ML090490105
| ML090490105 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/02/2009 |
| From: | Mckernon T Operations Branch IV |
| To: | Entergy Operations |
| References | |
| Download: ML090490105 (98) | |
Text
Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0003 Source: Modified Rev: 1 Rev Date: 12/7/00 Originator: GGiles TUOI: ANO-1-LP-RO-EFW Objective: 10 System Number: 003 System
Title:
Reactor Coolant Pump Section: 3.4 Type: RCS Heat Removal
==
Description:==
Knowledge of the effect that a loss or malfunction of the RCPS will have on the following:
Feedwater and Emergency Feedwater.
K/A Number: K3.03 CFR
Reference:
41.7 / 45.6 Point Value: 1 RO Imp: 2.8 SRO Imp: 3.1 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
The following plant conditions exist:
- Reactor is tripped
- All 4 RCPs are OFF
- RCS pressure is 1800 psig and rising
- "A" and "B" OTSG pressures are ~980 psig The CBOR reports that RT-5 (Verification of EFW Actuation and Control) is completed.
Which of the following describes the expected EFIC system response?
- a. OTSG levels will be rising at a rate of ~4 inches per minute to 370 to 410 inches.
- b. OTSG levels will be rising at a rate of ~6 inches per minute to 300 to 340 inches.
- c. OTSG levels will be rising at a rate of ~6 inches per minute to 370 to 410 inches.
- d. OTSG levels will be rising at a rate of ~4 inches per minute to 300 to 340 inches.
Answer:
- c. OTSG levels will be rising at a rate of ~6 inches per minute to 370 to 410 inches.
Notes:
RT-5 directs the operator to select REFLUX BOILING setpoint (~378 inches) if subcooling margin is less than adequate. The given plant conditions indicate a loss of subcooling margin. EFIC will automatically control OTSG levels at the REFLUX BOILING setpoint of ~378 inches. The rate of OTSG level rise when RCPs are OFF is variable from 2 to 8 inches per minute, depending on OTSG pressure (2 inches per minute at 800
- psig, 8 inches per minute at 1050 psig). Therefore, (c) is the only correct response. (a) is incorrect because the wrong rate is given. (b) is incorrect because the wrong setpoint is given. (d) is incorrect because the wrong rate and wrong setpoint are given.
History:
Developed for 1998 SRO Exam.
Modifed for use in 2001 RO/SRO Exam.
Difficulty: 4 Taxonomy: An
References:
1202.012 (Rev 004-01-0), Repetitive Tasks, RT5 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0005 Source: Direct Rev: 0 Rev Date: 6/17/98 Originator: GGiles TUOI: AA51003-006 Objective: 6.2
System Number: 024 System
Title:
Emergency Boration Section: 4.2 Type: Generic APEs
==
Description:==
Ability to determine and interpret the following as they apply to the Emergency Boration:
Correlation between boric acid controller setpoint and boric acid flow.
K/A Number: AA2.03 CFR
Reference:
43.5 / 45.13 Point Value: 1 RO Imp: 2.9 SRO Imp: 3.0 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
The reactor has tripped and 3 control rods failed to fully insert.
The CRS has instructed you to perform Emergency Boration in accordance with RT-12.
Which of the following best describes the initial setting on the batch controller?
- a. Set the batch controller to the batch size determined by the plant computer boron program to compensate for the reactivity worth of the stuck rods.
- b. Set the batch controller to the maximum batch size setting and commence adding boric acid to the make up tank.
- c. Set the batch controller to the batch size required to obtain a shutdown margin of 1.5% delta K/K as determined by a reactivity balance calculation.
- d. Set the batch controller to the batch size required to maintain make up tank level between 55 and 86 inches while maintaining pressurizer level >100 inches.
Answer:
- b. Set the batch controller to the maximum batch size setting and commence adding boric acid to the make up tank.
Notes:
RT-12 instructs the operator to commence emergency boration by setting the batch controller to the maximum batch size (999999 gals) and to begin adding boric acid via the batch controller if a boric acid pump is available. Therefore, answer (b) is correct. Answers (a) and (c) describe actions to determine the exact batch size after commencing emergency boration, the question is asking for the initial setting of the batch controller.
Answer (d) uses a variety of setpoints associated with emergency boration incorrectly.
History:
Developed for 1998 RO Exam.
Used in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1202.012 (Rev 004-01-0), Repetitive Tasks, RT-12, Emergency Boration Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0008 Source: Direct Rev: 0 Rev Date: 7/9/98 Originator: JCork TUOI: ANO-1-LP-AO-ICW Objective: 9 System Number: 026 System
Title:
Loss of Component Cooling Water (ICW at ANO)
Section: 4.2 Type: Generic APEs
==
Description:==
Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water: : The CCWS surge tank, including level control and level alarms, and radiation alarm.
K/A Number: AA1.05 CFR
Reference:
41.7 / 45.5 / 45.6 Point Value: 1 RO Imp: 3.1 SRO Imp: 3.1
Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
Given:
- Process Radiation Monitor RI-2236, Nuclear ICW, is in alarm.
- Nuclear ICW flow rate is >3100 gpm
- Local reports of Nuclear ICW Surge Tank overflowing A leak in which of the following components would be capable of causing these conditions?
- a. RCP Seal Return Coolers
- b. Spent Fuel Coolers
- c. Letdown Coolers
- d. Pressurizer Sample Cooler Answer:
- c. Letdown Coolers Notes:
"C" is correct since it is the only component with the piping size and differential pressure to cause the indications given.
All of the other choices have either small piping size or relatively low differential pressures.
History:
Developed for 1998 SRO Exam.
Used in 2001 RO/SRO Exam.
Difficulty: 4 Taxonomy: Ap
References:
STM 1-43, rev. 3 ch. 1, Intermediate Cooling Water System, page 27, 28 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0010 Source: Modified Rev: 1 Rev Date: 12/7/00 Originator: GGiles TUOI: ANO-1-LP-RO-AOP Objective: 4.2 System Number: 051 System
Title:
Loss of Condenser Vacuum Section: 4.2 Type: Generic APEs
==
Description:==
Ability to perform specific system and integrated plant procedures during all modes of plant operation.
K/A Number: 2.1.23 CFR
Reference:
45.2 / 45.6 Point Value: 1 RO Imp: 3.9 SRO Imp: 4.0 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
Given:
- Plant is operating at 60% power
- E-11A North Waterbox is OOS for maintenance
- Condenser vacuum is degrading rapidly Choose the appropriate operator actions:
- a. Trip the reactor and turbine if vacuum falls below 26.5 inches Hg.
- b. Trip the reactor and turbine if vacuum falls below 24.5 inches Hg.
- c. Trip the turbine if vacuum falls below 26.5 inches Hg.
- d. Trip the turbine if vacuum falls below 24.5 inches Hg.
Answer:
- b. Trip the reactor and turbine if vacuum falls below 24.5 inches Hg.
Notes:
(b) is the correct answer in accordance with 1203.016. 60% power is ~540 MW, therefore a turbine trip is required at 24.5 inches along with a reactor trip since power is >43%.
(a), (b) and (c) are incorrect because a reactor trip is required for a turbine trip above 43% and/or the wrong setpoint is given. A trip at 26.5" Hg is only required if power is 30% or less.
History:
Developed for 1998 RO/SRO Exam.
Modified for use in 2001 RO/SRO Exam Difficulty: 2 Taxonomy: K
References:
1203.016 (Rev 011-00-0), Loss of Condenser Vacuum, page 1 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0011 Source: Direct Rev: 0 Rev Date: 7/1/98 Originator: GGiles TUOI: AA51003-013 Objective: 13.5 System Number: 055 System
Title:
Loss of Offsite & Onsite Power (Station Blackout)
Section: 4.1 Type: Generic EPEs
==
Description:==
Knowledge of the operational implications of the following concepts as they apply to the Station Blackout: Natural circulation cooling.
K/A Number: EK1.02 CFR
Reference:
41.8 / 41.10 / 45.3 Point Value: 1 RO Imp: 4.1 SRO Imp: 4.4 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
Due to severe weather the plant is in a station blackout condition.
The steam driven emergency feedwater pump (P-7A) is feeding both steam generators.
Under what conditions would an "Emergency RCS Cooldown" be performed?
- a. Loss of subcooling margin, no HPI available, and reactor vessel head voids are indicated.
- b. Reactor vessel head voids are indicated and Core Exit Thermocouple temperature > 610 °F.
- c. Loss of subcooling margin and steam generator tube to shell delta-T
>100 °F (tubes colder).
- d. Core Exit Thermocouple temperature > 610 °F and steam generator tube to shell delta-T >100 °F (tubes colder).
Answer:
- a. Loss of subcooling margin, no HPI available, and reactor vessel head voids are indicated.
Notes:
Answer (a) is correct. Blackout EOP floating step states: If adequate subcooling margin is lost and RV head voids are indicated, then perform step 3 contingency. This contingency directs operators to perform emergency cooldown per step 49. A note preceeding step 49 states: This section is used for emergency RCS cooldown if adequate SCM is lost and no HPI is available and RV head voids are indicated. Answers (b), (c) and (d) are incorrect for the following reasons: CET temperature > 610 °F is indicative of an overheating condition but does not meet the criteria for establishing an emergency cooldown. Steam generator tube to shell delta-T should be maintained <100 °F when stabilizing RCS temperature and does not apply to the emergency cooldown criteria.
History:
Developed for 1998 RO/SRO Exam.
Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1202.008 (Rev6), Blackout, step 3 contingency Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0015 Source: Direct Rev: 0 Rev Date: 6/30/98 Originator: GGiles TUOI: ANO-1-LP-RO-AOP Objective: 4.3 System Number: A06 System
Title:
Shutdown Outside Control Room Section: 4.3 Type: B&W EPEs/APEs
==
Description:==
Knowledge of the interrelations between the (Shutdown Outside Control Room) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
K/A Number: AK2.1 CFR
Reference:
41.7, 45.7 Point Value: 1 RO Imp: 3.8 SRO Imp: 3.8 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
A fire in the control room forced an immediate evacuation.
An alternate shutdown is in progress and the crew is attempting to stabilize the plant at hot shutdown natural circulation conditions.
The TSC directs the CRS to raise reactor coolant system pressure.
How is this action accomplished in accordance with the Alternate Shutdown AOP?
- a. Energize all pressurizer heaters from their respective power supply breaker cubicles.
- b. Reduce steaming rate at the Atmospheric Dump Valves to raise RCS temperature and pressure.
- c. Manually initiate High Pressure Injection to compress the pressurizer steam bubble.
- d. Manually throttle open on the pressurizer makeup block valve to raise pressurizer level.
Answer:
- c. Manually initiate High Pressure Injection to compress the pressurizer steam bubble.
Notes:
The CRS will manually initiate HPI by manually closing in the breaker for a HPI pump as directed by the TSC, therefore (c) is the correct response. (a), [b], and (d) are not options covered in the alternate shutdown procedure.
History:
Developed for 1998 RO Exam.
Used in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1203.002 (Rev 015-01-0), Alternate Shutdown, page 8 Arkansas Nuclear One - Unit One
Initial RO/SRO Exam Question Data QID: 0019 Source: Direct Rev: 0 Rev Date: 7/6/98 Originator: GGiles TUOI: ANO-1-LP-RO-AOP Objective: 4.3 System Number: 076 System
Title:
High Reactor Coolant Activity Section: 4.2 Type: Generic APEs
==
Description:==
Ability to recognize abnormal indications for system operating parameters which are entry level conditions for emergency and abnormal operating procedures.
K/A Number: 2.4.4 CFR
Reference:
41.10 / 43.2 / 45.6 Point Value: 1 RO Imp: 4.0 SRO Imp: 4.3 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
Fuel pin leakage has caused higher than normal activity in the reactor coolant system.
Which of the following indications on the Failed Fuel Monitor (RI-1237) would be indicative of failed fuel and require power reduction?
- a. A marked rise by 20% in the IODINE/GROSS ratio.
- b. A marked rise by 40% in the GROSS/IODINE ratio.
- c. A marked drop by 20% in the IODINE/GROSS ratio.
- d. A marked drop by 40% in the GROSS/IODINE ratio.
Answer:
- d. A marked drop by 40% in the GROSS/IODINE ratio.
Notes:
The plant computer provides indication of the Calculated Failed Fuel Gross/Iodine Ratio (R1237R). Step 3.1 of 1203.019 specifies that if the failed fuel ratio drops by 40% as indicated by WCO logs or the Plant Computer, reduce reactor power by 50% of present power level, therefore, (d) provides the only correct response. Answers (a), (b) and (c) provide the wrong magnitude and/or direction of the change in failed fuel ratio.
History:
Developed for 1998 RO/SRO Exam.
Used in A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam Difficulty: 4 Taxonomy: Ap
References:
1203.019 (Rev 010-02-0), High Activity in Reactor Coolant, page 4 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0020 Source: Direct Rev: 0 Rev Date: 7/6/98 Originator: GGiles TUOI: ANO-1-LP-RO-NNI Objective: 6 System Number: A02 System
Title:
Loss of NNI-X Section: 4.3 Type: B&W EPEs/APEs
==
Description:==
Ability to determine and interpret the following as they apply to the (Loss of NNI-X): Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.
K/A Number: AA2.2 CFR
Reference:
41.7 / 45.5 / 45.6 Point Value: 1 RO Imp: 4.0 SRO Imp: 4.0
Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
Given the following indications/alarms:
- SASS Mismatch alarm (fast flash)
- SG BTU Limit alarm (slow flash)
- SG "B" FW Temp signal select switch selected to SASS Enable with the white indicating light out and the blue "Y" light on.
What operator action is procedurally required?
- c. No action necessary, SASS has automatically transferred to "Y" NNI.
- d. Place both FW loop demands in manual.
Answer:
Notes:
"a" is the correct response per procedure 1203.012F and 1105.006.
"b" is incorrect, this action is performed when resetting a failed signal.
"c" is incorrect although SASS has transferred to [Y], the response in "a" the only procedurally required action.
"d" is not necessary, the SASS system has transferred to a good signal, no ICS upset should occur, and this action should include placing the SG/Rx master in manual as well.
History:
Developed for 1998 RO/SRO Exam.
Modified QID 3127 Used in 2001 RO/SRO Exam.
Difficulty: 4 Taxonomy: An
References:
1203.012F, Rev. 026-00-0, Annunciator K07 Corrective Action, page 17, 20 1105.006 rev. 009-00-0 Reactor Coolant System NNI, page 13 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0030 Source: Direct Rev: 0 Rev Date: 7/8/98 Originator: GGiles TUOI: ANO-1-LP-RO-EOP10 Objective: 5 System Number: E08 System
Title:
LOCA Cooldown Section: 4.3 Type: B&W EPEs/APEs
==
Description:==
Knowledge of the interrelationships between the (LOCA Cooldown) and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
K/A Number: EK2.2 CFR
Reference:
41.7 / 45.7 Point Value: 1 RO Imp: 4.0 SRO Imp: 4.0 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
During a small break LOCA cooldown, which of the following criteria must be met before High Pressure
Injection may be throttled?
- a. Restoration of adequate subcooling margin.
- b. Total low pressure injection flow >2800 gpm.
- c. Restoration of forced flow cooling.
- d. At least one OTSG is available as a heat sink.
Answer:
- a. Restoration of adequate subcooling margin.
Notes:
Answer (a) is correct in accordance with B&W guidelines for HPI throttling/termination. Answer (b) is incorrect because it states incorrect LPI flow criteria, answers (c) and (d) are incorrect because they state other nonapplicable cooldown methods.
History:
Developed for 1998 RO/SRO Exam.
Used in RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
B&W Technical Document, Part VI, Specific Rules Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0031 Source: Direct Rev: 0 Rev Date: 7/9/98 Originator: GGiles TUOI: ANO-1-LP-RO-ELECD Objective: 7.19 System Number: E10 System
Title:
Post Trip Stabilization Section: 4.3 Type: B&W EPEs/APEs
==
Description:==
Knowledge of the interrelationships between the (Post-Trip Stabilization) and the following:
Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
K/A Number: EK2.1 CFR
Reference:
41.7 / 45.7 Point Value: 1 RO Imp: 3.5 SRO Imp: 4.0 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
Given:
- The unit is operating at 100% power.
- SU-1 transformer is INOPERABLE.
- The Startup Transformer Preferred Transfer Switches on C-10 for A1/H1 and A2/H2 are selected to SU-2 transformer.
- SU-2 feeder breakers to A1, A2, H1 and H2 are out of the pull-to-lock position.
Which of the following best describes the electrical system response to a reactor/turbine trip?
- a. With SU-1 inoperable, the load shed protective circuitry will require a manual transfer of electrical buses to SU-2 due to the limited capacity of SU-2.
- b. The reactor/turbine trip will actuate the generator lockout relay and result in a "slow transfer" of A1, A2, H1 and H2 electrical buses to SU-2 transformer.
- c. The tie breakers between the vital and non-vital buses will open, the emergency diesel generators will start and supply the vital buses and SU-2 will power A1, A2, H1 and H2.
- d. The generator lockout relay trip will cause a "fast transfer" to SU-2 transformer and applicable loads will be shed to limit the loading on SU-2 transformer.
Answer:
- d. The generator lockout relay trip will cause a "fast transfer" to SU-2 transformer and applicable loads will be shed to limit the loading on SU-2 transformer.
Notes:
Answer (d) is correct. The reactor/turbine trip will actuate the generator lockout relay trip which will cause a fast transfer to SU-2. The load shed circuitry will automatically shed applicable loads to limit the loading on SU-2. Answer (a) is incorrect since an automatic transfer will occur. Answer (b) is incorrect since the transfer will be a "fast transfer". Answer (c) is incorrect because tie breakers will not open and the EDGs will not start.
History:
Developed for 1998 RO/SRO Exam.
Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1107.001 (Rev 057-02-0), Electrical System Operation, page 4 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0039 Source: Direct Rev: 0 Rev Date: 7/10/98 Originator: GGiles TUOI: ANO-1-LP-RO-ICS Objective: 12 System Number: 035 System
Title:
Steam Generator System (S/GS)
Section: 3.4 Type: :Heat Removal from Reactor Core
==
Description:==
Knowledge of annunciator response procedures.
K/A Number: 2.4.10 CFR
Reference:
41.10 / 43.5 / 45.13 Point Value: 1 RO Imp: 3.0 SRO Imp: 3.1 Tier: 2 Group: 2 RO Select: Yes SRO Select: No Question:
ICS is in full automatic and the CBOR is verifying proper plant response to a trip of the "A" main feedwater pump from 100% power.
Which of the following should the CBOR expect to occur?
- a. The operating main feedwater pump demand will have a bias of 30% added.
- b. The pressurizer spray valve will come open when RCS pressure reaches 2205 psig.
- c. The control rods will start inserting immediately due to a crosslimit from feedwater.
- d. The main feedwater block valves will close in fast speed once power goes below 80%.
Answer:
- d. The main feedwater block valves will close in fast speed once power goes below 80%.
Notes:
Answer (d) is correct since the main feedwater block valve closure is inhibited above 80%, once <80% they will close in fast speed. Answers (a) is incorrect, no bias is applied to demand although it will be maximized due to the high power level and the MFP trip.
Answer (b) is incorrect, a bias of 125 psig is subtracted from the normal setpoint of 2205, the spray valves should open at 2080 psig.
Answer (c) is incorrect, control rods will start to insert but not due to a crosslimit from feedwater, rather the reactor system will be crosslimiting feedwater. A bias subtraction from the reactor demand signal will cause the rods to insert.
History:
Developed for 1998 RO Exam.
Used in 2001 RO Exam.
Difficulty: 2 Taxonomy: C
References:
1203.012F, Rev. 026-00-0, Annunciator K07 Corrective Action, page 9 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0040 Source: New Rev: 0 Rev Date: 12/06/00 Originator: S. Pullin TUOI: ANO-1-LP-RO-AOP Objective: 3 System Number: E04 System
Title:
Inadequate Heat Transfer Section: 4.3 Type: B&W EPEs/APEs
==
Description:==
Knowledge of the operational implications of the following concepts as they apply to the (Inadequate Heat Transfer): Normal, abnormal and emergency operating procedures associated with (Inadequate Heat Transfer).
K/A Number: EK1.2 CFR
Reference:
41.8 / 41.10 / 45.3 Point Value: 1 RO Imp: 4.0 SRO Imp: 4.2 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
A reactor trip due to a loss of electrical buses H1 and H2 has occurred.
While performing reactor trip follow-up actions, the CBOR reports that RCS temperature is rising.
When would transition to 1202.004, Overheating be required?
- a. CET temperature exceeds 580 degrees.
- b. CET temperature exceeds 610 degrees.
- c. Thot temperature exceeds 580 degrees.
- d. Thot temperature exceeds 610 degrees.
Answer:
b) CET temperature exceeds 610 degrees.
Notes:
(b) is the correct answer per 1202.004, Overheating, entry conditions.
(a) is incorrect, entry is required at CET temperature of 610 degrees.
(c) and (d) are incorrect because Thot indication is not valid with no RCPs running (H1 & H2 buses deenergized).
History:
Developed for 2001 RO/SRO NRC Exam.
Difficulty: 3 Taxonomy: C
References:
1202.004, Rev 4, Overheating, entry conditions.
Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0041 Source: Direct Rev: 0 Rev Date: 7/10/98 Originator: GGiles TUOI: ANO-1-LP-RO-AOP Objective: 4.2 System Number: 058 System
Title:
Loss of DC Power Section: 4.2 Type: Generic APEs
==
Description:==
Ability to determine and interpret the following as they apply to the Loss of DC Power: DC loads lost; impact on ability to operate and monitor plant systems.
K/A Number: AA2.03 CFR
Reference:
43.5 / 45.13 Point Value: 1 RO Imp: 3.5
SRO Imp: 3.9 Tier: 1 Group: 2 RO Select: No SRO Select: Yes Question:
A reactor trip has occurred from 100% power following a loss of D01.
Attempts to transfer 125V DC panel D11 to its emergency supply are unsuccessful.
Both OTSGs pressures are ~890 psig.
Which of the following is the procedurally required action?
- a. Crosstie D-11 and D-21 from panel C-10 in the control room to provide power to D11.
- c. Select main turbine control to TURBINE MANUAL and close the governor valves in fast speed.
- d. Dispatch an operator to manually trip the main turbine from the main turbine front standard.
Answer:
Notes:
Answer (b) is correct per reactor trip EOP contingency actions and Loss of D01 AOP follow up actions. (a) is not physically possible, (c) and (d) are not procedural options in the Loss of D01 procedure.
History:
Developed for 1998 Exam.
Used in 2001 SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1203.036, Rev. 005-00-0, Loss of 125V DC, page 2 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0044 Source: Modified Rev: 1 Rev Date: 12/7/00 Originator: GGiles TUOI: AA61002-006 Objective: 6.8 System Number: 060 System
Title:
Accidental Gaseous Radwaste Release Section: 4.2 Type: Generic APEs
==
Description:==
Knowledge of the reasons for the following responses as they apply to the Accidental Gaseous Radwaste Release: Implementation of the E-plan.
K/A Number: AK3.01 CFR
Reference:
41.5, 41.10 / 45.6 / 45.13 Point Value: 1 RO Imp: 2.9 SRO Imp: 4.2 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
Events are in progress which have resulted in an unplanned gaseous radioactive offsite release which is expected to exceed the EPA Protective Action Guideline exposure levels.
What emergency classification would you recommend to the Shift Manager?
- a. Notification of Unusual Event
- b. Alert
- c. Site Area Emergency
- d. General Emergency Answer:
- d. General Emergency Notes:
Answer (d) is correct per the definition of a General Emergency. (a) is incorrect since for an NUE no releases requiring offsite response or monitoring are expected. (c) is incorrect since for an SAE releases are not expected to exceed limits except near the site boundary. (b) is incorrect because for an Alert, releases are not expected to be more than just a fraction of the EPA guidelines.
History:
Developed for 1998 SRO Exam.
Modified for use in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1903.010 (Rev 036-02-0), Emergency Action Level Classification, page 3 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0052 Source: Modified Rev: 1 Rev Date: 12/7/00 Originator: GGiles TUOI: ANO-1-LP-RO-RXBAL Objective: 3 System Number: 001 System
Title:
Control Rod Drive System Section: 3.1 Type: Reactivity Control
==
Description:==
Ability to manually operate and/or monitor in the control room: Determination of an ECP.
K/A Number: A4.10 CFR
Reference:
41.7/ 45.5 to 45.8 Point Value: 1 RO Imp: 3.5 SRO Imp: 3.9 Tier: 2 Group: 1 RO Select: Yes SRO Select: No Question:
The plant had been operating at 100% power for 200 days.
Following a plant trip, preparations for startup are in progress.
The CBOR is performing a calculation for Estimated Critical Position (ECP).
At a given boron concentration, which of the following post-trip times would result in the lowest rod index due to the effects of Xenon?
- a. 4 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
- b. 8 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- c. 40 to 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />
- d. 70 to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> Answer:
- d. 70 to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> Notes:
Answer (d) is correct since Xenon is essentially depleted at ~80 hours following a reactor trip from 100%
power (thumb rule), thus adding essentially no reactivity for which rod withdrawal must compensate for. (a) is incorrect since xenon is still building in at this time, (b) is incorrect since xenon has peaked at it's highest concentration pre-trip and (c) is incorrect because at this time the core is still not xenon free.
History:
Developed for the 1998 Unit 1 RO/SRO Exam.
Modified for use in 2001 RO Exam.
Difficulty: 4 Taxonomy: Ap
References:
General Physics Corporation PWR / Reactor Theory, ch.6, p.25
Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0055 Source: Direct Rev: 0 Rev Date: 7/11/98 Originator: GGiles TUOI: ANO-1-LP-RO-MUP Objective: 10 System Number: 004 System
Title:
Chemical and Volume Control System (CVCS)
Section: 3.2 Type: Reactor Coolant System Inventory Control
==
Description:==
Ability to manually operate and/or monitor in the control room: Boration/dilution batch control.
K/A Number: A4.12 CFR
Reference:
41.7 / 45.5 to 45.8 Point Value: 1 RO Imp: 3.8 SRO Imp: 3.3 Tier: 2 Group: 1 RO Select: Yes SRO Select: No Question:
Batch Controller Flow Control Valve (CV-1249) should normally be open no more than 20% when feeding boric acid. Why?
- a. To prevent pump runout on the boric acid pumps.
- b. To prevent exceeding design flow through the batch controller.
- c. To prevent overheating of the boric acid pumps.
- d. To prevent inaccurate readings on the batch totalizer.
Answer:
- d. To prevent inaccurate readings on the batch totalizer.
Notes:
Answer (d) is correct per note in 1103.004, Soluble Poison Concentration Control.
Answer (a) will certainly not runout the pumps at only 20% open, [b] is incorrect for the same reason.
Answer [c] is a precaution against dead heading the pumps, 20% open on the batch controller meets minimum flow requirements.
History:
Developed for the 1998 RO exam Used in 2001 RO Exam.
Difficulty: 2 Taxonomy: K
References:
1103.004, (Rev. 016-00-0) Soluble Poison Concentration Control, page 12 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0059 Source: Direct Rev: 0 Rev Date: 7/12/98 Originator: GGiles TUOI: ANO-1-LP-RO-NNI Objective: 25 System Number: 017 System
Title:
In-Core Temperature Monitor System (ITM)
Section: 3.7 Type: Instrumentation
==
Description:==
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ITM system controls including: Core exit temperature.
K/A Number: A1.01 CFR
Reference:
41.5 / 45.7 Point Value: 1 RO Imp: 3.7 SRO Imp: 3.9 Tier: 2 Group: 1 RO Select: Yes
SRO Select: Yes Question:
During a large break LOCA the value on the ICCMDS CET Subcooling Margin Display is negative and flashing.
What does this indicate?
- a. An ICCMDS communications error.
- b. CET readings are invalid.
- c. ICCMDS ability to calculate SCM is lost.
- d. CET readings indicate superheat.
Answer:
- d. CET readings indicate superheat.
Notes:
(d) is correct. (a) would cause a trouble alarm, (b) is incorrect since invalid readings are automatically removed from the ICCMDS calculations, (c) is incorrect - the bar graph will display dashes for the value in this instance.
History:
Developed for the 1998 RO/SRO exam.
Modified existing QID 727.
Used in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1105.008 (Rev 012-01-0), Inadequate Core Cooling Monitor, page 7 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0060 Source: Direct Rev: 0 Rev Date: 7/12/98 Originator: GGiles TUOI: AA41002-010 Objective: 11 System Number: 022 System
Title:
Containment Cooling System (CCS)
Section: 3.5 Type: Containment Integrity
==
Description:==
Knowledge of power supplies to the following: Containment cooling fans K/A Number: K2.01 CFR
Reference:
41.7 Point Value: 1 RO Imp: 3.0 SRO Imp: 3.1 Tier: 2 Group: 1 RO Select: Yes SRO Select: Yes 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 Answer:
- a. B5, B6 and B7 Notes:
(a) is correct per 1107 series procedures. (b), (c) and (d) list combination of wrong load centers.
TUOI ANO-1-LP-RO-VENT, refers to TUOI ANO-1-LP-WCO-RBPUR (AA41002-010), for review of Reactor Building Ventilation System. Objective 10 covers system interrelationships including electrical distribution.
History:
Developed for the 1998 RO/SRO exam.
Used in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1107.002 (Rev 017-03-0), ES Electrical System Operation, p. 53-54 1107.001, Rev. 057-02-0, Electrical System Operation, p. 59 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0062 Source: Direct Rev: 0 Rev Date: 7/12/98 Originator: GGiles TUOI: ANO-1-LP-RO-ICS Objective: 12 System Number: A01 System
Title:
Plant Runback Section: 4.3 Type: B&W EOP/AOP
==
Description:==
Knowledge of abnormal condition procedures.
K/A Number: 2.4.11 CFR
Reference:
41.10 / 43.5 / 45.13 Point Value: 1 RO Imp: 3.4 SRO Imp: 3.6 Tier: 1 Group: 2 RO Select: No SRO Select: Yes Question:
Given the following plant conditions:
- 100% power
- Condensate Pump P-2A OOS
- K06-E7 "COND PUMP MTR WDG TEMP HI" is in alarm
- AO reports fire in P-2C motor The CRS instructs the CBOT to trip P-2C.
Which of the following describes the correct response?
- a. Trip P-2C, perform immediate actions per 1203.027, Loss of Steam Generator Feed and dispatch the fire brigade per 1203.034, Smoke, Fire or Explosion.
- b. Trip P-2C, monitor ICS runback to 40% power and dispatch the fire brigade per 1203.034, Smoke, Fire or Explosion.
- c. Trip P-2C and reduce power per 1203.045, Rapid Plant Shutdown, to maintain adequate main feed pump suction pressure and dispatch the fire brigade per 1203.034, Smoke, Fire or Explosion.
- d. Trip P-2C then trip the turbine and reactor and carry out immediate actions per 1202.001, Reactor Trip and dispatch the fire brigade per 1203.034, Smoke, Fire or Explosion.
Answer:
- b. Trip P-2C, monitor ICS runback to 40% power and dispatch the fire brigade per 1203.034, Smoke, Fire or Explosion.
Notes:
The plant is designed to survive a loss of 2 condensate pumps. ICS will run the plant back at 50%/min to 40%
power (360 MWe). Immediate action for fire is to dispatch the fire brigade, therefore (b) is the correct response. (a) is actions for a loss of a main feedwater pump which should not occur. (c) main feed pump suction pressure will go down but recover as ICS runs plant back. (d) a reactor/turbine trip should not be required.
History:
Used in 2001 SRO Exam.
Difficulty: 4 Taxonomy: Ap
References:
1105.004 Rev 013-02-0, Integrated Control System, p.11 1203.034, Rev. 012-01-0, Smoke, Fire, or Explosion, p. 4 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0065 Source: Modified Rev: 2 Rev Date: 12/7/00 Originator: GGiles TUOI: ANO-1-LP-RO-RMS Objective: 2 System Number: 068 System
Title:
Liquid Radwaste System (LRS)
Section: 3.9 Type: Radioactivity Release
==
Description:==
Ability to manually operate and/or monitor in the control room: Automatic isolation.
K/A Number: A4.04 CFR
Reference:
41.7 / 45.5 to 45.8 Point Value: 1 RO Imp: 3.8 SRO Imp: 3.7 Tier: 2 Group: 1 RO Select: Yes SRO Select: Yes Question:
A radioactive liquid release is in progress.
RE-4642, Liquid Radwaste Process Monitor detects high radiation.
What effect will this have on the release?
- a. This will cause a process monitor trouble alarm to alert the control room staff to manually terminate the release.
- b. An interlock will cause instrument air to be vented from radwaste flow control valve, CV-4642, terminating the release.
- c. An interlock will align instrument air to the radwaste flow control valve, CV-4642, and automatically terminate the release.
- d. An interlock will trip the running radwaste transfer pump to terminate the release.
Answer:
- b. An interlock will cause instrument air to be vented from radwaste flow control valve, CV-4642, terminating the release.
Notes:
A loss of power to RI-4642 will result in a process monitor trouble alarm and a process monitor high radiation alarm. In addition, the loss of power will cause a high radiation signal trip of CV-4642. CV-4642 is an air to open valve. Answer (b) is correct. Answer (a) is incorrect because manual termination of the release is not required. Answer (c) is incorrect because the loss of power will vent instument air off of CV-4642 vice align it to the valve. Answer (d) is incorrect since CV-4642 will fail closed and manual termination is not required.
History:
Developed for 1998 RO/SRO Exam.
Revised after 9/98 exam analysis review.
Modified for use in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1203.012I, Rev 038-02-0, Annunciator K10 Corrective Actions, p.9 1203.007, Rev. 8, Liquid Waste Discharge Line High Radiation Alarm, p.1,2 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0079 Source: Direct Rev: 0 Rev Date: 6/29/98 Originator: JCork
TUOI: ANO-1-LP-WCO-PRMS Objective: 2 System Number: 029 System
Title:
Containment Purge System (CPS)
Section: 3.8 Type: Plant Service Systems
==
Description:==
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment Purge System controls including: Containment pressure, temperature, and humidity.
K/A Number: A1.03 CFR
Reference:
41.5 / 45.5 Point Value: 1 RO Imp: 3.0 SRO Imp: 3.3 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
Given:
Unit operating at 100%
Reactor Building pressure is 15.9 psia and stable No other abnormal conditions exist What action should be taken to lower RB pressure?
- a. Open RB purge inlets first, then open outlets.
- b. Vent RB via H2 sample lines.
- d. Open RB purge outlets first, then open inlets.
Answer:
Notes:
The only procedural guidance (1104.033) for depressurizing the RB at power is via the RB Leak Detector, therefore (c) is correct. (a) & (d) are distractors that imply use of the RB Purge system which would not be done at power. (b) is incorrect because RB depressurization is not a function of the Hydrogen Sampling system.
History:
Developed for the 1998 RO/SRO Exam.
Used in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1104.033, Rev 058-00-0, Reactor Building Ventilation, p.10, 11 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0088 Source: Direct Rev: 0 Rev Date: 6/29/98 Originator: JCork TUOI: AA51002-016 Objective: 18 System Number: 064 System
Title:
Emergency Diesel Generators (ED/G)
Section: 3.6 Type: Electrical
==
Description:==
Knowledge of ED/G system design feature(s) and/or interlock(s) which provide for the following:
Incomplete-start relay.
K/A Number: K4.05 CFR
Reference:
41.7 Point Value: 1 RO Imp: 2.8 SRO Imp: 3.2 Tier: 2 Group: 2
RO Select: Yes SRO Select: No Question:
The CRS directs you to perform Supplement 1 of 1104.036, #1 EDG Monthly Test. You depress the start pushbutton on C10, nothing happens, then the "EDG 1 OVERCRANK" annunciator K01-B2 alarms. The CRS directs the inside AO to check the EDG out and then depress the local RESET pushbutton.
Which of the following would occur after the AO depresses the RESET pushbutton?
- a. EDG would be ready for another manual start.
- b. EDG will not manually or automatically start.
- c. EDG output breaker will be locked out.
- d. EDG will immediately start cranking.
Answer:
- d. EDG will immediately start cranking.
Notes:
Answer (d) is correct since the EDG will start after reset pushbutton is depressed if stop pushbutton is not depressed to reset the logic. Answer (a) is incorrect because the EDG start logic must be reset after an overcrank, (b) is incorrect since it will automatically start and (c) is incorrect since the breaker is not affected.
History:
Developed for 1998 RO/SRO Exam.
Used in A. Morris 98 RO Re-exam Used in 2001 RO Exam.
Difficulty: 4 Taxonomy: An
References:
STM-1-3,1 Rev.6, Emergency Diesel Generators, p.8 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0090 Source: Direct Rev: 1 Rev Date: 11/4/98 Originator: JCork TUOI: ANO-1-LP-RO-FPS Objective: 11 System Number: 086 System
Title:
Fire Protection System (FPS)
Section: 3.8 Type: Plant Service Systems
==
Description:==
Knowledge of design feature(s) and/or interlocks which for the following: Detection and location of fires.
K/A Number: K4.03 CFR
Reference:
41.7 Point Value: 1 RO Imp: 3.1 SRO Imp: 3.7 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
Given:
- The yellow trouble LED is ON on #1 EDG flame detector module on C463 and has been acknowledged on C463.
- Subsequently, a #1 EDG smoke detector spuriously actuates.
Besides the "FIRE PROT SYSTEM TROUBLE", K12-D1, which of the following annunciators would also alarm?
- a. "FIRE WATER FLOW", K12-A2
- b. "FIRE WATER PRESSURE LOW", K12-B1
- c. "FIRE", K12-A1
- d. "FIRE PUMP AUTO START", K12-B2
Answer:
- c. "FIRE", K12-A1 Notes:
The fire detection system for the EDGs require a smoke detector and a flame detector to actuate and trip the deluge, which will cause a FIRE alarm. However, because there is no fire, the system's fuseable heads will prevent firewater flow and the fire pump will not autostart. Therefore the only expected annunciator is ( is (c).
History:
Developed for the 1998 RO/SRO Exam.
Revised after 9/98 exam analysis review.
Used in 2001 RO/SRO Exam.
Difficulty: 4 Taxonomy: An
References:
1203.009, Rev. 020-04-0, Fire Protection System Annunciator Corrective Action, p.2, 20, 86-90 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0102 Source: Modified Rev: 1 Rev Date: 11/17/00 Originator: JCork TUOI: ANO-1-LP-RO-AOP Objective: 3 System Number: 079 System
Title:
Station Air System (SAS)
Section: 3.8 Type: Plant Service Systems
==
Description:==
Knowledge of SAS design feature(s) and/or interlock(s) which provide for the following:
Crossconnect with IAS.
K/A Number: K4.01 CFR
Reference:
41.7 Point Value: 1 RO Imp: 2.9 SRO Imp: 3.2 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
Given:
- Instrument Air pressure has fallen to 45 psig,
- Unit Two is in a refueling outage with Breathing Air in use.
Which of the following will be in use to restore or conserve Instrument Air pressure?
- a. Instrument Air to Service Air X-over valve, SV-5400
- b. Cross-connect with Unit Two Instrument Air
- c. Breathing Air to Instrument Air X-connection, HS-5503
- d. If ICW available, isolate Seal Injection by closing CV-1206 Answer:
- a. Instrument Air to Service Air X-over valve, SV-5400 Notes:
Answer [a] is correct, SV-5400 opens when IA <50 psig.
Answer [b] is incorrect, the Unit Two x-connect is isolated when IA pressure <60 psig.
Answer [c] is incorrect, Breathing Air to Inst Air X-Connect is used early in a loss of I.A. transient and is isolated if BA header pressure drops to <80 psig with personnel using BA.
Answer [d] is incorrect, CV-1206 is placed in Override and isolated only when ICW is not available.
History:
Developed for 1998 RO exam Used in A. Morris 98 RO Re-exam Modified for use in 2001 RO/SRO Exam.
Difficulty: 2
Taxonomy: C
References:
1104.024, Rev. 026-01-0, Instrument Air System, page 8 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0107 Source: Modified Rev: 1 Rev Date: 12/7/00 Originator: JCork TUOI: ANO-1-LP-RO-ICS Objective: 22 System Number: 059 System
Title:
Main Feedwater (MFW) System Section: 3.4 Type: Heat Removal From Reactor Core
==
Description:==
Ability to monitor automatic operation of the MFW, including: ICS.
K/A Number: A3.07 CFR
Reference:
41.7 / 45.5 Point Value: 1 RO Imp: 3.4 SRO Imp: 3.5 Tier: 2 Group: 3 RO Select: Yes SRO Select: No Question:
The plant is operating at 60% power with Delta Tc and SG/RX Master stations in Hand.
All other ICS stations are in Auto.
If one RCP has to be tripped due to high vibration, how will the ICS respond?
(Assume no operator action other than tripping the RCP.)
- a. The ICS will runback the plant to 45% load at 50%/min.
- b. No change to FW will occur since the SG/RX Master is in Hand.
- d. The RC flow difference will re-ratio the FW flow demand.
Answer:
- d. The RC flow difference will re-ratio the FW flow demand.
Notes:
Following an RCP trip Delta Tc will re-ratio feedwater demands, therefore answer (d) is correct. Answer (a) is incorrect since the plant is operating below the runback setpoint, while (b) and (c) are incorrect because they state that ICS will not re-ratio feedwater demands.
History:
Modified QID 4408 for use on 1998 RO/SRO Exam.
Modified for use in 2001 RO Exam.
Difficulty: 4 Taxonomy: Ap
References:
1203.012F, Rev. 026-00-0 Annunciator K07 Corrective Action, p.7 STM 1-64, Rev.6, Intergrated Control System page 42 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0110 Source: Modified Rev: 1 Rev Date: 12/1/00 Originator: JCork TUOI: ANO-S-LP-SRO-ADMIN Objective: 4 System Number: 2.1 System
Title:
Conduct of Operations Section: 2.0 Type: Generic K/As
==
Description:==
Ability to apply technical specifications for a system.
K/A Number: 2.1.12 CFR
Reference:
43.2 / 43.5 / 45.3 Point Value: 1 RO Imp: 2.9
SRO Imp: 4.0 Tier: 3 Group: G RO Select: No SRO Select: Yes Question:
The ONLY action statement in Tech Specs regarding hydrogen recombiner operability (3.14.2) states:
"With one hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 days or the reactor shall be placed in the HOT SHUTDOWN condition within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />."
Which of the following actions would be applicable if BOTH hydrogen recombiner were inoperable?
- a. Place the plant in Hot Shutdown within one hour.
- b. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> commence a plant shutdown to place the plant in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- c. Restore at least one recombiner to operable status in 15 days or be in Hot Standby within 6 hrs.
- d. Submit a report outlining the plan for restoring the system to OPERABLE status to the NRC within 7 days.
Answer:
- b. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> commence a plant shutdown to place the plant in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Notes:
Answer [b] is correct. Without an action statement covering two inoperable analyzers, this condition falls under the jurisidiction of T.S. 3.0.3.
Answer [a] is incorrect, although a shutdown must commence in one hour, the plant does not have to be in Hot SDN in one hour.
Answer [c] is incorrect, this is similar to LCO statements where only one train of two redundant trains is inoperable.
Asnwer [d] is incorrect, the H2 analyzers may not be considered to be essential to plant operations and this statement is similar to action statements for non-ESF equipment.
History:
Modified for use in 2001 SRO exam.
Difficulty: 4 Taxonomy: Ap
References:
Technical Specifications, 3.03 and 3.14 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0111 Source: Modified Rev: 1 Rev Date: 12/7/00 Originator: JCork TUOI: ANO-1-LP-RO-PROCS Objective: 1 System Number: 2.1 System
Title:
Conduct of Operations Section: 2.0 Type: Generic K/As
==
Description:==
Knowledge of conduct of operations requirements.
K/A Number: 2.1.1 CFR
Reference:
41.10 / 45.13 Point Value: 1 RO Imp: 3.7 SRO Imp: 3.8 Tier: 3 Group: G RO Select: Yes SRO Select: No Question:
A LOCA has occurred.
- RCS pressure 1650 psig and falling SLOWLY
- RB pressure is 2.0 psig and rising SLOWLY
- RT-2, Initiate HPI, is complete Which of the following operator actions is procedurally required for these conditions?
- c. Announce parameter and imminent ESAS actuation, then verify proper automatic actuation.
Answer:
Notes:
Since the parameters and trend given indicate that time is available to obtain CRS/SS concurrence answer (b) is the most correct per 1015.001, Conduct of Ops. Answer (a) is applicable if there is not enough time but the wrong RT is given, (c) is wrong because it waits for automatic actuation and (d) performs the wrong action, bypass vs. actuate.
History:
Developed for 1998 RO/SRO exam.
Used in A. Morris 98 RO Re-exam Modified for use in 2001 RO Exam.
Difficulty: 3 Taxonomy: C
References:
1015.001, Rev 052-05-0, Conduct of Operations, p. 49 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0113 Source: Direct Rev: 0 Rev Date: 7/13/98 Originator: JCork TUOI: AA61002-009 Objective: 5 System Number: 2.1 System
Title:
Conduct of Operations Section: 2.0 Type: Generic K/As
==
Description:==
Knowledge of less than one hour technical specification action statements for systems.
K/A Number: 2.1.11 CFR
Reference:
43.2 / 45.13 Point Value: 1 RO Imp: 3.0 SRO Imp: 3.8 Tier: 3 Group: G RO Select: Yes SRO Select: Yes Question:
Given:
- The unit is operating at 100% power.
- A system engineer enters the control room with a condition report stating the PZR code safety valve (PSV-1002), replaced during the last outage, was set by the vendor using out of calibration equipment.
- The condition report estimates the setpoint for PSV-1002 could be as high as 2790 psig.
- The system engineer recommends declaring PSV-1002 inoperable.
What action would you initiate?
- a. Restore PSV-1002 to operable status within 15 minutes or be in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- b. Within one hour initiate a shutdown to be in Hot Standby within 6 hrs and be in Hot Shutdown within another 6 hrs.
- c. Restore the safety to operable status within 6 hrs or place the unit in Hot Shutdown within the following 12 hrs.
- d. Restore safety to operable status within 12 hrs or place the unit in Hot Shutdown within the following 12 hrs.
Answer:
- a. Restore PSV-1002 to operable status within 15 minutes or be in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Notes:
(a) is the correct response per Tech Spec 3.1.1.3.A. (b), (c) and (d) provide erroneous time clocks.
History:
Developed for 1998 SRO exam Used in 2001 RO/SRO Exam.
Difficulty: 4 Taxonomy: An
References:
Technical Specification 3.1.3 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0116 Source: Direct Rev: 0 Rev Date: 7/14/98 Originator: JCork TUOI: ANO-1-LP-RO-NOP Objective: 7 System Number: 2.2 System
Title:
Equipment Control Section: 2.0 Type: Generic K/As
==
Description:==
Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.
K/A Number: 2.2.1 CFR
Reference:
45.1 Point Value: 1 RO Imp: 3.7 SRO Imp: 3.6 Tier: 3 Group: G RO Select: Yes SRO Select: No Question:
During an INITIAL approach to criticality, if criticality is NOT achieved within _________ of the ECP, insert ________ and ________ .
- a. plus or minus 1.0% delta k/k control rods to achieve 1.5% SD margin establish hot shutdown conditions
- b. plus or minus 1.0% delta k/k regulating groups notify Reactor Engineering
- c. plus or minus 0.5% delta k/k control rods to achieve 1.5% SD margin verify calculation
- d. plus or minus 0.5% delta k/k regulating groups verify calculation Answer:
- c. plus or minus 0.5% delta k/k control rods to achieve 1.5% SD margin verify calculation Notes:
Answer "C" is correct per 1102.008.
History:
Used in 1998 RO exam
Used in NRC developed RO exam 8/24/92, no. 88 Used in A. Morris 98 RO Re-exam Used in 2001 RO Exam Difficulty: 2 Taxonomy: K
References:
1102.008, Approach to Criticality, Rev. 018-0-0, page 12 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0117 Source: Direct Rev: 0 Rev Date: 6/29/98 Originator: JCork TUOI: ANO-S-LP-SRO-ADMIN Objective: 4 System Number: 2.2 System
Title:
Equipment Control Section: 2.0 Type: Generic K/As
==
Description:==
Knowledge of the process for controlling temporary changes.
K/A Number: 2.2.11 CFR
Reference:
41.10 / 43.3 / 45.13 Point Value: 1 RO Imp: 2.5 SRO Imp: 3.4 Tier: 3 Group: G RO Select: No SRO Select: Yes Question:
Which of the following would NOT require the use of a Temporary Alteration Package?
- a. Installation of a test gauge for a surveillance.
- b. Temporary power cables supplying temporary equipment.
- c. Installing a jumper to prevent a malfunctioning instrument loop from actuating equipment.
Answer:
- a. Installation of a test gauge for a surveillance.
Notes:
Answer (a) is correct in accordance with 1000.028, Control of Temporary Alterations. 4.15.2.h shows that activities using test gauges are NOT considered to be temporary alterations. The remaining answers are all covered by 4.15.1 as activities requiring a temporary alteration package.
History:
Developed for the 1998 SRO exam.
Used in 2001 SRO Exam.
Difficulty: 4 Taxonomy: Ap
References:
1000.028 (Rev 023-00-0), Control of Temporary Alterations, page 37 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0120 Source: Direct Rev: 0 Rev Date: 7/14/98 Originator: JCork TUOI: AA52001-009 Objective: 1 System Number: 2.3 System
Title:
Radiation Control Section: 2.0 Type: Generic K/As
==
Description:==
Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.
K/A Number: 2.3.10 CFR
Reference:
43.4 / 45.10 Point Value: 1 RO Imp: 2.9
SRO Imp: 3.3 Tier: 3 Group: G RO Select: Yes SRO Select: Yes Question:
Given:
- A Site Area Emergency has been declared on Unit 1.
- An Emergency Medical Team member must enter a 50 REM/hr area to rescue a critically injured employee.
Which of the following is the MAXIMUM time an individual team member can stay in this area?
- a. 15 minutes
- b. 30 minutes
- c. 45 minutes
- d. 60 minutes Answer:
- b. 30 minutes Notes:
The limit for life saving is 25 rem TEDE. 50R/hr means a 30 minute stay time, therefore "B" is correct.
History:
Modifiied for use in 1998 SRO exam Modified question from NRC developed SRO exam 2/6/95, no. 94 Used in A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam.
Difficulty: 4 Taxonomy: Ap
References:
1903.033, Protective Action Guidelines for Rescue, Rev. 017-01-0, p.5 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0121 Source: New Rev: 0 Rev Date: 12/06/00 Originator: S. Pullin TUOI: ANO-S-LP-RO-RADP Objective: 15 System Number: 2.3 System
Title:
Radiation Control Section: 2.0 Type: Generic K/As
==
Description:==
Knowledge of 10CFR20 and related facility radiation control requirements.
K/A Number: 2.3.1 CFR
Reference:
41.12 / 43.4 / 45.9, 45.10 Point Value: 1 RO Imp: 2.6 SRO Imp: 3.0 Tier: 3 Group: G RO Select: Yes SRO Select: Yes Question:
What is the federal occupational exposure limit to the skin of the whole body in accordance with 10CFR20?
- a. 5.0 rems/calendar year
- b. 15.0 rems/calendar year
- c. 25.0 rems/calendar year
- d. 50.0 rems/calendar year Answer:
d) 50.0 rems/calendar year
Notes:
(d) is the correct answer.
(a), (b), and (c) are incorrect values.
History:
New question developed for 2001 RO/SRO NRC Exam.
Difficulty: 3 Taxonomy: K
References:
1012.021, Exposure Limits and Controls, Rev. 004-01-0, page 5.
Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0125 Source: Direct Rev: 0 Rev Date: 7/13/98 Originator: JCork TUOI: ANO-1-LP-RO-AOP Objective: 5 System Number: 2.4 System
Title:
Emergency Procedures/Plan Section: 2.0 Type: Generic K/As
==
Description:==
Knowledge of RO tasks performed outside the main control room during emergency operations including system geography and system implications.
K/A Number: 2.4.34 CFR
Reference:
43.5 / 45.13 Point Value: 1 RO Imp: 3.8 SRO Imp: 3.6 Tier: 3 Group: G RO Select: No SRO Select: Yes Question:
During an Alternate Shutdown requiring an immediate control room evacuation, which of following is performed by a control board operator (RO#1 or #2) in accordance with 1203.002, Alternate Shutdown?
- a. Start and load EDGs to power vital components
- b. Start and stop HPI pump to maintain PZR level
- d. Strip 6900v H1 and H2 buses Answer:
Answer [c] is correct. The CRS performs all of the tasks with the exception of "c" which is shared between the RO's.
Answers [a], [b], [d] are performed during Alternate Shutdown by the CRS.
History:
Used in 2001 SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1203.002, Rev. 015-01-0, Alternate Shutdown p. 23 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0128 Source: Direct Rev: 1 Rev Date: 11/4/98 Originator: JCork TUOI: AA61002-006 Objective: 14 System Number: 2.4 System
Title:
Emergency Procedures/Plan Section: 2.0 Type: Generic K/As
==
Description:==
Knowledge of the emergency plan.
K/A Number: 2.4.29 CFR
Reference:
43.5 / 45.11 Point Value: 1 RO Imp: 2.6 SRO Imp: 4.0 Tier: 3 Group: G RO Select: No SRO Select: Yes Question:
Which of the following would be classified as a fission product barrier failure?
- a. RCS leakage indicates greater than 30 gpm.
- b. EOF Director determines the Reactor Building is breached.
- c. CNTMT radiation levels equal to Alert level from CNTMT Radiation EAL.
- d. The inability to monitor a Fission Product Barrier.
Answer:
Notes:
"B" is correct per EAL definition.
"A" is incorrect, leakage must be >50 gpm.
"C" is incorrect, rad levels must be equal to SAE level.
"D" is incorrect, the correct definition is two fission product barriers known to be breached with the inability to monitor the third.
History:
Developed for 1998 SRO exam.
Revised after 9/98 exam analysis review.
Used in 2001 SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1903.010, Emergency Action Level Classification, Rev. 036-02-0, pages 4 and 5.
Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0133 Source: Direct Rev: 0 Rev Date: 08/16/95 Originator: D. Walls TUOI: AA51002-001 Objective: 23 System Number: 010 System
Title:
Pressurizer Pressure Control System Section: 3.3 Type: RX Pressure Control
==
Description:==
Knowledge of the purpose and function of major system components and controls.
K/A Number: 2.1.28 CFR
Reference:
41.7 Point Value: 1 RO Imp: 3.2 SRO Imp: 3.3 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
Which of the following are the two purposes for maintaining a one gpm continuous spray flow bypassing the pressurizer spray valve?
- a. Assists in pressurizer level control and maintains pressurizer heaters at minimal firing rate.
- b. Minimizes surge and spray line temperature differentials and maintains pressurizer boron concentration near that in the RCS.
- c. Assists in maintaining pressurizer level and maintains pressurizer boron concentration near that in the RCS.
- d. Minimizes surge and spray line temperature differentials and raises the differential pressure across the spray valve.
Answer:
- b. Minimizes surge and spray line temperature differentials and maintains pressurizer boron concentration near that in the RCS.
Notes:
[b] is correct since the bypass flow will keep the spray line warm and the extra flow will cause some circulation through the surge line. The continous flow will also help to maintain the PZR boron close to the RCS boron concentration.
[a] and [c] are incorrect since the spray flow comes from the RCS, it will do nothing to maintain PZR leve.
[d] is incorrect, although the first portion is correct, the DP across the valve will be minimal.
History:
Taken from Exam Bank QID # 2182 Used in A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
STM 1-03, rev. 8 ch.1 Reactor Coolant System, page 14 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0134 Source: Direct Rev: 0 Rev Date: 11/30/98 Originator: B. Short TUOI: ANO-1-LP-RO-NI Objective: 3 System Number: 015 System
Title:
Nuclear Instrumentation System Section: 3.7 Type: Instrumentation
==
Description:==
Knowledge of the operational implications of the following concepts as they apply to the NIS:
Excore detector operation.
K/A Number: K5.10 CFR
Reference:
41.5 / 45.7 Point Value: 1 RO Imp: 2.8 SRO Imp: 3.0 Tier: 2 Group: 1 RO Select: Yes SRO Select: Yes Question:
The Gamma Metrics system uses discrimination circuitry to provide accurate indication of source range power levels.
Why is discrimination necessary in source range nuclear instrumentation and what is the result of incorrect discrimination?
- a. Discrimination separates out alpha and neutron pulses to provide a true gamma pulse.
With the discrimination set too high power will indicate higher than actual power.
- b. Discrimination separates out gamma and neutron pulses to provide a true alpha pulse.
With the discrimination set too low, power will indicate higher than actual power.
- c. Discrimination separates out gamma and alpha pulses to provide a true neutron pulse.
With the discrimination set too high, power will indicate lower than actual power.
- d. Discrimination separates out beta and gamma pulses to provide a true neutron pulse.
With the discriminator set too low, power will indicate lower than actual power.
Answer:
- c. Discrimination separates out gamma and alpha pulses to provide a true neutron pulse.
With the discrimination set too high, power will indicate lower than actual power.
Notes:
The discriminator module in Gamma Metrics passes only pulses produced from neutrons. Gamma and Alpha pulses are not above the minimum discrimination voltage. If the discrimination voltage was set too high, some of the neutron pulses would not be counted and thus indicated power would be less than actual power.
(c) is correct.
(a) (b) and (d) have various forms of the answer in incorrect applications.
History:
Developed for use on A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: K
References:
STM 1-67, Rev. 6, ch. 1, Nuclear Instrumentation, page 16, 17 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0137 Source: Direct Rev: 1 Rev Date: 04/15/93 Originator: G. Alden TUOI: ANO-1-LP-RO-ICS Objective: 28 System Number: 059 System
Title:
Main Feedwater Section: 3.4 Type: RCS Heat Removal
==
Description:==
Knowledge of the physical connections and/or cause-effect relationships between the MFW and the following systems: ICS.
K/A Number: K1.07 CFR
Reference:
41.2 to 41.9 / 45.7 to 45.8 Point Value: 1 RO Imp: 3.2 SRO Imp: 3.2 Tier: 2 Group: 1 RO Select: Yes SRO Select: Yes Question:
Which one of the following is NOT a function of the Rapid Feedwater Reduction feature of ICS?
- a. Low Load and Startup Control Valve demands are reduced to zero.
- b. Main Feedwater Pump speed goes to minimum.
- c. Both Main Feedwater Block Valves close in slow speed.
- d. Both Loop Feedwater demands are reduced to zero.
Answer:
- c. Both Main Feedwater Block Valves close in slow speed.
Notes:
[a], [b], & [d] are part of the RFR circuit and while [c] appears to be a logical component of this, the [c]
function is independent of RFR.
History:
Taken from Exam Bank QID # 3262 (modified answers slightly)
Used in A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
STM 1-64 Rev 6, Integrated Control System page 40
Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0143 Source: Modified Rev: 2 Rev Date: 12/7/00 Originator: D.Walls TUOI: ANO-1-LP-SRO-ADMIN Objective: 4 System Number: 2.1 System
Title:
Conduct of Operations Section: 2 Type: Generic K&As
==
Description:==
Knowledge of how to conduct and verify valve lineups.
K/A Number: 2.1.29 CFR
Reference:
41.10 / 45.1 / 45.12 Point Value: 1 RO Imp: 3.4 SRO Imp: 3.3 Tier: 3 Group: G RO Select: Yes SRO Select: No Question:
The plant is shut down for Refueling.
A Core Flood system valve alignment is in progress inside Controlled Access.
The primary sample room has become a high radiation area due to hydorgen peroxide cleanup.
The first check was made on CF-2, Core Flood Combined Sample Isolation, but the Shift Manager decided to waive the second check to reduce the exposure to high radiation.
Which one of the following statements most accurately describes why the Shift Manager's decison is acceptable or unacceptable?
- a. Acceptable, independent verifications are always waived for valve alignments inside High Radiation Areas.
- b. Unacceptable, independent verification cannot be waived if remote valve position indication is provided.
- c. Acceptable, independent verification can be waived for any valve with the Shift Manager's approval.
- d. Unacceptable, independent verifications cannot be waived for valve alignments without the approval of the Manager of Plant Operations.
Answer:
- b. Unacceptable, independent verification cannot be waived if remote valve position indication is provided.
Notes:
[b] meets the guidance of 1015.035.
[a] untrue, independent verifications are not always waived, they are only waived on a case by case basis.
[c] is untrue, the Shift Manager can only waive verification in specific situations.
[d] lists the wrong approval authority.
History:
Taken from Exam Bank QID # 3273 Used in A. Morris 98 RO Re-exam Modified for use in 2001 RO Exam.
Difficulty: 2 Taxonomy: Ap
References:
1015.035, Rev 010-00-0, Valve Operations, p. 11 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0145 Source: Direct Rev: 1 Rev Date: 11/16/94 Originator: G. Alden TUOI: ANO-1-LP-RO-CRD Objective: 16
System Number: 001 System
Title:
Control Rod Drive System Section: 3.1 Type: Reactivity Control
==
Description:==
Knowledge of system purpose and/or function.
K/A Number: 2.1.27 CFR
Reference:
41.7 Point Value: 1 RO Imp: 2.8 SRO Imp: 2.9 Tier: 2 Group: 1 RO Select: Yes SRO Select: No Question:
The purpose of the IN-LIMIT (LATCH) BYPASS switch on the Diamond panel is to:
- a. Apply power to the CRD motor which will engage the latching mechanism.
- b. Reset a fault condition provided the fault has cleared.
- d. Allow driving in Groups 1 thru 7 to engage roller nuts with lead screws.
Answer:
- d. Allow driving in Groups 1 thru 7 to engage roller nuts with lead screws.
Notes:
(a) is incorrect. The CRD motor has power applied at all times.
(b) is incorrect. This is accomplished with the fault reset pushbutton.
(c) is incorrect. This is accomplished with the Trip/reset pushbutton.
(d) is correct. The Latch pushbutton must be depressed to allow the in-limits to be bypassed when engaging lead scews on groups 1-7 after they have been deenergized.
History:
Taken from Exam Bank QID # 2684 Used in A. Morris 98 RO Re-exam Used in 2001 RO Exam.
Difficulty: 3 Taxonomy: K
References:
STM 1-02, Rev. 5, Control Rod Drive System, p. 24 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0156 Source: Direct Rev: 3 Rev Date: 06/28/97 Originator: M. Goad TUOI: ANO-1-LP-RO-EOP03 Objective: 7 System Number: E05 System
Title:
Excessive Heat Transfer Section: 4.2 Type: B&W EOP/AOP
==
Description:==
Ability to determine and interpret the following as they apply to the (Excessive Heat Transfer):
Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.
K/A Number: EA2.2 CFR
Reference:
43.5 / 45.13 Point Value: 1 RO Imp: 3.6 SRO Imp: 4.0 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
Given:
- A reactor trip has occurred.
- RCS pressure is 1800 psig,
- RCS T-cold is 532 degrees F,
- "A" OTSG pressure is 650 psig,
- "B" OTSG pressure is 970 psig,
- Reactor Building pressure is 6 psig.
Which emergency operating procedure contains the specific steps to mitigate the consequences of this event?
- a. ESAS 1202.010
- b. Overcooling 1202.003
- c. HPI Cooldown 1202.011
- d. Loss of Subcooling Margin 1202.002 Answer:
- b. Overcooling 1202.003 Notes:
The key to this question is in realizing that T-cold is lower than normal and that one OTSG is <900 psig. Also, RCS pressure is above ESAS actuation pressure but RB pressure is greater than ESAS actuation setpoint.
These are three of the five possible entry conditions for 1202.003, the Overcooling EOP, and all of these conditions are indicative of a steam line rupture inside the RB. Also, the floating steps for the Reactor Trip EOP send the user to the Overcooling EOP.
History:
Taken from Exam Bank QID # 556 Used in A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam Difficulty: 3 Taxonomy: Ap
References:
1202.010, Rev. 005-00-0, ESAS page 2 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0158 Source: Direct Rev: 0 Rev Date: 11/11/98 Originator: JCork TUOI: ANO-1-LP-RO-AOP Objective: 4.3 System Number: 069 System
Title:
Loss of Containment Integrity Section: 4.2 Type: Generic AOPs
==
Description:==
Knowledge of abnormal condition procedures.
K/A Number: 2.4.11 CFR
Reference:
41.10 / 43.5 / 45.13 Point Value: 1 RO Imp: 3.4 SRO Imp: 3.6 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
The plant is at 100% power.
The outside door of the personnel air lock was opened to replace a seal gasket 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ago.
How long does operations have to perform an LLRT on the personnel air lock before a loss of containment integrity will exist?
- a. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
- b. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- c. 6 days
- d. 13 days Answer:
- c. 6 days Notes:
[c] is the correct answer per 1203.005.
[a] is incorrect, this can occur but only one hour is allowed to perform the repairs.
[c] is incorrect, this is applicable to automatic containment isolation valves, but not the air lock.
[d] is an incorrect action since it states "within one hour" and is only necessary if the LLRT is unsuccessful.
History:
Developed for 1998 RO Re-exam Used in 1999 exam.
Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: Ap
References:
1203.005 [Rev 010-00-0], Loss of Reactor Building Integrity, page 1 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0160 Source: Direct Rev: 1 Rev Date: 08/05/94 Originator: E. Wentz TUOI: AA51003-014 Objective: 14.3 System Number: 041 System
Title:
Steam Dump System (SDS) and Turbine Bypass Control Section: 3.4 Type: RCS Heat Removal
==
Description:==
Knowledge of bus power supplies to the following: ICS, normal and alternate power supply.
K/A Number: K2.01 CFR
Reference:
41.7 Point Value: 1 RO Imp: 2.8 SRO Imp: 2.9 Tier: 2 Group: 1 RO Select: Yes SRO Select: No Question:
During a Rx trip transient, all nine of the + or - 24v DC NNI/ICS power supply status lights go out.
What would cause this condition?
- a. Loss of Offsite Power
- b. Loss of DC Bus D11
- c. Loss of Y02
- d. Loss of DC Bus D21 Answer:
- b. Loss of DC Bus D11 Notes:
The first step of EOP 1202.009 checks these indicating lights and a note gives the power supply as breaker 25 on D-11.
Therefore, [b] is the correct answer as long as no other abnormal indications are present.
[a] is incorrect, D11 is battery backed.
[c] is incorrect, although Y02 supplies power to ICS and NNI, neither of these power the indicating lights.
[d] is incorrect, the lights are only powered from D11, although the two are very similar in other functions.
History:
Taken from Exam Bank QID # 3202 Used in A. Morris 98 RO Re-exam Used in 2001 RO Exam.
Difficulty: 3 Taxonomy: K
References:
1203.047, Rev. 0, Loss of NNI Power, p.1 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0161 Source: Direct Rev: 0 Rev Date: 10/22/98 Originator: J. Cork TUOI: ANO-1-LP-RO-AOP Objective: 4.3 System Number: 2.1 System
Title:
Conduct of Operations Section: 2 Type: Generic K&As
==
Description:==
Ability to recognize indications for system operating parameters which are entry-level conditions for Technical Specifications.
K/A Number: 2.1.33 CFR
Reference:
43.2, 43.3 / 45.3 Point Value: 1 RO Imp: 3.4 SRO Imp: 4.0 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
Given:
- Power escalation is in progress following a shutdown.
- Reactor power is 30%.
- Rod 4 of Group 5 drops.
Which of the following actions should be taken?
- a. Insert all regulating rods in sequential mode.
- b. Trip the reactor and go to Reactor Trip, 1202.001.
- c. Verify plant stabilizes at 320 MWe after ICS runback.
- d. Refer to Tech Specs and recover the dropped rod.
Answer:
- d. Refer to Tech Specs and recover the dropped rod.
Notes:
[a] would only be performed if power was <2%.
[b] would not be done because only one rod dropped.
[c] power is <360 MWe so there wouldn't be any runback, the value given would require a power increase.
[d] is the correct answer per 1203.003, Tech Specs would be referenced, actions taken to ensure compliance, and the dropped rod recovered.
History:
Developed for A. Morris 98 RO Re-exam.
Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1203.003, Rev. 19 pc1, Control Rod Drive Malfunction Action, page 3 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0167 Source: Direct Rev: 0 Rev Date: 10/24/91 Originator: M. Cooper TUOI: ANO-1-LP-RO-AOP Objective: 4.2 System Number: 059 System
Title:
Accidental Liquid Radioactive-Waste Release Section: 4.2 Type: Generic AOP
==
Description:==
Ability to operate and/or monitor the following as they apply to the Accidental Liquid Radwaste Release: Radioactive-liquid monitor.
K/A Number: AA1.01 CFR
Reference:
41.7 / 45.5 / 45.6
Point Value: 1 RO Imp: 3.5 SRO Imp: 3.5 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
What action is required upon receipt of Liquid Radwaste Process Monitor (RI-4642) high alarm?
- a. Start another circ water pump to increase dilution flow.
- b. Verify no release in progress at Disch Flow to Flume (FI-4642) on C19.
- c. Verify with Unit 2 no other release is in progress.
- d. Have chemistry sample discharge flume for radionuclides.
Answer:
- b. Verify no release in progress at Disch Flow to Flume (FI-4642) on C19.
Notes:
Per 1203.007 immediate action for a liquid radwaste process monitor alarm is to verify that no release in progress at FI-4642.(answer b) All other actions are associated with radwaste discharges but are not the immediate action for the alarm.
History:
Modified from Exam Bank QID # 1725 Used in A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1203.007 (Rev 8), Liquid Waste Discharge Line High Radiation Alarm, p.1,2 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0169 Source: Direct Rev: 0 Rev Date: 11/19/98 Originator: J. Cork TUOI: ANO-1-LP-RO-NNI Objective: 19 System Number: 028 System
Title:
Pressurizer Level Malfunction Section: 4.2 Type: Generic AOP
==
Description:==
Ability to operate and/or monitor the following as the apply to the Pressurizer Level Control Malfunction: CVCS.
K/A Number: AA1.02 CFR
Reference:
41.7 / 45.5 / 45.6 Point Value: 1 RO Imp: 3.4 SRO Imp: 3.4 Tier: 1 Group: 3 RO Select: Yes SRO Select: Yes Question:
Given:
- Plant is at 100% power.
- PZR level transmitter LT-1001 selected via HS-1002 on C04.
- PZR temperature element TE-1001A selected via HS-1000 on C04.
The PZR temperature indicator, TI-1000, on C04 drops suddenly to 50°F (bottom of scale).
Without operator action, what will be the effect on the PZR Level Control System?
- a. PZR Level Control Valve, CV-1235, will open to establish a higher steady-state PZR level.
- b. PZR Level Control Valve, CV-1235, will maintain the same steady-state PZR level.
- c. PZR Level Control Valve, CV-1235, will close to establish a lower steady-state PZR level.
- d. PZR Level Control Valve, CV-1235, will fail open to continuously raise PZR level.
Answer:
- a. PZR Level Control Valve, CV-1235, will open to establish a higher steady-state PZR level.
Notes:
[a] is correct. A loss of temperature compensation will result which will appear as a low PZR level. This is the same reason which makes [b] & [c] incorrect.
(d) is incorrect. The loss of temperature compensation does not produce an indication that is similar to a high off scale indication.
History:
Developed for A. Morris 98 RO Re-exam.
Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: An
References:
STM 1-69, rev. 4, Non-Nuclear Instrumentation System, p. 20 thru 22 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0170 Source: Direct Rev: 0 Rev Date: 08/10/95 Originator: J. Haynes TUOI: AA51003-012 Objective: 12.2 System Number: 078 System
Title:
Instrument Air Section: 3.8 Type: Plant Service Systems
==
Description:==
Knowledge of bus power supplies to the following: Instrument air compressor.
K/A Number: K2.01 CFR
Reference:
41.7 Point Value: 1 RO Imp: 2.7 SRO Imp: 2.9 Tier: 1 Group: 3 RO Select: Yes SRO Select: Yes Question:
During a loss of offsite power with a SG tube leak, the A2 bus is re-energized from the A4 bus. The A4 bus is supplied by #2 EDG.
What is the key reason for this action?
- a. To start P-7B EFW pump and secure P-7A.
- b. To restart circ. water and re-establish condenser vacuum.
- c. To allow operation of the Aux Feedwater pump (P-75).
- d. To re-establish Instrument Air and ICW cooling.
Answer:
- d. To re-establish instrument air and ICW cooling.
Notes:
The strategy here, regardless of the tube leak, is to re-establish Instrument Air and ICW and ease complications of this transient by restoring RCP seal cooling and Letdown. Thus, answer [d] is correct.
[a] is a good idea during a tube leak but P-7B is powered from A3, making it unnecessary to energize A2.
[b] is also a good idea but procedural actions eliminate the need to re-establish condenser vacuum.
[c] incorrect, although the Aux Feedwater Pump is powered from A2, it is not the basis for performing this action.
History:
Taken from Exam Bank QID # 2791
Used in A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1202.007 [Rev 005-01-0], Degraded Power, pages 53 thru 55 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0178 Source: Direct Rev: 0 Rev Date: 11/21/98 Originator: J. Haynes TUOI: ANO-1-LP-RO-EOP05 Objective: 3 System Number: 074 System
Title:
Inadequate Core Cooling Section: 4.1 Type: Generic EOP
==
Description:==
Knowledge of symptom based EOP mitigation strategies.
K/A Number: 2.4.6 CFR
Reference:
41.10 / 43.5 / 45.13 Point Value: 1 RO Imp: 3.1 SRO Imp: 4.0 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
After entering Inadequate Core Cooling EOP (1202.005), RCS temperature and pressure indicate entry into Region 4 of Figure 4.
What RCP action is appropriate to restore core cooling?
- b. Restore RCP services and start one RCP/loop.
Answer:
Notes:
(a.) is correct. These are appropriate actions for Region 4.
(b.) is incorrect. These are apporpriate actions for Region 3.
(c.) is incorrect. RCPs are bumped in Region 2 only if RCP services are available.
(d) is incorrect. RCPs are never bumped without services restored.
History:
Developed for use in A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam.
Difficulty: 4 Taxonomy: C
References:
1202.005 Rev. 004-00-0, Inadequate Core Cooling, page 9 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0184 Source: Direct Rev: 0 Rev Date: 11/21/98 Originator: R. Fuller TUOI: ANO-1-LP-RO-AOP Objective: 4.2 System Number: 032 System
Title:
Loss of Source Range Nuclear Instrumentation Section: 4.2 Type: Generic AOP
==
Description:==
Knowledge of the reasons for the following responses as they apply to the Loss of Source Range Nuclear Instrumentation: Startup termination on source-range loss.
K/A Number: AK3.01 CFR
Reference:
41.5, 41.10 / 45.6 / 45.13 Point Value: 1 RO Imp: 3.2 SRO Imp: 3.6 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
Given:
- During a reactor startup with source range NI-2 and reactor power wide range recorder NR-502 inoperable, source range NI-1 fails to 10 E5.
- Intermediate range NI-3 indicates 5 E-11 amps
- Intermediate range NI-4 is off scale low.
What is required of the CBOR?
- a. Continue the startup utilizing NI-3 until NI-4 comes on scale.
- b. Perform a plant shutdown in accordance with normal operating procedures due to lack of proper overlap.
- c. Trip the reactor due to no on-scale indication of neutron flux available.
- d. Hold power constant and perform an NI calibration.
Answer:
- c. Trip the reactor due to no on-scale indication of neutron flux available.
Notes:
[c] is correct per guidance in 1203.021, if the recorder NR.502 is inoperable AND no SR channel is >10 E5 cps AND no IR channel is > 1 E-10 amps AND 3/4 PR instruments are <10% power, then no on-scale flux indication exists and the reactor must be tripped.
[a] is incorrect, although NI-4 might come on scale, the startup should not be continued without valid neutron flux indication.
[b] is incorrect, although shutting down is conservative, per procedure the reactor must be tripped immediately.
[d] is incorrect, this action sounds like it could rectify this situation, however, it would be impossible to calibrate the NI's at this point and would be contrary to procedural guidance.
History:
Developed for use in A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam.
Difficulty: 4 Taxonomy: An
References:
1203.021 (Rev 007-01-0), Loss of Neutron Flux Indication, page 7 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0192 Source: Direct Rev: 0 Rev Date: 11/23/98 Originator: J. Haynes TUOI: 51002-012 Objective: 4 System Number: 013 System
Title:
Engineered Safety Features Actuation System Section: 3.2 Type: RCS Inventory Control
==
Description:==
Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS:
Sensors and detectors.
K/A Number: K6.01 CFR
Reference:
41.7 / 45.5 to 45.8 Point Value: 1 RO Imp: 2.7 SRO Imp: 3.1
Tier: 2 Group: 1 RO Select: Yes SRO Select: Yes 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 come into alarm.
With the above conditions, a power loss to which of the following would cause an ESAS actuation?
- a. RS1
- b. B11
- c. RS3
- d. B21 Answer:
- a. RS1 Notes:
(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.
(b.) & (d.) are incorrect. Loss of power to B11 or B21 does not result in a loss of any ESAS functions.
(c.) is incorrect. Analog 3 would be tripped as a result of a power loss to B11. Analog 3 is already tripped.
History:
Developed for use in A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam.
Difficulty: 4 Taxonomy: An
References:
STM 1-65, ESAS, 3,4 and 8 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0193 Source: Direct Rev: 0 Rev Date: 11/23/98 Originator: R. Fuller TUOI: ANO-1-LP-RO-NI Objective: 8 System Number: 002 System
Title:
Reactor Coolant Section: 3.2 Type: Reactor Coolant System Invevtory Control
==
Description:==
Knowledge of the operational implications of the following concepts as they apply to the RCS:
Relationship between reactor power and RCS differential temperature.
K/A Number: K5.10 CFR
Reference:
41.5 / 45.7 Point Value: 1 RO Imp: 3.6 SRO Imp: 4.1 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
Several plant parameters can be monitored to ensure accurate indications of reactor power are available. Which of the following sets of parameters would be indicative of 60% reactor power?
- a. Tave 579 degrees, Thot 593 degrees, Tcold 564 degrees, total FW flow 6.5 million lbm/hr.
- b. Tave 580 degrees, Thot 599 degrees, Tcold 560 degrees, total FW flow 8.4 million lbm/hr.
- c. Tave 579 degrees, Thot 588 degrees, Tcold 570 degrees, total FW flow 6.5 million lbm/hr.
- d. Tave 581 degrees, Thot 590 degrees, Tcold 565 degrees,
total FW flow 9.8 million lbm/hr.
Answer:
- a. Tave 579 degrees, Thot 593 degrees, Tcold 564 degrees, total FW flow 6.5 million lbm/hr.
Notes:
(a.) is correct.
(b.) is incorrect. Parameters are indicative of >70% power.
(c.) is incorrect. Delta T is indicative of ~40% power.
(d.) is incorrect. Delta T is indicative of ~56% power, however, FW flow is indicative of ~90% power.
History:
Developed for use in A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: An
References:
1102.004 (Rev 039-03-0), Power Operations, pages 26, 31 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0202 Source: Direct Rev: 0 Rev Date: 11/23/98 Originator: R. Walters TUOI: AA51002-008 Objective: 8.9 System Number: 039 System
Title:
Main and Reheat Steam System Section: 3.4 Type: RCS Heat Removal
==
Description:==
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: Malfunctioning steam dump.
K/A Number: A2.04 CFR
Reference:
41.5 / 43.5 / 45.3 / 45.13 Point Value: 1 RO Imp: 3.4 SRO Imp: 3.7 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
A plant startup is in progress with the reactor critical below the point of adding heat.
'B' SG Turbine Bypass Valve (CV-6688) fails full open and is unable to be closed with the handjack.
Given the following plant conditions:
- Tave 526 degrees and dropping
- Pressurizer level 205 inches and dropping
- RCS pressure 2120 psig and dropping What is the proper course of action?
- a. Initiate MSLI for the 'B' SG and maintain the reactor critical using 'A' SG TBV to control RCS temperature and pressure.
- b. Continue the reactor startup maintaining startup rate <1 DPM while continuing to monitor primary and secondary plant parameters.
- c. Go directly to the Overcooling tab (1202.003) of the EOP for actions to mitgate the oversteaming of the 'B' SG.
- d. Trip the reactor and go to Reactor Trip tab (1202.001) of the EOP.
Answer:
- d. Trip the reactor and go to Reactor Trip tab (1202.001) of the EOP.
Notes:
(a.) is incorrect. You would not want to isolate a SG and maintain the reactor critical.
(b.) is incorrect. With the reactor below the point of adding heat with a stuck open TBV, this would not be
possible.
(c.) is incorrect. This will be the ultimate tab that you will end up in, however, it is necessary to trip the reactor first and progress through the Reactor Trip tab.
History:
Developed for use in A. Morris 98 RO Re-exam Used in 2001 RO/SRO Exam Difficulty: 3 Taxonomy: C
References:
1102.008 (Rev 018-00-0), Approach to Criticality, page 4 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0205 Source: Direct Rev: 0 Rev Date: 11/24/98 Originator: B. Short TUOI: ANO-1-LP-RO-MSSS Objective: 1.4 System Number: 075 System
Title:
Circulating Water System Section: 3.8 Type: Plant Services System
==
Description:==
Knowledge of circulating water system design feature(s) and interlock(s) which provide for the following: Heat Sink.
K/A Number: K4.01 CFR
Reference:
41.7 Point Value: 1 RO Imp: 2.5 SRO Imp: 2.8 Tier: 2 Group: 2 RO Select: Yes SRO Select: No Question:
During 3 circulating water pump operation, the 'A' circ water pump trips.
The standby circ pump was started and plant conditions have been stabilized.
It is noticed that the condenser waterbox discharge temperature is 10 degrees higher and plant efficency has dropped.
Which of the following is the cause of this condition?
- a. The stopping and starting of a circ pump caused fouling to be removed from the tube sheet promoting better heat transfer capabilities.
- b. The discharge valve on the tripped pump did not go completely closed and circulating water is short cycling.
- c. The debris on the bar grates of the circulating water bays was stirred up during the circ pump swap causing reduced flow.
- d. These are normal conditions following rotation of circulating pumps and temperatures will return to normal within 30 minutes.
Answer:
- b. The discharge valve on the tripped pump did not go completely closed and circulating water is short cycling.
Notes:
(a.) is incorrect. Although some fouling can be removed during pump rotations, it should not result in a 10 degree change in waterbox discharge temperature.
(b.) is correct. The discharge valve on an idle pump can allow a significant amount of backflow from the operating pumps if it is not closed completely.
(c.) is incorrect. This condition is normal for a circ pump swap and may contribute to waterbox fouling, however, the service water system would be affected by this condition as well.
(d.) is incorrect. There should not be such a large temperature difference even if only 3 CW pumps are in service.
History:
Developed for use in A. Morris 98 RO Re-exam Used in 2001 RO Exam.
Difficulty: 3 Taxonomy: C
References:
1104.008 (Rev 02-00-0), Circulating Water System, page11 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0213 Source: Modified Rev: 1 Rev Date: 11/20/00 Originator: S.Pullin TUOI: ANO-1-LP-WCO-VENT Objective: 8 System Number: 034 System
Title:
Fuel Handling Section: 3.8 Type: Plant Service Systems
==
Description:==
Knowledge of new and spent fuel movement procedures.
K/A Number: 2.2.28 CFR
Reference:
43.7 / 45.13 Point Value: 1 RO Imp: 2.6 SRO Imp: 3.5 Tier: 2 Group: 3 RO Select: Yes SRO Select: Yes Question:
Given:
- Refueling outage in progress.
- Fuel handlers are moving fuel from core to Spent Fuel Pool.
- You are assigned to perform Att. B, Refueling Boron, Temperature, and Level Check from 1502.004, Control of Unit 1 Refueling.
While performing Att. B you discover the Spent Fuel Ventilation Exhaust Fan VEF-14A flow rate is high outside of the allowable flow band.
What action should be promptly taken?
- a. Notify fuel handlers to stop fuel movement in the Spent Fuel Pool area.
- b. Stop Spent Fuel Ventilation Exhaust Fan VEF-14A.
- c. Adjust ventilation dampers to restore flow rate below maximum.
- d. No action is required when flow rate is high.
Answer:
- a. Notify fuel handlers to stop fuel movement in the Spent Fuel Pool area.
Notes:
Answer [a] is correct, per Att. B and 1502.004 precaution and limitation, ventilation must be within the normal flow band during fuel movements in Spent Fuel Pool.
Answer [b] is incorrect, stopping the fan will decrease flow but then there will be no ventilation during fuel movement.
Answer [c] is incorrect, this will possibly correct high flow, but this will take local manual action again, the primary response should be to stop fuel movement.
Answer [d] is incorrect, although high flow might not seem like a problem, action is required as stated above.
History:
Developed for use in A. Morris 98 RO Re-exam Modified for use in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1502.004, Rev 031-04-0, Control of Unit 1 Refueling, Att B page 25 Arkansas Nuclear One - Unit One
Initial RO/SRO Exam Question Data QID: 0216 Source: Direct Rev: 0 Rev Date: 11/18/98 Originator: J. Cork TUOI: AA51003-011 Objective: 11.3 System Number: 037 System
Title:
Steam Generator (S/G) Tube Leak Section: 4.2 Type: Generic APE's
==
Description:==
Knowledge of the operational implications of the following concepts as they apply to the Steam Generator Tube Leak:
Leak rate vs. pressure drop.
K/A Number: AK1.02 CFR
Reference:
41.8 / 41.10 / 45.3 Point Value: 1 RO Imp: 3.5 SRO Imp: 3.9 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
Given:
- A 15 gpm Steam Generator tube leak cooldown is in progress
- Normal cooldown limits are being used with the good OTSG.
- RCS pressure is 1000 psig, Tave is 405°F.
The CBOR is maintaining the RCS at about 140°F subcooled.
Why are the CBOR's actions incorrect for this accident?
- a. Tube to shell Delta T limits are being exceeded.
- b. A high primary to secondary Delta P is increasing primary coolant loss.
- c. Excessive thermal stresses are being imposed on the Rx vessel.
- d. Overfill could cause the ruptured SG main steam safeties to lift.
Answer:
- b. A high primary to secondary DP is increasing primary coolant loss.
Notes:
[b] is correct. The operators are directed to maintain RCS pressure low within the limits of Figure 3. This will result in subcooling margin close to the limit and as low as possible primary to secondary differential pressure to prevent loss of unrecoverable primary coolant.
[a] is incorrect, there is not enough information given to determine if tube to shell DT limits are being exceeded.
[c] is incorrect, excessive thermal stresses are not being imposed without a high pressure condition.
[d] is incorrect, overfill could not cause the safeties to lift at 405°F.
History:
Developed for A. Morris 98 RO Re-exam.
Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1202.006 [Rev 007-02-0) Tube Rupture page 11 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0221 Source: Modified Rev: 1 Rev Date: 11/16/00 Originator: B. Short TUOI: ANO-1-LP-RO-RPS Objective: 11 System Number: 012 System
Title:
Reactor Protection System Section: 3.7 Type: Instrumentation
==
Description:==
Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b)
based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of instrument power.
K/A Number: A2.02 CFR
Reference:
41.5 / 43.5 / 45.3 / 45.5 Point Value: 1 RO Imp: 3.6 SRO Imp: 3.9 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
Given:
- "A" RPS is in Channel Bypass due to RCS pressure transmitter failure.
- Plant is at 100% power and stable.
- Subsequently, a lightning strike on the Reactor Building has resulted in the trip of 120V Vital AC distribution panel RS-3.
What should your action be in response to this event?
- a. Go to 1202.001, Reactor Trip EOP in response to automatic reactor trip.
- b. Immediately remove "A" RPS channel from Channel Bypass.
- c. Perform 1203.045, Rapid Plant Shutdown, to comply with Tech Specs.
- d. Make a station log entry and take action to restore power to RS-3.
Answer:
- d. Make a station log entry and take action to restore power to RS-3.
Notes:
Answer [d] is correct, RPS is in 2/3 logic and only channel "C" is tripped with a loss of power to RS-3.
Answer (a) is incorrect, only one channel is tripped and it still takes two channels to trip the reactor in this condition.
Answer (b) is incorrect, removing "A" channel from bypass would result in an automatic reactor trip.
Answer (c) is incorrect, given conditions do not meet entry conditions for 1203.045 or TS LCO.
History:
Developed for A. Morris 98 RO Re-exam Modified for use in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: An
References:
STM 1- 63, Rev. 4, Reactor Protection System, page 13,14, & 17 1015.001, rev. 052-05-0, Conduct of Operations, page 27, 28, 29 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0225 Source: Modified Rev: 1 Rev Date: 12/1/00 Originator: B. Short TUOI: ANO-1-LP-RO-EDG Objective: 23 System Number: 064 System
Title:
Emergency Diesel Generators (ED/G)
Section: 3.6 Type: Electrical
==
Description:==
Ability to (a) predict the impacts of the following malfunctions or operations on the EDG System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Synchronization of the ED/G with other electrical power supplies.
K/A Number: A2.09 CFR
Reference:
41.5 / 43.5 / 45.3 / 45.13 Point Value: 1 RO Imp: 3.1 SRO Imp: 3.3 Tier: 2 Group: 2
RO Select: Yes SRO Select: No Question:
During the performance of the normal DG1 monthly surveillance test, while the CBOT is paralleling the diesel to the grid, he accidentally goes to closed on the output breaker with the synchroscope at the 15 minute before 12:00 position.
What would be the consequences of the CBOT's action?
- a. A protective feature will prevent the output breaker from closing in.
- b. The output breaker will close in and the diesel will pick up load.
- c. The output breaker will close in and immediately trip back open.
- d. Breaker will remain open due to a lockout relay trip.
Answer:
- a. A protective feature will prevent the output breaker from closing in.
Notes:
Answer (a) is correct. The sync check protective relay protects the diesel from tying on out of phase.
Answers (b) & (c) are incorrect. Although there are other features associated with the diesel circuitry that could cause these two conditions to occur, the sync check protective relay will prevent the breaker from closing in at all.
Answer (d) is incorrect. Paralleling the diesel to the grid out of phase does not directly result in a lockout relay trip.
History:
Modified for use on 2001 RO Exam.
Difficulty: 3 Taxonomy: C
References:
1104.036, Rev. 039-03-0, Emergency Diesel Generator Operation Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0231 Source: Modified Rev: 1 Rev Date: 11/20/00 Originator: J.Cork TUOI: AA30001-005 Objective: 3 System Number: 2.2 System
Title:
Equipment Control Section: 2.0 Type: Generic K/As
==
Description:==
Knowledge of tagging and clearance procedures.
K/A Number: 2.2.13 CFR
Reference:
41.10 / 45.13 Point Value: 1 RO Imp: 3.6 SRO Imp: 3.8 Tier: 3 Group: G RO Select: Yes SRO Select: Yes Question:
Which of the following conditions is correct with regard to preparation and installation authorization of a common unit tagout?
- a. Installation may be authorized by either the Unit 1 or the Unit 2 Operations Supervisor.
- b. Preparers and reviewers from both units must be licensed operators.
- c. Preparer and reviewer may be non-licensed if authorized by both Unit Operations Supervisors.
- d. Preparer and reviewer may be non-licensed if the opposite unit reviewer is licensed.
Answer:
- b. Preparers and reviewers from both units must be licensed operators.
Notes:
Answer [b] is correct, procedure requires both the preparer and the reviewer on the unit preparing the tagout have to be licensed.
Answer [a] is incorrect, a common unit tagout requires both Unit's Operations Supervisors to approve it.
Answer (c) is incorrect, both Unit Ops Supervisors must approve but the preparation & review must be done by licensed operators.
Answer [d] is incorrect, the preparation & review must be done by licensed operators on their respective units.
History:
Developed for use on A. Morris 98 RO Re-exam Modified for use in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1000.027, Rev. 026-01-0, Protective Tagging Control, Att. A Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0239 Source: Direct Rev: 0 Rev Date: 11/21/98 Originator: J. Haynes TUOI: ANO-1-LP-RO-EOP05 Objective: 3 System Number: 011 System
Title:
Large Break LOCA Section: 4.1 Type: Generic EOP
==
Description:==
Ability to determine the following as they apply to a Large Break LOCA: Actions to be taken based on RCS temperature and pressure - saturated and superheated.
K/A Number: EA2.01 CFR
Reference:
43.5 / 45.13 Point Value: 1 RO Imp: 4.2 SRO Imp: 4.7 Tier: 1 Group: 1 RO Select: No SRO Select: Yes Question:
Attempts to mitigate an ICC condition using 1202.005, Inadequate Core Cooling, are in progress.
Currently RCS CET's indicate 900°F and RCS pressure is 1400 psig.
Which of the following is the required action for the current conditions?
- a. Bypass start interlocks and start all RCPs.
- b. Restore RCP services and start one RCP/loop.
- c. Bump RCPs until primary to secondary heat transfer is established even if RCP services are not available.
- d. Depressurize the RCS by opening ERV and bypass ESAS if it has not already activated as RCS pressure drops below 1700 psig.
Answer:
- b. Restore RCP services and start one RCP/loop.
Notes:
(a.) & (d.) are incorrect. These are entry actions for Region 4 not Region 3 (b.) is correct. 1202.005 step 8 (c.) is incorrect. RCPs are bumped in Region 2 only if RCP services are available.
History:
Developed for use in previous exam.
Used in 2001 SRO Exam.
Difficulty: 4 Taxonomy: C
References:
1202.005, Rev 004-00-0, Inadequate Core Cooling Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0254 Source: Direct Rev: 0 Rev Date: 9-2-99 Originator: D. Slusher TUOI: ANO-1-LP-RO-EFW Objective: 13 System Number: 061 System
Title:
Auxiliary/Emergency Feedwater System Section: 3.4 Type: Heat Removal From Reactor Core
==
Description:==
Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: Pumps.
K/A Number: K6.02 CFR
Reference:
41.7 / 45.7 Point Value: 1 RO Imp: 2.6 SRO Imp: 2.7 Tier: 2 Group: 1 RO Select: Yes SRO Select: Yes Question:
Given:
- EFW started 10 minutes ago
- EFW pump P-7A speed is 900 RPM Which of the following would cause these indications?
- a. EFW Pump P-7A governor valve has lost power.
- b. EFW Pump P-7A trip/throttle valve does not indicate full open.
- c. EFW steam admission valve CV-2613 is closed.
- d. EFW Pump P-7A governor valve has an oil leak.
Answer:
- b. EFW Pump P-7A trip/throttle valve does not indicate full open.
Notes:
The trainee should conclude that EFW pump speed is below normal.
"a" is incorrect since the governor fails full open on a loss of power and this does not support the low speed condition.
"c" is incorrect since either steam admission valve will bring the pump up to full speed.
"d" is incorrect since a leak in the governor will only serve to reduce hydraulic pressure and increase speed.
"b" is the only correct answer since ramp initiate requires the trip/throttle valve to be full open and at least one steam admission valve greater than 90% open.
History:
Developed for 1999 exam.
Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: A
References:
1106.006 Rev 060-00-0 STM1-27 Rev 2 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0267 Source: Direct Rev: 0 Rev Date: 9-2-99 Originator: D. Slusher TUOI: ANO-1-LP-RO-MU Objective: 10 System Number: 003 System
Title:
Reactor Coolant Pump System Section: 3.4 Type: Heat Removal From Reactor Core
==
Description:==
Ability to monitor automatic operation of the RCPs, including: Seal Injection flow K/A Number: A3.01 CFR
Reference:
41.7 / 45.5
Point Value: 1 RO Imp: 3.3 SRO Imp: 3.2 Tier: 2 Group: 1 RO Select: Yes SRO Select: Yes Question:
The CBOR observes a change in seal injection flow rates and notes the following values:
"A" RCP 6.5 gpm "B" RCP 15.0 gpm "C" RCP 5.0 gpm "D" RCP 6.0 gpm Which of the following explains the seal injection flow indications?
- a. Reactor Coolant Pump P-32B trip due to a motor fault.
- b. Seal injection line break in the Upper North Piping Penetration Room .
- c. "B" Reactor Coolant Pump seal cooler leak.
- d. "B" seal injection flow transmitter failure.
Answer:
- c. "B" Reactor Coolant Pump seal cooler leak.
Notes:
"a" is incorrect because flow from B RCP would drop.
"b" is incorrect because all seal injection flows would drop to zero.
"c" is correct, a seal cooler leak lowers the seal pressure and "B" seal injection flow raises. The others lower because total seal injection flow is maintained at setpoint.
"d" is incorrect because a transmitter failure will not lower the other seal injection flows.
History:
Developed for 1999 exam.
Used in 2001 RO/SRO Exam.
Difficulty: 2.5 Taxonomy: An
References:
1203.039, Rev. 005-01-0, Excess RCS Leakage, pages 3,4 STM 1-03 rev. 8 ch.1 Reactor Coolant System, pages 34, 35, 36 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0271 Source: Direct Rev: 0 Rev Date: 9-2-99 Originator: D. Slusher TUOI: ANO-1-LP-RO-RMS Objective: 2 System Number: 073 System
Title:
Process Radation Monitoring System (PRM)
Section: 3.7 Type: Instrumentation
==
Description:==
Knowledge of the effect of a loss or malfunction of the PRM system will have on the following:
Radioactive effluent releases.
K/A Number: K3.01 CFR
Reference:
41.7 / 45.6 Point Value: 1 RO Imp: 3.6 SRO Imp: 4.2 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
Which of the following must be performed to release T-16A contents with the
Liquid Radwaste Process Monitor (RI-4642) inoperable?
- a. Chemistry personnel must estimate radiation level every four hours during the release.
- b. A Waste Control Operator must independently verify release path alignment prior to release.
- c. The release flow rate must be estimated at least once every three hours during the release.
- d. Discharge Flume process monitor RI-3618 must be checked for operability.
Answer:
- b. A Waste Control Operator must independently verify release path alignment prior to release.
Notes:
The requirements for release when the Liquid Radwaste Process Monitor is inoperable are
- a. An independent verification of the release path by a person qualified as Waste Control Operator.
- b. An independent sample and analysis of the tank contents
- c. Computer input data independently verified.
"b" is the correct answer.
History:
Used in 1999 exam.
Direct from ExamBank, QID# 2765 Used in 2001 RO/SRO Exam.
Difficulty: 2.5 Taxonomy: K
References:
1104.020 Rev 039-03-0, Clean Waste System Operation, pages 3 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0272 Source: Direct Rev: 0 Rev Date: 9-2-99 Originator: D. Slusher TUOI: ANO-1-LP-RO-AOP Objective: 3 System Number: 071 System
Title:
Waste Gas Disposal System (WGDS)
Section: 3.9 Type: Radioactivity Release
==
Description:==
Knowledge of the purpose and function of major system components and controls.
K/A Number: 2.1.28 CFR
Reference:
41.7 Point Value: 1 RO Imp: 3.2 SRO Imp: 3.3 Tier: 2 Group: 1 RO Select: Yes SRO Select: Yes Question:
When a high radiation condition occurs in the Waste Gas Discharge Header, the radiation monitor will cause what combination of automatic action(s) to occur?
- 1. Nitrogen is added for dilution.
- 2. The Aux. Building Vent Header diverts to the Waste Gas Surge Tank.
- 3. The Waste Gas Decay Tank effluent control valve (CV-4820) shuts.
- 4. The Aux. Building Vent Header diverts to the Waste Gas Decay Tank in service.
- a. 1 and 2
- b. 2 and 3
- c. 3 and 4
- d. 1 and 4 Answer:
- b. 2 and 3
Notes:
"b" is correct "a", "c", and "d" are incorrect because adding nitrogen is a manual operation and the ABVH is diverted to the Waste Gas Surge Tank.
History:
Used in 1999 exam.
Direct from ExamBank, QID# 1399 Used in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1203.006, Rev. 007-02-0, Waste Gas Discharge Line Radiation page 1 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0274 Source: Direct Rev: 0 Rev Date: 9-2-99 Originator: D. Slusher TUOI: ANO-1-LP-RO-RBS Objective: 5 System Number: 022 System
Title:
Containment Cooling System Section: 3.5 Type: Containment Integrity
==
Description:==
Ability to manually operate and/or monitor in the control room: Dampers in the CCS K/A Number: A4.03 CFR
Reference:
41.7 / 45.5 to 45.8 Point Value: 1 RO Imp: 3.2 SRO Imp: 3.2 Tier: 2 Group: 1 RO Select: Yes SRO Select: No Question:
What would be the consequences if the Reactor Building Cooler Chilled Water Bypass Dampers remained latched after an ESAS actuation?
- a. Damage to RB ventilation plenum from excessive pressure
- b. Excessive heat load on the Chilled Water System
- c. Inadequate air flow through the Service Water Cooling Coils
- d. Excessive current on the cooling fan motors Answer:
- c. Inadequate air flow through the Service Water Cooling Coils Notes:
"a" is incorrect, RB ventilation plenum has been analyzed for these conditions and it will withstand the pressures after an ESAS.
"b" is incorrect, the Chilled Water System is isolated on ESAS and therefore no additional heat load will be placed on it.
"d" is incorrect, the current on the motors is not a concern in this situation.
"c" 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.
History:
Developed for 1999 exam.
Used in 2001 RO Exam.
Difficulty: 1.5 Taxonomy: K
References:
1104.033, Rev. 058-00-0, Reactor Building Ventilation, pages 3, 4
Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0281 Source: Direct Rev: 0 Rev Date: 9-3-99 Originator: D. Slusher TUOI: ANO-1-LP-RO-MSSS Objective: 3 System Number: 062 System
Title:
Loss of Nuclear Service Water Section: 4.2 Type: Generic AOP
==
Description:==
Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: The automatic actions (alignments) within the nuclear service water resulting from the actuation of the ESFAS.
K/A Number: AK3.02 CFR
Reference:
41.4, 41.8 / 45.7 Point Value: 1 RO Imp: 3.6 SRO Imp: 3.9 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
Service Water Pumps P-4A, P-4B (supplied from A-4), and P-4C are running. An ES actuation coincident with a loss of off-site power occurs.
Which service water pumps will autostart when A-3 and A-4 are re-energized?
- a. P-4A, P-4B and P-4C
- b. P-4A and P-4B
- c. P-4B and P-4C
- d. P-4A and P-4C Answer:
- d. P-4A and P-4C Notes:
When ESAS actuates and the buses are re-energized the P-4A and P-4C handswitch position will interlock P-4B and keep P-4B from starting. Therefore, "a", "b", and "c" responses are incorrect.
History:
Developed for 1999 exam.
Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1203.012J, Rev. 034-00-0, Annunicator K11 Corrective Action, page 27 STM 1-42, Rev. 4, Ch. 3, Service and Auxiliary Cooling Water, page 13, 14 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0297 Source: New Rev: 0 Rev Date: 11/20/00 Originator: J.Cork TUOI: ANO-1-LP-RO-TURB Objective: 9E System Number: 045 System
Title:
Main Turbine Generator (MT/G) System Section: 3.4 Type: Heat Removal From Reactor Core
==
Description:==
Ability to manually operate and/or monitor in the control room: Turbine valve indicators (throttle, governor, control stop, intercept) alarms, and annunciators.
K/A Number: A4.01 CFR
Reference:
41.7 / 45.5 to 45.8 Point Value: 1 RO Imp: 3.1 SRO Imp: 2.9 Tier: 2
Group: 3 RO Select: Yes SRO Select: Yes Question:
While performing the Overspeed Test during Turbine startup, the operator mistakenly places the Overspeed Protection Controller (OPC) switch in the OPC Test position instead of the Overspeed Test position.
What will this mistake cause?
- b. Close the Throttle and Governor Valves only.
- c. Close the Governor Valves and the Reheat Intercept Valves.
- d. Close all Throttle, Governor, Reheat Intercept, and Reheat Stop Valves.
Answer:
- c. Close the Governor Valves and the Reheat Intercept Valves.
Notes:
Answer [c] is correct, the OPC Test switch simulates an OPC condition.
Answer [a] is incorrect since the Overspeed Protection Controller actuation is blocked when the the OPC test switch is placed in the Overspeed Test position.
Answer [b] is incorrect, these are the wrong valves.
Answer [d] is incorrect, this is essentially a Turbine Trip.
History:
New question for use in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1106.009, Rev. 029-03-0, Turbine Startup, Warmup, and Roll, page 28 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0309 Source: Direct Rev: 0 Rev Date: 9-5-99 Originator: R Cool TUOI: ANO-1-LP-RO-ICS Objective: 40 System Number: 016 System
Title:
Non-Nuclear Instrumentation System Section: 3.7 Type: Instrumentation
==
Description:==
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.
K/A Number: A2.01 CFR
Reference:
41.5 / 43.5 / 45.3 / 45.5 Point Value: 1 RO Imp: 3.0 SRO Imp: 3.1 Tier: 2 Group: 2 RO Select: Yes SRO Select: No Question:
Given:
- The plant is operating at 100% power.
- Loop "A" T-cold Narrow Range Temperature instrument fails HIGH.
If this instrument was hard selected by the SASS selector switch, what ICS HAND/AUTO stations should be placed in HAND?
- a. Reactor Demand and both Feedwater Loop Demands.
- b. SG/Rx Master and Reactor Demand.
- c. SG/Rx Master and both Feedwater Loop Demands.
Answer:
- a. Reactor Demand and both Feedwater Loop Demands.
Notes:
A cold leg temperature instrument failure causes the reactor demand signal to drive rods inward due to a high indicated Tave. Feedwater flows are changed to balance loop cold leg temperatures. Therefore reactor demand and feedwater loop demand stations must be taken to manual. "a" is the correct answer.
"b" is incorrect because feedwater is affected downstream of the SG/Rx Master.
"c" is incorrect because reactor demand is affected downstream of the SG/Rx Master.
"d" is incorrect because the turbine is not affected.
History:
Used in 1999 exam.
Direct from ExamBank, QID# 2871 Used in 2001 RO Exam.
Difficulty: 3 Taxonomy: An
References:
1105.006 rev. 009-00-0, Reactor Coolant System NNI, page 5, 6, 1105.004 rev. 013-02-0, Intergrated Control System, page 8 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0312 Source: Modified Rev: 1 Rev Date: 11/16/00 Originator: J Cork TUOI: ANO-1-LP-RO-SFC Objective: 2 System Number: 033 System
Title:
Spent Fuel Pool Cooling System Section: 3.8 Type: Plant Service Systems
==
Description:==
Knowledge of the effect that a loss or malfunction of the Spent Fuel Pool Cooling system will have on the following: Area ventilation systems.
K/A Number: K3.01 CFR
Reference:
41.5 / 45.5 Point Value: 1 RO Imp: 2.6 SRO Imp: 3.1 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
The WCO reports the Spent Fuel Pool level is +1.5 ft.
What problem could this level pose for Spent Fuel Pool operations or fuel handling in the SFP?
- a. SFP minmum water temperature limit will be exceeded.
- b. SFP ventilation ducts will be flooded.
- c. Area dose rates will rise.
- d. SFP must be sampled within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
Answer:
- b. SFP ventilation ducts will be flooded.
Notes:
Answer [b] is correct since normal level is 0 ft with a maximum allowable level of +1.0 ft which prevents water carryover into the ventilation ducts.
Answer [a] is incorrect because this answer is associated with SF cooling capacity which is largely unaffected by pool level.
Answer [c] is incorrect since this problem is associated with a low water level.
Answer [d] is incorrect but plausible since the time for sampling is correct but level is greater than maximum allowed.
History:
Developed for 1999 exam.
Modified for 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
STM 1-7, Rev. 2 Ch. 1, Spent Fuel Cooling System, page 2 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0315 Source: Modified Rev: 2 Rev Date: 11/16/00 Originator: R Soukup TUOI: ANO-1-LP-AO-VAC Objective: 17 System Number: 055 System
Title:
Condenser Air Removal System Section: 3.4 Type: RCS Heat Removal
==
Description:==
Knowledge of CARS design feature(s) and/or interlock(s) which provide for the following:
Turbine Startup.
K/A Number: K4.01 CFR
Reference:
41.7 Point Value: 1 RO Imp: 1.9 SRO Imp: 2.3 Tier: 2 Group: 2 RO Select: Yes SRO Select: No Question:
Given:
Plant is stablized at ~17% power in preps for placing the Main Turbine on line.
Condenser vacuum is 27.5 in Hg and slowly trending down.
CBOT is in the process of swapping from TV control to GV control.
What action would be taken when condenser vacuum reaches ~26.5 in Hg?
- a. Continue with TV/GV transfer. No action required.
- b. Trip the operating MFW Pump Turbine.
- c. Trip the Main Turbine and lower power.
- d. Adjust programmable alarm setpoint to ~25 in Hg.
Answer:
- c. Trip the Main Turbine and lower power.
Notes:
Answer [c} is correct, a manual trip of the turbine is required when vacuum reaches 26.5 in Hg when turbine load is < 270 Mwe.
Answer [a] is incorrect, continued operation in this condition can lead to turbine blading damage.
Answer [b] is incorrect, this action is not taken until vacuum drops to ~5" Hg.
Answer [d] is incorrect, this action would be taken if operating >270 Mwe to alert operator of approaching trip criteria of 24.5 in Hg.
History:
Developed for 1999 exam.
Modified for 2001 RO exam.
Difficulty: 2.5 Taxonomy: K
References:
1203.016 rev. 011-00-0, Loss of Condenser Vacuum, page 1 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0322 Source: Direct Rev: 0 Rev Date: 9-6-99 Originator: D. Slusher TUOI: ANO-1-LP-RO-ICS Objective: 14 System Number: 045 System
Title:
Main Turbine Generator
Section: 3.4 Type: RCS Heat Removal
==
Description:==
Knowledge of the MT/G system design feature(s) and/or interlock(s) which provide for the following: Automatic turbine runback.
K/A Number: K4.12 CFR
Reference:
41.7 Point Value: 1 RO Imp: 3.3 SRO Imp: 3.6 Tier: 2 Group: 1 RO Select: Yes SRO Select: Yes Question:
A reactor coolant pump trip has caused a plant runback.
What ensures ICS maintains power steady (does not return to its previous load demand) when the runback is complete?
- a. The Unit Master H/A station input tracks the Rate and Load Limited Megawatt demand signal.
- b. The ICS runback demand signal is fed directly into the input of the Unit Master H/A station.
- c. The input to the Unit Master H/A station is driven by cross limits to match the runback back demand signal.
- d. The ICS runback signal will clear only when the Unit Master H/A station output equals actual generated megawatts.
Answer:
- a. The Unit Master H/A station input tracks the Rate and Load Limited Megawatt demand signal.
Notes:
(a) is correct. The runback demand signal from the rate and load limit circuit is fed back into the input of the Unit Master to force the Unit Master to track the runback load demand signal.
(b) is incorrect because the runback limit is fed into the high load limit.
(c) is incorrect. While cross limits may come into effect during a runback, they do not drive the input of the Unit Master.
(d) is incorrect since the runback signal will clear when demand is less than the runback limit.
History:
Developed for 1999 exam.
Used in 2001 RO/SRO Exam.
Difficulty: 2.5 Taxonomy: C
References:
STM1-64 Rev 6, Intergrated Control System, page 23, 24 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0325 Source: Modified Rev: 1 Rev Date: 11/16/00 Originator: J. Cork TUOI: ANO-1-LP-RO-EOP01 Objective: 2 System Number: 011 System
Title:
Pressurizer Level Control Section: 3.2 Type: RCS Inventory Control
==
Description:==
?Ability to (a) predict the impacts of the following malfunctions or operations on the PZR LCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Isolation of letdown.
K/A Number: A2.07 CFR
Reference:
41.5 / 43.5 / 45.3 / 45.13 Point Value: 1 RO Imp: 3.0
SRO Imp: 3.3 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
Following a reactor trip, what pressurizer level value (dropping) requires isolation of letdown per 1202.001?
- a. 110 inches
- b. 90 inches
- c. 55 inches
- d. 30 inches Answer:
- c. 55 inches Notes:
Answer [c] is correct, this is the heater cutoff value and the value for isolation of letdown.
Answr [a] is incorrect, this is the upper limit of the normal control band prescribed in the Rx Trip EOP.
Answer [b] is incorrect, this is the lower limit of the normal control band prescribed in the Rx Trip EOP.
Answer [d] is incorrect, this setpoint is the off-scale low value and requires initiation of HPI to offset the loss of inventory and to attempt to sustain an adequate subcooling margin.
History:
Used in 1999 exam.
Modified from ExamBank, QID# 7742.
Modified for use in 2001 RO/SRO Exam.
Difficulty: 2.5 Taxonomy: K
References:
1202.001, Rev 27, Reactor Trip, page 14, and 19 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0328 Source: Direct Rev: 0 Rev Date: 9-6-99 Originator: D. Slusher TUOI: ANO-1-LP-RO-EOP01 Objective: 2 System Number: 029 System
Title:
Anticipated Transient Without Scram Section: 4.1 Type: Generic Emergency Plant Evolutions
==
Description:==
Knowledge of system setpoints/interlocks and automatic actions associated with EOP entry conditions.
K/A Number: 2.4.2 CFR
Reference:
41.7 / 45.7 / 45.8 Point Value: 1 RO Imp: 3.9 SRO Imp: 4.1 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
You are the CBOR and you observe the following indications:
"A" and "B" Main Feedwater Pumps are tripped CRD groups 1, 2, 3, and 4 are at the out limit.
CRD groups 5, 6, and 7 are at the in limit.
NI-3 indicates 1 E-8 and lowering.
What action should be performed FIRST?
- a. Depress the CRD Power Supply Breaker Trip Pushbuttons.
- c. Commence Emergency Boration per RT-12.
- d. Manually insert CRD groups 1, 2, 3, and 4.
Answer:
- a. Depress the CRD Power Supply Breaker Trip Pushbuttons.
Notes:
The conditions given are indicative of an ATWS since an anticipatory Rx Trip should have occurred (both MFW pumps tripped at 100% power) and a partial trip did occur (5,6,7 at in limit and Intermediate Range NI power decreasing), however all of the safety groups are at the out limit.
All of the answers given are correct contingency actions to a failure of the Reactor to trip, however, "a" is the preferred and most expedient action to drop the safety groups. Answers "b" and "d" are not performed unless "a" was unsuccessful.
History:
Developed for 1999 exam.
Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1202.001 Rev 27, Reactor Trip, page 2 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0329 Source: Modified Rev: 1 Rev Date: 11/7/00 Originator: D. Slusher TUOI: ANO-1-LP-RO-NI Objective: 10 System Number: 033 System
Title:
Loss of Intermediate Range Nuclear Instrumentation Section: 4.2 Type: Generic Abnormal Plant Evolutions
==
Description:==
Knowledge of the reasons for the following responses as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Guidance contained in EOP for loss of intermediate range instrumentation.
K/A Number: AK3.02 CFR
Reference:
41.5 / 41.10 / 45.6/ 45.13 Point Value: 1 RO Imp: 3.6 SRO Imp: 3.9 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
Given:
Plant startup in progress NI501 at 9 x E4 cps NI502 at 1 x E5 cps NR502 is operable and at 5 x E-2% power NI3 at 2 x E-10 amps NI4 at 5 x E-10 amps NI5 thru 8 at 0%
Subsequently NI3 fails low.
What action should be taken by control room operators?
- a. Maintain flux level in the source range
- b. Trip the reactor
- c. Continue with startup
- d. Stabilize power at 1 x E-8 amps Answer:
- c. Continue with startup Notes:
The Intermediate Range NI values are on scale and represent one decade of overlap with the Source Range indications. Therefore with IR NI3 failed, the startup may continue as in answer (c) in accordance with 1203.021.
(a) is incorrect, this action would only be taken if if less than one decade of overlap existed.
(b) is incorrect, although this action would be taken IAW 1203.021 if no on scale flux indication existed.
(d) is incorrect, power can continue up to 10-8 amps.
History:
Used in 1999 exam Direct from ExamBank, QID# 3099 used in class exam Modified for use in 2001 RO/SRO Exam.
Difficulty: 4 Taxonomy: Ap
References:
1203.021, Rev. 007-01-0, Loss of Neutron Flux Indication, page 5, 6 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0337 Source: Direct Rev: 0 Rev Date: 9-7-99 Originator: D Slusher TUOI: ANO-1-LP-RO-EOP10 Objective: 6 System Number: 011 System
Title:
Large Break LOCA Section: 4.1 Type: Generic Emergency Plant Evolutions
==
Description:==
Knowledge of the interrelations between the following and the Large Break LOCA: Pumps.
K/A Number: EK2.02 CFR
Reference:
41.7 / 45.7 Point Value: 1 RO Imp: 2.6 SRO Imp: 2.7 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
ESAS has actuated.
LPI/HPI flow rates for the past ten minutes have been as follows:
"A" LPI flow--2900 gpm "B" LPI flow--2850 gpm "A" HPI pump flow throttled to 100 gpm through CV-1220 "C" HPI pump flow throttled to 100 gpm through CV-1285 An overcurrent has resulted in an A-3 bus lockout and A-1 to A-3 tie breaker A-309 trip. The operator should:
- b. close A-308 to power A-3 from #1 EDG.
- c. energize bus B-5 from bus B-6
- d. start P-36B to supply 100 gpm train through CV-1220.
Answer:
Notes:
"a" is the only correct action with the loss of the A-3 bus and the "A" LPI pump. The criteria for throttling HPI is contingent upon both LPI pumps flow >2800 gpm OR one LPI pump >3200 gpm, therefore full HPI flow must be restored on the only running HPI pump. "b" and "c" are actions to restore proper electrical alignment but they do not address the immediate need of maintaining proper core cooling. "d" is incorrect since flow is only supplied at the 100 gpm rate.
History:
Used in 1999 exam Direct from ExamBank, QID# 4566 used in class exam Used in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1202.010 Rev. 005-00-0, ESAS, page 7, 8 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0347 Source: Direct Rev: 0 Rev Date: 9-7-99 Originator: G. Alden TUOI: ANO-1-LP-RO-FH Objective: 19 System Number: A08 System
Title:
Refueling Canal Level Decrease Section: 4.3 Type: B&W EOP/AOP
==
Description:==
Knowledge of the operational implications as they apply to the (Refueling Canal Level Decrease): Normal, abnormal, and emergency operating procedures associated with (Refueling Canal Level Decrease).
K/A Number: AK1.2 CFR
Reference:
41.8 / 41.10 / 45.3 Point Value: 1 RO Imp: 3.7 SRO Imp: 4.0 Tier: 1 Group: 3 RO Select: Yes SRO Select: Yes Question:
The main fuel bridge has a spent fuel assembly in route to the RB upender when a seal plate NI cover failure occurs.
Water level in the canal is falling at two inches per minute.
The main fuel bridge operator should:
- a. Continue to the upender and insert the assembly for transport to the SFP.
- b. Leave the fuel assembly in the mast and evacuate the area.
- c. Place the assembly in the fuel rack in the deep end of the canal.
- d. Return the assembly to any available location in the reactor vessel.
Answer:
- d. Return the assembly to any available location in the reactor vessel.
Notes:
In this scenario the fuel transfer canal level is decreasing rapidly and thus shielding for the spent fuel assembly will be decreasing rapidly and the fuel assembly must be placed in an area that will remain covered with water after the canal is drained.
Therefore, "d" is the only correct answer. "a" is incorrect as this is a time consuming maneuver and the transfer tube should be isolated anyway to prevent loosing level in the SFP. "b" is very incorrect since this will expose the assembly to atmosphere and dose rates will be lethal for quite some time. "c" is incorrect since the deep end will not contain enough water to keep the assembly covered.
History:
Used in 1999 exam.
Direct from ExamBank, QID# 4282 used in class exam Used in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1203.042 Rev 005-00-0, Refueling Abnormal Operations, page 7
Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0350 Source: New Rev: 0 Rev Date: 11/21/00 Originator: S.PULLIN TUOI: ANO-1-LP-RO-ESAS Objective: 6 System Number: 013 System
Title:
Engineered Safety Features Actuation System Section: 3.2 Type: Reactor Coolant System Inventory Control
==
Description:==
Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based ability on those predictions use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of instrument bus.
K/A Number: A2.04 CFR
Reference:
41.5 / 43.5 / 45.3 / 45.13 Point Value: 1 RO Imp: 3.6 SRO Imp: 4.2 Tier: 2 Group: 1 RO Select: No SRO Select: Yes Question:
Given, the plant is operating at 100% power.
ESAS Analog 2 RC pressure transmitter fails LOW due to loss of instrument power.
What operator action will allow continued plant operation at 100% power?
- a. Initiate administrative controls to document and correct the failure.
- b. Continued power operation is not allowed, plant shutdown is required.
- c. Immediately trip one of the two remaining operable channels.
- d. Test ES components associated with Analog Channel 2 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Answer:
- a. Initiate administrative controls to document and correct the failure.
Notes:
"a" is correct per Tech Spec table 3.5.1-1, Note 6.
"b" is incorrect, with only one inoperable channel, plant shutdown is not required.
"c" is incorrect, this action would result in ES actuation.
"d" is incorrect, testing of ES components is not required.
History:
Developed for 2001 SRO exam Difficulty: 3.5 Taxonomy: An
References:
Tech. Spec 3.5.1-1 table note 6 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0363 Source: New Rev: 0 Rev Date: 11/6/00 Originator: S.Pullin TUOI: ANO-1-LP-RO-AOP Objective: 3 System Number: E09 System
Title:
Natural Circulation Cooldown Section: 4.3 Type: B&W EPEs/APEs
==
Description:==
Knowledge of the reasons for the following responses as they apply to the (Natural Circulation Cooldown): Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.
K/A Number: EK3.1 CFR
Reference:
41.5 / 41.10, 45.6, 45.13 Point Value: 1 RO Imp: 3.2 SRO Imp: 3.4
Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
Given:
- Natural circulation cooldown in progress
- CETs at 550°F
- Reactor vessel head temperatures at 614°F
- Pressurizer level = 150 inches, then makes step change to 180 inches
- RCS pressure at 1700 psig and slowly dropping
- "A" OTSG pressure = 945 psig
- "B" OTSG pressure = 950 psig The required operator action for the above conditions is to pressurize the RCS slightly and reduce cooldown rate per 1203.013, Natural Circulation Cooldown.
What is the reason for this action?
- a. Reduce the thermal stresses on the Reactor Vessel.
- b. Restore adequate Subcooling Margin.
- c. Collapse a steam void in the Rx Vessel head.
- d. Comply with SG tube to shell delta-T limits.
Answer:
- c. Collapse a steam void in the Rx Vessel head.
Notes:
Answer (c) is correct since this action is due to a steam void in the upper head as evidenced by the sudden change in Pzr level with no RCS pressure increase.
Answer (a) would be the proper response for PTS concerns.
Answer (b) is not applicable, subcooling margin is adequate with the values given.
Answer (d) only applies when the OTSGs are not being used to cool the RCS.
History:
Developed for 2001 RO/SRO Exam Difficulty: 3 Taxonomy: C
References:
1203.013 (Rev. 016-02-0), Natural Circulation Cooldown, page 8 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0364 Source: New Rev: 0 Rev Date: 11/8/00 Originator: J.Cork TUOI: ANO-1-LP-RO-EOP06 Objective: 1 System Number: 038 System
Title:
Steam Generator Tube Rupture Section: 4.1 Type: Generic EPEs
==
Description:==
Ability to determine and interpret the following as they apply to the SGTR: Existence of an S/G tube rupture and its potential consequences.
K/A Number: EA2.02 CFR
Reference:
43.5 / 45.13 Point Value: 1 RO Imp: 4.5 SRO Imp: 4.8 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
After a reactor trip, the following indications are observed:
Makeup Tank level has lost 5 inches in the last 5 minutes
RB and Aux. Bldg. Sump levels are stable "A" EFIC level is 35 rising and "A" MFW Flow is .1 mlb/hr "B" EFIC level is 31 stable and "B" MFW Flow is .3 mlb/hr Which of the following actions would be required to minimize the threat of a potential radioactive release to the public?
- b. Cooldown and isolate the "B" SG
- c. Cooldown and isolate the "A" SG
- d. Commence a rapid RCS cooldown at 240 F/hr Answer:
- c. Cooldown and isolate the "A" SG Notes:
Answer [c] is correct, the SG level parameters indicate a rupture on the "A" SG and a cooldown should be commenced to reduce RCS temperature to <500 F to minimize the possibility of lifting a secondary safety on the "A" SG.
[a] is incorrect, the leak size is 30 gpm, this is within the capacity of normal makeup.
[b] is incorrect, a cooldown and isolation is required but not on this SG.
[d] is incorrect, a rapid cooldown at this rate is not required until overfilling of ruptured SG is imminent.
History:
Created for 2001 RO/SRO Exam.
Difficulty: 4 Taxonomy: An
References:
1202.006, Rev. 007-02-0, Tube Rupture, page 3 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0365 Source: New Rev: 0 Rev Date: 11/8/00 Originator: S.Pullin TUOI: ANO-1-LP-RO-EOP04 Objective: 3 System Number: 054 System
Title:
Loss of Main Feedwater Section: Type: Generic APEs
==
Description:==
Ability to operate and/or monitor the following as they apply to the Loss of Main Feedwater:
HPI, under total feedwater loss conditions.
K/A Number: AA1.04 CFR
Reference:
41.7 / 45.5 / 45.6 Point Value: 1 RO Imp: 4.4 SRO Imp: 4.5 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
Unit One is operating normally with the following equipment OOS for maintenance:
P7A, Steam Driven EFW Pump AACDG, Blackout Diesel Generator A tornado has touched down in the switchyard causing a Degraded Power event.
EDG #1 trips on low lube oil pressure.
4160v bus A3 lockout occurs and cannot be reset.
RCS pressure has risen to 2450 psig and the ERV is open.
Which of the following actions is required for these conditions?
- a. Close CV-1000, ERV Isolation valve.
- c. Crosstie A3 and A4 buses to restore EFW.
Answer:
B. Initiate HPI Cooling per RT-4.
Notes:
Answer [b] is correct. Without any source of feedwater to the SGs and RCS pressure at 2450 psig, HPI cooling is required to ensure adequate core cooling.
Answer [a] is incorrect since an open ERV is an essential component of HPI cooling.
Answer [c] is incorrect, A3 bus is locked out and cannot be restored quickly.
Answer [d] is incorrect as this is done only if HPI is not available or adequate to cool the core.
History:
Created for 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: An
References:
1102.007, Rev. 005-01-0, Degraded Power, page 29, 30 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0366 Source: New Rev: 0 Rev Date: 1/8/00 Originator: J.Cork TUOI: ANO-1-LP-RO-ESAS Objective: 5 System Number: 056 System
Title:
Loss of Offsite Power Section: 4.2 Type: Generic APEs
==
Description:==
Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Order and time to initiation of power for the load sequencer.
K/A Number: AK3.01 CFR
Reference:
41.5, 41.10 / 45.6 / 45.13 Point Value: 1 RO Imp: 3.5 SRO Imp: 3.9 Tier: 1 Group: 3 RO Select: Yes SRO Select: Yes Question:
An electrical storm has caused a Degraded Power situation with a spurious ES actuation of the even channels.
In which order will the following ES components be started automatically?
Answer [d] lists the correct order of load sequence with loss of offsite power and ES actuation.
The others are incorrect sequences of the correct components.
History:
Created for 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1305.006, Rev. 018-02-0, Integrated ES System Test, page 61 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data
QID: 0367 Source: New Rev: 0 Rev Date: 11/8/00 Originator: S.Pullin TUOI: ANO-1-LP-RO-AOP Objective: 10 System Number: 061 System
Title:
Area Radiation Monitoring (ARM) System Alarms Section: 4.2 Type: Generic APEs
==
Description:==
Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms: Need for area evacuation; check against existing limits.
K/A Number: AA2.05 CFR
Reference:
43.5 / 45.13 Point Value: 1 RO Imp: 3.5 SRO Imp: 4.2 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
You are assigned to move spent fuel for the Dry Fuel Storage Project.
A SFP bridge interlock fails and an assembly is damaged while you are moving it toward the cask loading pit.
The SFP area radiation monitor, RE-8009, begins alarming.
Which of the following is your required action for this event?
- a. Place the spent fuel assembly back in its original location.
- b. Notify the Radiation Protection Dept. immediately.
- c. Initiate a local evacuation of the Spent Fuel Pool area.
- d. Place the assembly in the cask loading pit.
Answer:
- c. Initiate a local evacuation of the Spent Fuel Pool area.
Notes:
Answer [c] is correct per AOP 1203.042.
Answers [a] and [d] are incorrect, an ARM in alarm means the area should be evacuated until damage and radiation levels can be assessed.
Answer [b] is incorrect, this would take too much time.
History:
Created for 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: C
References:
1203.042, Rev. 005-00-0, Refueling Abnormal Operations, page 3, 4 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0368 Source: Direct Rev: 0 Rev Date: 11/13/00 Originator: J.Cork TUOI: ANO-1-LP-RO-EOP02 Objective: 5 System Number: E03 System
Title:
Inadequate Subcooling Margin Section: 4.3 Type: B&W EPEs/APEs
==
Description:==
Knowledge of the reason for the following responses as they apply to the (Inadequate Subcooling Margin): RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and admendments are not violated.
K/A Number: EK3.4 CFR
Reference:
41.5 / 41.10, 45.6, 45.13 Point Value: 1 RO Imp: 3.2 SRO Imp: 3.5 Tier: 1
Group: 1 RO Select: Yes SRO Select: Yes Question:
A reactor trip has occurred from 100% power.
One minute later the following conditions exist:
- RCS temperature = 580 degrees F.
- RCS pressure = 1600 psig.
Which of the following operator actions will be performed?
- a. Trip one (1) RCP in each loop.
- b. Verify EFW flow to each Steam Generator is ~430 gpm.
- c. Verify Reflux Boiling setpoint is selected on both EFIC trains.
- d. Initiate 1202.001, Reactor Trip, and go to Overheating EOP.
Answer:
- c. Verify Reflux Boiling setpoint is selected on both EFIC trains.
Notes:
Answer [c] is correct since subcooling margin is lost and the Reflux Boiling setpoint is required to be selected in this situation.
Answer [a] is incorrect, this would be done for loss of subcooling margin but only if >2 minutes had expired without tripping the RCPs.
Answer [b] is incorrect, this is done for loss of subcooling margin but only if one SG is available.
Answer [d] is incorrect, this would not be done since the entry condition for Overheating have not been met.
History:
Used in 2001 RO/SRO Exam, regular exambank QID 3030.
Difficulty: 3 Taxonomy: An
References:
1202.012, Rev. 004-01-0, Repetitive Tasks, RT-5 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0369 Source: New Rev: 0 Rev Date: 11/13/00 Originator: R.Soukup TUOI: ANO-1-LP-RO-CRD Objective: 3 System Number: 003 System
Title:
Dropped Control Rod Section: 4.2 Type: Generic APEs
==
Description:==
Ability to operate and/or monitor the following as they apply to the Dropped Control Rod:
Controls and components necessary to recover rod.
K/A Number: AA1.02 CFR
Reference:
41.7 / 45.5 / 45.6 Point Value: 1 RO Imp: 3.6 SRO Imp: 3.4 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
Given:
- Plant is at 38% power.
- ICS is in full automatic.
- Rod 7 in Group 6 has dropped.
- Annunciator K08-C2 "CONTROL ROD ASYMMETRIC" is in alarm.
- All actions in response to the dropped rod have been completed.
Which of the following actions must be performed FIRST to recover the dropped rod?
- a. Depress FAULT RESET on the Diamond panel.
- b. Transfer dropped rod to its normal power supply.
- c. Latch the dropped rod using auxiliary power supply.
- d. Pull the dropped rod's motor power fuses.
Answer:
c] Latch the dropped rod using auxiliary power supply.
Notes:
Answer [c] is correct since an individual rod must be recovered using the auxiliary power supply and relatched.
Answer [a] is incorrect, this is not done until the rod is realigned with its group.
Answer [b] is incorrect, the rod's group is left on the normal supply.
Answer [d] is incorrect, this is only done for a rod with a high stator temperature condition.
History:
Created for 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: Ap
References:
1105.009. Rev. 016-02-0, CRD System Operating Procedure, page 30, 31 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0370 Source: Modified Rev: 0 Rev Date: 11/13/00 Originator: J.Cork TUOI: ANO-1-LP-RO-AOP Objective: 2 System Number: A04 System
Title:
Turbine Trip Section: 4.3 Type: B&W EPEs/APEs
==
Description:==
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.
K/A Number: AK3.2 CFR
Reference:
41.5 / 41.10 / 45.6 / 45.13 Point Value: 1 RO Imp: 3.4 SRO Imp: 3.6 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
A plant power escalation is in progress at 41% power.
The following conditions are observed:
- Rapid rise in RCS temperature
- Rapid rise in RCS pressure
- Rapid rise in PZR level
- Megawatt output = zero (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. 1203.001, "ICS Abnormal Operating"
- b. 1203.018, "Turbine Trip below 43% Power"
- c. 1203.027, "Loss of Steam Generator Feed"
- d. 1202.001, "Reactor Trip Procedure" Answer:
- b. 1203.018, "Turbine Trip below 43% Power" Notes:
Answer [b] is correct, with MW output at zero and power at 41%, then 1203.018 should be used.
Answer [a] is incorrect, this would be consulted if there was an indication of an instrument failure.
Answer [c] is incorrect, a loss of SG feed could produce these indications but other annunciators such as "RX
IS FW LIMITED", MFP trip, etc., would be in.
Answer [d] is incorrect, Reactor would not have tripped unless power was greater than or equal to 43%.
History:
Modified regular exambank QID #2786 for 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: C
References:
1203.018, Rev. 012-03-0, Turbine Trip Below 43% Power, page 1 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0371 Source: New Rev: 0 Rev Date: 11/13/00 Originator: J.Cork TUOI: ANO-1-LP-RO-NNI Objective: 4 System Number: 008 System
Title:
Pressurize Vapor Space Accident Section: 4.2 Type: Generic APEs
==
Description:==
Ability to determine and interpret the following as they apply to the Pressurize Vapor Space Accident: Effects on indicated Pressurizer pressure and/or level of sensing line leakage.
K/A Number: AA2.27 CFR
Reference:
43.5 / 45.13 Point Value: 1 RO Imp: 2.9 SRO Imp: 3.2 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
If 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 Answer:
d] rise, rise Notes:
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.
History:
Created for 2001 RO/SRO Exam.
Regular exambank QID #5470 used as inspiration.
Difficulty: 3 Taxonomy: Ap
References:
1304.022, Rev. 023-00-0, Unit 1 Pressurizer Level & Temperature Channel Calibration, page 6, 7 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0372 Source: Modified Rev: 0 Rev Date: 11/14/00 Originator: J.Cork TUOI: ANO-1-LP-RO-EOP02 Objective: 3 System Number: 009 System
Title:
Small Break LOCA Section: 4.1 Type: Generic EPEs
==
Description:==
Knowledge of the reasons for the following responses as they apply to the Small Break LOCA:
Actions contained in EOP for Small Break LOCA/leak.
K/A Number: EK3.21 CFR
Reference:
41.5 / 41.10 / 45.6 /45.13 Point Value: 1 RO Imp: 4.2 SRO Imp: 4.5 Tier: 1 Group: 2 RO Select: Yes SRO Select: Yes Question:
A LOCA is in progress concurrently with a Degraded Power condition.
- RCS pressure: 1000 psig
- RCS temperature: 535 degrees F
- HPI flows on C16: 120 gpm, 125 gpm, 115 gpm, 180 gpm
- P36B is inoperable due to maintenance.
- EDG #1 tripped and will not run.
Select the most appropriate action below for this situation:
- a. Close the HPI valve with 180 gpm flow to isolate break.
- b. Stop HPI pump P-36C to reduce break flow.
- c. Throttle HPI valve with 180 gpm flow until flow is 145 gpm.
- d. Close HPI recirc valve, CV-1300 or CV-1301.
Answer:
- c. Throttle HPI valve with 180 gpm flow until flow is 145 gpm.
Notes:
Answer [c] is correct since only one HPI pump is in service and although SCM has not been established, HPI flow must be throttled due to possiblity of an HPI line break robbing flow.
Answer [a] is incorrect, although high HPI flow thru one nozzle would be indicative of a break in this line, the line should not be isolated.
Answer [b] is incorrect although stopping HPI pump will stop possible LOCA pathway on that train, it is incorrect since only one pump is running.
Answer [d] would be correct if ESAS had not actuated but it has.
History:
Modified regular exambank QID #3116 for use in 2001 RO/SRO Exam..
Difficulty: 4 Taxonomy: An
References:
1202.012, Rev. 004-01-0, Repetitive Tasks, RT 3 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0373 Source: New Rev: 0 Rev Date: 11/14/00 Originator: J.Cork TUOI: ANO-1-LP-RO-RCS Objective: 5 System Number: 004 System
Title:
Chemical and Volume Control System Section: 3.2 Type: RCS Inventory Control
==
Description:==
Ability to monitor automatic operation of the CVCS: PZR level and pressure.
K/A Number: A3.10 CFR
Reference:
41.7 / 45.5 Point Value: 1 RO Imp: 3.9 SRO Imp: 3.9 Tier: 2 Group: 1 RO Select: Yes
SRO Select: Yes Question:
Unit One is operating at 100% power.
An ICS malfunction causes Tave to rise 2°F.
What is the effect on the Pressurizer level control system during this transient?
- a. Makeup flow will rise to restore Pressurizer level to setpoint.
- b. Makeup flow will drop to restore Pressurizer level to setpoint.
- c. PZR level will drop to a lower steady state level.
- d. PZR level will rise to a higher steady state level.
Answer:
- b. Makeup flow will drop to restore Pressurizer level to setpoint.
Notes:
Answer [b] is correct since an increase in RCS temperature will swell PZR level. This causes CV-1235 to close to maintain PZR level at setpoint, therefore Makeup flow will drop.
Answer [a], [c], [d] are options that the candidate might choose if he cannot recall proper system response.
History:
Created for 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: C
References:
1104.002, Rev. 053-02-0, Makeup and Purification System Operation, page 6 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0374 Source: New Rev: 0 Rev Date: 11/14/00 Originator: R.Soukup TUOI: ANO-1-LP-RO-RBS Objective: 7 System Number: 013 System
Title:
ESFAS Section: 3.2 Type: RCS Inventory Control
==
Description:==
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ESFAS controls including: Containment pressure, temperature, and humidity.
K/A Number: A1.02 CFR
Reference:
41.5 / 45.5 Point Value: 1 RO Imp: 3.9 SRO Imp: 4.2 Tier: 2 Group: 1 RO Select: Yes SRO Select: No Question:
A large break LOCA has ocurred causing the BWST to be emptied.
The shift to RB sump suction has been completed.
What short term effect will this have on RB pressure and temperature?
c RB pressure will trend up and RB temperature will trend up.
Answer:
c RB Pressure will trend up and RB Temperature will trend up.
Notes:
Answer [c] is correct due to the RB spray pumps now pumping hot RB sump water vs. the cooler BWST water.
Answer [a] is incorrect, RB sump temperature is hotter and initially cause pressure and temperature to rise.
Answer [b] is incorrect, this would be the trend if the RB Spray pumps continued to pump cool BWST water.
Answer [d] is incorrect, RB pressure and temperature will not remain constant in any case.
History:
Created for 2001 RO Exam.
Difficulty: 3 Taxonomy: C
References:
1202.012, Rev. 004-01-0, Repetitive Tasks, RT15 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0375 Source: Modified Rev: 0 Rev Date: 11/14/00 Originator: J.Cork TUOI: ANO-1-LP-RO-NNI Objective: 4 System Number: 017 System
Title:
In-Core Temperature Monitor System Section: 3.7 Type: Instrumentation
==
Description:==
Knowledge of the physical connections and/or cause- effect relationships between the ITM system and the following systems: RCS.
K/A Number: K1.02 CFR
Reference:
41.2 to 41.9 / 45.7 to 45.8 Point Value: 1 RO Imp: 3.3 SRO Imp: 3.5 Tier: 2 Group: 1 RO Select: Yes SRO Select: No Question:
With the plant operating at 100%, which condition would cause an "ICC Event Train A" or "B" alarm?
- a. Core exit thermocouple (one) fails high.
- b. Wide range RCS pressure instrument fails high.
- c. Hot leg level transmitter fails high.
- d. CET subcooling margin < 25 degrees F.
Answer:
- d. CET subcooling margin < 25 degrees F.
Notes:
Answer [d] is correct, if CET subcooling margin was less than the minimum, then an ICC alarm would be generated.
Answer [a] is incorrect, while this seemingly would generate the alarm, the alarm uses an average CET vs.
any CET.
Answer [c] is incorrect, while the hot leg level transmitters have an input to this alarm, a level transmitter failing high would be indicative of no voids and thus the ICC alarm would not be in alarm.
Answer [b] is incorrect, while wide range RCS pressure is compared to avg. CET temp to calculate SCM, a pressure transmitter failing high would provide more SCM, not less.
History:
Modified regular exambank QID #3685 for use in 2001 RO Exam.
Difficulty: 2 Taxonomy: C
References:
1203.012J, Rev. 034-00-0, Annunciator K11 Corrective Action, p.2,3 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0376 Source: New Rev: 0 Rev Date: 11/15/00 Originator: J.Cork TUOI: ANO-1-LP-RO-COND Objective: 6 System Number: 056 System
Title:
Condensate System Section: 3.4 Type: Heat Removal from Reactor Core
==
Description:==
Ability to manually operate and monitor in the control room: Condensate automatic makup valve controller.
K/A Number: A4.08 CFR
Reference:
41.7 / 45.5 to 45.8 Point Value: 1 RO Imp: 1.7 SRO Imp: 1.5 Tier: 2 Group: 1 RO Select: Yes SRO Select: No Question:
Why would the Condensate Makeup valve, CV-2873, require manual operation during condenser tube cleaning in a condenser E-11B waterbox?
- a. Differential pressure between the condensers could cause CV-2873 to open.
- b. The level switch that controls CV-2873 is on E-11B and could be isolated.
- c. Differential pressure between the condenser could prevent CV-2873 from opening.
- d. Differential pressure could cause reverse flow of condensate into Condensate tank, T-41.
Answer:
a) Differential pressure between the condensers could cause CV-2873 to "fail" open.
Notes:
Answer [a] is correct since an isolation of an E-11B waterbox could cause E-11B to have a higher pressure than E-11A causing erroneous low level indication on E-11B and thus CV-2873 would be open continuously.
Answer [b] is incorrect, although CV-2873's level switch is on E-11B, it would not be isolated.
Answer [c] is incorrect, the differential pressure would cause a high level indication only if a E-11A waterbox was isolated.
Answer [d] is incorrect, even with a waterbox isolated the condenser pressure would never drop low enough for reverse flow into T-41 to occur.
History:
Created for 2001 RO Exam.
Difficulty: 2 Taxonomy: C
References:
1104.008, Rev. 020-00-0, Circulating Water & Water Box Vacuum System, p.15 STM 1-20, Rev. 3, Condensate System, p.9 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0377 Source: Modified Rev: 0 Rev Date: 11/15/00 Originator: J.Cork TUOI: ANO-1-LP-RO-EOP04 Objective: 3 System Number: 059 System
Title:
Main Feedwater (MFW) System Section: 3.4 Type: Heat Removal from Reactor Core
==
Description:==
Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Feeding a dry S/G.
K/A Number: A2.04 CFR
Reference:
41.5 / 43.5 / 45.3 / 45.13 Point Value: 1 RO Imp: 2.9 SRO Imp: 3.4 Tier: 2 Group: 1 RO Select: Yes SRO Select: No
Question:
Given:
- RCS pressure 2350 decreasing
- RCS temperature 589 °F and slowly increasing
- ERV opened by CBOR
- "A" EFIC low range level = 18" and dropping slowly
- "B" EFIC low range level = 15" and steady
- HPI is in service
- EFW is NOT available Select the most correct statement from the following:
Answer:
Notes:
Answer [b] is correct since A OTSG level is 18" (<20" dry criteria) but is still dropping and is therefore NOT "dry."
Answer [a], [c], and [d] are incorrect and any could be chosen if candidate cannot determine what event is occurring and/or cannot recall "dry" OTSG criteria.
History:
Modified regular exambank QID #1323 for use in 2001 RO Exam.
Difficulty: 3 Taxonomy: Ap
References:
1202.004, Rev. 4, Overheating, p.3 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0378 Source: New Rev: 0 Rev Date: 11/15/00 Originator: J.Cork TUOI: ANO-1-LP-RO-EFIC Objective: 6 System Number: 061 System
Title:
Auxiliary/Emergency Feedwater (AFW) System Section: 3.4 Type: Heat Removal from Reactor Core
==
Description:==
Knowledge of the physical connections and/or cause-effect relationships between the AFW and the following systems: Main steam system.
K/A Number: K1.03 CFR
Reference:
41.2 to 41.9 / 45.7 to 45.8 Point Value: 1 RO Imp: 3.5 SRO Imp: 3.9 Tier: 2 Group: 1 RO Select: Yes SRO Select: Yes Question:
Both Steam Admission valves, CV-2613 and CV-2663, to the Turbine Driven EFW Pump, P-7A, have solenoid operated bypass valves, SV-2613 and SV-2663.
What is the purpose of these solenoid bypass valves?
- a. They open first to drain moisture from the steam lines thereby preventing an overspeed trip of P-7A.
- b. They open first to allow governor oil pressure to build up prior to opening the larger valves.
- c. They open to decrease the differential pressure across the steam admission valves.
- d. They open to ramp turbine speed up to 3700 rpm prior to opening of steam admission valves.
Answer:
- b. They open first to allow governor oil pressure to build up prior to opening the larger valves.
Notes:
Answer [b] is correct, the solenoids bypass the steam admission valves to spin the turbine via small diameter steam lines. The governor will then build up oil pressure and start to close prior to opening the larger diameter lines isolated by the steam admission valves.
Answer [a] is incorrect, although moisture in the steam lines definitely could cause an overspeed trip, there is a trap installed for this purpose.
Answer [c] is incorrect, this is a purpose of small bypass valves around larger valves but not in this case.
Answer [d] is incorrect, the turbine is ramped up to 3700 rpm but this is done by the governor.
History:
Created for use in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1106.006, Rev. 060-00-0, Emergency Feedwater Pump Operation, p.72 STM 1-27, Rev. 3, Emergency Feedwater System, p.14&22 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0379 Source: Modified Rev: 0 Rev Date: 11/15/00 Originator: J.Cork TUOI: Objective:
System Number: 072 System
Title:
Area Radiation Monitoring (ARM) System Section: 3.7 Type: Instrumentation
==
Description:==
Ability to manually operate and/or monitor in the control room: Alarm and interlock setpoint checks and adjustments.
K/A Number: A4.01 CFR
Reference:
41.7 / 45.5 to 45.8 Point Value: 1 RO Imp: 3.0 SRO Imp: 3.3 Tier: 2 Group: 1 RO Select: Yes SRO Select: Yes Question:
During your performance of 1305.001, Supplement 6, Area Radiation Monitor Monthly Alarm Check, you discover Relay Room Area Monitor, RI-8002, high alarm setpoint is greater than the maximum allowable value.
What are the required actions?
- a. Record the value found, notify the CRS, and initiate a Condition Report.
- b. Adjust the setpoint to less than or equal to max high alarm setpoint before recording the As-Left Setpoint.
- c. Record the value found, then have I&C make the required adjustment under a "blanket" MAI.
- d. Record the value found and continue, nothing else need be done since RI-8002 is not a Tech Spec required monitor.
Answer:
- b. Adjust the setpoint to less than or equal to max high alarm setpoint before recording the As-Left Setpoint.
Notes:
Answer [b] is correct per the procedure.
Answer [a] would be correct for discrepancies not governed by a procedural response.
Answer [c] is how this was handled in the past.
Answer [d] is how an incompetent operator might proceed.
History:
Modified regular exambank QID #2645 for use in 2001 RO Exam.
Difficulty: 2 Taxonomy: K
References:
1305.001, Rev. 014-02-0, Radiation Monitoring System Check and Test, p.37, 40-43 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0380 Source: New Rev: 1 Rev Date: 1/10/01 Originator: J.Cork TUOI: ANO-1-LP-RO-TS Objective: 4 System Number: 006 System
Title:
Emergency Core Cooling System (ECCS)
Section: 3.2 Type: RCS Inventory Control
==
Description:==
Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.
K/A Number: 2.1.33 CFR
Reference:
43.2 / 43.3 / 45.3 Point Value: 1 RO Imp: 3.4 SRO Imp: 4.0 Tier: 2 Group: 2 RO Select: No SRO Select: Yes Question:
Following a refueling outage, a plant "Heatup to 800 psig" is in progress.
- RCS temperature is 290°F.
- HPI pump P-36B is out of service for maintenance.
- HPI pump P-36A is in ES standby and P-36C is operable.
After reviewing the surveillance data for P-36C, System Engineering determines that P-36C is inoperable.
Which of the following is allowable in accordance with Unit One Tech Specs?
- a. Heatup may continue up to normal operating temperature and pressure.
- b. Heatup should be stopped and RCS cooled down to < 200°F.
- c. Heatup may continue up to, but not to exceed, an RCS temp of 350°F.
- d. Heatup should be stopped and RCS temp should not exceed 300°F.
Answer:
- c. Heatup may continue up to, but not to exceed, an RCS temp of 350°F.
Notes:
Answer [c] is correct, two HPI pumps are required to be operable with RCS temp > 350°F.
Answer [a] is incorrect, two HPI pumps must be operable from independent buses.
Answer [b] is incorrect. The RCS does not have to be cooled down, although most ECCS equipment must be operable prior to exceeding 200°F, HPI is not required for CNTMT integrity.
Answer [d] is incorrect, RCS temperature may exceed 300°F but not 350°F.
History:
Created for 2001 SRO Exam.
Difficulty: 4 Taxonomy: Ap
References:
Unit One Technical Specifications 3.3.2
Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0381 Source: Direct Rev: 0 Rev Date: 11/16/00 Originator: J.Cork TUOI: ANO-1-LP-RO-CRD Objective: 22 System Number: 014 System
Title:
Rod Position Indication System (RPIS)
Section: 3.1 Type: Reactivity Control
==
Description:==
Knowledge of the physical connections and/or cause-effect relationships between the RPIS and the following systems: CRDS.
K/A Number: K1.01 CFR
Reference:
41.3 to 41.9 / 45.7 to 45.8 Point Value: 1 RO Imp: 3.2 SRO Imp: 3.6 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
The CONTROL ROD ASYMMETRIC annunciator (K08-C2) in alarm indicates:
- a. a rod is greater than 7 inches from its group average as measured by Relative Position Indication (RPI).
- b. a rod is greater than 7 inches from its group average as measured by Absolute Position Indication (API).
- c. a rod is greater than 9 inches from its group average as measured by Relative Position Indication (RPI).
- d. a rod is greater than 9 inches from its group average as measured by Absolute Position Indication (API).
Answer:
- b. a rod is greater than 7 inches from its group average as measured by Absolute Position Indication (API).
Notes:
Answer [b] is correct since API is used to generate this alarm and this alarm does NOT indicate a runback.
Answer [a] is incorrect, the setpoint is correct but RPI is not utilized to generate this alarm.
Answer [c] is incorrect, this setpoint is for a plant runback and RPI is not utilized to generate this alarm.
Answer [d] is incorrect, although API is the system that generates this alarm, the >9" setpoint is for a plant runback.
History:
Direct from regular exambank QID #5339 for use in 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: K
References:
1203.012G, Rev. 032-01-0, Annunciator K08 Corrective Action, p.12 1203.003, Rev. 019, pc-1, Control Rod Drive Malfunction, p.2 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0382 Source: New Rev: 0 Rev Date: 11/16/00 Originator: R.Soukup TUOI: ANO-1-LP-RO-RBS Objective: 6 System Number: 026 System
Title:
Containment Spray System Section: 3.5 Type: Containment Integrity
==
Description:==
Knowledge of the effect that a loss or malfunction of the CSS will have on the following: CCS.
K/A Number: K3.01 CFR
Reference:
41.7 / 45.6 Point Value: 1 RO Imp: 3.9
SRO Imp: 4.1 Tier: 2 Group: 2 RO Select: Yes SRO Select: No Question:
During post-LOCA conditions in the Reactor Building, what percentage of containment cooling is provided with a failure of one RB Spray pump with no other failures?
- a. 100% Cooling
- b. 150% Cooling
- c. 200% Cooling
- d. 250% Cooling Answer:
- b. 150% Cooling Notes:
Answer [b] is correct, both trains of CNTMT coolers will provide 100% cooling and one train of RB Spray provides 50% cooling capacity.
Answer [a] is incorrect, the combination of CNTMT coolers provides 100% cooling but this answer doesn't consider the cooling capability of the RB Spray pump.
Answer [c] is incorrect, this percentage is based on BOTH RB Spray pumps and both trains of CNTMT coolers.
Answer [d] is incorrect, this percentage is not achievable.
History:
Created for 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: C
References:
STM 1-8, Rev. 5, Reactor Building Spray, p.10 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0384 Source: New Rev: 0 Rev Date: 11/17/00 Originator: R.Soukup TUOI: ANO-1-LP-RO-ELECD Objective: 14.f System Number: 063 System
Title:
DC Electrical Distribution Section: 3.6 Type: Electrical
==
Description:==
Ability to (a) predict the impacts of the following malfunctions or operations on the DC Electrical Systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Grounds.
K/A Number: A2.01 CFR
Reference:
41.5 / 43.5 / 45.3 / 45.13 Point Value: 1 RO Imp: 2.5 SRO Imp: 3.2 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
Plant operating at 100% power.
Annunciator K01-D7, DO1 TROUBLE, is in alarm.
The AO reports the local trouble annunciator indicates a DC ground.
How would you determine which component in the DC system is grounded?
- a. Electrical maintenance support is required to determine ground location.
- b. Check ground indication lamps on the DC distribution panels.
- c. At local panels, push TEST pushbutton, to determine grounded component .
- d. Transfer D11 to D21 and see if ground is still present on D01.
Answer:
- a. Electrical maintenance support is required to determine ground location.
Notes:
Answer [a] is correct in accordance with 1107.004.
Answer [b] is incorrect, individual ground detection lights are not available on DC distribution panels.
Answer [c] is incorrect, ground detection lights at local panel only indicate positive or negative ground, not the component.
Answer [d] is incorrect, transferring D11 to D21 is not allowed at power.
History:
New for 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
1107.004, Rev. 011-03-0, Battery And 125V DC Distribution, p.17 1203.012A, Rev. 033-02-0, Annunciator K01 Corrective Action, p.49, 163, 166 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0385 Source: New Rev: 0 Rev Date: 11/17/00 Originator: R Soukup TUOI: ANO-1-LP-RO-TS Objective: 3 System Number: 005 System
Title:
Residual Heat Removal System Section: 3.4 Type: RCS Heat Removal
==
Description:==
Knowledge of the operational implications of the following concepts as they apply to the RHRS:
Dilution and boration considerations.
K/A Number: K5.09 CFR
Reference:
41.5 / 45.7 Point Value: 1 RO Imp: 3.2 SRO Imp: 3.4 Tier: 2 Group: 3 RO Select: Yes SRO Select: Yes Question:
The plant is in Cold Shutdown.
According to Technical Specifications for the Reactor Coolant System, what MINIMUM conditions must exist before boron concentration can be reduced?
- a. At least 1 RCP and its associated generator are available for heat removal.
- c. Either a RCP or DHR pump in service circulating reactor coolant.
- d. Both RBS pumps available for emergency makeup.
Answer:
- c. Either a RCP or DHR pump in service circulating reactor coolant.
Notes:
Answer [c] is correct, must have either a RCP or DHR pump in service for mixing during boron concentration changes.
Answer [a] is incorrect, does not satisfy the spec when only available.
Answer [b] is incorrect, only 1 RCP required to be in service not 2 and all available not required.
Answer [d] is incorrect, RBS pumps not used to reciculate the RCS.
History:
New for 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
Technical Specification 3.1.1.1.B Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0386 Source: New Rev: 0 Rev Date: 11/20/00 Originator: S.Pullin TUOI: ANO-1-LP-RO-AOP Objective: 4 System Number: 007 System
Title:
Pressurizer Relief Tank / Quench Tank System Section: 3.5 Type: Containment Integrity
==
Description:==
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: Overpressurization of the PZR.
K/A Number: A2.03 CFR
Reference:
41.5 / 43.5 / 45.3 / 45.13 Point Value: 1 RO Imp: 3.6 SRO Imp: 3.9 Tier: 2 Group: 3 RO Select: Yes SRO Select: No Question:
Immediately following an "A" MFP trip at 100% power, the following indications are observed:
- Both MFW Block valves open
- RCS pressure 2155 and rising
- PZR Spray valve, CV-1008, closed
- MFW Cross-Over valve, CV-2827, going open
- "REACTOR IS FW LIMITED" K07-C1 in alarm
- Reactor power is 90% and dropping What operator actions should be taken in addition to the automatic actions occurring?
- a. Manually close MFW Block valves.
- b. Initiate EFW on both trains.
- c. Open PZR spray valve, CV-1008, in manual.
- d. Take manual control of Main Turbine.
Answer:
- c. Open PZR spray valve, CV-1008, in manual.
Notes:
Answer [c] is correct, with a MFP trip >80% power, the PZR spray valve is biased open with a 100 reduction in setpoint, therefore it should be open from 2080 and closes at 2030.
Answer [a] is incorrect, MFW Block valves will not start going closed until power is <80%.
Answer [b] is incorrect, no EFW setpoints have been met.
Answer [d] is incorrect, this is not necessary for this runback.
History:
New for 2001 RO Exam.
Difficulty: 3 Taxonomy: C
References:
1203.015, Rev. 010-01-0, Pressurizer Systems Failure, p. 11, 12 1203.027, Rev. 010-00-0, Loss of Steam Generator Feed, p.1,2 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0387 Source: New Rev: 0 Rev Date: 11/20/00 Originator: J.Cork TUOI: ANO-1-LP-RO-RBVEN Objective: 6
System Number: 028 System
Title:
Hydrogen Recombiner and Purge Control System HRPS Section: 3.5 Type: Contaiment Integrity
==
Description:==
Knowledge of the operations implications of the following concepts as they apply to HRPS:
sources of hydrogen within containment.
K/A Number: K5.03 CFR
Reference:
41.5 / 45.7 Point Value: 1 RO Imp: 2.9 SRO Imp: 3.6 Tier: 2 Group: 3 RO Select: Yes SRO Select: Yes Question:
Which one of the following is NOT a source of hydrogen production in the Reactor Building after a LOCA?
- a. Metal-to-water reaction (zirconium steam/water)
- b. Corrosion of aluminum
Answer:
Answer [c] is correct, the boric acid/carbon steel reaction is a corrosion concern not a major source of hydrogen.
Answers [a], [b], [d] are incorrect, because they are all sources of hydrogen production.
History:
Direct from regular exambank QID#1067, used in 2001 RO/SRO Exam.
Difficulty: 2 Taxonomy: K
References:
STM 1-9, Rev. 3, ch. 2, Reactor Building Ventilation, p.11 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0388 Source: New Rev: 0 Rev Date: 11/20/00 Originator: S. Pullin TUOI: ANO-1-LP-RO-ESAS Objective: 5 System Number: 076 System
Title:
Service Water System Section: 3.4 Type: RCS Heat Removal
==
Description:==
Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following:
Conditions initiating automatic closure of closed cooling water auxiliary building header supply and return valvels.
K/A Number: K4.01 CFR
Reference:
41.7 Point Value: 1 RO Imp: 2.5 SRO Imp: 2.9 Tier: 2 Group: 3 RO Select: Yes SRO Select: No Question:
ESAS has actuated on RB pressure alone.
The CRS is in the Overcooling EOP.
RT-10 directs you to stop all running RCP's.
Why is this action being performed?
- a. No cooling to the RCP motor and seal coolers.
- b. Loss of subcooling margin is imminent.
- c. RCP seal bleedoff path is isolated.
- d. RCP seal injection is isolated.
Answer:
- a. No cooling to the RCP motor and seal coolers.
Notes:
Answer [a] is correct since ESAS channels1-6 will actuate on RB pressure only and will isolate SW to ICW coolers and ICW to RCPs.
Answer [b] is incorrect, this is an overcooling event and SCM will be substantial.
Answer [c] is incorrect, RCP seal bleedoff is on alternate path and not isolated.
Answer [d] is incorrect, RCP seal injection does not close until flow <26 gpm and isn't closed by ESAS.
History:
New for 2001 RO Exam.
Difficulty: 3 Taxonomy: C
References:
1202.012, Rev. 004-01-0, Repetitive Tasks, RT-10 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0389 Source: Modified Rev: 1 Rev Date: 12/7/00 Originator: S.Pullin TUOI: ANO-S-LP-RO-PRCON Objective: 8 System Number: 2.1 System
Title:
Conduct of Operations Section: 2 Type: Generic
==
Description:==
Ability to obtain and verify controlled procedure copy.
K/A Number: 2.1.21 CFR
Reference:
45.10 / 45.13 Point Value: 1 RO Imp: 3.1 SRO Imp: 3.2 Tier: 3 Group:
RO Select: Yes SRO Select: No Question:
A job is in progress that will last for several weeks.
The procedure has been verified at the start of the job.
A pre-job brief has been completed for all participants.
How often should the procedure for this job be verified to be current?
- a. Every 7 days.
- b. Once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- c. Only prior to the start of the job.
- d. Every 14 days.
Answer:
- a. Every 7 days.
Notes:
Answer [a] is correct IAW 1000.006, all other choices are familiar frequencies of tasks.
History:
Modified regular exambank QID# 6054 for use in 2001 RO Exam.
Difficulty: 2 Taxonomy: K
References:
1000.006, Rev. 048-00-0, Procedure Control, step 12.8, p.32
Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0390 Source: Modified Rev: 1 Rev Date: 12/7/00 Originator: S. Pullin TUOI: ANO-1-LP-RO-CRD Objective: 5 System Number: 2.2 System
Title:
Control Rod Drive Section: 2.0 Type: Generic K & A's
==
Description:==
Knowledge of Control Rod programming.
K/A Number: 2.2.33 CFR
Reference:
43.6 Point Value: 1 RO Imp: 2.5 SRO Imp: 2.9 Tier: 3 Group: G RO Select: Yes SRO Select: No Question:
A startup is in progress.
All safety groups are 100% withdrawn.
The group 6 rods' RPI were not re-zeroed prior to startup.
The CBOR is continuing with the startup with group 5 rods.
Which of the following results?
- a. Auto inhibit.
- b. Asymmetric rod alarm.
- c. Out inhibit.
- d. Sequence inhibit.
Answer:
- d. Sequence inhibit.
Notes:
Answer [d] is correct, input to sequence inhibit is from RPI.
Answer [a] is incorrect, input from safety group out limit.
Answer [b] is incorrect, input from API (reed switches).
Answer [c] is incorrect, input from High SUR and API.
History:
Modified regular exambank QID # 3153 for use in 2001 RO exam.
Difficulty: 3 Taxonomy: C
References:
1105.009 Rev 016-02-0, CRD Operating Procedure, p.2,8,9,19 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0391 Source: New Rev: 0 Rev Date: 11/20/00 Originator: R. Soukup TUOI: ANO-1-LP-SRO-RAD Objective: 4 System Number: 2.3 System
Title:
Radiation Control Section: 2.0 Type: Generic K & A's
==
Description:==
Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized.
K/A Number: 2.3.4 CFR
Reference:
43.4 / 45.10 Point Value: 1 RO Imp: 2.5 SRO Imp: 3.1 Tier: 3 Group: G
RO Select: Yes SRO Select: Yes Question:
A worker arrives on site with 2.5 Rem accumulative dose for the calendar year.
The worker's NRC form 4 is on file.
The worker's expected exposure will be 1.8 Rem for his assigned job.
Whose authorization is required to extend the worker's TEDE exposure limit?
- a. The worker's Supervisor, Radiation Protection Manager, and General Manager, Plant Operations.
- b. The worker's Supervisor and Radiation Protection Manager.
- c. The worker's Supervisor, Radiation Protection Manager, General Manager, Plant Operations and Vice President, Operations.
- d. This exposure limit can not be authorized per ANO Admin Exposure Limits.
Answer:
- c. The worker's Supervisor, Radiation Protection Manager, General Manager, Plant Operations and Vice President, Operations.
Notes:
Answer [c] is correct IAW 1012.021 for doses >4 R but <4.5 R.
Answer [a] is incorrect, this is the authorization required for doses >3 R but <4 R.
Answer [b] is incorrect, this is the authorization required for doses >2 R but <3 R.
Answer [d] is incorrect, this is the authorization required for doses >4.5 R.
History:
New question created for 2001 RO/SRO Exam.
Difficulty: 3.5 Taxonomy: Ap
References:
1012.021, Rev. 004-01-0, Exposure Limits and Controls, p.7,9 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0392 Source: New Rev: 0 Rev Date: 11/20/00 Originator: J.Cork TUOI: ANO-1-LP-RO-FPS Objective: 10 System Number: 2.4 System
Title:
Emergency Procedures/Plan Section: 2 Type: Generic K & A's
==
Description:==
Knowledge of fire protection procedures.
K/A Number: 2.4.25 CFR
Reference:
41.10 / 45.13 Point Value: 1 RO Imp: 2.9 SRO Imp: 3.4 Tier: 3 Group: G RO Select: Yes SRO Select: No Question:
Who is the final authority on evaluating the fire loading of a safety related area for transient and insitu combustibles?
- a. Safety Coordinator
- b. Licensing Safety Engineer
- c. Fire Barrier Watch Supervisor
- d. Fire Protection Engineer Answer:
d) Fire Protection Engineer Notes:
Answer [d] is correct per 1000.152, all others have assigned responsibilities with fire protection or safety but not this one.
History:
New question created for 2001 RO Exam.
Difficulty: 2 Taxonomy: K
References:
1000.152, Rev. 003-01-0, Unit One and Two Fire Protection System Specifications, p.6 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0393 Source: New Rev: 0 Rev Date: 11/20/0 Originator: R.Soukup TUOI: ANO-1-LP-RO-AOP Objective: 3 System Number: 2.4 System
Title:
Emergency Procedures/Plan Section: 2 Type: Generic K & A's
==
Description:==
Knowledge of operator response to loss of all annunciators.
K/A Number: 2.4.32 CFR
Reference:
41.10 / 43.5 / 45.13 Point Value: 1 RO Imp: 3.3 SRO Imp: 3.5 Tier: 3 Group: G RO Select: Yes SRO Select: Yes Question:
Both AC and DC "Power Available" lamps have gone out for all Control Room annunciator panels.
Which of the following actions should be taken?
- a. Trip the reactor and enter 1202.001, Reactor Trip.
- b. Commence power reduction per 1203.045, Rapid Plant Shutdown.
- c. Commence normal plant shutdown per 1102.016, Power Reduction and Plant Shutdown.
- d. Notify the Shift Manager to implement 1903.010, Emergency Action Level Classification.
Answer:
- d. Notify the Shift Manager to implement 1903.010, Emergency Action Level Classification.
Notes:
Answer [d] is correct, power should be maintained steady and SM should consult 1903.010.
Answer [a] is incorrect, annunciators are vital when verifying plant conditions following a Rx trip.
Answers [b] and [c] are incorrect, steady state power should be maintained while annunciators are inoperable.
History:
New question created for 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1203.043, Rev. 001-00-0, Loss of Control Room Annunciators, p.1 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0394 Source: Modified Rev: 0 Rev Date: 11/21/00 Originator: S.PULLIN TUOI: AA-51001-001 Objective: 9.6 System Number: 005 System
Title:
Inoperable/Stuck Control Rod Section: 4.2 Type: Generic APEs
==
Description:==
Ability to apply Technical Specifications for a system.
K/A Number: 2.1.12 CFR
Reference:
43.2/43.5/45.3 Point Value: 1 RO Imp: 2.9 SRO Imp: 4.0
Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
Given:
- Rod 4 in Group 5 is inoperable, it did not pass rod drop time.
- Subsequently Rod 7 in Group 7 is found >9" from its group average.
What actions are required by Technical Specifications?
- a. Boron must be added to equal the worth of both rods.
- b. Power must be reduced to 60% of the maximum allowable power level for the number of RCP's running.
- c. The reactor must be brought to Hot Standby within one hour unless 1% availble shutdown margin can be verified.
- d. Plant shutdown is required since operation with more than one inoperable rod is not allowed.
Answer:
- d. Plant shutdown is required since operation with more than one inoperable rod is not allowed.
Notes:
Answer [d] is correct, operation is limited to only 1 inoperable rod per Technical Specifications 3.0.3.
Answer [a] is incorrect, this action does not comply with T.S. for inoperable rods.
Answer [b] is incorrect, this action does not comply with T.S. for inoperable rods.
Answer [c] is incorrect, this action does not comply with T.S. for inoperable rods.
History:
Modified QID 1823 for use in 2001 RO/ SRO exam.
Difficulty: 2 Taxonomy: K
References:
Technical Specifications 3.5.2.2 & 3.0.3.
Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0395 Source: Direct Rev: 0 Rev Date: 11/21/00 Originator: D.Slusher TUOI: ANO-1-LP-RO-NNI Objective: 14 System Number: 027 System
Title:
Pressurizer Pressure Control Malfunction Section: 4.2 Type: Generic APEs
==
Description:==
Knowledge of the interrelations between the Pressurizer Pressure Control Malfunction and the following: Controllers and positioners.
K/A Number: AK2.03 CFR
Reference:
41.7 / 45.7 Point Value: 1 RO Imp: 2.6 SRO Imp: 2.8 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
The plant is shutdown and cooled down.
RCS pressure is 220 psig.
I&C is performing calibration checks on "A" RPS channel.
Why will I & C request the Pzr Control Pressure Selector, HS-1038, be placed in the "Y" position?
- a. To allow remote indications to be checked during calibration.
- c. To maintain pressurizer heaters off during testing.
- d. To allow the ERV low setpoint to be checked.
Answer:
Notes:
Answer [b] is correct, testing will cause ERV to open since the LTOP setpoint is in effect.
Answer [a] is incorrect, the selector switch does not select between local and remote indications.
Answer [c] is incorrect, PZR heaters are in manual control for cooldown.
Answer [d] is incorrect, I&C verifies the setpoint, it is undesirable to operate ERV at this point.
History:
Direct from regular exambank QID#5545 for 2001 RO/SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1105.006, Rev. 009-00-0, Reactor Coolant System NNI, page 12, 13 STM 1-69 rev. 4, Non Nuclear Instrumentation system, page 14, 15 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0396 Source: New Rev: 0 Rev Date: 11/21/00 Originator: S.Pullin TUOI: ANO-1-LP-RO-ICS Objective: 4 System Number: 015 System
Title:
Reactor Coolant Pump (RCP) Malfunctions Section: 4.2 Type: Generic APEs
==
Description:==
Ability to operate and/or monitor the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): RCS flow.
K/A Number: AA1.05 CFR
Reference:
41.7 / 45.5 / 45.6 Point Value: 1 RO Imp: 3.8 SRO Imp: 3.8 Tier: 1 Group: 1 RO Select: Yes SRO Select: Yes Question:
The plant is steady state at 70% power per Dispatcher request.
Subsequently, you observe the following indications:
"A" MFW flow 4.7 e 6 lbm/hr "B" MFW flow 2.3 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:
- c. "A" RCP trip Notes:
Answer [c] 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 [d] is incorrect, this would cause FW to decrease to the "A" SG and increase to the "B" SG.
History:
New question created for 2001 RO/SRO Exam.
Difficulty: 3
Taxonomy: Ap
References:
STM 1-64, Rev. 6, Integrated Control System, page 34, 35 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0397 Source: New Rev: 0 Rev Date: 11/21/00 Originator: J.Cork TUOI: ANO-1-LP-RO-NOP Objective: 4 System Number: 001 System
Title:
Continuous Rod Withdrawal Section: 4.2 Type: Generic APEs
==
Description:==
Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:
Proper actions to be taken if automatic safety functions have not taken place.
K/A Number: AA2.03 CFR
Reference:
43.5 / 45.13 Point Value: 1 RO Imp: 4.5 SRO Imp: 4.8 Tier: 1 Group: 1 RO Select: No SRO Select: Yes Question:
Approach to criticality is in progress.
Reactor power is in the Source Range.
The CBOR commences sequential withdrawal of the regulating rods.
The following indications are observed:
- SR count rate rising
- Sustained SUR of 2.5 DPM
- Continued outward rod motion without a command.
What action is required to be taken?
- a. Trip the reactor and go to 1202.001, Reactor Trip.
- b. Commence Emergency Boration per RT-12 until SUR is negative.
- c. Select "JOG" on rod speed selector switch.
- d. No action required, Out Inhibit on high SUR will stop rod motion.
Answer:
- a. Trip the reactor and go to 1202.001, Reactor Trip.
Notes:
Answer [a] is correct, the Source Range rod hold on high SUR of 2.5 DPM should have stopped outward rod motion. A system failure has occurred, control of reactivity has been lost, and reactor should be tripped.
Answer [b] is incorrect, this will add negative reactivity but it is also slow.
Answer [c] is incorrect, attempts to stop rod motion with these actions will not work.
Answer [d] is incorrect, the SR High SUR inhibit should have stopped rod motion at 2 DPM and it would be imprudent to wait to see if the IR High SUR inhibit at 3 DPM would work.
History:
New created for 2001 SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1102.008, Rev. 018-00-0, Approach to Criticality, page 4, 11 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0398 Source: New Rev: 0 Rev Date: 11/21/00 Originator: J.Cork TUOI: ANO-1-LP-SRO-ADMIN Objective: 6 System Number: 067 System
Title:
Plant Fire On Site
Section: 4.2 Type: Generic APEs
==
Description:==
Ability to determine and interpret the following as they apply to the Plant Fire On Site:
Requirements for establishing a fire watch.
K/A Number: AA2.15 CFR
Reference:
43.5 / 45.13 Point Value: 1 RO Imp: 2.9 SRO Imp: 3.9 Tier: 1 Group: 1 RO Select: No SRO Select: Yes Question:
A fire in the T-57A Emergency Diesel Fuel Oil Storage vault has been in progress but is now extinguished. The intensity of the fire warped the room access door so badly that it will not close. The fire also contaminated the smoke detection in the room so that the smoke detectors are inoperable.
What are the MINIMUM compensatory measures required by 1000.152, Unit 1 & 2 Fire Protection Specifications, for this area?
a) Restore the inoperable door and detection to operable status within 14 days or initiate a condition report.
b) Within one hour establish an hourly fire watch since flame detection for this room is still operable.
c) Within one hour establish a continuous fire watch with standby suppression equipment.
d) Immediately test T-57B vault's fire detection and suppression equipment to prove operability.
Answer:
c) Within one hour establish a continuous fire watch with standby suppression equipment.
Notes:
Answer [c] is correct, the fire barrier is inoperable and no detection is available on either side of the door.
Answer [a] is incorrect, although this is similar to other 1000.152 actions, it is incorrect for this scenario.
Answer [b] is incorrect, no flame detection exists in the EDG F.O. vaults, so no detection exists and a continuous watch is required.
Answer [d] is incorrect, this is similar to Tech Spec action statements but is not applicable to Fire Systems.
History:
Created for 2001 SRO Exam.
Difficulty: 3 Taxonomy: An
References:
1000.152, Rev. 003-01-0, Unit 1 & 2 Fire Protection System Specifications, pages 11, 12, 21, 22 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0399 Source: New Rev: 0 Rev Date: 11/21/00 Originator: R.Soukup TUOI: ANO-1-LP-RO-EOP06 Objective: 4 System Number: E03 System
Title:
Loss of Subcooling Margin Section: 4.3 Type: B&W EOP/AOP
==
Description:==
Knowledge of how the event based emergency/abnormal operating procedures are used in conjunction with the symptom based EOPs.
K/A Number: 2.4.8 CFR
Reference:
41.10 / 43.5 / 45.13 Point Value: 1 RO Imp: 3.0 SRO Imp: 3.7 Tier: 1
Group: 1 RO Select: No SRO Select: Yes Question:
A plant shutdown is in progress due to a primary-to-secondary tube leak in accordance with 1203.023, Small Steam Generator Tube Leaks.
- Several tubes fail catastrophically
- RCS pressure drops to 1400 psig
- CET's indicate 560°F
- Reactor Building sump level is trending upward.
Which of the following procedures should be in use?
a) Reactor Trip and Loss of Subcooling Margin b) Tube Rupture and ESAS c) Loss of Subcooling Margin and ESAS d) Reactor Trip and Tube Rupture Answer:
d) Reactor Trip and Tube Rupture Notes:
Answer [d] is correct, the Tube Rupture procedure is used in conjunction with Reactor Trip when abnormal conditions other than a SGTR exist, i.e., RB sump level rising.
Answer [a] is incorrect, although a Loss of SCM has occurred, contingency actions for this exist within the Tube Rupture EOP.
Answer [b] is incorrect, although an ESAS would have occurred, it is not entered with a SGTR.
Answer [c] is incorrect, although both ESAS and loss of SCM has occurred, they are not entered with a SGTR.
History:
Created for 2001 SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1202.006, Rev. 007-02-0, Tube Rupture, page 1 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0400 Source: Direct Rev: 0 Rev Date: 11/21/00 Originator: R.Soukup TUOI: ANO-1-LP-RO-CRD Objective: 7 System Number: 001 System
Title:
Control Rod Drive Section: 3.1 Type: Reactivity Control
==
Description:==
Ability to explain and apply all system limits and precautions.
K/A Number: 2.1.32 CFR
Reference:
41.10 / 43.2 / 45.12 Point Value: 1 RO Imp: 3.4 SRO Imp: 3.8 Tier: 2 Group: 1 RO Select: No SRO Select: Yes Question:
Of the following CRD operating limits which one is NOT due to heat removal considerations?
- a. Maximum CRD travel is 420 inches per hour.
- b. Maximum CRD running time is 30 minutes per hour.
- c. No more than 2 phases are energized when movement is stopped.
- d. Latching control rods in JOG speed.
Answer:
- d. Latching control rods in JOG speed.
Notes:
Answer [d] is correct, this limit is to prevent damage to CRDM spider.
Answers [a], [b], [c] are all limits associated with heat removal from CRDM and are therefore incorrect.
History:
Direct from regular exambank, QID #2322, for 2001 SRO Exam.
Difficulty: 3 Taxonomy: C
References:
1105.009, Rev. 016-02-0, CRD System Operating Procedure, page 11 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0401 Source: New Rev: 0 Rev Date: 11/21/00 Originator: S.Pullin TUOI: ANO-1-LP-RO-RBS Objective: 8 System Number: 026 System
Title:
Containment Spray Section: 3.5 Type: Containment Integrity
==
Description:==
Knowledge of annunciator response procedures.
K/A Number: 2.4.10 CFR
Reference:
41.10 / 43.5 / 45.13 Point Value: 1 RO Imp: 3.0 SRO Imp: 3.1 Tier: 2 Group: 1 RO Select: No SRO Select: Yes Question:
A large break LOCA is in progress.
RCS pressure is ~ 50 psig.
RB pressure is 55 psia and trending down.
Annunciator K11-C6, "RB SPRAY P-35A ES FAILURE" is in alarm.
What condition brought in this alarm?
a) "A" RB Spray pump high motor winding temperature >300°F after ES Channel 7 actuation b) "A" RB Spray flow <1050 gpm 55 seconds after ES Channel 7 actuation.
c) "A" RB Spray flow >1650 gpm 55 seconds after ES Channel 7 actuation.
d) "A" RB Spray pump high bearing temperature >200°F after ES Channel 7 actuation.
Answer:
b) "A" RB Spray flow <1050 gpm 55 seconds after ES Channel 7 actuation.
Notes:
Answer [b] is correct, this is the correct logic for this annunciator.
Answer [a] is incorrect, high motor winding temperature does not input into this alarm.
Answer [c] is incorrect, a high flow condition will not actuate this alarm.
Answer [d] is incorrect, high bearing temperature does not input into this alarm.
History:
Created for 2001 SRO Exam.
Difficulty: 4 Taxonomy: Ap
References:
1203.012J, Rev. 034-00-0, Annunciator K11 Corrective Action, page 29 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0402 Source: Direct Rev: 0 Rev Date: 11/21/00 Originator: R.Soukup TUOI: ANO-S-LP-RO-CCM08 Objective: 1
System Number: 006 System
Title:
Emergency Core Cooling Section: 3.2 Type: RCS Inventory Control
==
Description:==
Knowledge of the process for determining the internal and external effects on core reactivity.
K/A Number: 2.2.34 CFR
Reference:
43.6 Point Value: 1 RO Imp: 2.8 SRO Imp: 3.2 Tier: 2 Group: 2 RO Select: No SRO Select: Yes Question:
Regarding void formation in the reactor coolant system during an accident, which area of void formation will have the greatest affect on the source range monitor response?
- a. Vessel head
- b. Vessel downcomer
- c. Core shroud and periphery
- d. Central core region Answer:
- b. Vessel downcomer Notes:
Answer [b] is correct, a voided downcomer will allow more neutron leakage.
Answer [a] is incorrect, voids in this region will not affect source range indication.
Answer [c] is incorrect, the downcomer region surrounds this area.
Answer [d] is incorrect, this region is surrounded by water.
History:
Direct from regular exambank QID#5305, used in 2001 SRO Exam.
Difficulty: 3 Taxonomy: C
References:
General Physics "Mitigating Reactor Core Damage" training material, p.4-89,90 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0403 Source: Direct Rev: 0 Rev Date: 11/21/00 Originator: D.Slusher TUOI: ANO-1-LP-RO-NNI Objective: 10 System Number: 016 System
Title:
Non-nuclear Instrumentation Section: 3.7 Type: Instrumentation
==
Description:==
Knowledge of annunciator response procedures.
K/A Number: 2.4.10 CFR
Reference:
41.10 / 43.5 / 45.13 Point Value: 1 RO Imp: 3.0 SRO Imp: 3.1 Tier: 2 Group: 2 RO Select: No SRO Select: Yes Question:
Initial Conditions:
SASS Mismatch Annunciator Rx is Feedwater Limited Annunciator Feedwater is Rx Limited Annunciator Unit Tave TI-1032 is 583 degrees F
Loop A Tave TI-1020 is 588 degrees F Loop B Tave TI-1043 is 578 degrees F Which of the following actions would clear the cross limits and return temperature indications to normal?
- a. Select the NNI-Y signal for RCS loop A hot leg temperature.
- b. Place the Controlling Tave Selector Switch in the Loop B position.
- d. Select the NNI-Y signal for RCS loop B cold leg temperature.
Answer:
- a. Select the NNI-Y signal for RCS loop A hot leg temperature.
Notes:
Answer [a] is correct, because cross limits indicate a problem with a false high temperature feeding into the unit Tave.
Answer [b] is incorrect, due to instrument failure, the controlling Tave would not be selected until ICS components were placed in Manual.
Answer [c] is incorrect, ICS would be trying to raise A FW flow due to the cross limits.
Answer [d] is incorrect, "B" T cold is not the problem, the loop with the higher temperature has the failed instrument.
History:
Direct from exambank 4574, used in 2001 SRO Exam.
Difficulty: 4 Taxonomy: An
References:
STM 1-64, rev. 6, Integrated Control System, page 44, 45 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0404 Source: New Rev: 0 Rev Date: 12/1/00 Originator: J.Cork TUOI: ANO-1-LP-RO-EOP02 Objective: 4 System Number: 035 System
Title:
Steam Generator Section: 3.4 Type: Heat Removal from Reactor Core
==
Description:==
Ability to (a) predict the impacts of the following malfunctions or operations on the SG; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Small break LOCA.
K/A Number: A2.06 CFR
Reference:
41.5 / 43.5 / 45.3 / 45.5 Point Value: 1 RO Imp: 4.5 SRO Imp: 4.6 Tier: 2 Group: 2 RO Select: No SRO Select: Yes Question:
Given:
- Small Break LOCA has occurred and Subcooling Margin was lost.
- ESAS actuated with all components operating properly.
- RCS pressure is now 1200 psig and rising slowly.
- CETs are 500°F and dropping slowly.
- Source of LOCA has been isolated.
Which of the following conditions will allow you to transition to the Reactor Trip EOP, 1202.001?
a) Uncontrolled RCS cooldown is occurring due to HPI/Break flow.
b) Primary to secondary heat transfer is NOT established.
c) Steam generator tube leakage is indicated.
d) Primary to secondary heat transfer is in progress.
Answer:
d) Primary to secondary heat transfer is in progress.
Notes:
Answer [d] is correct, the Reactor Trip EOP is only entered if the cause of the LOCA is isolated, SCM is restored, and primary to secondary heat transfer has been established.
Answer [a] is incorrect, this condition would lead to a transition to Small Break LOCA Cooldown, 1203.041.
Answer [b] is incorrect, the Loss of SCM EOP would still be in use.
Answer [c] is incorrect, the operator would transition to the Tube Rupture EOP if SCM was restored.
History:
New created for 2001 SRO Exam.
Difficulty: 3 Taxonomy: Ap
References:
1202.002, Rev. 004-00-0, Loss of Subcooling Margin, page 12 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0405 Source: New Rev: 0 Rev Date: 12/1/00 Originator: S.Pullin TUOI: ANO-1-LP-RO-MSSS Objective: 2 System Number: 076 System
Title:
Service Water Section: 3.4 Type: Heat Removal from Reactor Core
==
Description:==
Ability to analyze the affect of maintenance activities on LCO status.
K/A Number: 2.2.24 CFR
Reference:
43.2 / 45.13 Point Value: 1 RO Imp: 2.6 SRO Imp: 3.8 Tier: 2 Group: 3 RO Select: No SRO Select: Yes Question:
Given:
- Plant is at 100% power.
- P-4B SW pump is inoperable due to high vibrations on last surveillance.
- P-4A SW pump is supplying Loop 1 SW
- P-4C SW pump is supplying Loop 2 SW
- Maintenance is complete on P-4B.
- Pre-evolution brief has been conducted and the surveillance is ready to be performed to prove operability of P-4B.
Are any compensatory measures needed to be in place while running the P-4B SW pump on Loop 1?
a) No actions needed due to P-4A and P-4C operable and powered from independent buses.
b) A dedicated licensed operator stationed to monitor SW, with no other duties.
c) No actions needed since P-4A will start on ESAS signal.
d) Enter Tech Spec 3.3.6 for inoperable SW loop during the surveillance.
Answer:
b) A dedicated licensed operator stationed to monitor SW, with no other duties.
Notes:
Answer [b] is correct, per 1104.029 operability section which states that an inoperable Service Water pump can be run on an operable loop of service water as long as the following compensatory measures are taken:
pre-job brief and dedicated operator stationed to monitor SW, with no other duties.
Answer [a] is incorrect, due to if ESAS acutates and no operator action is taken an inoperable pump will be supplying loop 1 SW.
Answer [c] is incorrect, due to P-4A will not start on ESAS signal due to P4B was the pump running prior to the
acutation.
Answer [d] is incorrect, if operations chooses to run P-4B without taken any measures then all equipment supplied by SW Loop 1 would be inoperable making LCO 3.0.3 the approriate LCO to enter.
History:
New created for 2001 SRO Exam.
Difficulty: 3 Taxonomy: Ap
References:
1104.029, Rev. 054-00-0, Service Water and Auxiliary Cooling System, page 29 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0406 Source: New Rev: 0 Rev Date: 12/1/00 Originator: J.Cork TUOI: ANO-S-LP-SRO-ADMIN Objective: 3 System Number: 2.1 System
Title:
Conduct of Operations Section: 2 Type: Generic Knowledges and Abilities
==
Description:==
Knowledge of shift turnover practices.
K/A Number: 2.1.3 CFR
Reference:
41.10 / 45.13 Point Value: 1 RO Imp: 3.0 SRO Imp: 4.0 Tier: 3 Group:
RO Select: No SRO Select: Yes Question:
You are the oncoming Shift Manager for night shift following your days off after a training week.
How many days of Station Logs are you required to review?
a) three b) four c) six d) seven Answer:
d) seven Notes:
Answer [d] is correct, per 1015.001 the SM is required to review the station logs for the last 7 days or the last time on shift, whichever is shorter.
Answers [a], [b], [c] are other intervals the candidate might choose, [a] being the most probable since three days had expired since the SM was last at the plant but the SM was not on shift.
History:
New created for 2001 SRO Exam.
Difficulty: 2 Taxonomy: C
References:
1015.001, Rev. 052-05-0, Conduct of Operations, page 38 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0407 Source: New Rev: 0 Rev Date: 12/1/00 Originator: J.Cork TUOI: ANO-S-LP-SRO-ADMIN Objective: 3 System Number: 2.1 System
Title:
Conduct of Operations Section: 2 Type: Generic Knowledges and Abilities
==
Description:==
Knowledge of shift staffing requirements.
K/A Number: 2.1.4 CFR
Reference:
41.10 / 43.2
Point Value: 1 RO Imp: 2.3 SRO Imp: 3.4 Tier: 3 Group:
RO Select: No SRO Select: Yes Question:
On New Year's Eve night shift, the on-duty CRS has a heart attack and must be transported to St. Mary's at 0210.
What is the latest time at which a replacement CRS must be in the Control Room BEFORE the requirement of 1015.001, Conduct of Operations is violated?
a) 0300 b) 0400 c) 0500 d) 0600 Answer:
b) 0400 Notes:
Answer [b] is the correct answer since the maximum time the shift can be below the minimum complement is two hours.
Answers [a], [c], [d] are one hour increments around the correct answer.
History:
New created for 2001 SRO Exam.
Difficulty: 2 Taxonomy: C
References:
1015.001, Rev. 052-05-0, Conduct of Operations, page 36 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0408 Source: New Rev: 0 Rev Date: 12/1/00 Originator: S.Pullin TUOI: ANO-S-LP-SRO-ADMIN Objective: 3 System Number: 2.1 System
Title:
Conduct of Operations Section: 2 Type: Generic Knowledges and Abilities
==
Description:==
Ability to manage short term information such as night and standing orders.
K/A Number: 2.1.15 CFR
Reference:
45.12 Point Value: 1 RO Imp: 2.3 SRO Imp: 3.0 Tier: 3 Group:
RO Select: No SRO Select: Yes Question:
During a refueling outage, the RP Manager approaches the Shift Manager about issuing a Night Order he has prepared concerning radiological precautions during fuel handling.
The Shift Manager takes the Night Order and reviews it with his crew.
Do you agree or disagree with his actions?
a) Agree, the Shift Manager has the authority to issue Night Orders.
b) Disagree, a procedure change is required to implement guidance.
c) Agree, any Manager has the authority to issue Night Orders.
d) Disagree, only the Ops Manager or his designee may issue Night Orders.
Answer:
d) Disagree, only the Ops Manager or his designee may issue Night Orders.
Notes:
Answer [d] is correct per 1015.001.
Answers [a], [b], [c] do not comply with 1015.001.
History:
New created for 2001 SRO Exam.
Difficulty: 2 Taxonomy: C
References:
1015.001, Rev. 052-05-0, Conduct of Operations, page 33 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0409 Source: New Rev: 0 Rev Date: 12/1/00 Originator: Jcork TUOI: ANO-S-LP-SRO-ADMIN Objective: 3 System Number: 2.2 System
Title:
Equipment Control Section: 2 Type: Generic Knowledges and Abilities
==
Description:==
Knowledge of process for making changes in the facility as described in the Safety Analysis Report.
K/A Number: 2.2.5 CFR
Reference:
43.3 / 45.13 Point Value: 1 RO Imp: 1.6 SRO Imp: 2.7 Tier: 3 Group:
RO Select: No SRO Select: Yes Question:
Which of the following would require a 10CFR50.59 Evaluation per 1000.131, 10CFR50.59 Review Program?
a) Replacing Emergency Diesel Generators' Air Receiving Tanks with larger tanks.
b) A change to a different manufacturer for a "like-for-like" valve operator in a safety system.
c) A change in the title of Shift Superintendent to Shift Manager.
d) A drawing change to correct a HPI injection valve number on a P&ID used in the SAR.
Answer:
a) Replacing EDG's Air Receiving Tanks with larger tanks.
Notes:
Answer [a] is correct, any time you are changing out Plant Equipment, you must perform a 50.59 evaluation.
Answers [b], [c], [d] require 50.59 determinations but are excluded from evaluation as listed in Att. 1 in 1000.131.
History:
Created for 2001 SRO Exam.
Difficulty: 2 Taxonomy: C
References:
1000.131, Rev. 003-04-0, 10CFR50.59 Review Program page.29-30, Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data
QID: 0410 Source: New Rev: 0 Rev Date: 12/1/00 Originator: S.Pullin TUOI: ANO-S-LP-SRO-ADMIN Objective: 4 System Number: 2.2 System
Title:
Equipment Control Section: 2 Type: Generic Knowledges and Abilities
==
Description:==
Knowledge of the process for managing maintenance activities during power operation.
K/A Number: 2.2.17 CFR
Reference:
43.5 / 45.13 Point Value: 1 RO Imp: 2.3 SRO Imp: 3.5 Tier: 3 Group:
RO Select: No SRO Select: Yes Question:
Given:
- Plant is at 100% power.
- Mechanic wants to sign in a refueling work package for Main Turbine.
- The reason for doing this is to draw parts from warehouse for prep work.
What would the CRS-Admin use (if possible) to allow MM to draw the parts they need without actually starting work?
a) This is not allowed prior to the start of the refueling outage.
b) CRS must initiate a separate MAI for parts acquisition only.
c) Sign in work package and place an Operations Hold in package.
d) Sign in work package and conduct a pre-job brief.
Answer:
c) Sign in work package and place an Operations Hold in package.
Notes:
Answer [c] is correct IAW 1000.024.
Answer [a] is incorrect, the Ops Hold allows this type of activity.
Answer [b] is incorrect, a separate MAI sounds good but is not needed.
Answer [d] is incorrect, a pre-job brief is used in most cases but would not prevent other MM personnel from starting work.
History:
Created for 2001 SRO exam.
Difficulty: 3 Taxonomy: C
References:
1000.024, Rev. 047-00-0, Control of Maintenance, p.13 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0411 Source: Modified Rev: 0 Rev Date: 12/1/00 Originator: E-Plan TUOI: ANO-S-LP-EP-A0082 Objective: 4 System Number: 2.4 System
Title:
Emergency Procedures/Plan Section: 2 Type: Generic Knowledges and Abilities
==
Description:==
Knowledge of which events related to system operations/status should be reported to outside agencies.
K/A Number: 2.4.30 CFR
Reference:
43.5 / 45.11 Point Value: 1 RO Imp: 2.2 SRO Imp: 3.6 Tier: 3 Group:
RO Select: No SRO Select: Yes Question:
A fire was reported at 0844 in the vicinity of the Old Radwaste Building.
It is now 0920 and the fire is still burning.
What is the requirement for notification to the NRC?
a) Within 15 minutes of declaration of an emergency class.
b) Immediately following notification of the state and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the declaration of an emergency class.
c) Immediately following notification of the local agencies.
d) Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of declaration of an emergency class.
Answer:
b) Immediately following notification of the state and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the declaration of an emergency class.
Notes:
Answer [b] is correct since this is the procedural requirement.
Answer [a], [c], [d] are incorrect, these are not in accordance with 1903.011.
History:
Modified E-Plan exambank QID#61 for use in 2001 SRO Exam.
Difficulty: 2 Taxonomy: C
References:
1903.011, Rev. 025-04-0, Emergency Response/Notifications, page 67 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0412 Source: New Rev: 0 Rev Date: 12/5/00 Originator: S.Pullin TUOI: ANO-1-LP-RO-RPS Objective: 6 System Number: 057 System
Title:
Loss of Vital AC Inst. Bus Section: 4.2 Type: Generic APEs
==
Description:==
Knowledge of EOP entry conditions and immediate action steps.
K/A Number: 2.4.1 CFR
Reference:
41.10 / 43.5 / 45.13 Point Value: 1 RO Imp: 4.3 SRO Imp: 4.6 Tier: 1 Group: 1 RO Select: No SRO Select: Yes Question:
Given:
- Plant at 100% power.
- RPS Channel "C" inoperable and in the tripped state.
- I&C are performing RPS Channel "B" monthly calibration.
- Inverter Y-11 has a fault and the static switch fails to operate.
Which of the following procedures should be in use?
a) Loss of NNI Power, 1203.047 b) Reactor Trip, 1202.001 c) Loss of 125 VDC, 1203.036 d) ESAS, 1202.010 Answer:
b) Reactor Trip, 1202.001 Notes:
Answer [b] is correct since a loss of Y11 will cause a loss of RS-1 which will trip "A" RPS channel. This will
cause a reactor trip since "C" is in the tripped state.
Answer [a] is incorrect, NNI is powered from RS-1 but will transfer to alternate power.
Answer [c] is incorrect, although Y11 is normally powered from a DC source, the Rx Trip EOP has the highest priority.
Answer [d] is incorrect, although ESAS is partially powered from RS-1, no actuations will result.
History:
New created for 2001 SRO Exam.
Difficulty: 3 Taxonomy: An
References:
1105.001, Rev.019-00-0, NI & RPS Operating Procedure, pages 7,8 Arkansas Nuclear One - Unit One Initial RO/SRO Exam Question Data QID: 0413 Source: New Rev: 0 Rev Date: 12/06/00 Originator: S. Pullin TUOI: ANO-1-LP-AO-ELECD Objective: 4 System Number: 062 System
Title:
AC Electrical Distribution System Section: 3.6 Type: Electrical
==
Description:==
Knowledge of AC distribution system design feature(s) and/or interlock(s) which provide for the following: Bus Lockouts.
K/A Number: K4.01 CFR
Reference:
41.7 Point Value: 1 RO Imp: 2.6 SRO Imp: 3.2 Tier: 2 Group: 2 RO Select: Yes SRO Select: Yes Question:
The plant is in Hot Shutdown with a normal electrical alignment.
Due to an electrical fault, K02-A6 "A1 L.O. RELAY TRIP" comes into alarm.
How will the electrical system respond?
- a. #1 EDG will auto-start and will supply bus A3.
- b. Bus A1 will auto-transfer to transformer SU#1.
- c. Bus A1 will auto-transfer to transformer SU#2.
- d. #1 EDG will auto-start, but will not supply bus A3.
Answer:
- a. #1 EDG will auto-start and will supply bus A3.
Notes:
(a) is the correct answer, the A1 bus lock-out will cause a loss of A3, EDG#1 will auto-start and supply A3.
(b) and (c) are incorrect, the A1 lockout will prevent any transformer from supplying the bus.
(d) is incorrect, the A1 lockout will not prevent #1 EDG from supplying bus A3.
History:
New Question for 2000 RO/SRO NRC Exam.
Difficulty: 2 Taxonomy: C
References:
1203.012B, Rev. 025-02-0, page 36.