Annunciator D01 TROUBLE (K01-D7) in alarm
- Annunciator D01 CHARGER TROUBLE (K01-E7) in alarm
- The reactor has tripped
- The turbine trip solenoid light is off
- Breaker position lights on the LEFT side of C10 are off Per Procedure 1203.036, Loss of 125VDC, which of the following is the FIRST required action?
A. Trip the main generator output breakers B. Transfer D11 to emergency supply D02 C. Trip any running reactor coolant pumps D. Transfer D12 to emergency supply D02 Proposed Answer: ___B___
Explanation (Optional):
Answer B is correct per procedure Answer A and D are not performed in Procedure 1203.036 Answer C is performed only if both RCP seal injection and seal cooling are lost Technical Reference(s): _1203.036, Loss of 125VDC, Revision 10, Page 2 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # ___X____ (Note changes or attach parent)
New _____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis ___X_
10 CFR Part 55 Content: 55.41 __10__
55.43 _____
Comments:
Modified from Bank Question 0187 QID: 0187 Rev: 1 Rev Date: 4/25/2002 Source: Direct Originator: S.Pullin TUOI: A1LP-RO-AOP Objective: 4.5 Point Value: 1 Section: 4.2 Type: Generic APE System Number: 058 System
Title:
Loss of DC Power
==
Description:==
Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation.
K/A Number: AK1.01 CFR
Reference:
41.8 / 41.10 / 45.3 Tier: 1 RO Imp: 2.8 RO Select: No Difficulty: 3 Group: 1 SRO Imp: 3.1 SRO Select: No Taxonomy: C Question: RO: SRO:
Given the following indications at 100% power:
- Annunciator D02 UNDERVOLTAGE (K01-A8) in alarm.
- Annunciator D02 TROUBLE (K01-D8) in alarm.
- Annunciator D02 CHARGER TROUBLE (K01-E8) in alarm.
- The reactor has tripped.
- The turbine trip solenoid light is on.
- Breaker position lights on the RIGHT side of C10 are off.
What are the actions required of the CBOT?
A. Trip the main generator output breakers.
B. Transfer D11 to emergency supply D01.
C. Trip all RCPs.
D. Transfer D21 to emergency supply D01.
Answer D. Transfer D21 to emergency supply D01.
Notes:
[d] is correct per 1203.036 as the conditions are indicative of a loss of D02.
[a] and [b] are incorrect due to this a loss of D02 not D01 these are actions for the loss of D01.
[c] is incorrect due to we have not loss seal injection and seal cooling, this is an action in this procedure section if both of the before mentioned system functions are lost
References:
1203.036, Chg. 08 History:
Developed for use in 98 RO Re-exam Selected for use in 2002 RO/SRO exam, revised slightly.
Selected for 2005 RO re-exam.
Selected for the 2010 RO/SRO exam
Question: 16 Examination Outline Cross-
Reference:
Level RO Tier # __1__ _____
Group # __1__ _____
K/A # __062AK3.01______
Importance Rating _3.2 _ _____
AK3. Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: AK3.01 The conditions that will initiate the automatic opening and closing of the SWS isolation valves to the nuclear service water coolers.
Proposed Question:
The plant is operating normally at rated thermal power with the following conditions:
- Service water pumps P4A, P4B, and P4C are in service
- The service water cross-tie valves are all open and in AUTO
- The P4B MOD breaker is aligned to provide power from the A4 bus Following an Engineered Safeguards Actuation System (ESAS) actuation; A. All the crosstie valves will close and P4B will trip in order to maintain train separation B. The crosstie valves between P4B and P4C will remain open to prevent deadheading the pump C. All four crosstie valves will close to establish divisional separation and P4B will continue to provide cooling water to the Aux Cooling heat exchanger D. All four crosstie valves will remain open as long as off-site power is not lost in order to maximize system reliability Proposed Answer: ___B__
Explanation (Optional):
With the system aligned as described in the stem, the cross tie valves between P4B and C will remain open making A and C incorrect. The cross tie valves between P4A and B will close making D incorrect. Distractor B is the correct answer.
Technical Reference(s): _ __STM 1-42, Service and Auxiliary Cooling Water, page 19, Rev 21 _____________________ ______________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___XX _
Question History: Last NRC Exam _____NA_____
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __7__
55.43 _____
Comments:
Question 17 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # __065 AK3.04_____
Importance Rating __3.0 _____
AK3.04 - Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: Cross-over to backup air supplies Proposed Question:
Which of the following describes why AOP 1203.024, Loss of Instrument Air, directs closing the Breathing Air to Instrument Air cross-connect when Breathing Air pressure drops to less than 80 psig?
A. Prevent the likelihood of personnel overexposure to airborne radiation B. Prevent automatic alignment of the Service Air system to the IA system.
C. Ensure operation of Critical Air Operated Components identified in Attachment A.
D. Ensure proper operation of the BA to IA cross connect line three solenoid operated valves (SV-5503A, SV-5503B and SV-5503C)
Proposed Answer: __A___
Explanation (Optional):
Technical Reference(s): _AOP 1203.024, Loss of Instrument Air, Rev 13 STM 1-48, Compressed Air Systems, Rev 13_______
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ___None_________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New __XX___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __XX _
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 _____
55.43 _____
Comments:
Question: 18 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # BE04EG2.4.20 ____
Importance Rating _3.8__ _____
K/A: (Generic) Inadequate Heat Transfer - Loss of Secondary Heat Sink: Knowledge of the operational implications of EOP warnings, cautions, and notes.
Proposed Question:
Unit 1 was at 100 percent power, normal lineup.
Given:
- A Loss of Offsite Power occurred and both EFW pumps could not be started.
- The crew has just entered 1202.005, Inadequate Core Cooling.
The following conditions exist:
- A3 and A4 are energized by their respective Diesel Generators
- EFW pump P-7B is now available
In accordance with 1202.005, which of the following actions is required next?
A. Align P-36B to A4, start P-36B AUX Lube Oil pump and start P-36B B. Start P-36B AUX Lube Oil Pump and verify automatic start of P-36B C. Verify RCP Seal Injection is aligned, start LPI pumps and open ERV (PSV-1000)
D. Start LPI pumps, open ERV (PSV-1000) and start P-7B Proposed Answer: __A__
Explanation (Optional):
A. (Correct) Because P-36C is the ES STBY pump, P-36B would be aligned to A3 in a normal lineup. Per the CAUTION statement in 1202.005, P-36B Motor-Operated Disconnect needs to be aligned to A4. The reason for this is that P-7B (Motor Operated EFW pump powered from A3) will be started, and A3 may be overloaded when carrying 2 HPI pumps and the EFW pump.
B. (Incorrect) This sequence is correct, but not the next, which is to align HPI B to A4 first C. (Incorrect) Correct sequence, but not the next.
D. (Incorrect) Correct sequence, but not the next.
Technical Reference(s): 1202.005, STM 1-04 ______________________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___N/A______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __10_
55.43 _____
Comments:
Question: 19 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __2__ _____
K/A # __024 AA1.02______
Importance Rating __3.7_ _____
K/A Statement: [024 Emergency Boration] Ability to operate and / or monitor the following as they apply to Emergency Boration: Boric acid pump Proposed Question:
The plant has the following conditions:
- The reactor has been tripped
- 3 Control Rods failed to fully insert
- The Control Room Supervisor (CRS) directs the At the Controls (ATC) operator to perform Emergency Boration per Repetitive Task 12 (RT 12)
- Boric Acid Pump P39A and P39B have been started
- The RUN key has been depressed on the Batch Controller
- Batch Controller Flow Control Valve CV-1249 is fully open The ATC operator observes that the Batch Controller output rate is 3 gpm. Per procedure 1202.012, Repetitive Tasks, the ATC operator should FIRST:
A. Close CV-1250, Batch Controller Outlet B. Open CV-1251, Condensate to Batch Controller C. Secure P39A and P39B Boric Acid Pumps D. Adjust pressurizer level control set point to 220 Proposed Answer: __C___
Explanation (Optional): Per 1202.012, Repetitive Tasks, RT 12, Step 1.I., if the Batch Controller output rate is less than 5 gpm, the first action is to secure running Boric Acid pumps.
C is correct.
A is incorrect because although it would be plausible to adjust flow rate with this, it conflicts with direction in RT 12.
CV-1251, the Condensate to Batch Controller valve, is closed in RT 12, Step 1.C. B is incorrect.
Adjusting the pressurizer level control set point is a step completed later in the procedure if the Batch Controller output rate was equal to or greater than 5 gpm (RT 12, Step 1.J). D is incorrect.
Technical Reference(s): ___1201.012 (Change 010)____________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _________________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __7___
55.43 _____
Comments:
Question 20 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __2__ _____
K/A # _____059AA1.03___
Importance Rating __3.0 _____
AA1. Ability to Operate and/or Monitor the following as they apply to the Accidental Liquid Radwaste Release: Flow rate controller Proposed Question:
The plant is operating at full power. No planned releases are in progress when the following alarms occur:
- PROC MONITOR RADIATION HI (K10-B2)
- Liquid Radwaste Process Monitor (RI-4642), (C25, Bay 2)
Per Procedure 1203.007, Liquid Waste Discharge Line High Radiation Alarm, which of the following actions is required to be taken FIRST?
A. Verify that no release is in progress by monitoring discharge flow to flume (FI-4642) on C19 B. Verify that no release is in progress by verifying RADIATION MONITOR TROUBLE (K10-C1) has also annunciated C. Ensure Treated Waste Discharge to Circulating Water (CW) Flume (CZ-58) Manual Valve is closed D. Ensure Filtered Waste Monitoring Tank Discharge to CW Flume (DZ-25) Manual Valve is closed Proposed Answer: ___A___
Explanation (Optional):
Answer A is the correct answer Answer B could be thought correct if problem was thought to be a faulty rad monitor Answer C incorrect but contingency action in 1203.007 Answer D incorrect but contingency action in 1203.007 Technical Reference(s): _1203.007, Liquid Waste Discharge Line High Radiation Alarm, Revision 008-01-0, Page 2 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis ___X_
10 CFR Part 55 Content: 55.41 __13__
55.43 _____
Comments:
Question # _21__
Examination Outline Cross-
Reference:
Level RO SRO Tier # _1___ _____
Group # _2___ _____
K/A # 060 AK3.03________
Importance Rating _3.8__ _____
060 Accidental Gaseous Radwaste Release: AK3.03 Knowledge of the reasons for the following responses as they apply to the Accidental Gaseous Radwaste: Actions contained in EOP for accidental gaseous-waste release.
Proposed Question:
Given:
- The unit is in Mode 1
- The following alarms are received:
o PROC MONITOR RADIATION HI (K10-B2) o RADWASTE GAS PANEL TROUBLE (K09-D5) o PROC MONITOR RE-4830 (Gaseous Radwaste) High Alarm AOP 1203.006, WASTE GAS DISCHARGE LINE RADIATION HIGH, directs the crew to verify the automatic actuations for this condition have occurred as designed. Which of the following describes a purpose of this direction?
A. To ensure that Gas Collection Header discharge to Station Vent Plenum is secured B. To ensure that any running Waste Gas Compressors have tripped C. To ensure that Reactor Building Vent Header Isolation Valves CV-4803 and CV-4804 close D. To ensure that Auxiliary Building Vent Header gases are diverted to the Waste Gas Surge Tank T-17 Proposed Answer: __D___
Explanation (Optional):
When high radiation is sensed on Waste Gas Process Monitor RE-4830, three actions occur:
- 1) CV-4830 shuts to secure RBVH and ABVH discharge to the Station Vent Plenum; 2) CV-4820 shuts to secure any Waste Gas Decay Tank discharges to Station Vent Plenum; and 3)
CV-4806 opens to divert ABVH gases from Station Vent Plenum to Waste Gas Surge Tank T-17.
A) Incorrect. No automatic actions occur on the Gas Collection Header.
B) Incorrect. No automatic actions occur on the Waste Gas Compressors.
C) Incorrect. CV-4803 & CV-4804 close on ESAS, not high waste gas disch line radiation.
D) Correct. CV4830 shuts and CV-4806 opens to divert ABVH gases to T-17.
Technical Reference(s): STM 1-54 Rev. 6; 1203.006 Rev 10-02; SAR Fig. 11-3 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None_______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____No______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 .10/.11/.13 55.43 _____
Comments:
Question 22 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __2__ _____
K/A # _____074EA2.06 ___
Importance Rating __4.0_ __4.6_
EA2. Ability to determine or interpret the following as they apply to Inadequate Core Cooling: Changes in pressurizer level due to pressurizer steam bubble transfer to the reactor coolant system during inadequate core cooling.
Proposed Question:
The following plant conditions exist:
-The reactor is tripped
-Reactor Coolant Pumps are tripped
-Core Exit Thermocouples read 800 Degrees F and INCREASING
-RCS Pressure is 1500 psig and steady
-Pressurizer level is INCREASING Which of the following procedures should be used to mitigate the most urgent problem?
A. Procedure 1202.004, Overheating B. Procedure 1202.010, ESAS C. Procedure 1202.002, Loss of Subcooling Margin D. Procedure 1202.005, Inadequate Core Cooling Proposed Answer: ___D___
Explanation (Optional):
Answer D the correct answer.
Answer A is incorrect. 1202.004 does not address inadequate core cooling Answer B is incorrect. 1202.010 does not address inadequate core cooling Answer C is incorrect. While a loss of subcooling margin exists, the more urgent problem is the inadequate core cooling Technical Reference(s): _1202.002, Loss of Subcooling Margin, Revision 6, Page 12 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge ____
Comprehension or Analysis _X__
10 CFR Part 55 Content: 55.41 _10__
55.43 _____
Comments:
Question: 23 Examination Outline Cross-
Reference:
Level RO Tier # __1__ _____
Group # __2__ _____
K/A # __BA01AK2.2______
Importance Rating _3.5 _ _____
AK2. Knowledge of the interrelations between the (Plant Runback) and the following:
AK2.2 - Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
Proposed Question:
Which of the following reactor coolant system related signals input to the Integrated Control System (ICS) as part of the plant runback logic?
A. Reactor Coolant Pump breaker positions B. Reactor Coolant Pump differential pressures C. Reactor Coolant System loop flows D. Reactor Coolant System total flow Proposed Answer: ___A__
Explanation (Optional):
A is correct since plant runback is based upon the number of RCPs as determined by breaker position B will cause an alarm, but not a runback C will cause feedwater to be re-ratioed, and is indicative of an RCP trip, but will not cause a runback D will cause a BTU limit alarm, but not a runback Technical Reference(s): STM 1-64, Integrated Control System, Rev 13,page 22__
(Attach if not previously provided)
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ___None__________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___NA__
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __7__
55.43 _____
Comments:
Question 24 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __2__ _____
K/A # __BW/A04 AK3.2____
Importance Rating __3.4_ _____
AK3.2 - Knowledge of the reasons for the following responses as they apply to the (Turbine Trip): Normal, abnormal and emergency operating procedures associated with (Turbine Trip).
Proposed Question:
A plant power escalation is in progress at 39% power. The following conditions are observed:
- Rapid rise in RCS temperature
- Rapid rise in RCS pressure
- Rapid rise in PZR level
- Megawatt output = 0
- MSSV open alarm No other annunciators in alarm except for those expected for the above conditions.
What procedure contains the required mitigating operator actions?
A. 1202.001, Reactor Trip B. 1203.001, ICS Abnormal Operation C. 1203.018, Turbine Trip Below 43% Power D. 1203.027, Loss of Steam Generator Feed Proposed Answer: __C___
Explanation (Optional):
Technical Reference(s): ___AOP 1203.018, Rev 13_______________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None_______
Learning Objective: _________________________ (As available)
Question Source: Bank # _QID 0370_
Modified Bank # _______ (Note changes or attach parent)
New _______
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __ __
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 _____
55.43 _____
Comments:
Question: 25 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __2__ _____
K/A # BA06AG2.4.34 ____
Importance Rating _4.2__ _____
K/A: (Generic) Shutdown Outside the Control Room: Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.
Proposed Question:
Due to an event, Unit 1 has entered 1203.002, Alternate Shutdown. All of the actions in the control room have been completed and you have been assigned RO #1 follow-up actions.
While opening CV-1408, BWST Outlet Valve, you notice significant flow indicated by throttling noise and vibration.
Based on this, what is happening, and what action should you perform FIRST in accordance with 1203.002?
A. HPI is injecting into the reactor vessel. Close HPI Pump Recirculation Block (CV-1300).
B. BWST is draining to the Reactor Building Sump. Close BWST Outlet Valve (CV-1408).
C. BWST is draining to the Reactor Building Sump. Immediately proceed to the Decay Heat vaults and close the RB Sump - Line A and B Outlet valves (CV-1405 and CV-1406).
D. HPI is injecting into the reactor vessel. Open the other BWST Outlet (CV-1407).
Proposed Answer: __B__
Explanation (Optional):
Opening 1408 is the first local valve manipulation of RO #1 A (Incorrect) HPI Blocks are still closed B (Correct)
C (Incorrect) Closing 1405 and 1406 are done after closing 1408 first.
D (Incorrect) HPI Blocks are still closed Technical Reference(s): 1203.002 ______________________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___N/A______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __10_
55.43 _____
Comments:
Question: 26 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __2__ _____
K/A # __BW/E08 G 2.2.39___
Importance Rating __3.9_ _____
K/A Statement: [BW/E08 LOCA Cooldown] Knowledge of less than or equal to one hour Technical Specification action statements for systems.
Proposed Question:
The plant has the following conditions:
- A LOCA has occurred
- An RCS cooldown is in progress
- Calculations preparing for needed boration during the cooldown per procedure 1103.015 Attachments 4 and 5 show that the SDM at the beginning of the cooldown is -1.3% k/k
- At 1430, RCS temperature is 440F, RCS Pressure is 1100 psig Based on the given conditions, what action is required to be taken FIRST and what is the MAXIMUM completion time for this action per Technical Specifications?
A. Initiate boration to restore SDM to within COLR limits within 30 minutes B. Adjust the RCS cooldown rate to meet rate restrictions within 30 minutes C. Initiate boration to restore SDM to within COLR limits within one hour D. Adjust the RCS cooldown rate to meet rate restrictions within one hour Proposed Answer: __B___
Explanation (Optional): Per LCO 3.4.3, RCS cooldown rates shall be maintained within the limits specified in Figure 3.4.3-2. Note 4 on Figure 3.4.3-2 states the RCS cooldown rate and step cooldown rate limits. With RCS temperature greater than or equal to 280F, the maximum cooldown rate is 100F/hour, or a maximum step change of 50F in any 1/2 hour. From 1400 to 1430, the step cooldown rate change was 60F, and from 1430 to 1445 the change was 13F, which is still cooling down at a rate which exceeds the cooldown limits. Therefore, LCO 3.4.3, Condition B requires that the cooldown parameters be restored to within limits within 30 minutes. B is correct.
Answers A and C are required to be addressed per plant procedure 1103.015, Step 7.5.1, to restore the SDM to equal to or more negative than -1.5% k/k. However, the Technical Specifications require that the SDM be maintained at -1.0% k/k per procedure 1103.015, Step
5.5. Therefore, no action on this is required by the Technical Specifications with this condition.
A and C are incorrect.
The action in answer D is correct, but the completion time stated is not in accordance with LCO 3.4.3, Condition B.
Technical Reference(s): __1103.015 (Change 052), ANO Unit 1 Technical Specifications, LCO 3.4.3 and 3.1.1 (Rev. 245)______
(Attach if not previously provided) _1103.004 (Change 023), 1202.001 (Change 032),
1202.013 (Change 004), 1203.013 (Change 018)______
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _________________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X___
10 CFR Part 55 Content: 55.41 __7, 10___
55.43 _____
Comments:
Question # _27__
Examination Outline Cross-
Reference:
Level RO SRO Tier # __1___ _____
Group # __2___ _____
K/A # __E14 EK1.3 _____
Importance Rating __3.2__ _____
K/A Statement: (E14 EOP Enclosures) Knowledge of the operational implications of the following concepts as they apply to the (EOP Enclosures): Annunciators and conditions indicating signals, and remedial actions associated with the (EOP Enclosures).
Proposed Question:.
Given the following:
- SCM has been lost
- ESAS has actuated
- HPI has been stopped
- LPI is injecting
- 1202.010 ESAS has been entered
- RCS pressure has just dropped below 140 psig
- Boric Water Storage Tank has just reached six feet in level What is the NEXT required operator action and the MAXIMUM allowed time to perform this action?
A. Isolate Core Flood Tank outlet valves; ten minutes B. Align LPI pump suction to RB sump; ten minutes C. Isolate Core Flood Tank outlet valves; three minutes D. Align LPI pump suction to RB sump; three minutes Proposed Answer: __D__
Explanation: Isolating CFT outlet valves is not a time critical operator action and is not the NEXT required operator action when BWST reaches six feet. BWST level is the driving parameter and must be accomplished within three minutes once the level has reached six feet.
RCS pressure dropping below 140 psig does drive isolating CFT outlet valves but with LPI injecting, the cooldown and depressurization can continue no matter the status of these valves.
Technical Reference(s): 1202.002 (change 006), Loss of Sub Cooling Margin, 1202.010 (change 007), ESAS, and 1015.050 (change 001), Time Critical Operator Actions Program (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: None Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X __
Comprehension or Analysis __ ___
10 CFR Part 55 Content: 55.41 _(7)__
55.43 _____
Comments:
Question # _28__
Examination Outline Cross-
Reference:
Level RO SRO Tier # _2___ _____
Group # _1___ _____
K/A # 003 A3.05____
Importance Rating _2.7__ _____
003 Reactor Coolant Pump System (RCPS)
A3.05: Ability to monitor automatic operation of the RCPS, including: RCP lube oil and bearing lift pumps Proposed Question:
RCP 32A is running with the Normal and Emergency HP Lift Oil Pumps both stopped, NOT in pull-to-lock.
If RCP 32A hand switch is taken to STOP with no other manual actions, (1) How will the HP Lift Oil Pumps respond?
(2) Once RCP 32A shaft is stopped, what is the MINIMUM action required to maintain the Normal HP Lift Oil Pump stopped?
A. (1) The Normal and Emergency HP Lift Oil Pumps (P-63A & P-80A) should both start automatically.
(2) The Normal HP Lift Oil Pump P-63A hand switch must be taken to STOP B. (1) The Normal and Emergency HP Lift Oil Pumps (P-63A & P-80A) should both start automatically.
(2) The Normal HP Lift Oil Pump hand switch P-63A must be taken to STOP THEN PULL-TO-LOCK C. (1) The Normal HP Lift Oil Pump P-63A should start automatically. The Emergency HP Lift Oil Pump P-80A should start automatically only if the Normal HP Lift Oil Pump P-63A fails to start.
(2) The Normal HP Lift Oil Pump P-63A hand switch must be taken to STOP D. (1) The Normal HP Lift Oil Pump P-63A should start automatically. The Emergency HP Lift Oil Pump P-80A should start automatically only if the Normal HP Lift Oil Pump P-63A fails to start.
(2) The Normal HP Lift Oil Pump P-63A hand switch must be taken to STOP THEN PULL-TO-LOCK Proposed Answer: __B__
Explanation (Optional):
See FIGURE 03.28A, RC PUMP OIL LIFT PUMPS LOGIC, in STM 1-03 rev 19.
A) Incorrect. Both pumps will start automatically, but to maintain either pump stopped the hand switch must be taken to PULL-TO-LOCK because of the interlock which starts both pumps when the associated RCP breaker is tripped open.
B) Correct.
C) Incorrect. Both HP Lift Oil Pumps start automatically when the RCP breaker is open and the Oil Pump hand switches are not in PULL-TO-LOCK.
D) Incorrect. Both HP Lift Oil Pumps start automatically when the RCP breaker is open and the Oil Pump hand switches are not in PULL-TO-LOCK.
Technical Reference(s): STM 1-03 rev 19; 1103.006 Rev 38________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ____None_____
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam _____NO_____
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 .3___
55.43 _____
Comments:
Question 29 revision 2 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __1__ _____
K/A # _____004 K1.02___
Importance Rating __3.5_ _____
K1. Knowledge of the physical connections and/or cause and effect relationships between the CVCS and the following systems: pressurizer and RCS temperature and pressure relationships.
Proposed Question:
Unit One is operating at 100% power.
An ICS malfunction causes Tave to rise 2°F.
What is the effect on the Pressurizer, including level and pressure, during this transient?
- a. PZR level will rise to a higher steady state level. RCS pressure will drop to a lower value initially.
- b. PZR level will rise to a higher steady state level. RCS pressure will rise to a higher value initially.
- c. Makeup flow will rise to restore Pressurizer level to setpoint. RCS pressure will drop to a lower value initially
- d. Makeup flow will drop to restore Pressurizer level to setpoint. RCS pressure will rise to a higher value initially.
Answer:
- d. Makeup flow will drop to restore Pressurizer level to setpoint. RCS pressure will rise to a higher value initially.
Explanation:
Answer [d] is correct since an increase in RCS temperature will swell PZR level and RCS pressure. This causes CV-1235 to close to maintain PZR level at setpoint, therefore Makeup flow will drop.
Answer [a], [b], [c] are options that the candidate might choose if he cannot recall proper system response.
Technical Reference(s): _ASLPSTM 1-03, Reactor Coolant System, Pages 11-12 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis ___X_
10 CFR Part 55 Content: 55.41 __5__
55.43 _____
Comments:
Question: 30 Examination Outline Cross-
Reference:
Level RO Tier # __1__ _____
Group # __2__ _____
K/A # __005A4.03 ______
Importance Rating _2.8 _ _____
A4 Ability to manually operate and/or monitor in the control room: A4.03 RHR temperature, PZR heaters and flow, and nitrogen . .
Proposed Question:
The reactor is in MODE 5, the Decay Heat Removal System (DHRS) in service, and reactor pressure is at atmospheric pressure. RCS temperature is 85 degrees and slowly rising.
An operator is in the process of increasing DHRS and service water flow rates, however, in order to minimize the potential for vibration induced piping damage, DHRS total flow should be limited to no more than __________ gpm.
A. 1600 B. 2000 C. 3550 D. 4000 Proposed Answer: ___B___
Explanation (Optional):
All four distractors are valid DHRS numbers; 1600 gpm is the minimum flow or damage to the service water piping can occur, 2000 gpm is correct, 3550 gpm is the high DHRS flow alarm setpoint, and 4000 gpm is the system design flowrate.
Technical Reference(s): _Normal Operating Procedure 1104.004, Decay Heat Removal, Change 096, page 66_____________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _NONE___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___XX__
Question History: Last NRC Exam ___NA__
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __XX Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __7__
55.43 _____
Comments:
Question 31 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __1__ _____
K/A # __005 K2.01_____
Importance Rating __3.1_ _____
K2.01 - Knowledge of bus power supplies to the following: RHR pumps Proposed Question:
Given the following:
- Plant is in Mode 6
- Only P-34B Decay Heat Pump is running Which of the following would cause a loss of Decay Heat Removal?
A. A-1 voltage of 3150 volts B. A-2 voltage of 3150 volts C. B-5 voltage of 428 volts D. B-6 voltage of 428 volts Proposed Answer: __D___
Explanation (Optional):
A incorrect, this is the red train, P-34B is green train B incorrect, setpoint is 2450 volts for A4 C incorrect, this is the red train, P-34B is green train D correct, B6 undervoltage setpoint is 429 volts Technical Reference(s): __STM 1-32, Electrical Distribution, Rev 36___________
STM 1-05, Decay Heat Removal Sys, Rev 18 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None_______
Learning Objective: _________________________ (As available)
Question Source: Bank # _QID 0786__
Modified Bank # _______ (Note changes or attach parent)
New _______
Question History: Last NRC Exam ___2010______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 _____
55.43 _____
Comments:
Question: 32 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __1__ _____
K/A # 006K4.05 ____
Importance Rating _4.3__ _____
K/A: K4.05 Knowledge of ECCS design features and/or interlocks which provide for:
Autostart of HPI/LPI/SIP.
Proposed Question:
The following is a diagram of analog channels for ESAS:
I&C Technicians have been authorized to perform maintenance on Analog Channel 1, and the following occur over the next two minutes (no operator action):
- 2. K11-E7 ANALOG LOGIC BUFFER REMOVED alarms
- 3. Power is lost to Analog Channel 1
A. Only LPI initiated B. Only HPI initiated C. Both HPI and LPI initiated D. Neither HPI nor LPI initiated Proposed Answer: __A__
Explanation (Optional):
A. (Correct) Because the HPI logic buffer is removed, 2-of-2 coincidence exists. The valid signal from channel A power loss is blocked, and the only remaining signal for HPI is channel 2. LPI receives 2 signals, one from channel A due to power failure and one from channel B due to PT-2406 failing high B(Incorrect)
C(Incorrect)
D(Incorrect)
Technical Reference(s): STM 1-65 ______________________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination:
__1203.012T ACA for K11-E7 and Vendor dwgs 6600-M1Q-26-2, 38-3, 39-4___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___N/A______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __7_
55.43 _____
Comments:
Question: 33 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __1__ _____
K/A # __007 A1.01___
Importance Rating __2.9_ _____
K/A Statement: [007 Pressure Relief Tank/Quench Relief Tank (PRTS)] Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including: Maintaining quench tank water level within limits.
Proposed Question:
Quench Tank level is high The CRS directed transfer of water from the Quench Tank to the Clean Liquid Radwaste System per Section 13.0 of procedure 1103.005, Pressurizer Operations, and during this evolution the following occurs:
-K09-B4, Quench Tank Level HI/LO alarms
-Quench Tank level LIS-1051 reads 3800 gallons In reference to design limitations, which of the following operational problems should this present?
A. Inadequate NPSH for the transfer pump B. Insufficient cooling water during relief operations C. Inadequate water level for level indication D. Loss of a loop seal on the relief lines Proposed Answer: __B____
A is incorrect because the NPSH requirement is lower than the minimum water level C is incorrect because the transmitter reference leg is external to the tank D is incorrect because the relief valves relieve to a header that is aligned in the bottom of the tank Technical Reference(s): _1103.005 9 (Change 038) ______
(Attach if not previously provided) _1203.012H (Change 041)_______________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _________________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __5___
55.43 _____
Comments:
Question # _34__
Examination Outline Cross-
Reference:
Level RO SRO Tier # __2___ _____
Group # __1___ _____
K/A # __007 A2.06 _____
Importance Rating __2.6__ _____
K/A Statement: (007 Pressurizer Relief Tank/Quench Tank System ) Ability to (a) predict the impacts of the following malfunctions or operations on the PS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Bubble formation in the PZR.
Proposed Question:
Given the following conditions:
- Plant startup is in progress
- RCS pressure is 150 psig
- RCS temperature is 225 F
- Pressurizer level is 90 inches
- 1103.005, Pressurizer Operation, is being utilized to draw a steam bubble in the pressurizer A three minute blow through the ERV resulting in a Quench Tank pressure rise of __________
psig in conjunction with saturation conditions in the __________ is indication that a steam bubble has now formed in the pressurizer.
A. at least three; RCS B. less than or equal to one; pressurizer.
C. At least three; pressurizer D. less than or equal to one; RCS Proposed Answer: __B___
Explanation: IAW with 1103.005, Pressurizer Operations, a three minute blow through the ERV resulting in a Quench Tank pressure rise of 1 psig combined with saturation temperature/pressure conditions in the pressurizer is the indication to the operators that a bubble has now formed in the pressurizer.
Technical Reference(s): 1103.005, Pressurizer Operations (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: None Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X___
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 _(5)__
55.43 _____
Comments:
Question # _35__
Examination Outline Cross-
Reference:
Level RO SRO Tier # 2____ _____
Group # 1____ _____
K/A # 008 K4.01_________
Importance Rating 3.1___ _____
008 K4.01 Component Cooling Water System (CCWS). Knowledge of CCWS design feature(s) and/or interlock(s) which provide for the following: Automatic start of standby pump.
Proposed Question:
The Intermediate Cooling Water system is aligned as follows:
- P-33A is in service supplying the Non-Nuclear loop;
- P-33B is in service supplying the Nuclear loop;
- P-33C is in standby.
Which of the following correctly describes the operation of the Intermediate Cooling Water System in this alignment?
A. P-33C should start immediately when P-33B discharge pressure drops below 35 psig.
P-33B to P-33C Suction / Discharge Cross-Connect valves CV-2241 and CV-2239 should automatically shut.
B. P-33C should start when P-33B discharge pressure drops below 35 psig for >10 seconds. P-33B to P-33C Suction / Discharge Cross-Connect valves CV-2241 and CV-2239 should automatically shut.
C. P-33C should start immediately when P-33B discharge pressure drops below 35 psig.
P-33B to P-33C Suction / Discharge Cross-Connect valves CV-2241 and CV-2239 should remain open.
D. P-33C should start when P-33B discharge pressure drops below 35 psig for >10 seconds. P-33B to P-33C Suction / Discharge Cross-Connect valves CV-2241 and CV-2239 should remain open.
Proposed Answer: __B__
Explanation (Optional):
With P-33C as the standby pump, suction cross-connect valve CV-2241 and discharge cross-connect valve 2239 are open. With P-33B in service on either of the ICW loops and P-33B discharge pressure switch, PS-2231 reaches alarm setpoint of <35 psig for > 10 seconds then the standby pump P-33C will auto start. The associated loop suction and discharge cross-connect valves will close isolating P-33B from that loop. If supplying Nuclear ICW flow, then CV-2241 and CV-2239 close on P-33B low discharge pressure.
A) Incorrect. Discharge pressure must be <35 psig for >10 seconds to start the standby pump.
B) Correct.
C) Incorrect. CV-2241/-2239 will auto-shut, and discharge pressure must be <35 psig for
>10 seconds to start the standby pump.
D) Incorrect. CV-2241/-2239 will auto-shut.
Technical Reference(s): _STM 1-43 Rev. 13 ___________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___x____
Question History: Last NRC Exam ___no_________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __x__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __.7_
55.43 _____
Comments:
Question 36 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __1__ _____
K/A # _____010A1.08___
Importance Rating __3.2_ _____
A1. Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including: Spray nozzle differential temperature Proposed Question:
To prevent thermal stress across the pressurizer spray valve, Manual Bypass Valve RC-4 provides a flow path for bypass flow to limit spray valve differential temperature. RC-4 is operated to bypass flow across the A. Auxiliary Spray Line Tie-in B. Motor Isolation Valve CV-1009 C. Spray Flow Control Valve CV-1008 D. Manual Isolation Valve RC-3 Proposed Answer: ___C___
Explanation (Optional):
Answer C is the correct answer with the other distracters being plausible as they are connections to the pressurizer.
Technical Reference(s): _STM1-03, Reactor Coolant System, Revision 19, Page 16 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge ___X__
Comprehension or Analysis _______
10 CFR Part 55 Content: 55.41 __3__
55.43 _____
Comments:
Question: 37 Examination Outline Cross-
Reference:
Level RO Tier # __2__ _____
Group # __1__ _____
K/A # __012K6.04 ______
Importance Rating _3.3 _ _____
Knowledge of the effect of a loss or malfunction of the following will have on the RPS:
K6.04 Bypass-block circuits . .
Proposed Question:
The reactor is in the process of being started up with the following conditions:
- All four RCPs are running.
- The power range instruments indicate reactor power is approximately 12%.
- The high startup rate rod hold for Intermediate Range Detector IR 3 has failed to trip.
Given these conditions; A. A rod hold will occur if the Source Range Instruments sense a startup rate of 2 decades per minute.
B. A rod hold will occur if the Intermediate Range Instruments sense a startup rate of 3 decades per minute.
C. The high start up rate rod hold has been bypassed.
D. The high start up rate rod hold will inhibit rod withdrawal until IR 3 is repaired or bypassed.
Proposed Answer: ___C__
Explanation (Optional):
The RPS high startup rate hold interlock prevents rod withdrawal on a high startup rate. This block is bypassed if one of two conditions exist; if reactor power, as sensed by the PR detectors, is greater than 10% or if IR indicated power is greater than 1 x E-9 amps. Even though the IR high startup rate bistable failed to trip, since PR reactor power is greater than 10%, the high startup rate rod hold interlock has been bypassed making distractor C correct.
Distractors A and B would both be correct if PR power was less than 10%. Distractor D is incorrect because a single failure will not actuate the rod hold interlock.
Technical Reference(s): __STM 1-63, Reactor Protection System, Rev 9, page 26 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination:_None______________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___NA__
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __7__
55.43 _____
Comments:
Question 38 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __1__ _____
K/A # _013 K5.02_ ____
Importance Rating __2.9_ _____
K5.02 - Knowledge of the operational implications of the following concepts as they apply to the ESFAS: Safety system logic and reliability Proposed Question:
Reactor Building Pressure Transmitter (PT-2407) has failed high causing an ES CH3 Trip (analog 3) and the ESAS Partial Trip annunciator (K11-F6) to alarm.
With the above conditions, a power loss to which of the following would cause an ESAS actuation?
A. RS1 B. B71 C. RS3 D. B72 Proposed Answer: __A__
Explanation (Optional):
A is correct. A loss of power to RS1 will trip Analog Channel 1 which would then complete a 2 out of 3 analog trip causing an ESAS actuation. Loss of power to B71 or B72 does not result in a loss of ESAS functions. Analog 3 would be tripped as a result of a power loss to RS3 (Analog 3 is already tripped).
Technical Reference(s): __1105.003 (Change 016)____________
(Attach if not previously provided) __ _____________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _________________
Learning Objective: _________________________ (As available)
Question Source: Bank # ___X____
Modified Bank # _______ (Note changes or attach parent)
New ___ ___
Question History: Last NRC Exam ____2007________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis ____
10 CFR Part 55 Content: 55.41 __7___
55.43 _____
Comments:
Question 39 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __1__ _____
K/A # __K2.01__________
Importance Rating __2.9 _____
K/A 022K2.01: Knowledge of power supplies to the following: containment cooling fans.
Proposed Question:
Which of the following load centers supply power to the five (5) Reactor Building Ventilation Fans?
A. B5, B6 and B7 B. B3, B4 and B2 C. B3, B4 and B7 D. B5, B6 and B2 Proposed Answer: __A___
Explanation (Optional):
a) is correct per 1107 series procedures. (b), (c) and (d) list combination of wrong load centers.
Technical Reference(s):
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _________________
Learning Objective: _________________________ (As available)
Question Source: Bank # ___X___
Modified Bank # _______ (Note changes or attach parent)
New _______
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __5__ 55.43 _____
Question: 40 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __1__ _____
K/A # __022 K3.01___
Importance Rating __2.9_ _____
K/A Statement: [022 Containment Cooling System (CCS)] Knowledge of the effect that a loss or malfunction of the CCS will have on the following: Containment equipment subject to damage by high or low temperature, humidity and pressure Proposed Question:
Given:
Reactor has tripped.
RCS pressure 1200 psig Reactor Building pressure 20 psia What would be the consequences if the Reactor Building Cooler Chilled Water Bypass Dampers remained latched?
- a. RB cooling fan motors would overheat
- b. Excessive heat load on the Chilled Water System
- c. Damage to RB ventilation plenum from excessive pressure
- d. Inadequate cooling of the RB atmosphere Answer:
- d. Inadequate cooling of the RB atmosphere Explanation:
"a" is incorrect, the current on the motors, with subsequent overheating, is not a concern in this situation.
"b" is incorrect, the Chilled Water System is isolated on ESAS and therefore no additional heat load will be placed on it.
"c" is incorrect, RB ventilation plenum has been analyzed for these conditions and it will withstand the pressures after an ESAS.
"d" is the correct answer, the bypass dampers drop to allow more flow through the Service Water coils by bypassing the Chilled Water coils and thus more cooling to the RB atmosphere.
Technical Reference(s): __1104.033, pages 3, 4
__STM 1-09__
(Attach if not previously provided) _______________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # __274__ (Note changes or attach parent)
New _______
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __7___
55.43 _____
Comments:
Parent question:
What would be the consequences if the Reactor Building Cooler Chilled Water Bypass Dampers remained latched after an ESAS actuation?
- a. Inadequate air flow through the Service Water Cooling Coils
- b. Excessive heat load on the Chilled Water System
- c. Damage to RB ventilation plenum from excessive pressure
- d. Excessive current on the cooling fan motors Answer:
- a. Inadequate air flow through the Service Water Cooling Coils
Question # _41__
Examination Outline Cross-
Reference:
Level RO SRO Tier # __2___ _____
Group # __1___ _____
K/A # __026 K1.02 _____
Importance Rating __4.1__ _____
K/A Statement: (026 Containment Spray System) Knowledge of the physical connections and/or cause-effect relationship between the CSS and the following systems: Cooling water Proposed Question:
Which of the following are the correct cooling water source and bearing cooler for Reactor Building Spray Pump, P-35B?
A. CCW, E-47B B. SW, VUC-1C and 1D C. SW, E-47B D. CCW, E-47A Proposed Answer: __C__
Explanation:
A. Incorrect - CCW is the incorrect source, but E-47B is the correct cooler for P-35B.
B. Incorrect - Correct cooling water source, but VUC-1C and 1D are the room coolers for Decay Heat Vault B in which P-35B sits and not the bearing cooler.
C. Correct D. Incorrect - Incorrect cooling water source and incorrect cooler. Plausible because E-47A is the correct bearing cooler for P-35A Reactor Building Spray Pump.
Technical Reference(s): STM 1-08, Reactor Building Spray and Containment Building, Rev. 17 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: None Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X __
Comprehension or Analysis __ ___
10 CFR Part 55 Content: 55.41 _(7)__
55.43 _____
Comments:
Question 42 Examination Outline Cross-
Reference:
Level RO SRO Tier # 2_____ _____
Group # 1_____ _____
K/A # 039 A2.03__
Importance Rating 3.4_____ _____
039 Main and Reheat Steam System (MRSS). A2.03: Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Indications and alarms for main steam and area radiation monitors (during SGTR).
Proposed Question:
Given that a Steam Generator Tube Rupture has occurred on A OTSG, answer the following:
(1) Per EOP 1202.006, Tube Rupture, when is the crew required to start monitoring for the need to isolate the bad OTSG?
(2) If A OTSG is isolated, which radiation monitor should be used to monitor the activity of the affected OTSG?
A. (1) When RCS pressure is < 700 psig (2) A OTSG N-16 Detector (RI-2691)
B. (1) When RCS pressure is < 700 psig (2) Steam Line A High Range Rad Monitor (RI-2682)
C. (1) When RCS T-hot is < 490 F (2) A OTSG N-16 Detector (RI-2691)
D. (1) When RCS T-hot is < 490 F (2) Steam Line A High Range Rad Monitor (RI-2682)
Proposed Answer: __D__
Explanation (Optional):
A) Incorrect. RCS Press < 700 psig is the point where CFTs are isolated per step 57. The N-16 detectors are downstream of the MSIVs and will be isolated when OTSGs are isolated, and will therefore not be representative of OTSG activity.
B) Incorrect. RCS Press < 700 psig is the point where CFTs are isolated per step 57.
C) Incorrect. The N-16 detectors are downstream of the MSIVs and will be isolated when OTSGs are isolated, and will therefore not be representative of OTSG activity.
D) Correct. Step 54 states WHEN RCS T-hot is < 490F, THEN monitor for need to isolate bad SG as follows: . Steam Line High Range Rad Monitors are located upstream of the MSIVs and can be used to track OTSG activity even if the OTSG is isolated.
Technical Reference(s): _ EOP 1202.006 Rev 12, Tube Rupture__________
(Attach if not previously provided) _ STM 1-15 Rev.12, Main Steam_____________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ___None__________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___x___
Question History: Last NRC Exam ____no_____
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __x__
10 CFR Part 55 Content: 55.41 __.11_
55.43 _____
Comments:
Question 43 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __1__ _____
K/A # _____039K5.05 ___
Importance Rating __2.7_ _3.1_
K1. Knowledge of the operational implications of the following concepts as they apply to the main and reheat steam system: Basis for RCS cooldown limits Proposed Question:
Per the Technical Specification Bases, the PRIMARY reason for maintaining Technical Specification Limiting Condition for Operation cooldown limits for the reactor coolant system (RCS) from hot shutdown conditions using the main steam system IS:
A. To ensure RCS pressure/temperature limits remain within the design basis accident analysis B. To ensure RCS pressure/temperature limits remain within ASME code requirements C. To avoid pressurizer thermal shock and possible pressurizer failure D. To avoid conditions that might cause brittle fracture of the Reactor vessel Proposed Answer: ___D___
Explanation (Optional):
Answer D the correct answer.
Answer A is incorrect. The Tech Spec Bases state that the P/T limits are not derived from Design Basis Accident (DBA) analyses.
Answer B is incorrect. The Tech Spec bases only state that the ASME Code,Section XI, Appendix E (Ref. 10) provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
Answer C is incorrect. The Tech Spec bases specifically indicate that the limits do not apply to the pressurizer.
Technical Reference(s): _Tech Spec Bases, Amendment 215, Pages B 3.4.3-3 and B 3.4.3-4 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis ___X_
10 CFR Part 55 Content: 55.41 __5__
55.43 _____
Comments:
Although Tech Spec Bases questions are traditional an SRO only question, this particular K/A specifically refers to the basis for RCS cooldown, and has a rating factor of >2.5 (2.7) and is therefore applicable.
Question: 44 revision 1 Examination Outline Cross-
Reference:
Level RO Tier # __2__ _____
Group # __1__ _____
K/A # __059A1.03 ______
Importance Rating _2.7 _ _____
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW controls including: A1.03 Power level restrictions for operation of MFW pumps and valves. .
Proposed Question:
The reactor is at rated thermal power when one of the Reactor Coolant Pumps (RCPs) trips.
According to Procedure 1102.004, Power Operations, with only three RCPs in operation, the MAXIMUM allowed feedwater flow to each steam generator is __________ lbm/hr due to feedwater flow induced vibration concerns at the resultant reactor power level.
A. 1.2 x E6 B. 4.0 x E6 C. 5.4 x E6 D. 5.7 x E6 Proposed Answer: ___D__
Explanation (Optional):
All the distractors are valid feedwater flow values; A is the limit for transferring FW to auto with 3 RCPs running, B is the expected FW flow following the loss of RCP induced main turbine runback, and D is the correct limit contained in the referenced procedure for the reason given in the stem.
Technical Reference(s): __Procedure 1102.004, Power Operations, Rev 52, page 9 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _________________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___XX__
Question History: Last NRC Exam ___NA__
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __5__
55.43 _____
Comments:
Question: 45 Examination Outline Cross-
Reference:
Level RO SRO Tier # _2___ _____
Group # _1___ _____
K/A # _059 A3.06________
Importance Rating _3.2_ _____
A3.06 Main Feedwater (MFW) System: Ability to monitor automatic operation of the MFW, including: Feedwater isolation Proposed Question:
A Steam Line Break has occurred. The following conditions exist:
- The Reactor is tripped.
- A OTSG pressure = 620 psig.
- B OTSG pressure = 550 psig.
- All RCPs are running.
- ESAS Channel 3 and 4 have not actuated.
If all safety systems function as designed, what are the expected statuses of the OTSGs?
A. Both A and B OTSGs are being fed by Main Feed B. Both A and B OTSGs are being fed by Emergency Feed C. A OTSG is being fed by Main Feed, B OTSG is being fed by Emergency Feed D. A OTSG is being fed by Main Feed, B OTSG is not being fed Proposed Answer: __D___
Explanation (Optional):
Both the Main Steam Line Isolation signal and EFW actuation signal have a setpoint of 600 psig. Therefore, A OTSG has not received either signal and B OTSG has received both.
The Vector module logic looks at both OTSGs; if one OTSG is above 600 psig and the other OTSG is below 600 psig, EFW would be isolated to the bad OTSG, in this case, B.
Technical Reference(s): STM 1-66 Rev 11, Emergency Feedwater Initiation and Control_________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None_________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____No______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __x__
10 CFR Part 55 Content: 55.41 _.7__
55.43 _____
Comments:
Question: 46 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # 061K2.02 ____
Importance Rating _3.8__ _____
K/A: Emergency Feedwater: Knowledge of bus power supplies to AFW electric pumps.
Proposed Question:
Given:
- Both units have experienced a degraded power condition and the crew is performing 1202.007 (Degraded Power)
- All feedwater is lost
- EDG #1 will not start
- EDG #2 is running
- P-7A tripped and will not reset
- Adequate subcooling margin exists
- Overheating condition exists In accordance with 1202.007, what action should operators perform next to restore maximum heat removal?
A. Run two HPI pumps for HPI cooling per RT-4 B. Establish atmospheric dump pressure control on both SGs C. Cross-tie buses A3 and A4 D. Restore Instrument Air for ADV control Proposed Answer: __C__
Explanation (Optional):
A(Incorrect)
B(Incorrect)
C(Correct)
D(Incorrect)
Technical Reference(s): 1202.007______________________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___N/A______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __10_
55.43 _____
Comments:
Question: 47 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __1__ _____
K/A # __062 K3.02___
Importance Rating __4.1_ _____
K/A Statement: [062 AC Electrical Distribution System] Knowledge of the effect that a loss or malfunction of the ac distribution system will have on the following: ED/G Proposed Question:
Isolation of bus B51 for maintenance would remove AC power to which load associated with EDG 1?
A. Soak Back Pump (P-106A-1)
B. Oil Circulating Pump (P-106A-2)
C. Starting Air Compressor (C-4A1)
D. C-4A1 Aftercooler Blower/Dryer/Motorized Drain Valve Proposed Answer: ___B___
Explanation (Optional): Power supplies are indicated by breaker numbers in procedure 1104.036 (Change 060), Attachment C, DG1 Breaker and Switch Lineup. The P-106A-2 oil circulating pump is powered by bus B51. Soak Back pump P-106A-1 is a DC powered pump, supplied by D11 (A is incorrect). Starting Air Compressor C-4A1 is powered by bus B31 (C is incorrect). The C-4A1 drain valve is powered by bus Y4 (D is incorrect).
Technical Reference(s): ____STM 1-32-2 (Rev. 1); STM 1-32-3 (Rev. 1); ____
(Attach if not previously provided) __1104.036 (Change 060)____________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____E-15, sh 1__
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ____X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X___
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __7___
55.43 _____
Comments:
Question # _48__
Examination Outline Cross-
Reference:
Level RO SRO Tier # __2___ _____
Group # __1___ _____
K/A # __063 A4.01 _____
Importance Rating __2.8__ _____
K/A Statement: (063 D.C. Electrical Distribution) Ability to manually operate and/or monitor in the control room: Major breakers and control power fuses.
Proposed Question:
- An electrical fault has made the normal supply to D11 inoperable
- D11 should be transferred to its emergency supply Which of the following is a concern per procedure 1107.004, Battery and 125 VDC Distribution?
A. The discharge rate of the battery cannot be monitored in this configuration.
B. The static guard circuit is disabled and could cause inadvertent grounds on the bus.
C. A single fault may disable both trains of safety equipment such as EDGs and ES pumps.
D. It will prevent EDG#2 from starting, should it be required in an emergency situation.
Proposed Answer: __C__
Explanation:
AIncorrect BIncorrect CCorrect DIncorrect Technical Reference(s): 1107.004, Battery and 125V DC Distribution, Change 019 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: None Learning Objective: _________________________ (As available)
Question Source: Bank # ___QID 0529___
Modified Bank # _______ (Note changes or attach parent)
New ___ ____
Question History: Last NRC Exam ____2006_______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X __
Comprehension or Analysis __ ___
10 CFR Part 55 Content: 55.41 _(7)__
55.43 _____
Comments:
Question 49 Examination Outline Cross-
Reference:
Level RO SRO Tier # _2__ _____
Group # _1__ _____
K/A # _063 K3.01________
Importance Rating _3.7__ _____
063 D.C. Electrical Distribution, K3.01 Knowledge of the effect that a loss or malfunction of the DC electrical system will have on the following: ED/G Proposed Question:
Given:
- A Station Blackout condition exists.
- Battery Chargers are not loaded.
Answer the following questions:
(1) How is generator output voltage adjusted in this condition?
(2) What EDG protection is functional?
A. (1) Use the manual voltage adjuster on Exciter Control Panel.
(2) Mechanical Overspeed.
B. (1) Use the manual voltage adjuster on Exciter Control Panel.
(2) Local Emergency Stop Pushbutton.
C. (1) Adjust the setpoint for the automatic voltage adjuster.
(2) Mechanical Overspeed.
D. (1) Adjust the setpoint for the automatic voltage adjuster.
(2) Local Emergency Stop Pushbutton.
Proposed Answer: __A__
Explanation (Optional):
Per 1104.036 NOTE before step 13.10:
From Loss of DC power to field flashing circuit will cause loss of automatic voltage regulator setpoint adjustment (loss of power to motor-operated rheostat). Automatic voltage regulation should still function.
Step 13.10.1 then states:
IF voltage falls outside 3800-4500 volts, THEN perform the following:
A. Place DG1 Excitation control selector switch on (E11) in MANUAL.
B. Adjust manual voltage adjust dial to maintain ~ 4160 V.
Also, the CAUTION before step 13.8 states:
With loss of control power, the only functional DG protection is the mechanical overspeed device.
A) Correct.
B) Incorrect. Local Emergency Stop pushbutton energizes K-11. With no DC power available though, the Emergency Trip Relay will not trip the EDG if it is running.
C) Incorrect. With no DC power there is no power to motor-operated rheostat on automatic voltage adjuster.
D) Incorrect. See explanations B and C above.
Technical Reference(s): __ Emergency Diesel Generators STM-1-31 Rev. 12__
(Attach if not previously provided) 1104.036 Rev 60, EDG Operation (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None_________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # 2011 RO #48_ (Parent attached)
New _______
Question History: Last NRC Exam ___None_________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 _.7__
55.43 _____
Comments:
Question 50 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ ___2__
Group # __1__ ___1__
K/A # _____064K6.08 ___
Importance Rating __3.2_ __3.3_
K6. Knowledge of the effect of a loss or malfunction of the following will have on the EDG system: Fuel oil storage tanks Proposed Question:
Per the Technical Specifications, which of the following is the MAXIMUM fuel oil storage tank inventory which would still REQUIRE the associated emergency diesel generator to be declared INOPERABLE IMMEDIATELY?
A. 18,000 gallons B. 17,500 gallons C. 17,000 gallons D. 16,500 gallons Proposed Answer: ___C___
Explanation (Optional):
Answer C the correct answer.
Answer A is incorrect. Still operable for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Answer B is incorrect. Still operable for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Answer D is incorrect. Not the maximum value of the offered choices.
Technical Reference(s): _Tech Spec 3.8.3, Amendment 215 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis ____
10 CFR Part 55 Content: 55.41 __2__
55.43 _____
Comments:
Question: 51 Examination Outline Cross-
Reference:
Level RO Tier # __2__ _____
Group # __1__ _____
K/A # __073G2.1.27______
Importance Rating _3.9 _ _____
Process Radiation Monitoring (PRM) System - 2.1.27 Knowledge of system purpose and/or function.
Proposed Question:
One function of the Process Radiation Monitoring (PRM) System associated with monitoring the
_____________ is to automatically isolate the associated process flow in the event of a high radiation signal.
A. Intermediate Cooling Water System B. Steam Generators N-16 activity C. Main Condenser Air Discharge System D. Waste Gas System Proposed Answer: __D___
Explanation (Optional):
All the systems listed have process radiation monitors however only the Waste Gas System will automatically isolate on a high rad signal. The remainder will only alarm.
Technical Reference(s): _STM 1-62, Radiation Monitoring, Revision 12, Page 16 ff (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __11__
55.43 _____
Comments:
Question 52 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __1__ _____
K/A # __073 G2.4.45_____
Importance Rating __4.1_ _____
073 - Process Radiation Monitoring G2.4.45 - Emergency Procedures/Plan: Ability to prioritize and interpret the significance of each annunciator or alarm.
Proposed Question:
During plant operation at 100% power:
- Annunciator alarm K07-A5 ("A" OTSG N-16 TROUBLE) is received with RI-2691 reading 5.5 x 10³ cpm and stable
- Annunciator alarm K07-A6 ("B" OTSG N-16 TROUBLE) is received with RI-2692 reading 4 x 10³ cpm and slowly rising After operators have taken both RI-2691 and RI-2692 from GROSS to ANALYZER per 1203.012F, ANNUNCIATOR K07 CORRECTIVE ACTION, the following readings are observed:
- RI-2691 is reading 5.5 x 10³ cpm and stable
- RI-2692 is reading 3.5 x 10³ cpm and slowly rising Which of the following explains the observations above?
A. RI-2692 has a low voltage condition as confirmed by a drop in count rate following completion of 1203.012F.
B. RI-2691 is operating properly and a steam generator tube leak is confirmed as detected by a steady 5.5 x 10³ cpm reading.
C. RI-2691 has failed to 5.5 x 10³ cpm as confirmed by no change in count rate.
D. Both RI-2691 and RI-2692 are operating properly with a confirmed steam generator tube leak present in both steam generators with different leak rates.
Proposed Answer: __C__
Explanation (Optional):
A. Incorrect. A drop in count rate is expected after completion of 1203.012F where the rate meter mode is changed from Gross to Analyzer.
B. Incorrect. RI-2691 would read lower after completion of 1203.012F.
C. 1203.012F directs the rate meter mode to be changed from Gross to Analyzer if the reactor is critical and RI-2691/2692 in Alert or High alarm. This will screen out all activity except N-16 thus the reading would be lower after the change.
D. Incorrect. RI-2691 would read lower after completion of 1203.012F.
Technical Reference(s): ___1203.012F, Change 029_______________________
STM 1-62, Radiation Monitoring, Rev 12 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None__________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New __X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 _____
55.43 _____
Comments:
Question: 53 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ _____
Group # __1__ _____
K/A # 076G2.1.7 ____
Importance Rating _4.4__ _____
K/A: Service Water System: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
Proposed Question:
Given:
- ESAS actuation has occurred on channels 1 through 6
- HPI Pump P-36B is aligned to the A4 bus Which components' temperatures should be closely monitored until P-4B can be re-aligned?
A. Circ Water Pumps P-3A & P-3B, HPI pump P-36B B. Circ Water Pumps P-3C & P-3D, HPI pump P-36A C. Circ Water Pumps P-3A & P-3B, HPI pump P-36A D. Circ Water Pumps P-3C & P-3D, HPI pump P-36C Proposed Answer: __C__
Explanation (Optional):
"C" is correct, P-4A supplies cooling water to Loop I which serves the components listed.
"A" is incorrect, Loop I does not cool P-36B.
"B" is incorrect, Loop I does not cool Circ Water pumps P-3C & P-3D.
"D" is incorrect, this is a list of Loop II cooled components.
Technical Reference(s): 1203.030 STM 1-42 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # ___X___
Modified Bank # _______ (Note changes or attach parent)
New ___ ___
Question History: Last NRC Exam ___2011______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __7_
55.43 _____
Comments:
Question: 54 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __1__ _____
K/A # __078 A4.01___
Importance Rating __3.1_ _____
K/A Statement: [078 Instrument Air] Ability to manually operate and/or monitor in the control room: Pressure gauges Proposed Question:
Which of the following Instrument Air (IA) indications in the control room require the initiation of a reactor trip?
A. PMS Point P5409 (Unit 1 IA pressure) reads 58 psig B. INST AIR HEADER PRESS LO alarm comes in C. PMS Point P3013 (Unit 2 IA pressure) reads 40 psig D. INSTRUMENT AIR HEADER PRESS (PI-5409) reads 25 psig Proposed Answer: ___D___
Explanation (Optional): Per procedure 1203.024, Section 3, a reactor trip is required when Instrument Air header pressure is equal to or less than 35 psig. PMS Point P5409 does provide indication of IA header pressure in the control room, but the pressure provided is greater than 35 psig (A is incorrect). The receipt of the INST AIR HEADER PRESS LO alarm occurs when IA header pressure is less than 75 psig. By itself, it drives the operators to implement procedure 1203.024 (per procedure 1203.012K, K12-B3), but the indication itself does not require a reactor trip (B is incorrect). PMS Point P3013 is an indication of ANO Unit 2 IA header pressure. Since the Unit 1 and Unit 2 headers are normally cross-connected, it is plausible that this could be a valid indication. However, the pressure indicated is not equal to or below 35 psig (C is incorrect). PI-5409 reading less than 35 psig is the only indication which meets the criteria for a reactor trip. Therefore, D is correct.
Technical Reference(s): _STM 1-48 (Rev. 13), 1203.024 (Change 013)
(Attach if not previously provided) _1104.024 (Change 040), 1203.012K (Change 040)_
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _________________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ____X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __7___
55.43 _____
Comments:
PI-5409 can be seen on C19 in picture C19_050.
Question # _55__
Examination Outline Cross-
Reference:
Level RO SRO Tier # __2___ _____
Group # __1___ _____
K/A # __103 A2.03 _____
Importance Rating __3.5__ _____
K/A Statement: (103 Containment System) Ability to (a) predict the impacts of the following malfunctions or operations on the containment system - and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Phase A and B isolation Proposed Question:
A LOCA is in progress with the following conditions;
- RCS Pressure is 1090 psig and slowly lowering
- ESAS has actuated
- RB pressure is 17.8 psia and steady
- 1202.005, INADEQUATE CORE COOLING, has been entered Which of the following correctly describes the ESAS channels that have actuated AND the appropriate repetitive task to be performed?
A. Channels 7, 8, 9, 10 - RT 9, MAXIMIZE RB COOLING B. Channels 1, 2, 3, 4 - RT 2, INITIATE HPI C. Channels 7, 8, 9, 10 - RT 2, INITIATE HPI D. Channels 1, 2, 3, 4 - RT 9, MAXIMIZE RB COOLING Proposed Answer: __D__
Explanation:
E. Incorrect - Channels 7, 8, 9, 10 actuate on high RB pressure of 44.7 psia. Correct procedure F. Incorrect - Correct channels, wrong procedure. Even though Initiate HPI sounds like the logical thing to do in this situation, RT 3, Initiate Full HPI is re-performed per 1202.005 but not RT 2.
G. Incorrect - Wrong channels, wrong procedure.
H. Correct - Channels 1, 2, 3, 4 will trip on low RCS pressure less than 1590 psig. Per procedure 1202.005, the operators would verify RB cooling maximized via RT-9 since RB pressure is less than 18.7 psia.
Technical Reference(s): 1202.005, INADEQUATE CORE COOLING, Change 6. STM 1-65, ENGINEERED SAFEGUARDS ACTUATION SYSTEM, Rev. 5 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: None Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ___________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __ __
Comprehension or Analysis __ X___
10 CFR Part 55 Content: 55.41 _(7)__
55.43 _____
Comments:
Question 56 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2___ _____
Group # __2___ _____
K/A # __002 K4.07_____________
Importance Rating __3.1 _____
002 Reactor Coolant System (RCS) K4.07 Knowledge of RCS design feature(s) and/or interlock(s) which provide for the following: Contraction and expansion during heatup and cooldown.
Proposed Question:
When a steam bubble exists in the Pressurizer, an increase in RCS temperature causes a Pressurizer ____(1)____ to occur. The resultant pressure change is mitigated by ____(2)____.
A. (1) out-surge (2) pressurizer spray flow quenching a portion of the steam bubble.
B. (1) out-surge (2) water in the pressurizer flashing to steam.
C. (1) in-surge (2) water in the pressurizer flashing to steam.
D. (1) in-surge (2) pressurizer spray flow quenching a portion of the steam bubble.
Proposed Answer: __D___
Explanation (Optional):
If RCS temperature is raised, the change in coolant density causes an in-surge to the pressurizer. As pressurizer level increases the steam space volume becomes smaller causing RCS pressure to increase. The pressurizer spray system will compensate for the change in pressure by quenching a portion of the steam in the steam space.
A) Incorrect. An out-surge occurs when RCS temperature lowers.
B) Incorrect. Same as A), and water in pressurizer flashes to steam when pressure drops.
C) Incorrect. Water in pressurizer flashes to steam when pressure drops.
D) Correct.
Technical Reference(s): _Reactor Coolant System, STM1-03 Rev.19___________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None_________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam __No__________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 _.5__
55.43 _____
Comments:
Question 57 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ ___2__
Group # __2__ ___2__
K/A # _____016K3.07 ___
Importance Rating __3.6_ __3.7_
K6. Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: ECCS Proposed Question:
With the plant operating at 100% power, DC power is lost to non nuclear instrumentation channel X, and AC power is lost to non nuclear instrumentation channel Y.
Per Procedure 1203.047, Loss of NNI Power, which of the following actions is required to be taken?
A. Manually trip the reactor and maintain pressurizer level using a high pressure injection pump B. Place main feedwater, reactor demand, and diamond in HAND and operate the turbine bypass valves manually to control steam generator pressure C. Manually trip the reactor and close Reactor Coolant Pump Seal Injection Block Valve CV-1206 D. Place main feedwater, reactor demand, and diamond in HAND and operate pressurizer heater banks as necessary to maintain reactor coolant system pressure Proposed Answer: ___A___
Explanation (Optional):
Answer A the correct answer.
Answer B is incorrect. The operator places the equipment in HAND if only DC power lost to X and Y still has AC power.
Answer C is incorrect. CV-1206 is closed only if DC power IS available to X Answer D is incorrect. The operator places the equipment in HAND if only DC power lost to X and Y still has AC power.
Technical Reference(s): _1203.047, Loss of NI Power, pages 2, 6, and 7 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge ____
Comprehension or Analysis _X__
10 CFR Part 55 Content: 55.41 _10__
55.43 _____
Comments:
Question: 58 Examination Outline Cross-
Reference:
Level RO Tier # __2__ _____
Group # __2__ _____
K/A # __028K2.01 ______
Importance Rating _2.5 _ _____
K2 Knowledge of bus power supplies to the following: K2.01 Hydrogen recombiners . . .
Proposed Question:
An Auxiliary Operator (AO) has been sent to check the breaker position for Hydrogen Recombiner 55A.
The AO calls the control room and asks where he would find the breaker.
The control room should direct the AO to:
A. A3 B. B33 C. B53 D. Y3 Proposed Answer: ___C__
Explanation (Optional):
A. Incorrect - This is a safety related 4160 V ESF power supply for Division I and is credible if the applicant thinks the H2 recombiner is a 4160 load.
B. Incorrect -This is a 480 V non-safety MCC and is credible since it is a 480V power supply.
C. Correct - This is a Division I 480 V ESF MCC and it provides power to the A H2 Recombiner D. Incorrect - This is a 120 V Division I instrument bus and is credible if the candidate doesnt know the electrical nomenclature or what the power requirements are for the recombiner.
Technical Reference(s): __SD-1-2-32, Figure 32-2-11, Rev 1, and STM 1-09, (Attach if not previously provided) ___ page 17, Rev 12._ ________________________
(including version/revision number) ____________________________________________
Proposed references to be provided to applicants during examination: _None_________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New _XXX_
Question History: Last NRC Exam ____NA______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _XX_
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __7__
55.43 _____
Comments:
Question: 59 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __2__ _____
K/A # _035 A1.01________
Importance Rating _3.6__ _____
035 A1.01, Steam Generator System (S/GS): Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the S/GS controls including: S/G wide and narrow range level during startup, shutdown, and normal operations.
Proposed Question:
Given:
- The plant is operating at 80% power.
This malfunction would affect the output of A OTSG NNIX ____(1)___ level. This indicated level would read ____(2)____ than actual OTSG level.
(1) (2)
A. Startup Higher B. Operate Higher C. Startup Lower D. Operate Lower Proposed Answer: __D___
Explanation (Optional):
Of the three level ranges (Startup, Operate, and Full), only the Operate Level is temperature compensated. Startup and Full Level ranges would be unaffected by a failed OTSG Down comer Temperature instrument.
As water temperature decreases, density increases and specific volume decreases. Therefore for the same sensed water mass at a lower temperature, indicated level would be less than actual level.
Technical Reference(s): _STM 1-69 Rev 15, Non-Nuclear Instrumentation System.
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _None________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___x___
Question History: Last NRC Exam _None_______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __x___
10 CFR Part 55 Content: 55.41 _.7__
55.43 _____
Comments:
Question: 60 Examination Outline Cross-
Reference:
Level RO SRO Tier # _2___ _____
Group # _2___ _____
K/A # __045 A4.01______
Importance Rating _3.1__ _____
045 A4.01 Main Turbine Generator (MT/G) System: Ability to manually operate and/or monitor in the control room: Turbine valve indicators (throttle, governor, control, stop, intercept), alarms, and annunciators.
Proposed Question:
The crew is performing a turbine startup in accordance with procedure 1106.009, TURBINE STARTUP (WARMUP & ROLL).
The CBOT performs the following step to latch the turbine:
7.11.3 Latch by one of the following methods:
( ) At panel C01, depress LATCH pushbutton and hold until PB backlight turns on and UNIT TRIP light turns off.
What valves should be open when this step is complete?
A. All Governor Valves and Intercept Valves only B. All Intercept Valves and Reheat Stop Valves only C. All Reheat Stop Valves and Throttle Valves only D. All Throttle Valves and Governor Valves only Proposed Answer: __B__
Explanation (Optional):
When the LATCH pushbutton is depressed on panel C01, it causes all Intercept Valves and all Reheat Stop Valves to open, per step 7.11.6. The Throttle Valves and Governor Valves should remain closed, per step 7.11.7 of 1106.009.
A) Incorrect.
B) Correct.
C) Incorrect.
D) Incorrect.
Technical Reference(s): 1106.009 R-45, TURBINE STARTUP (WARMUP & ROLL)
(Attach if not previously provided) STM 1-24 Rev. 26 Turbine Controls And Auxiliaries
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam __No________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _X_
Comprehension or Analysis ____
10 CFR Part 55 Content: 55.41 .7___
55.43 _____
Comments:
Question: 61 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ _____
Group # __2__ _____
K/A # __055 G 2.1.20___
Importance Rating __4.6_ _____
K/A Statement: [055 Condenser Air Removal System (CARS)] Ability to interpret and execute procedure steps.
Proposed Question: During plant startup, the Condenser Vacuum System is being started up per procedure 1106.010, Vacuum System Operations. Preparations for both Condenser Vacuum Pumps C-5A and C-5B, including setting the position of the pump inlet valves, are in progress.
- C-5A is started, with C-5B in auto standby
- During the initial run of C-5A, the pumps amperage is 165 A
- The pumps air flow is 33 scfm
- Pump inlet valve V-5A is currently 2 turns open Per procedure 1106.010, what action should be taken?
A. Open V-5A an additional turn B. Fully open V-5A C. Close V-5A a part turn D. Fully close V-5A Proposed Answer: __C____
Explanation (Optional): Per STM 1-22, Section 2.6.5, the inlet valves to the Condenser Vacuum Pumps are throttled open to limit the amount of current drawn during initial high flow conditions.
The throttle position of the inlet valve is adjusted open during drawing of Main Condenser vacuum to maintain either air flow less than 30 scfm or pump amperage less than 150 A. Steps 7.10 and 7.11 of procedure 1106.010 address how to control the inlet valve position during startup.
In the given conditions, the system air flow is not within specification (less than 30 scfm), and the pump amperage is greater than the maximum rating (150 A). Therefore, the pumps inlet valve needs to throttled closed in order to reduce the pump amperage to less than 150 A.
Therefore, answer C is correct.
Answers A and B are plausible if the applicant believes the pumps air flow and amperage are within specification, or if he/she believes there are no specific controls on air flow/amperage. If that would be the case, the operator would continue to open the inlet valve until it is fully open.
However, opening the inlet valve partially or to full open in this case would cause the pump amperage to increase further above the maximum rating. Therefore, they are incorrect.
Answer D is plausible if the applicant believes that the amperage needs to be reduced as quickly as possible. However, procedure 1106.010, Step 7.2, aligns C-5A for start up with a minimum one-half turn open on valve V-5A. Closing the valve completely would cause the pump to lose all suction, and is contrary to procedure. Therefore, it is incorrect.
Technical Reference(s): _____STM 1-22 (Rev. 9);_______________
(Attach if not previously provided) _____1106.010 (Change 16)____________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __10___
55.43 _____
Comments:
Question # _62__
Examination Outline Cross-
Reference:
Level RO SRO Tier # __2___ _____
Group # __2___ _____
K/A # __071 K5.04 _____
Importance Rating __2.5__ _____
K/A Statement: (071 Waste Gas Disposal) Knowledge of the operational implication of the following concepts as they apply to the Waste Gas Disposal System: Relationship of hydrogen/oxygen concentrations to flammability.
Proposed Question:
The following conditions exist in the plant;
- Waste Gas System is aligned to compress WGDT T-18A
- H2/O2 Analyzer C119A is inoperable
- Waste Gas Surge Tank, T-17, H2 concentration has risen to 40%
- O2 concentration is 7%
To what piece of equipment should H2/O2 Analyzer C119 be aligned to?
A. Gas Collection Header B. Aux Building Vent Header C. Inservice WGDT T-18 D. Waste Gas Surge Tank T-17 Proposed Answer: __C__
Explanation:
A. Incorrect B. Incorrect C. Correct D. Incorrect Technical Reference(s): 1203.010, ABOVE NORMAL H202 CONCENTRATION, CHANGE 010 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: 1203.010 Attachment A Learning Objective: _________________________ (As available)
Question Source: Bank # ___0437____
Modified Bank # _______ (Note changes or attach parent)
New ___ ____
Question History: Last NRC Exam ____2002_______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __ X__
Comprehension or Analysis __ ___
10 CFR Part 55 Content: 55.41 _(5)__
55.43 _____
Comments:
Question 63 Examination Outline Cross-
Reference:
Level RO SRO Tier # _2____ _____
Group # _2____ _____
K/A # _072 A3.01________
Importance Rating _2.9___ _____
072 Area Radiation Monitoring (ARM) System, A3.01 Ability to monitor automatic operation of the ARM system, including: Changes in ventilation alignment.
Proposed Question:
Which of the following Area Radiation Monitors will cause a ventilation realignment when it alarms?
A. RE-8060, Reactor Building High Range Radiation Monitor B. RE-8001, Control Room Area Radiation Monitor C. RE-8009, Spent Fuel Pool Radiation Monitor D. RE-8017, Fuel Handling Area Radiation Monitor Proposed Answer: __B__
Explanation (Optional):
The Control Room Monitor is the only ARM that actually causes automatic system manipulation, in this case Control Room Isolation. The others, while required by Tech Specs, cause no automatic realignments.
Technical Reference(s): _ Radiation Monitoring STM 1-62 Rev. 12____________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ___None__________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___x____
Question History: Last NRC Exam ___No_________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __x__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 _.11____
55.43 _____
Comments:
Question 64 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ ___2__
Group # __2__ ___2__
K/A # _____078 K1.05 ___
Importance Rating __3.4_ __3.5_
K1. Knowledge of the physical connections and/or cause-effect relationships between the IAS and the following systems: MSIV air.
Proposed Question:
What is the design purpose of the backup air accumulators for the MSIVs?
A. To ensure air from the MSIV backup accumulators operate the Atmospheric Dump Valve for 30 minutes due to loss of Instrument air B. To ensure air from the MSIV backup accumulators will maintain Main Steam Isolation valve open for 30 minutes due to loss of Instrument air C. To ensure air from the MSIV backup accumulators to operate the Atmospheric Dump Valve for 60 minutes due to loss of Instrument air D. To ensure air from the MSIV backup accumulators will maintain Main Steam Isolation valve open for 60 minutes due to loss of Instrument air Proposed Answer: ___B___
Explanation (O ptional):
Answer A is incorrect. The purpose is to ensure the MSIVs have enough air to hold them open for 30 minutes not for ADV control this can be done locally with no Instrument air pressure although the accumulators supply air to the ADVs .
Answer B is correct. The purpose is to ensure the MSIVs have enough air to hold them open for 30 minutes. They are spring to close valves and air to open and maintain open.
Answer C is incorrect. Due to reasons above and the time is too long.
Answer D is incorrect Due to reasons above and the time is too long.
Technical Reference(s): _STM1-15 page 20 rev 12 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # ___
Modified Bank # _______ (Note changes or attach parent)
New ____X___
Question History: Last NRC Exam _________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge ____
Comprehension or Analysis _X__
10 CFR Part 55 Content: 55.41 _4__
55.43 _____
Comments:
Question: 65 Examination Outline Cross-
Reference:
Level RO Tier # __2__ _____
Group # __2__ _____
K/A # __086A2.04 ______
Importance Rating _3.3 _ _____
Ability to (a) predict the impacts of the following malfunctions or operations on the Fire Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: A2.04 Failure to actuate the FPS when required, resulting in fire damage.
Proposed Question:
The plant is at 100% power when the following annunciator alarms:
-K16-C2, TURB/MAIN GEN CO2 SYS TROUBLE Simultaneously, the Inside AO reports leaking oil has ignited on the Main Turbine #3 bearing housing.
The Inside AO also reports that there are NO INDICATING LIGHTS lit on the C-420 Turbine Cardox panel.
In ADDITION to tripping the main turbine and reactor, which of the following actions should be performed to limit damage from this fire?
A. Actuate the Cardox system via the applicable OPERATE switch on the C-463 panels B. Depress the TRIP button for bearing #2/3 to the right of C-420 C. Break glass and rotate the manual handle for bearing #2/3 to OPEN D. Depress the SPURT button for bearing #2/3 to the right of C-420 Proposed Answer: ___C__
Explanation (Optional):
Distractor A is incorrect because although the suppression systems can be manually actuated from the C-463 panels, the CO2 system can only be actuated locally.
Distractor B is incorrect because with no indicating lights lit on C-420 there is no power available for the trip button to function.
Distractor C is correct because the master pilot solenoid will fail open on loss of power, so the Inside AO only has to manually operate the solenoid for the affected zone.
Distractor D is incorrect because with no indicating lights lit on C-420 there is no power available for the spurt button to function.
Technical Reference(s): __ Procedure 1203.009 (Change 27).
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _None____________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam _____NA_____
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge ____
Comprehension or Analysis __X___
10 CFR Part 55 Content: 55.41 __5__
55.43 _____
Comments:
Question 66 Examination Outline Cross-
Reference:
Level RO SRO Tier # __3__ _____
Group # _____ _____
K/A # __G2.1.13_________
Importance Rating __2.5 _ _____
G2.1.13 - Knowledge of facility requirements for controlling vital / controlled access.
Proposed Question:
You are assigned as an escort for visiting contractors.
In accordance with OP-1000.019, Station Security Requirements, which of the following describes the MAXIMUM number of visitors that you may provide escort for at one time in a vital area, and the process required for entry into a Vital Area?
A. Three (3) visitors; visitors present badge to card reader first prior to entering a Vital Area.
B. Three (3) visitors; escort presents badge to card reader first prior to entering a Vital Area.
C. Five (5) visitors; visitors present badge to card reader first prior to entering a Vital Area.
D. Five (5) visitors; escort presents badge to card reader first prior to entering a Vital Area.
Proposed Answer: __C__
Explanation (Optional):
A. Incorrect. Up to 5 visitors may be escorted at one time. Second part is incorrect since visitor present badge to card reader first prior to entering Vital Area.
B. Incorrect. Up to 5 visitors may be escorted at one time. Second part is incorrect as escort presents badge to card reader last prior to entering Vital Area.
C. Correct. Up to 5 visitors may be escorted at one time, second part is correct as visitors present badges to card reader first prior to entering Vital Area.
D. Incorrect. Up to 5 visitors may be escorted at one time but escort presents badge to card reader last prior to entering Vital Area.
Technical Reference(s): __OP-1000.019, Station Security Requirements _____
__ASLP-RO-SEC, Rev 6__________________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None__________
Learning Objective: _________________________ (As available)
Question Source: Bank # ___X___
Modified Bank # _______ (Note changes or attach parent)
New _______
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 _____
55.43 _____
Comments:
Question: 67 Examination Outline Cross-
Reference:
Level RO SRO Tier # __3__ _____
Group # __ __ _____
K/A # G2.1.31 ____
Importance Rating _4.6__ _____
K/A: (Generic) Ability to locate control room switches, controls, and indications, and to determine they correctly reflect the desired plant lineup.
Proposed Question:
Given:
The crew has entered 1203.019, "High Activity in Reactor Coolant" Due to dose rates you are directed to isolate Letdown and swap RCP seal bleed-off from the Normal path to the Alternate path (Quench Tank.)
Which of the following combinations of indications are correct for the desired plant line-up?
A. CV-1221, "L/D Cooler Outlet," green light SV-1270, "Seal Bleedoff Alternate Path," red light CV-1270, "Seal Bleedoff Normal Path," green light B. CV-1221, "L/D Cooler Outlet," green light SV-1270, "Seal Bleedoff Alternate Path," green light CV-1270 ,"Seal Bleedoff Normal Path," red light C. CV-1221, "L/D Cooler Outlet," red light SV-1270, "Seal Bleedoff Alternate Path," red light CV-1270, "Seal Bleedoff Normal Path," green light D. CV-1221, "L/D Cooler Outlet," red light SV-1270, "Seal Bleedoff Alternate Path," green light CV-1270, "Seal Bleedoff Normal Path," red light.
Proposed Answer: __A__
Explanation (Optional):
A is correct in that it has the correct indications.
B is incorrect in that it has the incorrect indications.
C is incorrect in that it has the incorrect indications.
D is incorrect in that it has the incorrect indications.
Technical Reference(s): 1203.019, "High Activity in Reactor Coolant" (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # ___X___
Modified Bank # _______ (Note changes or attach parent)
New ___ ___
Question History: Last NRC Exam ___N/A______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __10_
55.43 _____
Comments:
Question: 68 Examination Outline Cross-
Reference:
Level RO SRO Tier # __3__ _____
Group # ____ _____
K/A # __G 2.1.38___
Importance Rating __3.7_ _____
K/A Statement: [Conduct of Operations] Knowledge of the station's requirements for verbal communications when implementing procedures.
Proposed Question:
Per procedure 1015.001, Conduct of Operations, who is the operations staff required to contact when any annunciator is removed from service for malfunctioning alarms?
A. Operations Work Liaison B. Design Engineering C. Engineering Programs D. System Engineering Proposed Answer: __D___
Explanation (Optional):
Per Section 10.0 of 1015.001, page 56 of 220, System Engineering (SYE) shall be contacted about any annunciator modified or removed from service for malfunctioning alarms. Therefore, D is the correct answer.
Answer A is incorrect because the Operations Work Liaison (OWL) is not required to be contacted per Section 10.0 of this procedure. The OWL is to be contacted for arranging unplanned non-emergency maintenance activities per Section 8.2.2 of the procedure.
Answer B is incorrect because the group isnt stated in Section 10.0. They are to be contacted to coordinate Category E valve position alignment checks per Attachment D.2, Section 2.0.
Answer C is incorrect because the group isnt stated in Section 10.0. They are supposed to be contacted when containment isolation valves are inoperable to perform reactor building leakage assessments per Sections 15.6 and 15.7 of the procedure.
Technical Reference(s): _____1015.001, Conduct of Operations (Change 090)__
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ___None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __10___
55.43 _____
Comments:
Question # _69__
Examination Outline Cross-
Reference:
Level RO SRO Tier # __3___ _____
Group # _____ _____
K/A # __G2.2.15 _____
Importance Rating __4.3__ _____
K/A Statement: (Equipment Control) Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.
Proposed Question:
It is necessary to manipulate a valve within the boundary of an existing tag-out. What is the requirement per the Configuration Control Program?
A. Manipulation of component must be less than one hour and must be returned to its normal position prior to next Operations shift turnover B. Manipulation of component must be less than one hour and must be returned to its normal position within twenty four hours of the manipulation C. Manipulation of component must be less than thirty minutes and must be returned to its normal position prior to next Operations shift turnover D. Manipulation of component must be less than thirty minutes and must be returned to its normal position within twenty four hours of the manipulation Proposed Answer: __A__
Explanation:
A. Correct B. Incorrect C. Incorrect D. Incorrect Technical Reference(s): 1015.001, CONDUCT OF OPERATIONS, CHANGE 090 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination:
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X ____
Question History: Last NRC Exam ___________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __ X__
Comprehension or Analysis __ ___
10 CFR Part 55 Content: 55.41 _(10)__
55.43 _____
Comments:
Question 70 Examination Outline Cross-
Reference:
Level RO SRO Tier # _3___ _____
Group # _____ _____
K/A # __G2.2.23________
Importance Rating _3.1_ _____
2.2.23 Equipment Control: Ability to track Technical Specification limiting conditions for operations.
Proposed Question:
A number of tools are available to track the status of equipment.
Some of these tools are:
- 1. LCO Tracking Record
- 2. eSOMS Configuration Control Record
- 3. Maintenance Configuration Tracking Log Sheet
- 4. Component Out of Position Log
- 5. Condition Report
- 6. Safety System Status Board The Surveillance Coordinator calls you, the Control Room Supervisor, and states that Hydrogen Recombiner, M-55B, surveillance frequency was exceeded.
Which of the above items should you use to document entry of the LCO?
A. 2, 3, 5 B. 1, 5, 6 C. 1, 3, 5 D. 2, 4, 5 Proposed Answer: B. 1, 5, 6 Explanation:
The only answer with all of the items required is answer "B".
Technical Reference(s): 1015.001 Chapter 8, Conduct of Operations: Plant/System Configuration Changes (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ___None_______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # __488___ (Note changes or attach parent)
New ___ ___
Question History: Last NRC Exam ____No________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __x__
10 CFR Part 55 Content: 55.41 .10__
55.43 _____
Comments:
Parent Exam Bank Question QID 488:
A number of tools are available to track the status of equipment.
Some of these tools are:
- 1. Inoperable Equipment Checklist
- 2. System Configuration Controls Tracking Log
- 3. Maintenance Configuration Tracking Log Sheet
- 4. Component Out of Position Log
- 5. Condition Report
- 6. Safety System Status Board
- 7. Control Room Log
- 8. Station Log The Surveillance Coordinator calls you, the Control Room Supervisor, and states that Hydrogen Recombiner, M-55B, surveillance frequency was exceeded.
Which of the above items will you use to document entry of the LCO?
A. 2, 3, 5, 7, & 8 B. 1, 2, 4, 6, & 8 C. 1, 5, 6, 7, & 8 D. 2, 4, 5, 6, & 8
Question 71 Examination Outline Cross-
Reference:
Level RO SRO Tier # __3__ _____
Group # _____ _____
K/A # _____G2.2.44 ___
Importance Rating __4.2_ __4.4_
G2. Equipment Control: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions affect plant and system conditions.
Proposed Question:
The following plant conditions exist:
-The reactor is tripped
-Adequate subcooling margin does NOT exist
-Reactor Coolant Pumps are tripped
-Emergency Feedwater is supplying both Steam Generators
-Reactor Coolant System Pressure is 2450 psig Per Procedure 1202.002, Loss of Subcooling Margin, which of the following actions should be performed FIRST?
A. Attempt to restore main feedwater flow to augment emergency feedwater flow to the steam generators B. Bump a reactor coolant pump in the loop with the lowest steam generator level C. Lower steam generator pressure until secondary saturated temperature is 20 - 40°F less than core exit thermocouple temperature D. Open the electromatic relief valve and lower pressurizer pressure to less than or equal to 1650 psig Proposed Answer: ___D___
Explanation (Optional):
Answer D the correct answer.
Answer A is incorrect. 1202.002 does not prescribe restoring main feedwater Answer B 1202.002 later directs bumping all RCPs and has nothing to do with SG level Answer C is incorrect. It is in the contingency action in 1202.002 but it requires 40 to 60°F temperature difference.
Technical Reference(s): _1202.002, Loss of Subcooling Margin, Revision 6, Page 7
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge ____
Comprehension or Analysis _X__
10 CFR Part 55 Content: 55.41 _10__
55.43 _____
Comments:
Question: 72 Examination Outline Cross-
Reference:
Level RO Tier # __3__ _____
Group # __ __ _____
K/A # __G2.3.15 ______
Importance Rating _2.9 _ _____
2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
Proposed Question:
At ANO there are several highly sensitive radiation detectors, such as the Main Steam N-16 Radiation Monitors, that use photomultiplier tubes as part of their radiation detection process.
Radiation monitors that use photomultiplier tubes are ____________ type detectors.
A. Scintillation B. Geiger-Mueller C. Ion Chamber D. Proportional Proposed Answer: ___A__
Explanation (Optional):
A is the correct answer because the N-16 radiation monitors use scintillation detectors and scintillation detectors use photomultipliers as part of the detection process.
B is incorrect because while many of the radiation monitors use Geiger-Mueller type detectors, the N-16 monitors do not. Also, Geiger-Mueller detectors do not use photomultipliers.
C is incorrect because none of the radiation monitors use an ion chamber type detector. Also, ion Chambers detectors do not use photomultipliers.
D is incorrect because none of the radiation monitors use a proportional type detector. Also, Proportional detectors do not use photomultipliers.
Technical Reference(s): _______STM 1-62, Radiation Monitors, Rev 12, page 25 _
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _________________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X___
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __12_
55.43 _____
Comments:
Question 73 Examination Outline Cross-
Reference:
Level RO SRO Tier # __3__ _____
Group # _____ _____
K/A # ____G2.3.7_______
Importance Rating __3.5_ _____
G2.3.7 - Ability to comply with radiation work permit requirements during normal or abnormal conditions.
Proposed Question:
In accordance with EN-RP-105, Radiological Work Permits, a Radiation Work Permit would identify An area, accessible to individuals, in which radiation levels external to the body could result in an individual receiving a deep dose equivalent greater than or equal to 1 Rem (10mSv) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm (~12 inches) from the radiation source or from any surface that the radiation penetrates. as a __________.
A. Radiation Area B. High Radiation Area C. Locked High Radiation Area D. Very High Radiation Area Proposed Answer: __C___
Explanation (Optional):
A. RA: in excess of 5 mR/hr at 30 cm from the radiation source B. HRA: in excess of 100 mR/hr at 30 cm from the radiation source C. LHRA: in excess of 1000 mR/hr at 30 cm but less than 500 Rads/hr at 1 m D. VHRA: in excess of 500 Rads/hr at 1 m from the radiation source Technical Reference(s): ___EN-RP-105, Radiological Work Permits____________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ___None_________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ____X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 _____
55.43 _____
Comments:
Question: 74 Examination Outline Cross-
Reference:
Level RO SRO Tier # __3__ _____
Group # __ __ _____
K/A # G2.4.14 ____
Importance Rating _3.8__ _____
K/A: (Generic) Knowledge of general guidelines for EOP usage.
Proposed Question:
Which of the following upsets in heat transfer has the HIGHEST priority per 1015.043, ANO-1 EOP/AOP Users Guide?
A. Overcooling B. Overheating C. Loss of Adequate SCM D. Steam Generator Tube Rupture Proposed Answer: __C__
Explanation (Optional):
A and B are the second and third highest priorities per 1015.043. D is the lowest priority per 1015.043.
Technical Reference(s): 1015.043 ANO-1 EOP/AOP Users Guide (Change 006)
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # ______
Modified Bank # _______ (Note changes or attach parent)
New ___ X___
Question History: Last NRC Exam ___N/A______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis ____
10 CFR Part 55 Content: 55.41 __10_
55.43 _____
Comments:
Question: 75 Examination Outline Cross-
Reference:
Level RO SRO Tier # __3__ _____
Group # ____ _____
K/A # __G 2.4.40___
Importance Rating __2.7_ _____
K/A Statement: [Emergency Procedures/Plans] Knowledge of SRO responsibilities in emergency plan implementation.
Proposed Question:
- ANO Unit 1 is operating in Mode 1
- The Unit 1 Shift Manager passes out, and medical assistance is requested
- A replacement Shift Manager has been contacted at home and will report onsite to take the watch within the hour During the time period before the replacement takes watch, per procedure 1903.010, Emergency Action level Classification, who has the FIRST responsibility of Emergency Direction and Control if an event were to occur at Unit 1?
A. The Unit 1 Shift Technical Advisor B. The Unit 2 Shift Manager C. The Unit 1 Control Room Supervisor D. The Technical Support Center Director Proposed Answer: __C__
Explanation (Optional): Per procedure 1903.010, Section 5.2, if the unit Shift Manager is not available to assume his/her Emergency Direction and Control responsibility, the unit Control Room Supervisor (CRS) will assume this responsibility until a replacement Shift Manager arrives. Therefore, answer C is correct.
Answer A is plausible because the person filling this position assists the Shift Manager in EAL classifications and notifications, and may be an individual who is licensed as a SRO. However, per procedure 1903.010, Section 5.2, this is incorrect.
Answer B is plausible because the unit Shift Managers both implement actions in procedure 1903.010 during events, so it may be perceived that they are an acceptable replacement while the incoming Shift Manager heads to the site. However, this is incorrect.
Answer D is plausible because this is one of the positions that is filled when the Emergency Response Organization (ERO) is activated, and this individual may assume Emergency Direction and Control responsibilities during the course of an event. However, this individual is not manning the TSC on a continuous basis to perform this role whenever an event occurs.
Therefore, the responsibility falls to the on shift CRS, as directed in procedure 1903.010.
Technical Reference(s): ___1903.010 (Change 044)__________________
(Attach if not previously provided) _____________________
(including version/revision number) ________
Proposed references to be provided to applicants during examination: _None_________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X___
10 CFR Part 55 Content: 55.41 __10___
55.43 _____
Comments:
Question # _76__
Examination Outline Cross-
Reference:
Level RO SRO Tier # _____ __1__
Group # _____ __1__
K/A # _____ 008 AA2.20 Importance Rating ____ _3.6__
K/A Statement: (008 Pressurizer Vapor Space Accident) Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident: The effect of an open PORV on code safety, based on observation of plant parameters.
Proposed Question:
Given the following:
- Reactor has tripped
- Pressurizer level is 285 inches and rising slowly
- CET temperatures 590F and rising
- RCS pressure is 135 psig and lowering slowly
- HPI/LPI have initiated
- BWST suction has been shifted to RB sump suction Subsequently the RCS has returned to Region 1 of Figure 4.
Which procedure would initially be entered with the given plant conditions and what is the correct transition?
A. 1202.002 (Loss of Subcooling Margin) with a transition to 1203.041 (Small break LOCA cooldown)
B. 1202.005 (Inadequate Core Cooling) with a transition to 1202.010 (ESAS)
C. 1202.002 (Loss of Subcooling Margin) with a transition to 1202.010 (ESAS)
D. 1202.005 (Inadequate Core Cooling) with a transition to 1203.041 (Small break LOCA cooldown)
Proposed Answer: __B__
Explanation:
A. Incorrect B. Correct - Stuck open PORV, facilitating a LOCA, with superheated conditions is entry condition to 1202.005. With RCS pressure less than 150 psig and the RCS returning to Region 1 of Figure 4, the correct transition is to 1202.010.
C. Incorrect D. Incorrect Technical Reference(s): 1202.005, INADEQUATE CORE COOLING, CHANGE 006
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination:
Learning Objective: _________________________ (As available)
Question Source: Bank # ______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ___________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __ __
Comprehension or Analysis __ X_
10 CFR Part 55 Content: 55.41 _____
55.43 _(5)__
Comments:
Question 77 Examination Outline Cross-
Reference:
Level RO SRO Tier # __ __ ___1__
Group # __ __ ___1__
K/A # _____015AA2.01___
Importance Rating __3.5______
AA2. Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): Cause of RCP failure. 10 CFR 43.5/45.13 Proposed Question:
The plant is at 90% with current identified RCS leakage of 1.0 gpm.
All other plant parameters are normal and as expected.
Subsequently the following occur:
- RCP VIBRATION HI (K08-B6) alarms.
- The plant computer shows a low oil reservoir for RCP P32B.
- P32B indicates motor bearing temperature is 168ºF.
- All other RCP motor bearing temperatures are normal.
As Control Room Supervisor you direct operators to Reactor Coolant Pump and Motor Emergency, 1203.31, Section _____________ and the correct action for plant conditions is A. 5, Motor/Bearing Trouble; initiate reactor building entry to add oil to the reservoir B. 2, Seal Failure; trip RCP 32B and verify proper ICS response C. 5, Motor/Bearing Trouble; trip RCP 32B and verify proper ICS response D. 2, Seal Failure; trip RCP 32B and perform immediate actions for reactor trip Proposed Answer: ___A___
Explanation (Optional):
A. Correct. Entry conditions for Section 5 are met (Annunciator RCP VIBRATION HI (K08-B6), and per step 2, 2. IF RCP motor bearing temperature exceeds 190°F (167°F for P32B) AND associated RCP oil reservoir is low, THEN initiate reactor building entry to add oil to reservoir.
B. Incorrect. Entry conditions are not met for Section 2. Plausible because one of the required conditions for entry to section 2 is 10 gpm AND a change in seal cavity pressure behavior. But RCS leakage is 1.0 gpm in the given conditions, TS 3.4.13 allows 10 gpm of identified leakage and the initial condition of all other plant parameters are normal means that section 2 entry conditions are not met. In addition, at 95% power
if a RCP were to be tripped iaw section 2, the reactor likely would not trip. Therefore, part 2 of this answer is also plausible.
C. Incorrect. Plausible because Section 5 is correct. However, trip RCP 32B and verify proper ICS response is from section 2. The candidate would have to know that for a high vibration alarm additional criteria must be met to trip the RCP.
D. Incorrect. Section 2 is not correct (see answer B explanation). Also, Note in section 3 states that if 1 RCP were tripped from 92% power, then a reactor trip may trip on high power/flow/imbalance. If the candidate does not know this note then tripping from 91%
power is plausible. Therefore, the actions for reactor trip will seem plausible.
Technical Reference(s): _Technical Specification 3.4.13 RCS Operational Leakage; Reactor Coolant Pump and Motor Emergency, 1203.31 Change 22, sections 2,3, and 5; Lesson_ASLP-RO-PROBM R3, Enabling Objectives 1,4,5, and 8 Proposed references to be provided to applicants during examination: _____None______
Learning Objective:_Lesson ASLP-RO-PROBM R3, Enabling Objectives 1,4,5, and 8; Lesson A1LP-RO-RCS_R13, Terminal Objective TERMINAL OBJECTIVE: When given a written examination,Trainee's will, from memory, identify or describe the Design Bases, Purpose and Operation of the RCS and it's component parts, instruments and connecting systems with 80% accuracy.________________________
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis ___X_
10 CFR Part 55 Content: 55.41 __ __
55.43 __x__
Comments:
Question 78 Examination Outline Cross-
Reference:
Level RO SRO Tier # __3__ __3__
Group # _____ _____
K/A # _____025AG2.4.50___
Importance Rating __4.2_ __4.0_
025A2.4. Loss of Residual Heat Removal System Emergency Procedures: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.
Proposed Question:
The plant is in Mode 4 following a normal reactor shutdown from full power with the following conditions:
-Decay Heat Removal (DHR) Pump P-34A is in service
-P-34 Suction Temperature on TI 1404 is 283 degrees F
-Annunciator K09-C8 DH PUMP A/B SUCT TEMP HI is in alarm
-Service Water is in service and operating normally
-There is NO leakage from the Decay Heat System Per Procedure 1203.012H, Annunciator K09 Corrective Action, which of the following actions is REQUIRED to be taken?
A. Place DHR Pump P-34B in service and secure DHR Pump P-34A B. Throttle open DHR Supply to Makeup Suction Valve CV-1276 C. Throttle SW to Decay Heat Cooler SW-22A to scribed T position.
D. Throttle closed E-35A Cooler Bypass Valve CV-1433 Proposed Answer: ___D___
Explanation (Optional):
Answer D the correct answer.
Answer A is incorrect. Not part of 1203.012H Answer B is incorrect. Not part of 1203.012H Answer C is incorrect. Not part of 1203.012H Technical Reference(s): _1203.012H, Annunciator K09 Corrective Action, Revision 41, Pages 59-60 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge ____
Comprehension or Analysis _X__
10 CFR Part 55 Content: 55.41 _____
55.43 ___5_
Comments:
Question: 79 Examination Outline Cross-
Reference:
Level SRO Tier # __1__ _____
Group # __1 __ _____
K/A # __054AG2.4.35____
Importance Rating _4.0 _ _____
Loss of Main Feedwater (MFW): 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.
Proposed Question:
Reactor power is 15% and Main Feedwater Pump (MFWP) P-1A is out of service while the thrust bearing wear detector is being replaced. Then the following sequence of events takes place:
- The MFW P-1B Main Oil Pump trips and both standby oil pumps auto start
- Both MFW P-1B standby oil pumps trip about a minute later
- The B MFP TURBINE TRIP (K07-A8) alarm annunciates
- The control room operators report MFW P-1B has failed to trip and they are unable to trip it from C02.
- Indicated total feed flow is zero.
- S/G levels are now 30 and lowering.
- The Inside Auxiliary Operator (IAO) is directed to locally trip MFWP 1B.
Based on this sequence of events and assuming all the other plant equipment operated as designed, the CRS should:
A. Immediately trip the reactor and implement EOP 1202.001, Reactor Trip.
B. Trip the reactor when S/G level reaches 15 and enter EOP 1202.001, Reactor Trip.
C. Manually actuate EFW per 1203.027, Loss of S/G Feed, and stabilize S/G level.
D. Enter Procedure 1203.027, Loss of S/G Feed, when the IAO locally trips MFW P 1B Proposed Answer: __A____
Explanation (Optional):
Note: Because feedpump 1B is not tripped, a reactor trip will not occur until the IAO trips the pump locally even though both pumps are functionally failed.
A. Per AOP 1203.027, if both feedpumps are lost and power is greater than 7%,
immediately trip the reactor and enter EOP 1202.001.
B. This is also contained in AOP 1203.027, however on the loss of both feedpumps the reactor should be tripped immediately.
C. This step is also in AOP 1203, however the reactor should be tripped immediately and the AOP is exited before reaching this step.
D. When the IAO locally trips the feedpump, the reactor will automatically trip and the crew would then enter the EOP. However, with both feedpumps functionally failed, the reactor should be tripped immediately.
Technical Reference(s): _AOP 1203.027, Loss of SG Feed, Rev 14_____________
(Attach if not previously provided) _EOP 1202.001, Reactor Trip, Rev 32 _______________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ___None__________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____NA_
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __ _
55.43 ___5_
Comments:
Question 80 Examination Outline Cross-
Reference:
Level RO SRO Tier # _____ __1__
Group # _____ __1__
K/A # _058 AG 2.2.40_____
Importance Rating _____ __4.7_
058 GENERIC 2.2.40, Loss of DC Power: Ability to apply Technical Specifications for a system.
Proposed Question:
[Provide Tech Specs 3.8.6 and 3.8.4 as references]
Given:
- The unit is operating at 100% power.
- At 0900 on February 18, 2013, it was discovered that one connected cells float voltage on D06 was determined to be 2.02 V.
- At 1000 on February 18, 2013, it was discovered that one connected cells float voltage on D07 was determined to be 2.07 V.
Based on the above conditions, Technical Specifications require the unit to be in MODE 3 no later than what time?
A. 0500 on February 20, 2013 B. 0500 on February 19, 2013 C. 1700 on February 18, 2013 D. 1600 on February 18, 2013 Proposed Answer: __C___
Explanation (Optional):
SR 3.8.6.1 is to Verify battery cell parameters meet Table 3.8.6-1 Category A limits.
Table 3.8.6-1 Category A has a limit of >= 2.13 V, which both batteries violate. Further, Category C has a limit for connected battery cells > 2.07 V. Both D06 and D07 therefore violate this limit as well.
At 0900 the crew enters LCO 3.8.6.B.1, and declares D06 INOP. This also causes an entry into LCO 3.8.4.A.1 for one DC train INOP, which allows 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to restore D06 to operable status if no other failures were to occur.
At 1000 the crew separately enters LCO 3.8.6.B.1 for D07 as allowed by the Action Note, and declares D07 INOP. Tech Spec 3.8.4 has no LCO for multiple DC trains Inoperable, therefore at 1000 the crew also enters LCO 3.0.3, which requires the unit to be in MODE 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
The unit must be in MODE 3 no later than 1700 with both DC electrical power subsystems inoperable.
A) Incorrect. An applicant would select this if they incorrectly allowed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> between entry into LCO 3.8.6.A and entry into LCO 3.8.6.B, and incorrectly evaluated D07 as operable (D07 must be GREATER THAN 2.07V to meet Category C limits).
B) Incorrect. An applicant would select this if they did not recognize that an LCO 3.0.3 entry was required due to 2 DC trains INOP.
C) Correct.
D) Incorrect. An applicant would select this if they incorrectly calculated the LCO 3.0.3 entry time starting at 0900 instead of 1000.
Technical Reference(s): _ANO1 Tech Specs 1.3, 3.0.3, 3.8.4, & 3.8.6._______
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: Tech Specs 3.8.4 & 3.8.6 Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam __None______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 _____
55.43 .2___
Comments:
Question: 81 Examination Outline Cross-
Reference:
Level RO SRO Tier # ____ __1__
Group # ____ __1__
K/A # BE04EA2.2 ____
Importance Rating _ __4.4 K/A: Inadequate Heat Transfer - Ability to determine and interpret the following as they apply to the (Inadequate Heat Transfer): Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.
Given:
- Reactor tripped and the crew is operating in Loss of Subcooling Margin (1202.002).
- P-7A tripped.
- P-7B is out for maintenance.
- CET temperatures are 610°F and rising.
- Offsite power is subsequently lost.
- Both DGs are supplying their emergency buses.
- RCS pressure is 1800 psig.
Based on the above current conditions, the correct procedure transition is to _______.
A. Remain in Loss of Subcooling Margin (1202.002)
B. Overheating (1202.004)
C. Inadequate Core Cooling (1202.005)
D. Degraded Power (1202.007)
Proposed Answer: __D_
Explanation (Optional):
A. (Incorrect) Even though loss of SCM is the highest priority, the lack of offsite power requires entry into 1202.007 which has steps for loss of SCM.
B. (Incorrect) Even though CETs are rising above 610, with all feedwater is lost, 1202.007 has steps for mitigating an overheating condition.
C. (Incorrect) Conditions given could cause entry into ICC if actions are not taken but entry into the ICC EOP is not required yet.
D. (Correct) This is the correct procedure since offsite power is lost.
Technical Reference(s): 1202.004 ______________________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___N/A______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __10_
55.43 _____
Comments:
Question: 82 Examination Outline Cross-
Reference:
Level RO SRO Tier # ____ __1___
Group # ____ __2___
K/A # __028 G 2.4.47___
Importance Rating ___ _4.2___
K/A Statement: [028 Pressurizer (PZR) Level Control Malfunction] Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
Proposed Question:
On March 3, 2013, the plant is preparing for startup following refueling.
RCS temperature is 525°F.
RCS pressure is 2170 psig and slowly rising.
Preparations are being made to withdraw control rods.
At 1300, the following conditions occur:
- PZR LEVEL HI/HI alarm is in
- LT-1001 (controlling channel) reads 280, increasing at 2/minute With no operator action taken, what Technical Specification LCO action would be required using the MAXIMUM amount of time allowed to restore the Pressurizer to operable status?
A. Restore pressurizer level within limits by 1400 B. Place the plant in Mode 3 by 1920 C. Place the plant in Mode 3 by 2000 D. Restore pressurizer level within limits by 1420 Proposed Answer: ___D___
Explanation (Optional):
Per Technical Specification 3.4.9, Condition A, when the pressurizer level is equal to or exceeds its limit (320), action is required to restore pressurizer level to less than the limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
At the rate of level increase stated (2/minute), with no operator action, level would reach 320 in 20 minutes. Therefore the LCO action statement would be in effect at 1320, meaning that the plant has until 1420 to address restoring pressurizer level (application of LCO 3.0.2). Therefore, answer D is correct.
Answer A is plausible because of the rising pressurizer level. However, the Tech Spec pressurizer level hasnt been exceeded yet, so the time requirement is incorrect.
If actions to restore the pressurizer level to within specification were unsuccessful, there is a requirement to 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 />. However, the Tech Specs require the efforts to restore pressurizer level first. In addition, the time stated would be incorrect (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS requirement completed by 1420, if not, Mode 3 by 2020). Therefore, answer B is incorrect.
Answer C is plausible because there was a failure associated with the pressurizer heaters given in the stem. However, this results in the loss of HUB 13 heaters, which are not required to meet Tech Spec 3.4.9 requirements (Banks 1, 2, and HUB 14). Therefore, it is incorrect.
With the statement in the stem with no operator action, the intent is to keep the applicant from assuming that a reactor trip would be manually initiated at 290 pressurizer level per procedure 1203.012H, Alarm K09-B3 response.
Technical Reference(s): ___STM 1-03 (Rev. 19); Tech Spec 3.4.9, _
(Attach if not previously provided) ___ K09 1203.012H _________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _________________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis ___X__
10 CFR Part 55 Content: 55.41 __10___
55.43 ___5__
Comments:
Question # _83__
Examination Outline Cross-
Reference:
Level RO SRO Tier # ______ __1__
Group # ______ __2__
K/A # ______ 037 AA2.01 Importance Rating ______ _3.4__
K/A Statement: (037 Steam Generator (S/G) Tube Leak) Ability to operate and/or monitor the following as they apply to the Steam Generator Tube Leak: Unusual readings of the monitors; steps needed to verify readings.
Proposed Question:
Given the following:
- Reactor is at 100% power
- Pressurizer level has reached 205 inches and is now dropping slowly
- 1202.006, TUBE RUPTURE, has been entered Which of the following indications CONCLUSIVELY shows on which S/G the tube rupture exists and what is the required next step with the given plant conditions?
A. Steam Line High Range RAD Monitors, Begin controlled shutdown at > 5%/minute B. Plant Monitoring System N16T screen, Trip the reactor C. Steam Line High Range RAD Monitors, Trip the reactor D. Plant Monitoring System N16T screen, Begin controlled shutdown at > 5%/minute Proposed Answer: __D__
Explanation:
A. Incorrect - Procedure states that due to shielding between MS lines this reading may be inconclusive. Second part is correct with pressurizer level greater than 200 inches, a controlled shutdown is begun.
B. Incorrect C. Incorrect D. Correct - Plant Monitoring System N16Tscreen has many diverse indications that can be used to identify the S/G with the leak. This is correct with the given plant conditions.
Technical Reference(s): 1202.006, TUBE RUPTURE, CHANGE 012 (Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination:
Learning Objective: _________________________ (As available)
Question Source: Bank # ______
Modified Bank # _0694__ (Note changes or attach parent)
New _______
Question History: Last NRC Exam ___________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __ __
Comprehension or Analysis __ X_
10 CFR Part 55 Content: 55.41 _____
55.43 _(5)__
Comments:
Question 84 Examination Outline Cross-
Reference:
Level RO SRO Tier # __ __ ___1__
Group # __ __ ___2__
K/A # _____BE03EG2.4.6___
Importance Rating __4.7______
G.2.4.6. Knowledge of EOP mitigation strategies. 10 CFR 41.10/43.5/45.13 Proposed Question:
A reactor trip and ESAS actuation occurred 60 seconds ago. Containment temperature and pressure are at their alarm set points and rising. The reactor operator announces Loss of Subcooling Margin.
As Control Room Supervisor, you direct the crew to __________ RCPs per 1202.002, Loss of Subcooling Margin, step 1. The reason for this is ____________________.
A. Trip RCPs; to prevent damage to the pump motors when the RCS flashes to steam B. Leave RCPs running; to prevent uncovering the core C. Trip RCPs; to prevent the void fraction from exceeding design limits.
D. Leave RCPs running; to maintain forced flow through the core for decay heat removal Proposed Answer: ___C___
Explanation (Optional):
A. Incorrect. Plausible because trip RCPs is correct but the reason is incorrect.
B. Incorrect. RCPs should be tripped. Plausible because the procedure does have a note that says if RCPs are not tripped within 2 minutes then leave them running to prevent uncovering the core. The stem clearly states that time elapsed is 60 seconds or less.
C. Correct. From GEOP Vol 2. III.B.2, The RCPs are tripped immediately (within 2 minutes) to prevent possible core damage. If the RCPs are tripped when void fraction is greater than 70% the peak cladding temperature from exceeding the maximum temperature allowed by 10CFR50.46.
D. Incorrect. RCPs should be tripped. Plausible because decay heat needs to be removed from the core.
Technical Reference(s): _GEOG Vol 2. III.B.2 Loss of Subcooling Margin, 12002.002 change 6.
Proposed references to be provided to applicants during examination: _____None______
Learning Objective:_Lesson Plan A1LP-RO-EOP2 Enabling Objectives B and I.___________
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis ___X_
10 CFR Part 55 Content: 55.41 __ __
55.43 __x__
Comments:
Question 85 Examination Outline Cross-
Reference:
Level RO SRO Tier # __1__ __1__
Group # __2__ __2__
K/A # _____BE13EA2.1___
Importance Rating __3.4_ __4.0_
BE13. EOP Rules: Facility conditions and selection of appropriate procedures during abnormal and emergency Proposed Question:
The following plant conditions exist:
- Reactor tripped due to a loss of both MFWPs approximately 15 minutes ago.
- Annunciator K02-B6 "A3 L.O. RELAY TRIP" is in alarm.
- AFW pump, P-75, is tagged out for maintenance.
- Steam Driven EFW Pump, P-7A, has tripped on overspeed.
- RCS pressure is 1800 psig.
- CETs are 600°F.
- Both OTSG levels are 30".
Which of the following procedures should be in use for the above conditions?
A. 1202.002, Loss of Subcooling Margin B. 1202.004, Overheating C. 1202.011, HPI Cooldown D. 1203.037, Abnormal ES Bus Voltage Proposed Answer: ___A___
Explanation (Optional):
Answer A the correct answer.
Answer B is incorrect. CETs are <610 Answer C is incorrect. 1202.002 Exit criteria not met Answer D is incorrect. This procedure is used when ES bus voltage is low but not de-energized Technical Reference(s): _1202.002, Loss of Subcooling Margin, Revision 6, Pages 1-4 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # ___X___ (Note changes or attach parent)
New ______
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge ____
Comprehension or Analysis _X__
10 CFR Part 55 Content: 55.41 _____
55.43 ___5_
Comments:
Modified QID 0588. Changed CET temperature from 612 degrees F to 600 degrees F. This changes the correct answer from B to A
Question: 86 Examination Outline Cross-
Reference:
Level SRO Tier # __2__ _____
Group # __1 __ _____
K/A # __006G2.2.42 ____
Importance Rating _4.6 _ _____
006 Emergency Core Cooling System (ECCS): 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
Proposed Question:
Following an outage, the reactor has been started up and is now operating at rated thermal power.
A system engineer calls the control room and informs the crew three Emergency Core Cooling System (ECCS) valves in one ECCS train were not tested during the outage as required by Technical Specification (TS) Surveillance Requirement 3.5.2.3 (provided).
The three valves were last tested 20 months ago and one of the valves is inaccessible with the reactor at power.
A mid-cycle outage is scheduled to occur in 12 months.
The control room crew:
A. Is required to immediately enter TS 3.5.2.A.
B. Is required to enter TS 3.5.2.A when the 25% grace period expires.
C. Can delay entering TS 3.5.2.A for a maximum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the time of discovery if a risk assessment is performed and the risk is managed. Entry is then required.
D. Can delay entering TS 3.5.2.A until the mid-cycle outage if a risk assessment is performed and the risk is managed.
Proposed Answer: __D___
Explanation (Optional):
Per TS 3.0.3, if a required surveillance is missed, entry into the applicable TS can be delayed for an additional (in this case) 18 months from the original due date if a risk assessment is done and the risk is managed. Because the mid-cycle outage is scheduled to fall within this window, distractor D is correct. Based on this discussion, distractors A, B, and C are all incorrect because entry is not required for any of the conditions listed.
Technical Reference(s): __TS 3.5.2, SR 3.0.1, SR 3.0.2, and TS 3.0.3__________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _TS 3.5.2 ________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___NA__
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 __ _
55.43 ___2_
Comments:
Question 87 Examination Outline Cross-
Reference:
Level RO SRO Tier # _____ __2__
Group # _____ __1__
K/A # __007 A2.03_______
Importance Rating ______ _3.6__
007 - Pressurizer Relief / Quench Tank A2.03 - Ability to (a) predict the impacts of the following malfunctions or operations on the PS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Over pressurization of the PZR Proposed Question:
Given the following:
- RCS pressure is 2050 psig and lowering
- All Pressurizer heaters are on
- Pressurizer level is 235 in and rising
- Quench Tank (T-42) level is rising
- Annunciator alarm RELIEF VALVE OPEN (K09-A1)
- There is a rise on acoustic Relief Valve Monitor (VYI-1000A)
Which of the following identifies the correct action and procedure required for these conditions?
A. Take manual control of Pressurizer Level Control Valve (CV-1235) and open per OP-1203.026, Loss of Reactor Makeup.
B. Close the Spray Line Isolation Valve (CV-1009) per OP-1203.015, Pressurizer Systems Failure, Section 6, Pressurizer Spray Valve (CV-1008) Failure.
C. Trip the reactor and refer to EOP series (1202.XXX) per OP-1203.015, Pressurizer Systems Failure, Section 2, Leaking Pressurizer Code Safety Valve.
D. Close the ERV Isolation Valve (CV-1000) per OP-1203.015, Pressurizer Systems Failure, Section 1, Electromatic Relief Valve (PSV-1000) System Failure or Leak.
Proposed Answer: __D___
Explanation (Optional):
A. A pressurizer steam space leak will cause level to swell. Conditions do not require opening the level control valve.
B. This would be correct if the spray valve were leaking, but its not per the given conditions.
C. Per Section 2 of OP-1203.015, a reactor trip is only required IF code safety valve leakage exceeds capability to maintain RCS pressure. For the given conditions, a code safety is not leaking.
D. Given conditions indicate a stuck open ERV. Therefore, it is correct to isolate the ERV using CV-1000 per OP-1203.015, Section 1.
Technical Reference(s): __OP-1203.015, Pressurizer Systems Failure, Section 1__
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ___None_______
Learning Objective: _________________________ (As available)
Question Source: Bank # _QID 842 (AN1)_
Modified Bank # _______ (Note changes or attach parent)
New _______
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X___
10 CFR Part 55 Content: 55.41 _____
55.43 _____
Comments:
Question: 88 Examination Outline Cross-
Reference:
Level RO SRO Tier # ____ __2__
Group # ____ __1__
K/A # 061A2.08 ____
Importance Rating _ __3.8 K/A: Emergency Feedwater - Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Pump failure or improper operation Given:
- Reactor tripped and the crew is operating in Overheating (1202.004)
- Subcooling is inadequate
- CET temperatures are lowering
- Adequate primary to secondary heat transfer cannot be established due to problems with EFW flow control
- RCP services cannot be restored Based on this, the crew should _______.
A. Go to HPI Cooldown (1202.011)
B. Remain in Overheating (1202.004) until heat transfer can be established C. Go to Inadequate Core Cooling (1202.005)
D. Go to Reactor Trip (1202.001)
Proposed Answer: __A__
Explanation (Optional):
Technical Reference(s): 1202.004 ______________________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___N/A______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 ___
55.43 __5___
Comments:
Question: 89 Examination Outline Cross-
Reference:
Level RO SRO Tier # ____ __2___
Group # ____ __1___
K/A # __063 A 2.2.42___
Importance Rating ___ _4.6___
K/A Statement: [063 D.C. Electrical Distribution] Ability to recognize system parameters that are entry-level conditions for Technical Specifications Proposed Question:
Which of the following conditions requires entry into Technical Specification 3.8.4, "DC Sources, Operating and what is the bases for Technical Specification 3.8.4?
A. D04A, "Battery Charger" inoperable and D06, "Battery" operable.
Bases is to insure reactor coolant pressure boundary limits are not exceeded as a result of abnormalities B. D04B, "Battery Charger" inoperable and D03B, "Battery Charger" inoperable.
Bases is to insure reactor coolant pressure boundary limits are not exceeded as a result of abnormalities C. D04A, "Battery Charger" inoperable and D04B, "Battery Charger" inoperable.
Bases is to insure adequate core cooling is provided, and reactor building operability and other functions are maintained in the event of a postulated DBA D. D03B, "Battery Charger" inoperable and D07, "Battery" operable.
Bases is to insure adequate core cooling is provided, and reactor building operability and other functions are maintained in the event of a postulated DBA Answer C. D04A, "Battery Charger" inoperable and D04B, "Battery Charger" inoperable.
Bases is to insure adequate core cooling is provided, and reactor building operability and other functions are maintained in the event of a postulated DBA Notes:
C is correct, with both battery chargers on the same train being inoperable, the subsystem is inoperable requiring entry into TS 3.8.4. The bases for TS 3.8.4 is to insure adequate core cooling is provided, and reactor building operability and other functions are maintained in the event of a postulated DBA A is incorrect, Only one of the two charges being inoperable does not affect the operability of the subsystem. The bases used for this option is partially correct.
B is incorrect, two battery chargers are inoperable but since they are on different trains they do not affect the operability of either subsystem. The bases used for this option is partially correct.
D is incorrect, Only one of the two charges being inoperable does not affect the operability of the subsystem. The bases used for this option is partially correct.
References:
T.S. 3.8.4 Amendment 215 History:
Bank Q from 2010 SRO exam Learning Objective: _________________________ (As available)
Question Source: Bank # ____X___
Modified Bank # _______ (Note changes or attach parent)
New _______
Question History: Last NRC Exam ____2010________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 _____
55.43 __2___
Comments:
Question# 90 Examination Outline Cross-
Reference:
Level RO SRO Tier # _____ __2 _
Group # _____ __1 _
K/A # __064G2.2.22___
Importance Rating _____ ____4.7_
Knowledge of Limiting Condition for Operations and Safety Limits: EDG Proposed Question:
The plant is in Mode 4. During the performance of DG Surveillance 3.8.1.8, to demonstrate operability for a Loss of Offsite Power event, all DG loads are properly sequenced on in 14 seconds, except one non-essential load that failed to trip off. During the surveillance, the DG supplied connected loads for 5 minutes.
As Control Room Supervisor, you decide the EDG is ___________ because _____________.
A. Operable; the tripping of non-essential loads does not affect operability.
B. Inoperable; the tripping of non-essential loads affects operability.
C. Operable; the EDG is capable of meeting the 15 second time requirement.
D. Inoperable; the EDG did not supply connected loads for more than 5 minutes.
Proposed Answer: ___B___
Explanation (Optional):
A. Incorrect. TS Bases page 3.8.1-4 (and page 3.8.1.15) states that non-essential loads tripping off is required for Operability. In addition, TS SR 3.1.8 lists Load Shedding as a surveillance requirement.
B. Correct. See A.
C. Incorrect. The 15 second time requirement was met (plausible) but the Load Shedding requirement was not.
D. Incorrect. Plausible, but the actual requirement is equal to or greater than 5 minutes, not greater than 5 minutes.
Technical Reference(s): _TS 3.1.8, SR 3.1.8, and TS Bases 3.1.8___________________________________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ___None__________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ____X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X___
10 CFR Part 55 Content: 55.41 _____
55.43 __X___
Comments:
Question 91 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ __2__
Group # __2__ __2__
K/A # _____016A2.01__
Importance Rating __3.0_ __3.1_
A2.01. Non Nuclear Instrumentation System: Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector Failure Proposed Question:
Plant power 100%
TE-1014 A Loop Narrow Range T-hot SLOWLY starts to fail low.
What procedure should be used for this condition?
A. 1203.001, ICS Abnormal Operation B. 1203.012F, Annunciator K07 Corrective Action (K07-B4 SASS Mismatch)
C. 1202.001, Reactor Trip D. 1105.006, Reactor Coolant System NNI Proposed Answer: ___A___
Explanation (Optional):
Answer A is the correct answer.
Answer B is incorrect. Would be correct if TE1014 instantly failed low.
Answer C is incorrect. Trip would not occur if SRO enters correct procedure Answer D is incorrect. Although this is an NNI system, only recovery actions exist for this instrument Technical Reference(s): _1203.001, ICS Abnormal Operation, Revision 11, Pages 7-8 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # ___X__ (Note changes or attach parent)
New ______
Question History: Last NRC Exam ____2009 Retake_____
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge ____
Comprehension or Analysis __X_
10 CFR Part 55 Content: 55.41 _____
55.43 __5__
Comments:
Bank QID 0744 - Changed stem from instantly failing to slowly failing, changing the correct answer from B to A
Question 92 Examination Outline Cross-
Reference:
Level RO SRO Tier # __2__ __2__
Group # __2__ __2__
K/A # _____017G2.1.30__
Importance Rating __4.4_ __4.0_
017G2.1. In-Core Temperature Monitoring System Conduct of Operations: Ability to locate and operate components, including local controls.
Proposed Question:
Given:
The plant computer is unavailable.
Core EFPD is 200 Reactor power is 80%
NI-7 is out of service Calculated quadrant tilt is 5.90% in quadrant WX.
If these indications continue, which of the following actions is required to be taken in accordance with Technical Specifications?
A. Reduce applicable RPS trip setpoints 1.5%
B. Reduce applicable RPS trip setpoints 3%
C. Reduce applicable RPS trip setpoints 6%
D. Reduce applicable RPS trip setpoints 8%
Answer:
D. Reduce applicable RPS trip setpoints 8%
Explanation:
Answer D is correct per the COLR and T.S. 3.2.4. With the plant computer unavailable and NI-7 out of service then Quadrant Tilt limits become the Minimum In-core Detector System Setpoint of 1.90 if >60% power. With QPT at 5.90% (4% above limit), then the RPS trip setpoints must be reduced by 8% per TS 3.2.4 (2% for each 1% of limit exceeded).
The other answers are incorrect possibilities.
Technical Reference(s): _COLR page 24 and T.S. 3.2.4.
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination:
COLR page 24 and T.S. 3.2.4 Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # __QID ANO-OPS1-6624a ____ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis ___
10 CFR Part 55 Content: 55.41 _____
55.43 __6__
Comments:
Parent question QID ANO-OPS1-6624a.
The plant computer is unavailable and excore nuclear instrumentation is being used to calculate quadrant tilt.
The following conditions exist:
Core EFPD is 200 Reactor power is 80%
Calculated quadrant tilt is 4.96% in quadrant WX.
If these indications continue, which of the following actions is required to be taken in accordance with Technical Specifications?
A. Reduce applicable RPS trip setpoints 1.5%
B. Reduce applicable RPS trip setpoints 3%
C. Reduce applicable RPS trip setpoints 6%
D. Reduce applicable RPS trip setpoints 8%
Answer:
C. Reduce applicable RPS trip setpoints 6%
Explanation:
Answer C is correct per the COLR and T.S. 3.2.4. With the plant computer unavailable and excore NIs used to calculate QPT, then Quadrant Tilt limits are 1.96 if >60% power. With QPT at 4.96% (3% above limit), then the RPS trip setpoints must be reduced by 6% per TS 3.2.4 (2%
for each 1% of limit exceeded).
Question: 93 Examination Outline Cross-
Reference:
Level SRO Tier # __2__ _____
Group # __2 __ _____
K/A # __068G2.4.45 ____
Importance Rating _4.3 _ _____
068 Liquid Radwaste System (LRS): 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.
Proposed Question:
While discharging liquid radwaste to the flume, the PROC MONITOR RADIATION HI (K10-B2) alarms and there is also an alarm on the Liquid Radwaste Process Monitor (RI-4642), (C25, Bay 2). An operator reports one of the liquid radiation monitors has failed high.
The CRS should find the operability requirements for the failed liquid radwaste radiation monitor, in:
A. Technical Specifications B. Technical Requirements Manual (TRM)
C. Off-Site Dose Calculation Manual (ODCM)
D. Updated Safety Analysis Report (USAR)
Proposed Answer: ___C__
Explanation (Optional):
The ODCM has the operability requirements and required actions for the liquid radwaste radiation monitors. While TS (A) and the TRM (B) address a variety of station instrumentation, they do not address the liquid radwaste system. The USAR (D) discusses the radiation monitors but does not contain the operability requirements.
Technical Reference(s): ____ODCM, L2.1.1, Rev 17________________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___NA__
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 __ _
55.43 ___1_
Comments:
Question 94 Examination Outline Cross-
Reference:
Level RO SRO Tier # _____ __3__
Group # _____ _____
K/A # __G2.1.20_________
Importance Rating ______ __4.6__
G2.1.20 - Ability to execute procedure steps.
Proposed Question:
Per OP-1015.043, ANO-1 EOP/AOP User Guide, which of the following identifies one of the criteria to bring steps forward from another EOP/AOP or steps within an EOP/AOP to be performed out of order:
A. The CRS and the Operations Managers approval are required before performing the steps out of sequence.
B. The actions are necessary to mitigate the event and maintain the safety function.
C. The actions to be performed will affect the ability to assess the status of the safety functions.
D. The CRS and the Assistant Operations Managers approval are required before performing the steps out of sequence.
Proposed Answer: __B___
Explanation (Optional):
A. Shift Manager AND Control Room Supervisor approval is required.
B. Correct.
C. The actions to be performed will NOT affect the ability to assess the status of the safety functions.
D. Shift Manager AND Control Room Supervisor approval is required.
Technical Reference(s): __OP-1015.043, ANO-1 EOP/AOP User Guide, Chg 5, Step 5.3.2_________________________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: ___None_______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New __X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis _____
10 CFR Part 55 Content: 55.41 _____
55.43 _____
Comments:
Question: 95 Examination Outline Cross-
Reference:
Level RO SRO Tier # ____ __3__
Group # ____ __ __
K/A # 2.2.36 ____
Importance Rating _ __4.2 K/A: Generic - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.
Proposed Question:
While in Mode 5, a supporting SSC for a Technical Specification applicable system has been removed from service for maintenance activities. Prior to this, the applicable system was not out of service for any other reason.
Which of the following would be used to determine and track Maximum Out-of-Service Time?
A. LCO 3.0.6 B. 1015.045, Safety Function Determination Program C. 1015.049, Configuration Control Program D. 1015.048, Shutdown Operations Protection Plan Proposed Answer: __B__
Explanation (Optional):
Technical Reference(s): 1015.045 ______________________________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: __None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___N/A______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge ___X__
Comprehension or Analysis ____
10 CFR Part 55 Content: 55.41 __ _
55.43 __2__
Comments:
Question: 96 Examination Outline Cross-
Reference:
Level RO SRO Tier # ____ __3___
Group # ____ _____
K/A # __G 2.2.35___
Importance Rating ___ _4.5___
K/A Statement: [Equipment Control] Ability to determine Technical Specification Mode of Operation.
Proposed Question:
The plant has the following initial conditions:
- NI-3 and -4 read 2x10-11 A
- RCS pressure is 415 psig
- RCS temperature is 270 degrees F
- Reactor Coolant Pumps 32A and 32B are running
- Reactor Coolant Pumps 32C and 32D are off What is the current Mode of operation in accordance with the plants Technical Specifications?
A. Mode 2 B. Mode 3 C. Mode 4 D. Mode 5 Proposed Answer: __C___
Explanation (Optional):
RCS temperature is less than 280F. Based on Technical Specifications Table 1.1-1, the plant is in Mode 4. Answer C is correct. RCS temperature is not equal to or greater than 280F, so B is incorrect. Reactor power is less than 5% RTP, so but RCS temperature is lower than the transition point from Mode 2 to Mode 3, so A is incorrect. Since RCS temperature is greater than 200F, the plant is not in Mode 5. Therefore, D is incorrect.
Technical Reference(s): _STM 1-67, Nuclear Instrumentation, Revision 13____
(Attach if not previously provided) ANO Technical Specifications, Amendment 245, Table 1.1-1 (including version/revision number) 1102.010, Plant Shutdown and Cooldown, Change 067 Proposed references to be provided to applicants during examination: _________________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X____
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X__
10 CFR Part 55 Content: 55.41 _____
55.43 __2___
Comments:
Question#_97 Examination Outline Cross-
Reference:
Level RO SRO Tier # _____ __3_
Group # _____ __ _
K/A # __G2.3.12___
Importance Rating _____ ____3.7_
Knowledge of Radiological Safety Principles pertaining to licensed operator duties.
Proposed Question:
The plant is in Mode 6 and refueling is in progress. The Spent Fuel Pool (SFP) level is 23.5 feet over the top of the irradiated fuel assemblies in the storage racks.
Is the SFP inoperable or operable, and why?
A. The SFP is Operable; 23 feet is the minimum level allowed by Technical Specifications because this level will provide adequate shielding for a fuel element failure following a fuel handling accident if the storage racks are 95% full.
B. The SFP is Inoperable; the minimum level allowed by Technical Specifications is >23 feet because that level is required to provide adequate shielding against beta particles from the decay of Thorium-232 (decay heat).
C. The SFP is Operable; 23 feet is the minimum level allowed by Technical Specifications because this level meets the assumptions of iodine decontamination factors following a fuel handling accident.
D. The SFP is Inoperable; the minimum level allowed by Technical Specifications is >23 feet because this level is required to provide adequate shielding against decay heat and gammas released through decay chains following a fuel handling accident.
Proposed Answer: ___C___
Explanation (Optional):
A. Incorrect. Plausible because 23 feet is the TS minimum, but fuel element failure is incorrect; the basis for min level also assumes the storage racks are completely full.
B. Incorrect. Plausible because >23 feet is close to 23 feet and Th-232 does undergo beta decay. Wrong because the level requirement is not for Th decay.
C. Correct. 23 feet is correct as is the level is for meeting assumptions of iodine decontamination factors following a fuel handling incident.
D. Incorrect. Plausible because >23 feet is close to 23 feet but still incorrect. Decay chains occur independent of a fuel handling accident.
Technical Reference(s): _TS 3.7.13 and TS B3.7.13______________
(Attach if not previously provided) _______________________________________________
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination:
___None____
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ____X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis __X___
10 CFR Part 55 Content: 55.41 _____
55.43 __X___
Comments:
Question 98 Examination Outline Cross-
Reference:
Level RO SRO Tier # __3__ __3__
Group # ____ ____
K/A # _____G2.3.7__
Importance Rating __3.2_ __3.7_
G2.3.7. Radiation Control: Knowledge of radiation exposure limits under normal or emergency conditions Proposed Question:
- Large break LOCA has occurred
- P-34A, LPI pump operation is degrading
- No offsite release is in progress The OSC is dispatching a Repair Team to attempt repair of CV-1405.
What is the MAXIMUM dose each member of the team is allowed to receive, and whose authorization is required?
A. Planned dose shall not exceed 10 Rem TEDE with TSC Directors authorization B. Planned dose shall not exceed 10 Rem TEDE with OSC Directors authorization C. Planned dose shall not exceed 25 Rem TEDE with TSC Directors authorization D. Planned dose shall not exceed 25 Rem TEDE with OSC Directors authorization Proposed Answer: ___A___
Explanation (Optional):
Answer A is the correct answer Answer C is incorrect. Wrong Limit Answer B is incorrect. Wrong authorization Answer D is incorrect. Wrong Limit and authorization Technical Reference(s): _1903.033 (Change 022)
(including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # ____X___
Modified Bank # ______ (Note changes or attach parent)
New ______
Question History: Last NRC Exam ____2007________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge ____
Comprehension or Analysis __X_
10 CFR Part 55 Content: 55.41 _____
55.43 __4__
Comments:
Question 99 Examination Outline Cross-
Reference:
Level RO SRO Tier # __3__ __3__
Group # ____ ____
K/A # _____G2.4.13__
Importance Rating __4.0_ __4.6_
G2.4.13. Emergency Procedures/Plan: Knowledge of crew roles and responsibilities during EOP usage.
Proposed Question:
In the Emergency Operating Procedures, procedural steps marked with a bullet (*) are considered A. Non-sequential steps B. Floating steps C. Time Dependent steps D. Immediate Action steps Proposed Answer: ___A___
Explanation (Optional):
Answer A is the correct answer.
Answer B is incorrect. Floating steps do not have bullets Answer C is incorrect. Time dependent steps do not have bullets Answer D is incorrect. Immediate Action steps do not have bullets Technical Reference(s): _1015.043, ANO-1 EOP/AOP User Guide, Revision 005, Page 13 (including version/revision number) _______________________________________________
Proposed references to be provided to applicants during examination: _____None______
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # ______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ____________
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge __X__
Comprehension or Analysis ___
10 CFR Part 55 Content: 55.41 _____
55.43 __5__
Comments:
Question: 100 Examination Outline Cross-
Reference:
Level SRO Tier # __3__ _____
Group # __ __ _____
K/A # __G2.4.16 ____
Importance Rating _4.4 _ _____
2.4.16 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.
Proposed Question:
The reactor is operating at 80 percent power, there is a PROC MONITOR RADIATION HI (K10-B2) alarm in, and the control room crew has determined there is a 7 gpm Steam Generator Tube Leak. The leak rate is increasing approximately 1 gpm every five minutes. Based on this, the CRS should; A. Direct the reactor be tripped and then enter EOP 1202.001, Reactor Trip B. First enter AOP 1203.023, Small Steam Generator Tube Leak, and then transition to EOP 1202.001, Reactor Trip C. First enter AOP 1203.023, Small Steam Generator Tube Leak, and then transition to EOP 1202.006, Tube Rupture D. First enter AOP 1203.045, Rapid Plant Shutdown, then transition to EOP 1202.001, Reactor Trip Proposed Answer: __C___
Explanation (Optional):
The PROC MONITOR RADIATION HI is a direct entry condition into AOP1203.023 so this procedure would be entered first. This AOP directs entry into EOP 1202.006 if leak rate exceeds 10 gpm. With the given information, this will occur in about 15 minutes. According to procedure 1015.043, ANO 1 AOP/EOP User Guide, given a SG tube rupture, the reactor should not be tripped but a controlled shutdown is preferred making A and B incorrect. The rapid plant shutdown is performed using EOP 1202.006, not AOP 1203.045 making D incorrect.
Technical Reference(s): __AOP 1203.023, Small Steam Generator Tube Leak, (Attach if not previously provided) __Rev 20, 1015.043, ANO 1 AOP/EOP User Guide, Rev (including version/revision number) __5, EOP 1202.006, Tube Rupture, Rev 12___________
Proposed references to be provided to applicants during examination: _None___________
Learning Objective: _________________________ (As available)
Question Source: Bank # _______
Modified Bank # _______ (Note changes or attach parent)
New ___X___
Question History: Last NRC Exam ___NA_______
(Optional: Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge _____
Comprehension or Analysis ___X_
10 CFR Part 55 Content: 55.41 __ _
55.43 ___5_
Comments: