ML032390332
| ML032390332 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/21/2003 |
| From: | Berry P Entergy Nuclear Northeast |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| Shared Package | |
| 127 | List: |
| References | |
| 50-333/03-301 | |
| Download: ML032390332 (127) | |
Text
Examination Outline Cross-reference: Level SRO Tier # 1 Partial or Complete Loss of AC / 6 Group # 1 Ability to determine and/or interpret the following W A # 295003 AA2.05 as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: (CFR: 41.10&$$;5/45.13)
Whether a partial or complete loss of AC. power has occurred Importance Rating 4.2 Proposed Question: The plant is shutdown for a Refueling outage. Site electrical power is being provided from the 115 KV system. The only deviation from the normal alignment is that disconnect 10017, North-South Bus Disconnect, is currently OPEN. From this condition, circuit breaker 10022, (LHH-FITZ 115 KV LINE 3 BKR), trips.
Which one of the following identifies the expected procedural response?
a) AOP-16, Loss of 10300 Bus and AOP-18, Loss of 10500 Bus RO/SRO b) AOP-17, Loss of 10400 Bus and AOP-19, Loss of 10600 Bus SI c) AOP-57, Recovery from Residual Bus Transfer d) AOP-49A, Station Blackout In Cold Condition Proposed Answer: b) AOP-I 7, Loss of 10400 Bus and AOP-19, Loss of 10600 Bus Explanation (Optional):
Technical Reference(s): OP-44, AOP-17 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-71D, EO-1.05.a, 1.06, 1.09 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
L Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Partial or Complete Loss of Forced Core Flow Group # 1 1 Circulation 1I& 4 Knowledge of facility ALARA program. W A # 295001 2.3.2 2.3.2 (CFR: 41.12143.4145.9145.10)
Importance Rating 2.5 2.9 Proposed Question: During 100% power, the Shift Manager authorizes an entry into the Steam Affected Area to find the source of a new steam leak. The Operator is given a time limit for the search based on expected dose rates. Just as the Operator enters the Steam Affected Area, an announcement is made that A Recirculation Pump has tripped.
The Operator should recognize ... ...
a) dose rates are less than expected and contact RP to request a verbal time extension.
RO1SRO b) dose rates are the same as expected and immediately leave the area due to the change in plant conditions.
112 c) dose rates are the same as expected and leave the area when the time limit is expi red.
d) dose rates are less than expected and immediately leave the area due to the change in plant survey conditions.
Proposed Answer: d) dose rates are less than expected and immediately leave the area due to the change in plant survey conditions.
v Explanation (Optional): Distractor A -AP-7.01, RWP Program would be violated - no provision for verbal changes to a RWP, would require new survey data &time to route for approval.
Requires recognition that SAA dose rates are power dependant & drop as power drops, also understand requirement to immediately leave area due to change from survey conditions.
Technical Reference(s): AP-7.01, AP-7.03, AP-7.06 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAP-7.01, EO-26.02 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 4
Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Partial or Complete Loss of AC / 6 Group # 1 1 Ability to determine andlor interpretthe following WA # 295003 AA2.01 AA2.01 as they apply to PARTIAL O R COMPLETE LOSS 0FA.C. POWER: (CFR: 41.10/43.5/45.13)
Cause of partial or complete loss of A.C. power Importance Rating 3.4 3.7 Proposed Question: From a normal full power operating condition, a complete and instantaneous loss of bus 10500 occurs. All equipment functions as designed. Several minutes after the loss, the bus is still deenergized.
Which one of the following is the cause?
a) Loss of DC Control Power to bus 10500 RO/SRO b) Actuation of the bus 10500 Degraded Bus Voltage timer 213 c) Ground fault trip of circuit breaker 10514 d) Overcurrent condition on CRD pump A motor Proposed Answer: c) Ground fault trip of circuit breaker 10514 Explanation (Optional):
Technical Reference(s): AOP-18 (Attach if not previously provided)
~~ ~
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-71E, EO-1.05.C, 1.10, SDLP-710, EO-1.23 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO L
Tier # 1 1 Partial or Total Loss of DC Pwr I6 Group ## 1 1 Knowledge of the interrelations between KIA# 295004 AK2.01 AK2.01 PARTIAL OR COMPLETE LOSS OF D.C. POWER and the following: (CFR: 41.7 / 45.8)
Battery charger Importance Rating 3.1 3.1 Proposed Question: A reactor startup is in progress at 5 % power when the following indications occur:
09-8-1-1, UPS INPUT DC VOLT LO 09-5-2-28, MAIN TURBINE EHC DC POWER LOSS TRIP 0 Loss of breaker position indication on; 10100, 10300, 10500 and 10700 busses Which one of the following is consistent with the above indications?
a) 09-8-1-22, 125 VDC BATT CHGR B AC SUPP TROUBLE and 09-4-3-10, RWR MG B GEN LOCKOUT b) 09-8-1-19, 125 VDC BATT CHGR A A C SUPP TROUBLE and RO/SRO 09-8-1-20, 125 VDC BATT A VOLT LO 314 c) 09-8-1-19, 125 VDC BATT CHGR A AC SUPP TROUBLE and 09-4-3-1, RWR MG A GEN LOCKOUT d) 09-8-1-22, 125 VDC BATT CHGR B AC SUPP TROUBLE and 09-8-1-23, 125 VDC BATT B VOLT LO Proposed Answer: b) 09-8-1-19, 125 VDC BATT CHGR AAC SUPP TROUBLE and 09-8-1-20, 125 VDC BATT A VOLT LO Explanation (Optional): The correct response, 'b)" are direct symptoms of AOP-45, LOSS OF DC POWER SYSTEM A.
The incorrect responses require stem evaluation to conclude that only the "A" Division is affected and recognition that "A" Busses are powered by 115 KV and therefore will NOT de-energize.
Technical Reference(s): AOP-45 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-71B, EO-1.05.A.2, 1.13A (As available)
Question Source: Bank #
Modified Bank .f: (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X
10 CFR Part 55 Content: 55.41 8 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Main Turbine Generator Trip I 3 Group # 1 1 Knowledge of the interrelations between MAIN K/A # 295005 AK2.07 AK2.07 TURBINE GENERATOR TRIP and the following:
(CFR: 41.7 I45.8)
Reactor pressure control Importance Rating 3.6 3.7 Proposed Question: The following conditions exist while performing a shutdown for a refueling outage:
Reactor power is 29%
Recirculation flow is at minimum.
A Main Turbine Trip occurs.
Which of the following is the correct plant and procedural response?
a) Turbine Bypass Valves may be able to control RPV pressure. Supplement RPV pressure control by opening Main Steam Line Drains.
b) Turbine Bypass Valves are not able to control RPV pressure. A manual scram ROISRO is required unless opening available steam drains is capable of controlling RPV pressure.
4/5 c) Turbine Bypass Valves may be able to control RPV pressure. A manual SCRAM is required due to the loss of feedwater heating.
d) Turbine Bypass Valves are not able to control RPV pressure. Insert Cram Groups per Reactor Analyst Instructions.
Proposed Answer: c) Turbine Bypass Valves may be able to control RPV pressure. A manual SCRAM is required due to the loss of feedwater heating.
Explanation (Optional): AOP-2 requires an immediate manual SCRAM if Power is 2 29% CTP, making "C" the correct response. With 25% BPV capacity and 3-5 % steam loads for auxiliary uses (RFP's, SJAE's, Steam Seals, etc.), the BPV's may be capable of controlling RPV pressure. Although possible, the BPV's may NOT be capable of controlling RPV pressure, the AOP-2 manual SCRAM requirement is not conditional on actions to control RPV pressure at 3 29%. At c 29% "B" may be correct. "D" is always incorrect in that no CRAM Groups exits at e 70 % Rod Line. "A" is not correct because a Rx SCRAM is required.
Technical Reference(s): OP-9, AOP-2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-94C, EO-I. 10.K (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge
Comprehension or Analysis X L 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 SCRAM I 1 Group # 1 1 Knowledge of the operational implications of the K/A # 295006 AKI .03 AKI .03 following concepts as they apply to SCRAM :
(CFR: 41.8 to 41.10)
Reactivity control Importance Rating 3.7 ' 4.0 Proposed Question: The Control Room Supervisor orders you to insert a manual scram because power is unexpectedly rising. Which of the following responses indicates that the scram has successfully controlled reactivity under all conditions?
a) Reactor power dropping rapidly through the IRM and SRM ranges ROJSRO b) 6 rods indicate position 02, remaining rods indicate position 00.
516 c) 1 rod indicates 48, 1 rod at 10, remaining rods indicate position 00.
d) Annunciators, 09-51-13, RPS A MAN SCRAM and 09-51-14, RPS B MAN SCRAM are in alarm.
Proposed Answer: b) 6 rods indicate position 02, remaining rods indicate position 00.
Explanation (OptionaI):
Technical Reference(s): AOP-1, EP-1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None v
Learning Objective: LP-AOP, EO-2.01, EOP2LP, EO-1.07 (As available)
Question Source: Bank # Dresden 2 INPO Bank# 6558 (Modified to JAF)
~
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 311 111996 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Control Room Abandonment / 7 Group # 1 1 Knowledge of the interrelations between WA # 295016 AK2.03 AK2.03 CONTROL ROOM ABANDONMENT and the following: (CFR: 41.7 /45.8)
Control room HVAC Importance Rating 2.9 3.1 Proposed Question: During normal full power steady-state operation, a 55-gallon drum of diesel fuel is inadvertently spilled in the Ventilation Equipment Room just west of the Control Room.
In a short time the Control Room Operators sense a strong concentration of gaseous fumes, which are becoming progressively more irritating to them.
Which of the following sequences would be expected by the Crew?
Place the Control Room Ventilation ISOL & Purge CNTRL switch in ; if conditions continue to worsen, consider entry into a) ISOLATE; AOP-43, Plant Shutdown From Outside the Control Room ROlSRO b) ISOLATE; AOP-28, Operation during Plant Fires 6/? c) PURGE; AOP-43, Plant Shutdown From Outside the Control Room d) PURGE; AOP-28, Operation during Plant Fires Proposed Answer: c) PURGE; AOP-43, Plant Shutdown From Outside the Control Room Explanation (Optional): Although a flammable liquid, AOP-28 is only entered for a confirmed fire. As a Toxic Gas is entering the Control Room environment, AOP-43 entry is warranted. Operating Control Room Ventilation in the ISOLATE Mode will trap the toxic gas in the Control Room while the PURGE Mode will turnover the Control Room atmosphere volume to outside air.
Technical Reference(s): AOP-43, OP-55B (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAOP, EO-I .03.A, SDLP-70, EO-I .06 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since ?0/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 v
Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Partial or Total Loss of CCW / 8 Group # 1 1 Knowledge of the operational implications of the WA # 295018 AK1.01 AK1.O1 following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : (CFR: 41.8 to 41.10)
Effects on componenffsystem operations Importance Rating 3.5 3.6 Proposed Question: The plant is operating at 90% power with one Reactor Building Closed Loop Cooling (RBCLC) pump tagged out of service. An electrical problem causes the two running RBCLC pumps to trip.
Operators have the ability to restore cooling via Emergency Service Water to EACH of the followinq EXCEPT:
a) RWCU Non- Regenerative Heat Exchanger RO/SRO b) Drywell Cooling Assemblies 718 c) Recirculation Pump Seal Coolers d) Drywell Equipment Sump Cooler Proposed Answer: a) RWCU Non- Regenerative Heat Exchanger Explanation (Optional):
Technical Reference(s): AOP-11 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-15, EO-I .09, SDLP-468, EO-1.06.B (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Partial or Total Loss of Inst. Air / 8 Group # 1 1 WA # 295019 24a3 ??a26 2.1.7 2.1.7 (RANDOMLY RESELECTED)
Ability to evaluate plant performance and make Operational Judgements based on operating characteristics/reactorbehavior and instrument interpretation.
(CFR: 43.5/45.12/45.13)
Importance Rating ?&
3.7 4.4 Proposed Question: The plant is in a refueling outage. The spent fuel pool gates are removed. Thirty, (30) minutes after a complete loss of Instrument Air occurs, you note that 09-4 Refuel Water Level (02-3Ll-86) indicates that RPV Level has risen several inches over the last hour.
Which of the below is the probable cause?
a) RWCU Blowdown Flow Control Valve (12FCV-55) failed closed.
RO/SRO b) In-Service CRD Flow Control Valve (03FCV-19NB) failed open.
8/9 c) In-Service Fuel Pool Filter/ Demineralizer has isolated d) Feedwater Low Flow Control Valve, (34FCV-137), loss of air signal.
Proposed Answer: a) RWCU Blowdown Flow Control Valve (12FCV-55) failed closed.
Explanation (Optional): Question forces conclusion that CRD is in service and the RWCU Blowdown Mode is being used for level control. CRD FCVs fail closed on loss of air while FW Low Flow Control Valve fails As IS. FPCC has no effect on level.
Technical Reference(s) AOP-12, SDLP-39 1.09f,j (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAOP, EO-1.10 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 4,7,10 55.43 5
Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Loss of Shutdown Cooling I 4 Group # 1 1 Ability to operate andlor monitor the following as WA # 295021 AA1.04 AA1.04 they apply to LOSS OF SHUTDOWN COOLING :
(CFR: 41.7 145.6)
Alternate heat removal methods Importance Rating 3.7 3.7 Proposed Question: The plant is in an outage with the RBCLC System out of service when a loss of shutdown cooling occurs. The cavity is flooded and the spent fuel pool gates are installed. The current decay heat load of the core and spent fuel pool is 1 . 8 ~O6 1 BTU/hr.
Which decay heat removal lineup listed below will provide sufficient decay heat removal?
a) RWCU in blowdown mode - leave gates installed RO/SRO b) Fuel pool cooling system - remove gates 9110 c) Decay heat removal system - leave gates installed.
d) RWCU in recirculation mode - remove gates.
Proposed Answer: a ) RWCU in blowdown mode - leave gates installed Explanation (Optional):
Technical Reference(s): AOP-30 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: AOP-30, Attachment 3 Learning Objective: SDLP-10, EO 1.15.a (As available)
Question Source: Bank # JAF LOR 20004206802C Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Refueling Acc Cooling Mode I 8 Group # 1 1 Knowledgeof the reasons for the following WA# 295023 AK3.02 AK3.02 responses as they apply to REFUELING ACCIDENTS : (CFR: 41.5 / 45.6)
Interlocks associated with fuel handling equipment Importance Rating 3.4 3.8 Proposed Question: Fuel Handling operations are in progress. The following conditions exist:
0 Mode switch in REFUEL 0 Fuel Grapple NOT loaded 0 Fuel Grapple full up 0 One (1) control rod selected and fully withdrawn Bridge over the Spent Fuel Pool From these conditions, which one (1) of the following restrictions occurs and why?
a) Bridge motion near or over the core will not be permitted to prevent bridge operator overexposure.
Bridge motion near or over the core will not be permitted to prevent inadvertent criticality.
A second control rod selected will cause a rod block to prevent bridge operator overexposure.
A second control rod selected will cause a rod block to prevent inadvertent criticality.
Proposed Answer: d) A second control rod selected will cause a rod block to prevent inadvertent criticality.
Explanation (Optional): JAF Safety Evaluation-96-013 R.3 FSAR Section 14.6.1.4 Refueling Accidents. "The refueling interlocks, which impose restrictions on the movements of refueling equipment and control rods, prevent inadvertent criticality during refueling operations.. ..
Technical Reference(s): ST-20F (Attach if not previously provided)
JAF Safety Evaluation-96-013 R.3 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O8B, EO-1.02, 1.05.B (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5
55.43 u
I Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Drywell Pressure / 5 Group # 1 1 Knowledge of the operational implications of the WA # 295024 EK1.01 EK1.01 following concepts as they apply to HIGH DRYWELL PRESSURE : (CFR: 41.8 to 41.10)
Drywell integrity: Plant-Specific Importance Rating ' 4.1 4.2 Proposed Question: From full power operation, a SCRAM occurs due to a sudden loss of coolant in the primary containment, Following the initial SCRAM response, Operators note the following:
Drywell pressure............ 6 psig (rising slowly)
Drywell temperature...... 280 a F (rising slowly)
Torus pressure ............. 4 psig (rising slowly)
Torus level ................... 14.1' (steady)
Several Crew members recommend initiating Drywell Sprays to restore Containment temperature and pressure.
Which of the following, identifies the correct response to this recommendation and why?
Drywell Sprays.. ...
a) Should be initiated to prevent Containment damage due to chugging of the downcommers.
b) Should NOT be initiated because sprays could cause chugging of the ROlSRO downcommers.
11/12 c) Should be initiated to prevent excessive differential pressure between the Drywell and Torus.
d) Should NOT be initiated because sprays could cause excessive differential pressure between the Drywell and Torus.
Proposed Answer: d) Should NOT be initiated because sprays could cause excessive differential pressure between the Drywell and Torus.
Explanation (Optional): Requires recognition that Sprays are warranted on Containment temperature leg and NOT warranted on Containment pressure leg.
Requires application of DWSIL Curve and understanding of the basis of the curve.
Chugging is a concern on pressure leg but are not correct because Sprays are warranted on the temperature leg.
Not allowed on either leg based upon DWSIL Curve which is based upon sufficient Non-condensibles in Torus.
Distradors, a & b- Chugging is NOT a concern until after Torus violates DW Spray Limit when pressure rises to a higher value than 4 psig.
Technical Reference(s): EOP-4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOP's Learning Objective: MIT-301.11E, E04.07 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW
, Question History: Last NRC Exam v
(Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Reactor Pressure 1 3 Group # 1 1 Ability to operate andlor monitor the following as WA # 295025 EA1.07 EA1.07 they apply to HIGH REACTOR PRESSURE:
(CFR: 41.7 145.6)
ARIIRPTIAWS: Plant-Specific Importance Rating 4.1 4.1 Proposed Question: Which ONE of the following describes the effect a reactor vessel pressure signal of 1170 psig will have on the reactor recirculation pumps and alternate rod insertion (ARI) system?
The Recirculation motorIgenerator...
a) drive motor breakers will trip and the ARI solenoid valves will energize.
ROISRO b) generator field breakers will trip and the ARI solenoid valves will energize.
12/13 c) drive motor breakers will trip and the ARI solenoid valves will de-energize d) generator field breakers will trip and the ARI solenoid valves will de-energize Proposed Answer: a) drive motor breakers will trip and the ARI solenoid valves will energize.
Explanation (Optional):
Technical Reference(s): ITS-3.3.4.1/SR-3.3.4.1.4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-02H EO 1.05.C.2, SDLP-03C E01.05.C.2 (As available)
Question Source: Bank # Quad Cities 1 INPO Bank # 16832 (Modified for JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 3/16/1998 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 2 Comments:
Examination Outline Cross-reference: Level SRO Tier # 1 Control Room Abandonment I7 Group # 1 Ability to perform specific system and integrated W A # 295016 2.1.23 plant procedures during different modes of plant oDeration. (CFR: 45.2 / 45.6)
Importance Rating 4.0 Proposed Question: A Plant shutdown is in progress after a year of continuous full power operation. The following conditions currently exist:
All Control Rods fully inserted e Mode switch is in SHUTDOWN e RPV pressure at 900 psig, controlled by EHC e A normal forced cooldown to < 212 F is commenced using EHC.
O At this point, a verbal report is received in the Control Room of a significant fire in the Cable Spreading Room.
Simultaneously, an un-expected half SCRAM occurs on RPS A.
Based on these conditions, which one (1) of the following methods should be utilized to depressurize and cooldown the RPV?
a) AOP-55, Alternate Shutdown Cooling Due To Plant Fires.
RO/SRO b) AOP-43, Plant Shutdown From Outside The Control Room.
S I4 c) EP-11, Alternate Depressurization using SRVs From 02ADS-71.
d) OP-65, Startup And Shutdown Procedure.
Proposed Answer: b) AOP-43, Plant Shutdown From Outside The Control Room.
Explanation (Optional): a) AOP-55 is only used with AOP-28, which is not applicable.
b) AOP-43 is required in Mode 3 (Mode Switch in Shutdown and 212 F), the Cable Spread Area is in AOP-43 Areas. Verbal report, significant fire, un-expected equipment operation are in decision tree for AOP-43.
c) EP-11 is only used when in the EOPs.
d) OP-65 is used for a Normal Shutdown and Cooldown.
Technical Reference(s): AOP-43 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AP, EO-46.03 (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge
Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6,7,10 55.43 5 Comments:
Examination Outline Cross-reference: Level SRO W
Tier # 1 Partial or Total Loss of Inst. Air I 8 Group # 1 Ability to determine andlor interpret the following WA # 295019 AA2.01 as they apply to PARTIAL OR CO OF INSTRUMENT AIR :(CFR: 41.10 Instrument air system pressure Importance Rating 3.6 Proposed Question: Given the following conditions at 100% power:
Condenser Vacuum at 26" and slowly worsening The Blue SCRAM lights are ON for Control Rods 26-27 and 42-19 09-6-2-38, Breathing Air HDR Press Lo alarm is in Condensate Pump Minimum Flow valve is OPEN 09-5-2-3, Rod Drift Alarm is in Which of the below describes the expected plant condition and appropriate mitigative procedure?
a) Air header pressure above 65 psig. AOP-42, Feedwater Malfunction, Lowering Feedwater Flow and AOP-27, Control Rod Drift.
RO/SRO b) Air header pressure below 85 psig. AOP-12, Loss of Instrument Air and AOP-31, Loss of Condenser Vacuum.
SI5 c) Air header pressure above 65 psig. AOP-12, Loss of Instrument Air and AOP-27, Control Rod Drift.
t d) Air header pressure below 85 psig. AOP-31, Loss of Condenser Vacuum and AOP-42, Feedwater Malfunction, Lowering Feedwater Flow.
Proposed Answer: b) Air header pressure below 85 psig. AOP-12, Loss of Instrument Air and AOP-31, Loss of Condenser Vacuum.
Explanation (Optional): Pressure is not above 65 psig. Must conclude < 65 psig because a rod has drifted due to Scram Valve opening. Scram Air Header low is set at 65 psig to alarm before any rods drift. Pressure is < 85 psig= Breathing Air isolates at 85 psig. AOP-12 &31 have appropriate mitigative actions. AOP-42 is a symptom of the true problem and loss of feedwater flow does not truly exist. AOP-27 provides no guidance if the AOP-12 required Manual SCRAM is inserted.
Technical Reference(s): AOP-12 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-39, EO-1.15.A (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge
Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Reactor Pressure / 3 Group # 1 1 Ability to determine and/or interpret the following K/A # 295025 EA2.06 EA2.06 as they apply to HIGH REACTOR PRESSURE:
(CFR: 41.10 143.5 f 45.13)
Reactor water level Importance Rating 3.7 3.8 Proposed Question: Following a reactor SCRAM and MSlV isolation, HPCl is injecting into the reactor. RPV level on narrow Range is 200" and rising. Reactor pressure is 800 psig and rising.
The HPCl turbine will trip ....
a) At a lower indicated NR level at 800 psig than at 1100 psig.
RO/SRO b) At a higher indicated NR level at 800 psig than at 1100 psig 13/16 c) At the same indicated NR level at 800 psig and at 1100 psig d) When NR level indication reaches 222.5".
Proposed Answer. a) At a lower indicated NR level at 800 psig than at 1100 psig Explanation (Optional):
Technical Reference(s): OP-15, attachment 3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-23, EO-1.05.C.1, 1.13 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Suppression Pool High Water Group # 1 1 Temp. / 5 Knowledge of the operational implications of the WA # 295026 EK1.O1 EK1.O1 following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE : (CFR: 41.8 to 41.10)
Pump NPSH Importance Rating 3.0 3.4 Proposed Question: The following plant conditions exist:
Torus Pressure-I .O psig Torus Level- 11.92 feet What is the maximum Torus water temperature that two (2) RHR Pumps can operate at 8,000 gpm each without exceeding NPSH limitations?
a) 173°F RO/SRO b) 182°F 14/17 c) 200" F d) 206' F Proposed Answer: c) 200 F Explanation (Optional):
Technical Reference(s): OP-I 3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: OP-l3A, Attachment # 1 Learning Objective: SDLP-13, EO-I .13.A (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Drywell Temperature / 5 Group # 1 1 Ability to determine andlor interpret the following WA # 295028 EA2.01 EA2.01 as they apply to HIGH DRYWELL TEMPERATURE (CFR: 41.10/43.5/45.13)
Drywell temperature Importance Rating 4.0 4.1 Proposed Question: EOP4, PRIMARY CONTAINMENT CONTROL, has been entered due to a low Torus water level condition. Simultaneously, EPIC power is lost and no Control Room computer screens are available.
The following indications exist:
0 DW TEMP A, 16-1TR-108 reading 140 O F 0 DW TEMP B, 16-1TR-107 reading 132 F 0 DW COOLER A TEMP, 68TE-100 reading 160 In, 120 Out 0 DW COOLER B TEMP, 68TE-100 reading 140 In, 120 Out Which of the below describes the expected operator action, if any?
a) EOP-4 reentry on DW Cooler A and B average inlet temperature.
ROlSRO b) EOP-4 reentry is not required on subsequent entry conditions 15118 c) EOP-4 reentry on DW TEMP A and B average temperature.
d) No additional EOP-4 entry condition exist.
Proposed Answer: c) EOP-4 reentry on DW TEMP A and B average temperature.
Explanation (Optional): Question tests application of EP-1 to determine alternative indications on loss of computer. Also tests EP-1 expectation to reenter EOP's (4.2.1.A). Correct answer is "C". "A" does not use correct indications. "B"& "D" violate 4.2.1.A.
Technical Reference(s): EP-1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EP-1 (excluding section 4.7)
Learning Objective: (As available)
Question Source: Bank # JAFLOR20005204B06C Rev.1
~
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Low Suppression Pool Wtr Lvl I 5 Group # 1 1 Ability to interpret control room indicationsto K/A# 295030 2.4.48 2.4.48 verify the status and operation of system I and understand how operator actions and directives affect plant and system conditions.
(CFR: 43.5 / 45.12)
Importance Rating 3.5 3.8 Proposed Question: While experiencing torus water level control problems, an operator opens an ADS valve with torus water level at 5.2 ft.
Opening the SRV under these conditions will result in:
a) direct suppression chamber pressurization.
ROlSRO b) excessive hydrodynamic loading of SRV Tailpipe.
16/19 c) valve seat damage from the excessive flowrates.
d) drawing water up into the tailpipe.
Proposed Answer: a) direct suppression chamber pressurization Explanation (Optional):
Technical Reference(s): EOP-2 (Attach if not previously provided)
W Proposed references to be provided to applicants during examination: None Learning Objective: MIT 301.11E- EO 4.03 (As available)
Question Source: Bank # Dresden 1INPO # 6483 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 912611998 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge F Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Reactor Low Water Level I2 Group # 1 1 Knowledge of the reasons for the following W A # 295031 EK3.01 EK3.01 responses as they apply to REACTOR LOW WATER LEVEL : (CFR: 41.5 / 45.6)
Automatic depressurization system actuation Importance Rating 3.9 4.2 Proposed Question: From full power and with HPCl inoperable, a SCRAM occurs from a small primary leak in the drywell simultaneous with a loss of offsite power. EDG's start and reenergize vital busses. RClC initiates, but insufficient injection results in RPV water level continuing to lower.
Which of the following is correct assuming operator action?
a) SRVs should open on their automatic pressure relief setpoints and lower reactor pressure to permit level recovery injection with low pressure ECCS.
b) SRV's assigned to ADS should open when RPV level lowers to an assigned RO/SRO setpoint to permit level recovery injection with low pressure ECCS.
17/20 c) A residual bus transfer will result in automatic start and injection by the condensate booster pumps.
d) SRVs should cycle on their automatic pressure relief setpoints and together with the RClC injection will provide steam cooling with injection.
Proposed Answer: b) SRV's assigned to ADS should open when RPV level lowers to an assigned setpoint to permit level recovery injection with low pressure ECCS.
Explanation (Optional): a) incorrect, automatic pressure relief will reset and SRV's will close before pressure lowers enough to enable low pressure ECCS injection.
b) The correct response is intended to backup HPCl B enable low pressure ECCS.
c) Incorrect- A residual transfer locks out the CBP's and they can only be started by operator action.
d) Incorrect-Steam Cooling with or without injection requires core uncovery.
Additionally, ADS SRV's will open on ADS function rather than cycle on over pressure.
Technical Reference(s) OP-68 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O2J, EO-1.01, 1.05.A, 1.05.C (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43
Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 SCRAM Condition Present and Power Above APRM Group # 1 1 Downscale or Unknown / 1 Ability to determine andlor interpret the following WA # 295037 EA2.02 EA2.02 as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN :
(CFR: 41.10 / 43.5 / 45.13)
Reactor water level Importance Rating 4.1 4.2 Proposed Question: As directed by EOP-3, the current RPV level band is -19 to 110 inches and being controlled at 80-100 inches with Feedwater.
Which of the following is the preferred instrumentation for maintaining the 80-100 inches band?
a) Narrow Range.
RO/SRO b) Wide Range.
18/21 c) Refuel Zone.
d) Fuel Zone.
Proposed Answer: b) Wide Range.
Explanation (Optional): a) Normal range is off scale low b) Wide Range is located at Panel 09-5 & hot calibrated c) Refuel Zone is cold calibrated and located remote to FW control (Panel 09-3 & 4) d) Fuel Zone is cold calibrated and located remote to RW control (Panel 09-3 & 4)
Technical Reference(s): SDLP-OZB, Table IV (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O2B, EO-I .05.A.3 (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Off-site Release Rate / 9 Group # 1 1 Knowledge of the interrelations between HIGH W A # 295038 EK2.02 EK2.02 OFFSITE RELEASE RATE and the following:
(CFR: 41.7 I45.8)
Offgas system Importance Rating 3.6 3.8 Proposed Question: While operating at full power, a large fuel leak develops.
Which of the following automatic responses from a high radiation signal will occur to limit off-site release rates?
a) Condenser Vacuum Pump trip.
ROfSRO b) Off gas System isolation 19/22 c) Hydrogen Addition System trip.
d) Reactor SCRAM.
Proposed Answer: b) Off gas System isolation.
Explanation (Optional): a) Condenser Vacuum Pump will still isolate at 3 X NFPB MSL Radiation, but is not allowed to be in service >5% CTP.
b) On a high setpoint and after a 15 minute delay, SJAE Off-Gas will isolate SJAE discharge to stack.
c) Although Hydrogen Injection flowrate is directly related to radiation levels throughout the plant, it has no MSL Hi Radiation signal trip.
d) MSL Radiation no longer provides an automatic Reactor SCRAM function.
Technical Reference(s) OP-24A (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-OIA, EO-I .05.C.1 (As available)
Question Source: Bank #
~ ~~~
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Plant Fire On Site / 8 Group # 1 1 Ability to operate and /or monitor the following WA # 600000 AA1.06 AA1.06 as they apply to PLANT FIRE ON SITE:
Fire alarm Importance Rating 3.0 3.0 Proposed Question: With the Fire Protection System in a normal standby lineup, which one ( I ) of the following Fire Protection Panel Alarms would you expect t o result in the start of one or more Fire Pumps?
a) Heat detection actuation in the West Cable Tunnel ROlSRO b) Heat detection actuation in the North EDG Switchgear Room 20/23 c) Ionization detector actuation in the Reactor Building 272 Drywell Entrance d) Ultraviolet Flame detector in the Recirculation M/G Room Proposed Answer: a) Heat detection actuation in the West Cable Tunnel Explanation (Optional):
Technical Reference(s): OP-33 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-76 EO 1 . 0 5 ~ (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge F Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Loss of Main Condenser Vac I 3 Group # 2 2 Ability to determine andlor interpret the following K/A # 295002 AA2.01 AA2.01 as they apply to LOSS OF MAIN CONDENSER VACUUM : (CFR: 41.10143.5145.13)
Condenser vacuumlabsolute pressure Importance Rating 2.9 3.1 Proposed Question: Reactor power is 38%, on the APRM's when annunciator 09-6-1-29, CONDENSER VAC LOW, alarms. Condenser vacuum, as read on control room meters, indicates 24.8" and lowering slowly. If vacuum continues to lower, WHICH ONE (1) of the following automatic protective actions would occur first?
a) Reactor Feed Pump Turbine Trip ROlSRO b) Main Turbine Trip 2 1I24 c) Bypass Valve Closure d) MSlV Closure Proposed Answer: b) Main Turbine Trip Explanation (Optional):
Technical Reference(s): OP-9, OP-2A, OP-I, AOP-31 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AOP, EO-I .02 (As available)
Question Source: Bank # Nine Mile Point 1 lNPO# 11813 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1/20/1998 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Reactor Pressure 1 3 Group # 2 2 Knowledge of the interrelations between HIGH W A # 295007 AK2.04 AK2.04 REACTOR PRESSURE and the following:
(CFR: 41.7 I45.8)
LPCS Importance Rating 3.2 3.3 Proposed Question: An Emergency Depressurization is to be performed from 700 psig to permit low pressure ECCS injection into the reactor. The only ECCS available is Core Spray System A. CS Pump A is running on minimum flow and all other components are in a normal standby condition. When SRV's are operated, only one (1) SRV responds. Reactor pressure lowers at approximately 10 psi/minute.
The Core Spray Injection Valve opens when reactor pressure goes below RPV injection immediately.
a) 450 psig: occurs RO/SRO b) 450 psig: does NOT occur 22/25 c) 310 psig: occurs d) 310 psig: does occur Proposed Answer: b) 450 psig: does NOT occur Explanation (Optional):
Technical Reference(s): OP-14 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-14, EO-1.13e, 1.14b (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Inadvertent Reactivity Addition I 1 Group # 2 2 Knowledge of the interrelationsbetween W A # 295014 AK2.07 AK2.07 INADVERTENT REACTIVITY ADDITION and the following: (CFR: 41.7 145.8)
Reactor power Importance Rating 3.9 3.9 Proposed Question: From normal full power operation, which ofthe following will result in a stable higher power level?
a) Inadvertently isolating the Reactor Water Cleanup System.
RO/SRO b) Raising 10100 Bus frequency.
23/26 c) Main Condenser Circulating Pump Trip.
d) Closing the manual extraction steam valve for Feed Heater 6B Proposed Answer: d) Closing the manual extraction steam valve for Feed Heater 6B.
Explanation (Optional): Explanation: 4 3 &bl A-A. /AI A<. i.a<fcw L a) Inadvertently isolating the Reactor Water Cleanup System results in higher feedwater temperature- therefore a lower power level.
b)Raising 10100 bus frequency will momentarily raise Recirc MG speed. Speed vs.
Speed demand will reduce it back down.
c)A Main Condenser Circulating Pump Trip will result in higher condensate and therefore feedwater temperature resulting in a lower power level.
d)The manual extraction steam valve for Feed Heater 6B closing will prevent the heating of the feedwater in the 6B heater, thereby, causing colder feedwater to enter the vessel and drive reactor power up.
Technical Reference(s): AOP-62, AOP-32, OP-3A (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AOP EO 1.02 (As available)
.Question Source: Bank # Clinton INPO #20412 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 7/23/200 1 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level SRO Tier # 1 Refueling Acc Cooling Mode / 8 Group # 1 Knowledge of the operational implications of the WA # 295023 AK1.01 following concepts as they apply to REFUELING ACCIDENTS :(CFR: 41.8 to 41.10) rds Importance Rating 4.1 Proposed Question: Core Alterations are in progress.
An irradiated fuel bundle being moved from the reactor cavity to the Spent Fuel Pool becomes ungrappled and falls into the reactor vessel downcomer area. (Between the vessel wall and the shroud). Bundle integrity is maintained.
Which of the below describes the person at greatest risk?
a) Mechanic working on Torus to Drywell Vacuum Breaker.
RO/SRO b) Refuel SRO on the Bridge.
S27 c) 18C Technician at SLC Skid.
d) Mechanic working on SRVs.
Proposed Answer: d) Mechanic working on SRVs.
Explanation (Optional):
Technical Reference(s): RAP-7.1.1.048 (Attach if not previous.,' provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AP, RAP-7.1.04873.03 (As available)
Question Source: Bank # Clinton INPO # 20401 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 7/23/2001 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 9 55.43 4 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Secondary Containment Group # 2 2 Area Temperature I 5 Ability to perform procedures to reduce W A # 295032 2.3.10 2.3.10 excessive levels of radiation and guard against personnel exposure. (CFR: 43.4 / 45.10)
Importance Rating 2.9 3.3 Proposed Question: The plant was operating at 100°/b power when a un-isolable steam leak occurred inside the Reactor Building. EOP-5, SECONDARY CONTAINMENT CONTROL, had been initially entered on an area temperature > MAX NORMAL. When the first full set of data were collected, Reactor Building conditions were as follows:
0 HPCI drywell entrance temperature.................... ,253 F.
RClC drywell entrance temperature.................... ,240 F.
R.B. 272 ft. elevation southwest temperature......... 130 F.
A RHR Heat Exchanger Room temperature...........110 F.
West HCU area radiation level........................... ,100 mr/hr.
R.B. Access 272 R. elev. area radiation level.. ........10 mr/hr.
CRD Removal Hatch Area radiation level...............250 mr/hr.
Which one of the following is correct for these plant conditions?
a) Evacuate the Protected Area, enter EOP-2, open the Bypass Valves fully b) Evacuate the Protected Area, enter EOP-2, commence an orderly Plant cool-RO/SRO down using the SRVs and/or the Bypass Valves.
24/28 c) Evacuate the Reactor Building, exit EOP-5 and enter EOP-2; perform an Emergency RPV Depressurization per EOP-2.
d) Evacuate the Reactor Building, concurrently with EOP-5, enter EOP-2, and perform an Emergency RPV Depressurization.
Proposed Answer: d) Evacuate the Reactor Building, concurrently with EOP-5, enter EOP-2, and perform an Emergency RPV Depressurization.
Explanation (Optional): EOP-5 is initially entered, the R.B. 272 R. elevation southwest temperature is above it's maximum normal value, the HPCl drywell entrance temperature and RClC drywell entrance temperature are greater than their maximum safe operating values and the West HCU area radiation level & CRD Removal Hatch Area radiation levels are above their maximum normal levels. EOP-5 then directs the operator to enter EOP-2, which directs the operator to shutdown the reactor. EOP-2 directs leaving the RPVIP leg and entering Emergency RPV Depressurization.
Correct answer is to evacuate the affected area, enter EOP-5 which directs an Emergency Depressurization due to > Max Safe Temperatures and entry into EOP-2, EOP-2 directs leaving RPV/P leg and using Emergency Depressurization leg. As a result- anticipating ED and using BPV's and SRV's for either anticipating ED or performing a shutdown and a normal cool down are not correct. EOP-5 and 2 are to be performed concurrently.
Evacuation of PA not required at this point as the problem is localized. Reactor Building evacuation is required.
Technical Reference(s): EOP-5, EOP-2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOP's
Learning Objective: EOP5LP, EO-1.07 (As available)
Question Source: Bank # JAF LOR # 0877 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 12 55.43 4 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Secondary Containment Group # 2 2 Area Radiation Levels / 9 Ability to operate andlor monitor the following as WA # 295033 EA1.05 EA1.05 they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS : (CFR: 41.7I45.6)
Affected systems so as to isolate damaged portions Importance Rating 3.9 4.0 Proposed Question: During power operation, the Area Radiation Monitor (ARM) for the CRD Removal Hatch area alarms, together with receipt of a Fire Protection System ionization detector alarm in the Southwest Drywell Entrance Area.
Which system(s) should be considered for manual isolation?
a) HPCl and RWCU RO/SRO b) RClC and Main Steam 25/29 c) Main Steam and RWCU d) HPCl and RClC Proposed Answer: d) HPCl and RClC Explanation (Optional):
Technical Reference(s): EOP-5 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: EOPSLP, EO-1.07 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level SRO Tier # 1 High Reactor Pressure I 3 Group # 1 Ability to determine andlor interpret the following WA# 295025 EA2.05 as they apply to HIGH REACTOR PRESSURE:
(CFR 41 1014!?1.5/45 13)
Decay heat generation Importance Rating 3.6 Proposed Question: After a long period of full power operation, an instantaneous loss of all AC power occurs and is not corrected. The HPCl System failed to start.
Without initial operator action, over the next hour you would expect (1) and subsequent Operator action to be (2) ?
a) (1) SRVs to open and close periodically on mechanical overpressure.
(2) Commencing a cooldown at 100 F/ hr.
O b) (1) SRV's initial operation with RClC operation precluding the need for further RO/SRO SRV operation.
(2) Commencing a cooldown at 100 F/ hr.
S30 c) (1) SRV's to open and close periodically on mechanical overpressure.
(2) Commencing a cooldown at less than 20 F/ hr.
d) (1)SRV's initial operation with RClC operation precluding the need for further SRV operation.
(2) Commencing a cooldown at less than 20 F/ hr.
O Proposed Answer: c) (1) SRV's to open and close periodically on mechanical overpressure (2) Commencing a cooldown at less than 20 F I hr.
Explanation (Optional): Must recognize that RClC Steam demand is far less than decay heat steam generation. Must also recognize that stem conditions describe AOP-49 applicability thus limiting cooldown rate to < 20 F/ hr.
Technical Reference(s): AOP-49 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AOP, EO-1.10 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Secondary Containment High Sump/Area Water Group # 2 2 Level 1 5 Knowledge of the reasons for the following W A # 295036 EK3.01 EK3.01 responses as they apply to SECONDARY CONTAINMENT HIGH SUMPlAREA WATER LEVEL : (CFR: 41.5 145.6)
Emergency depressurization Importance Rating 2.6 2.8 Proposed Question: While operating at full power, an earthquake has resulted in the following:
A severe piping crack between the CSTs and the Torus results in a rapid addition of water to the Torus Room and both Crescent Areas.
A small, un-isolable leak in the RWCU Pump suction piping in the Reactor Building.
Crescent Area water levels are 19 rising Highest Reactor Building Area (RB 300 Southwest) temperature is 103°F Why must an Emergency Depressurization be performed for these conditions?
a) A loss of CST inventory will result in total loss of HPCl and RClC for inventory control.
RO/SRO b) Operability of equipment located in the Crescents is threatened by Crescent water level rise.
2613 1 c) Primary Containment integrity is threatened by Torus Room water level rise.
d) Operability of RPV Water Level instruments located on Reactor building 300 is challenged.
Proposed Answer: b) Operability of equipment located in the Crescents is threatened by Crescent water level rise.
Explanation (Optional):
Technical Reference(s): EOP-5 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOPs Learning Objective: EOP5LP, EO-1.07 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High CTMT Hydrogen Conc. / 5 Group # 2 2 Ability to operate and monitor the following as W A # 500000 EA1.06 EA1.06 they apply to HIGH CONTAINMENT HYDROGEN CONTROL: (CFR: 41.7 145.6)
Drywell sprays Importance Rating 3.3 3.4 Proposed Question: A LOCA has occurred with the following conditions:
Drywell Hydrogen........... .6.03O h Drywell Oxygen ............... 5.40 O h Torus Hydrogen.. .............5.70 O h Torus Oxygen ..................3.00 O h Torus Water Level.. .......... 13.92 Feet Reactor Pressure.............. 0 psig Offsite release rates will not exceed the release rate LCO What actions are required to control containment gas?
a) Establish Torus Sprays. When Drywell Hydrogen is less than .6 %, Vent and Purge the Drywell.
b) Establish Torus Sprays. Vent and Purge the Drywell until'Drywell Hydrogen is ROlSRO less than .6 %.
27132 c) Establish Drywell Sprays. When Torus Hydrogen is less than .6%, Vent and Purge the Torus.
d) Establish Drywell Sprays. Vent and Purge the Torus until Torus Hydrogen is less than .6 %.
Proposed Answer: d) Establish Drywell Sprays. Vent and Purge the Torus until Torus Hydrogen is less than .6 YO.
Explanation (Optional): The 2 legs directed by the conditions are DWIG-3 and T/G-I, making 'D' the correct response. Torus Sprays are not required by the stem conditions, eliminating ' A and W. Venting and Purging of Torus is required to reduce hydrogen to <.6 % - not contingent on being below .6%.
Technical Reference(s): EOP-4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOP's Learning Objective: EOP4LP, EO-4-03 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge
Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 , 10 55.43 5 Comments:
Examination 0ut1ine Cross-reference: Level RO SRO Tier # 2 2 RHRILPCI: Injection Mode Group # 1 1 Ability to predict andlor monitor changes in W A # 203000 A I .05 A I .05 parameters associated with operating the RHWLPCI INJECTION MODE (PLANT SPECIFIC) controls including: (CFR: 41.5 / 45.5)
Suppression pool level Importance Rating 3.8 3.7 Proposed Question: A Design Basis LOCA has occurred. ECCS systems are injecting into the reactor. Torus Level is at 12.8 feet and lowering.
Which one of the following is the expected response of Low Pressure Coolant Injection (RHR)?
a) The RHR pumps will continue to operate regardless of Torus Level until the pumps trip on motor overload.
ROISRO b) The RHR pumps will automatically trip when Torus Level drops to the RHR Pump Vortex Limit.
28/33 c) The RHR pumps will continue to operate regardless of Torus Level due to automatic bypass of all trip signals.
d) The RHR Pump Torus Suction valves will automatically close. The RHR pumps will trip on Interlock.
Proposed Answer: a) The RHR pumps will continue to operate regardless of Torus Level until the pumps trip on motor overload.
Explanation (Optional):
Reference:
ESK-5BU. Pump trips are:
L 0 No Suction Path 0 Breaker electrical protection interlocks 0 EDG programmed restart sequence 0 09-3 and breaker mounted control switches The RHR Suction path MOVs have no auto stroke provisions. They only have open permissive interlock to prevent cross connecting the suction flow path.
Technical Reference(s): OP-13A (Attach if not previously provided)
ESK-5BU Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-10, EO-1.1O.f (As available)
Question Source: Bank # Grand Gulf 1 INPO # 16342 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 411ROO0 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5
55.43 Comments:
Examination Outline Cross-reference: Level SRO Tier # 1 Low Suppression Pool Wtr Lvl / 5 Group # 1 Ability to determine andlor interpret the following K/A # 295030 EA2.04 a s they apply to LOW SUPP ON POOL WATER LEVEL :(CFR: 41.10 / 45.13)
DrywelU suppression chamber differential pressure:
Mark-l&Il Importance Rating 3.7 Proposed Question: Two hours into the shift, the SNO reports that Torus water level has dropped from 14.0 ft to 13.91 ft while Drywell to Torus DIP has dropped from 1.8 psid to 1.6 psid and Torus pressure has remained constant at 0.0 psig. You have confirmed the indications on EPIC-LOGI.
Your required actions are__.._..._.
a) Enter EOP-4, Primary Containment Control, and immediately makeup nitrogen to the Drywell restore DIP.
b) Enter EOP-4, Primary Containment Control, and immediately makeup water to RO/SRO the Torus to restore Torus level.
s34 c) Enter AOP-9, Loss of Primary Containment Integrity, and dispatch Operators to search for Primary Containment leakage.
d) Enter AOP-9, Loss of Primary Containment Integrity, and dispatch Operators to determine why RBCLC temperature has risen.
Proposed Answer: c) Enter AOP-9, Loss of Primary Containment Integrity, and dispatch Operators to search for Primary Containment leakage.
Explanation (Optional): 0 EOP-4 Entry conditions of 13.88 ft or 2.7 psig DW pressure do not exist, therefore EOP-4 Entry is not required.
0 Adequate AOP-9 symptoms do exist warranting entry 0 The stem conditions are symptomatic of Primary Containment leakage. A RBCLC temperature rise will cause the reverse of the indications based on higher DW temperatures caused by less heat removal by the DW cooling units.
Technical Reference(s1: OP-37 (Attach if not previously provided)
Proposed references t o be provided to applicants during examination: EOP's Learning Objective: SDLP-166, EO-1.09d (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10
55.43 5 Comments:
Examination Outline Cross-reference: Level SRO W
Tier ## 2 SLC Group ## 1 Ability to direct personnel activities inside the WA # 211000 2.1.9 control room.
(CFR: 45.5 / 45.12 / 45.13)
I rnportance Rating . .
4.0 Proposed Question: A failure to SCRAM has occurred with the following conditions:
RPV water level is being controlled 80 to 110 inches using HPCI.
RPV pressure is being controlled 800 to 1000 psig using SRVs.
Initial SLC Tank level - 80 O h .
0 Current SLC Tank level - 58 %.
Reactor Power - 10 O h .
Torus water level - 14.0 feet.
0 Torus water temperature - 120 "F.
0 There is indication of a Steam Line break.
0 The 10300, 10400 and 10700 busses are de-energized.
Based upon these indications, you should order:
Terminate and Prevent all injection except SLC, RClC and CRD until APRMs are downscale, RPV level is at TAF, or SRV's remain closed.
Irrespective of the resulting RPV cooldown rate, maintain RPV pressure below v
the Heat Capacity Temperature Limit.
Equalize and reopen the MSIV's per EP-9. Bypass the MSlV Low Water Level Isolation per EP-2.
Using HPCI, restore and maintain RPV Level between 177 and 222.5 inches.
Be cautious of rapid level changes.
Proposed Answer: d) Using HPCI, restore and maintain RPV Level between 177 and 222.5 inches. Be cautious of rapid level changes.
Explanation (Optional): Question was rewritten as SRO Only swapped original ROlSRO question 35/47 to make it a SRO Only- S35, SRO original question S35 was made RO/SRO question 35/47.
BllT at 10% power is 125 O F . 120 "F is not yet violating BIIT, Terminate and Prevent is not required until BllT is violated.
At a Torus level of 14 ' and RPV pressure of 1000 psig, the HCTL will be violated at 172 "F. Torus temperature of 120 "F is far from violating the HCTL.
Without the 10300, 10400 & 10700 Busses- there are no Circ Water Pumps, thus the Main Condenser is not available.
This is the correct response- When SLC Tank Level drops by 22 % (80 to 58%) level restoration is required with caution that rapid level changes may cause a reactivity excursion.
Technical Reference(s): EOP-3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOP's
Learning Objective: EOPBLP, EO-I .07 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 6 55.43 Comments:
-- Examination Outline Cross-reference:
RHRILPCI: Injection Mode Level Tier #
Group #
RO 2
1 SRO 2
1 Ability to manually operate and/or monitor in the K/A # 203000 A4.11 A4.11 control room: (CFR: 41.7 145.5 to 45.8)
Indicating lights and alarms Importance Rating 3.7 3.5 Proposed Question: Given the following alarms and indications:
09-3-1-8, RHR SYS A LOGIC ACTUATED 09-3-1-27, RHR & CORE SPRAY INJ VLV PERM e 09-3-1-34, RWR INJ VLV PERM e RPV pressure 200 psig and lowering e LPCl A Inboard and Outboard Injection Valves OPEN e EPIC is unavailable e 09-3, RHR System Flow Indications unavailable Which of the following indications can be used to help verify that LPCl A is functioning as designed and injecting water into the RPV?
RHR PUMP RHR PUMP 1OMOV-I6A( B)
MTR AMPS DISC PRESS POSITION INDICATION a) lowering rising closed RO/SRO b) rising lowering closed 29/36 c) rising rising open d) lowering rising open Proposed Answer: b) rising lowering closed Explanation (Optional):
Technical Reference(s): OP-13A (. ttach if not previously pr vided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-10, EO 1.05.a.l.b (As available)
Question Source: Bank ## J A F LOR 20505001RHRC19 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Shutdown Cooling Group # 1 1 Knowledge of the effect that a loss or W A # 205000 K3.04 K3.04 malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: (CFR: 41.7 I45.4)
Recirculation loop temperatures Importance Rating 3.7 3.7 Proposed Question: The Plant is in Mode 4. A loss of Shutdown Cooling has occurred.
Which of the below is an acceptable Operator action to provide for reliable Reactor Coolant temperature monitoring?
a) Place at least one Recirculation Pump in service.
RO/SRO b) Establish an RPV Water Level Band of 177 - 234.5 inches.
30/37 c) Open either loop, Recirculation Pump suction and discharge valves.
d) Average the 02-3TR-89, RPV Vessel Metal Temperatures Recorder on Panel 09-21.
Proposed Answer: a) Place at least one Recirculation Pump in service.
Explanation (Optional): a) Correct response- restores recirculation loop temperature indications.
b) Incorrect- raising level to 234.5-270" will promote natural circulation. Further action will be necessary to restore temperature indication.
c) incorrect- action will accomplish nothing unless RPV level is raised to 234.5
- 270". (Normal is 177- 234.5") This is not a procedurally supported action unless it results from Recirculation loop startup.
d) Incorrect- action is a twist on AOP-30 options to monitor FDWfR Nozzle N4B INBD temperature on RPV Vessel Temperature Recorder, 02-3TR-89 at Panel 09-21.
Technical Reference(s): OP-l3D, AOP-30, ITS Definitions (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AOP, EO-1.03, 1.04 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 2 Comments:
Examination Outline Cross-reference: Level SRO Tier # 1 Reactor Low Water Level l 2 Group # 1 Ability to operate andlor monitor the following as K/A# 295031 EA 1.08 they apply to REACTOR LOW WATER LEVEL:
(CFR: 41.7 145.6)
Alternate injection systems: Plant Specific Importance Rating 3.9 Proposed Question: A startup is in progress at 20% CTP when an RPS electrical malfunction results in the following:
0 HPCIIRCIC & MSIV Isolation on High Temperature 0 Full Reactor SCRAM 0 One ( I ) rod remains at position 40 and one (1) other rod is at position 02. All other rods are Full In.
0 RPV water level is 150 inches, slowly trending down.
RPV pressure is 1000 psig, slowly trending up.
The correct course of action is to:
a) Enter EOP-3, stabilize RPV pressure, and maintain RPV level with FeedlCondensate.
b) Enter EOP-2, commence a normal cooldown, and maintain RPV level with ROlSRO FeedlCondensate.
S38 c) Enter EOP-3, commence a normal cooldown, and maintain RPV level with SLCICRD.
Iv d) Enter EOP-2, Emergency Depressurize, and maintain RPV level with SLCICRD.
Proposed Answer: b) Enter EOP-2, commence a normal cooldown, and maintain RPV level with FeedlCondensate.
Explanation (Optional):
Technical Reference(s): EOP-2, EP-1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOP Learning Objective: EOPZLP, EO-1.07 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since IOl95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 HPCl Group # 1 1 Knowledge of 10 CFR: 20 and related facility W A # 206000 2.3.1 2.3.1 radiation control reauirements. (CFR: 41 . I 2 / 43.4.
45.9 /45. IO)
Importance Rating 2.6 3.0 Proposed Question: The plant is at full power and HPCl is being operated in the Test (CST to'CST) lineup.
HPCl flow control is in automatic.
Which of the following will the general area Dose Rates for personnel in the vicinity of the HPCl Turbine?
a) Throttle Open 23MOV-21, Test Valve to CST.
RO/SRO b) Align HPCl to Torus Suction.
31/39 c) Place RHR System ' A into Torus cooling.
d) Close 23MOV-24, HPCl & RClC Test Valve to CST Proposed Answer: a) Throttle Open 23MOV-21, Test Valve to CST.
Explanation (Optional): In the Test Mode, throttle closed on 23MOV-21, raises discharge pressure. HPCl must raise speed to maintain flowrate. Raising speed requires more turbine steam flow and therefore higher dose rates.
a) Correct- Reduces Turbine Steam Flow.
b) If anything, Torus suction may raise dose rates.
c) No Impact on Dose Rates- opposite side of Building.
d) Cause same effect as closing 23MOV-21.
Technical Reference(s): OP-15, Step C.2.9, AP-07.03 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-23, EO-1.13.A, LPAP-28.03 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 12 55.43 4 Comments:
Examination Outline Cross-reference: Level SRO Tier # 1 Inadvertent Reactivity Addition / 1 Group # 2 Knowledge of the operational implications of the WA # 295014 AK1.01 following concepts as they apply to INADVERTENT REACTIVITY ADDITION :
(CFR: 41.8 to 41.10)
Prompt criiical
~~~~~~~~~~~~~~,~~~~
Importance Rating 3.8 Proposed Question: The plant is in a startup and control rods are being withdrawn to bring the Reactor critical. The selected control rod is two (2) notches from the ECP's predicted criticality when a control rod drop occurs. The control rod blade that dropped went from position 4 to 48.
Assuming no further Operator action, which of the following barriers are in place to prevent this type of event AND what is a potential impact?
a) ST-ZOA, Rod Worth Minimizer Functional Test, the Reactor will heat up until alpha T turns power.
ROlSRO b) ST-20A, Rod Worth Minimizer Functional Test, the Reactor will go critical until full SCRAM on IRM HI-HI trip.
S40 c) ST-Z3B, Control Rod Coupling Integrity Test, the Reactor will heat up until alpha T turns power.
d) ST-23B, Control Rod Coupling Integrity Test, the Reactor will go critical until full SCRAM on IRM HI-HI trip.
Proposed Answer: d) ST-23B, Control Rod Coupling lntegriiy Test, the Reactorwill go critical until full SCRAM on IRM HI-HI trip.
Explanation (Optional):
Technical Reference(s): ST-23B, FSAR-14.5.4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O3F, EO-1.13 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam
~
(Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 6 Comments:
Examination Outline Cross-reference: Level SRO Tier # 1 Loss of CRD Pumps / 1 Group # 2 Ability to determine andlor interpret the following WA # 295022 AA2.02 ly to LOSS OF CRD PUMPS : (CFR:
Importance Rating 3.4 Proposed Question: While at full power, the following alarms and indications are received:
09-5-143, CRD ACCUM PRESS LO or LVL HI 09-5-1-9, CRD CHARGING WTR PRESS LO 03PDI-303, DRV W R DlFF PRESS, indicates 0 psid.
03Fl-306, CLG WTR FLOW, indicates 0 gpm.
03FI-310, CRD FLOW CNTRL, indicates 0 gpm.
0 Several Yellow Accumulator lamps are lit on the Full Core Display.
Which of the following is the cause and the appropriate mitigating procedure?
a) 03CRD-56, CRD Charging Water Supply Header Isolation Valve, has been closed, ARP-09-51-9, CRD Charging WTR Press Lo.
RO/SRO b) OBFCV-lSA(B), in-service CRD Drivewater Flow Control Valve, has failed closed, AOP-69, Control Rod Drive Trouble.
S4 1 c) 03 MOV-22, CRD Cooling Water Pressure Control Valve, has been closed, ARP-09-51-9, CRD Charging WTR Press Lo.
d) 03P-I6A(B), in-service CRD Drive Water Pump has failed, AOP-69, Control Rod Drive Trouble.
Proposed Answer: d) 03P-I6A(B), in-service CRD Drive Water Pump has failed, AOP-69, Control Rod Drive Trouble.
Explanation (Optional):
Technical Reference(s): OP-25, AOP-69 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O3C, EO-1.12.8 (As available)
Question Source: Bank # Fermi 2 INPO # 8900 (Modified to JAF)
Modified Bank ## (Note changes or attach parent)
New Question History: Last NRC Exam 4/6/1998 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5
Comments:
Examination Outline Cross-reference: Level SRO Ld Tier # 1 Secondary Containment High Differential Pressure / Group # 2 5
Knowledge of the proces W A # 295035 2.3.9 containment purge. (CFR Importance Rating 3.4 Proposed Question: While at full power, a unisolable leak has developed in the RWCU suction piping in the Reactor Building. Secondary Containment pressure has risen due tQthe leak into the Reactor Building but is still slightly negative.
Which of the following will minimize the radiation hazard and control the Secondary Containment pressure?
a) Initiate SGT System and manually isolate Reactor Building Ventilation.
ROlSRO b) Ensure that SGT starts and Reactor Building Ventilation isolated when High D/P Setpoint is reached.
S42 c) Place all Crescent Area Unit Coolers in service.
d) Operate RWCU in the Blowdown Mode to the Main Condenser.
Proposed Answer: a) Initiate SGT System and manually isolate Reactor Building Ventilation.
Explanation (Optional):
Technical Reference(s): OP-20, OP-51A (Attach if not previously provided) v Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-OlB, EO-1.14.E (As available)
Question Source: Bank # JAF LOR 20005214BOlC Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 12 55.43 4 Comments:
Examination Outline Cross-reference: Level SRO LJ Tier # 1 Secondary Containment High SumpIArea Water Group # 2 Level / 5 Ability to determine andlor interpret the following WA # 295036 EA2.01 as they apply to SECONDARY CONTAINMENT WATER LEVEL:
nents within the affected area.
Importance Rating 3.2 Proposed Question: Thirty (30) minutes after an earthquake, the followin 0 RPV water level is 185 inches increasing RPV Pressure is 1000 psig increasing All control rods are at position 00, except for one (1) rod at position 22 0 One (1) foot of water is on the Crescent Floors due to a leaking Torus drain flange.
0 The MSlVs are Closed.
Reactor Scram has been reset.
0 Torus level is 10.75 feet and slowly lowering.
Based on these conditions, which statement below correctly states the procedure to be used and the basis for the action?
a) EOP-3 action is based upon ensuring Reactor remains shutdown without Boron Injection.
RO/SRO b) EOP-4 action is based upon preserving HPCl Injection capability.
L s43 c) EOP-5 action is based upon a loss of the Core Spray Hold Pumps.
d) There are NO EOP entry conditions. Plant is controlled by AOP-1.
Proposed Answer: c) EOP-5 action is based upon a loss of the Core Spray Hold Pumps.
Explanation (Optional): a) Incorrect- Stem indicates all rods are at 00 except 1 at 22. EOP-3 is not applicable.
b) HPCl is tripped at 10.75 feet Torus level to preclude HPCl operation from pressurizing the Containment. HPCl exhaust to Torus is uncovered.
C) Correct- EOP-5 action protects equipment in the secondary containment.
The CS Hold Pumps are the lowest elevation equipment of concern in the Crescents.
EOP-4 on low Torus Level and EOP-5 on high crescent level both exist.
Technical Reference(s): EOP-5, OP-14 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPEOP-5, EO-1.07 (As available)
Question Source: Bank # Monticello 1 lNPO# 15350 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 8/23/1999 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X LJ Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 HPCl Group # 1 1 Knowledge of the physical connections and/or WA # 206000 K1.07 K1.07 cause effect relationships between HIGH PRESSURE COOLANT INJECTION SYSTEM and the following: (CFR: 41.2 to 41.9/ 45.7 to 45.8)
D.C. power: BWR-2,3,4 Importance Rating 3.7 3.8 Proposed Question: Which of the following would render HPCl incapable of automatically injecting into the RPV from a Normal Standby Lineup?
a) Loss of the 10500 Bus RO/SRO b) Loss of 125 VDC Bus B 32/44 c) Loss of Condensate Storage Tank level d) Both Standby Gas Treatment Trains Out of Service Proposed Answer: b) Loss of 125 VDC Bus B Explanation (Optional):
Technical Reference(s): OP-15 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-23, EO-1.10. E (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7,8 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 LPCS Group # 1 1 Knowledge of tagging and clearance procedures. WA # 209001 2.2.13 2.2.13 (CFR: 41.10 / 45.13)
Importance Rating 3.6 3.8 Proposed Question: Which one (1) of the following requirements must be met in order to permit operation of Core Spray Hold Pump 'A' under a Striped Tag?
a) Tag Holder for the CS Hold Pump must be designated by position such as "Electrical Supervisor".
ROlSRO b) A procedure or Work Request with Step Text must exist to provide CS Hold Pump operation guidance.
33/45 c) Tag Holder for the CS Hold Pump with concurrence from the SNO directs CS Hold Pump operation.
d) If the CS Hold Pump is out of it's protected position for > one (1) shift, Tagout control must shift to the Work Week Manager.
Proposed Answer: b) A procedure or Work Request with Step Text must exist to provide CS Hold Pump operation guidance.
Explanation (Optional):
Technical Reference@): AP- 12.01 (Attach if not previously provided)
W Proposed references to be provided to applicants during examination: None Learning Objective: LP-AP44.10 (As available)
Question Source: Bank ##
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 LPCS continued Group # 1 1 Knowledge of the effect that a loss or WA# 209001 K3.02 K3.02 malfunction of the LOW PRESSURE CORE SPRAY SYSTEM will have o n following: (CFR:
41.7 145.4)
ADS logic Importance Rating 3.8 . 3.9 Proposed Question: While at full power, a small break LOCAW-thHPCl inoperable has occurred.
ADS has initiated.
The only Low Pressure ECCS in sewice is Core Spray B which subsequently trips.
ADS valves will ?
a) Remain open.
RO/SRO b) Close immediately.
34/46 c) Close after a two (2) minute delay.
d) Remain open until RPV level reaches 2 59.5.
Proposed Answer: b) Close immediately.
Explanation (Optional):
Technical Reference(s): OP-68, OP-14 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-14, EO-I .09.B (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
\ -l Downscale or Examination Outline ross-reference:
Tier Level# +
1 I
SRO 1
Unknown I 1 Ability to determine EA2 05 EA2.05 DOWNSCALE OR UNKNO (CFR: 41.10 f43.5/45.13)
Control rod position 4.2 4.3 Proposed Question: a complete loss of UPS.
is directed by / 111 and SCRAM Rod (2) n the Full Core Display RO/SRO n the Full Core Display 35147 CRAM Lamps on the Full Core Display ull In Lamps on the Full Core Display Proposed Answer:
e Full Core Display Explanation (Optional): n as SRO Only swapped original ROlSRO question 35/47
- S35, SRO original question S35 was made ROlSRO the UPS. Per AOP-21, the SCRAM is verified by confirming the Blue & mps lit on the Full Core Display. This indication only opened and the accumulator ng Rod Digital Position Indication, the ition is Unknown, thus entry into EOP-3 is required.
EOP-3 (Attach if not previously provided)
\
vided to applicants during exami EOPs LP-AOP, EO-1.03, EOP3L (As available)
Bank #
\
Modified Bank # (Note changes or attach parent)
New NEW Last NRC Exam
\
\
(Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensionor Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RPS Group # 1 1 Knowledge of the operational implications of the WA # 212000 K5.02 K5.02 following concepts a s they apply to REACTOR PROTECTION SYSTEM : (CFR: 41.5 / 45.3)
Specific logic arrangements Importance Rating 3.3 3.4 Proposed Question: While at 20% power, what possible Reactor Protection System (RPS) response(s) can occur if the Inboard and Outboard MSIV's on any two (2) Main Steam Lines are closed?
a) No response a full SCRAM RO/SRO b) No response half SCRAM 36/48 c) Half SCRAM always d) Full SCRAM always Proposed Answer: b) No response OR half SCRAM Explanation (Optional):
Technical Reference(s): ST-1I, OP-I (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-29, EO 1.09.f, 1.13.C (As available)
.-.-- Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 IRM Group # 1 1 Ability to monitor automatic operations of the WA# 215003 A3.04 A3.04 INTERMEDIATE RANGE MONITOR (IRM) SYSTEM including: (CFR: 41.7 / 45.7)
Control rod block status Importance Rating 3.5 3.5 Proposed Question: An IRM HI Flux Control Rod Block is automatically bypassed when ?
a) The Reactor Mode Switch is placed in RUN.
RO/SRO b) The IRM is on Range 1.
37/49 c) The IRMs companion APRM is downscale.
d) The SRMs are fully inserted.
Proposed Answer: a) The Reactor Mode Switch is placed in RUN.
Explanation (Optional):
Technical Reference(s): OP-16 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O7B, EO- 1.05.C.2 (As available)
Question Source: Bank #
~ ~ ~~
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Source Range Monitor Group # 1 1 Ability to monitor automatic operations of the WA # 215004 A3.04 143.04 SOURCE RANGE MONITOR (SRM) SYSTEM includina: (CFR: 41.7 / 45.7)
Control rgd block status Importance Rating 3.6 3.6 Proposed Question: The following plant conditions exist:
Reactor Mode Switch is in STARTUP/HOT STBY.
Intermediate Range Monitors (IRM's) all on Range 3.
Source Range Monitor (SRM) A is reading 0.5 cps SRM's B and C are reading 8.3 x I O 4 SRM D mode switch is in STANDBY A rod block signal has been generated.
Which one of the following has caused the rod block?
a) SRM Inoperable ROlSRO b) SRM not full in 38/50 c) SRM Downscale d) SRM Upscale Proposed Answer: a) SRM Inoperable Explanation (Optional):
Technical Reference(s): OP-16 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O7B, EO 1.05.b.1, EO 1.05.~,EO 1.14.~ (As available)
Question Source: Bank # Perry 1 I N P W 21837 (Modified to JAF)
~-
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1/1R001 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 APRM I LPRM Group # 1 1 Ability to monitor automatic operations of the KIA # 215005 A3.02 A3.02 AVERAGE POWER RANGE MONITOWLOCAL POWER RANGE MONITOR SYSTEM including:
(CFR: 41.7 / 45.7)
Full core display Importance Rating 3.5 3.5 Proposed Question: A reactor startup is being performed following a planned outage.
Annunciator, 09-52-33, LPRM Downscale, clears.
The SNO can confirm that this is expected and correct by verifying?
a) All APRM Downscale alarms are clear.
ROISRO b) All Full Core Display LPRM downscale lights are out.
39151 c) All IRM Range Switches are above Range 1.
d) Reactor Mode Switch is in RUN.
Proposed Answer: b) All Full Core Display LPRM downscale lights are out.
Explanation (Optional):
Technical Reference(5): OP-16, ARP- 09-5-2-33 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None v
Learning Objective: SDLP-O7C, EO-I. 12.D,1.05.C.1.B (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level SRO
\--
Tier # 2 SLC Group # 1 Knowledge of the effect that a loss or W A # 211000 K3.02 malfunction of the STANDBY LIQUID CONTROL SYSTEM will have on following: (CFR: 41.7 / 45.4)
Core sp tem: Plant-Specific Importance Rating 3.2 Proposed Question: Engineering has just informed the Shift Manager of where flow induced vibration has breached the integrity of the in-vessel section of the Standby Liquid Control piping at the vessel wall.
In addition to SLC, which one (1) of the following systems Technical Specification operability could be threatened by such an occurrence?
a) A Loop of Core Spray only.
RO/SRO b) Both Loops of Core Spray.
s52 c) A Loop Recirc Pump Trip function only.
d) Both Loops Recirc Pump Trip functions.
Proposed Answer: b) Both Loops of Core Spray.
Explanation (Optional): 0 Above Core Plate section of pipe is one side of Core Spray Sparger Leak Detection System AP cell- it is connected as high side of AP.
This pipe is outer part of pipe- within - a pipe. Inner pipe is SLC Injection pipe and exits outer pipe below core plate.
0 If flow induced vibrations caused wearing away of pipe where SLC (inner) exits outer, then the breach would put Core Spray sensing point below the core plate, making high side AP correction value even higher.
Therefore, if break did occur in CS Injection piping, there would be no guarantee that the resulting AP change, calibrated for high side on above core plate would shift enough to cause associated alarm warning of sparger injection problem.
0 Until Engineering could prove/disprove or calculate exact effect, then TRM requirement 3.3.H (Table 3.3.H-1) could not be assured to be met. The TRM Action is to monitor parameters within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If this could not be done, then the system would be declared in-operable.
0 Distradors- C and D are incorrect in that the pressure sensing for ARVRPT comes from RPV level Sensing Reference Legs. (3ABB Condensing Chambers).
Technical Reference(s): OP-14, OP-11, ARP-09-03-1-1 (Attach if not previously provided)
~~
TRM 3.3.H (Table 3.3.H-I)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-14, EO-1.07.C, 1.16, 1.05.A.13 (As available)
Question Source: Bank #
~~ ~~
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam
(Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge F Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 2 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RClC Group # 1 1 Knowledge of the effect that a loss or WA # 217000 K3.01 K3.01 malfunctionof the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) will have on following:
(CFR: 41.7 145.4)
K3.01 Reactor water level Importance Rating 3.7 3.7 Proposed Question: From a normal full power condition, the Station has lost all AC electrical power. RPV level is being controlled using RClC system operation.
Which statement below describes the effect, if any, that the loss of A Station Battery will have on level control?
a) HPCl system will have to be used to control RPV level.
RO/SRO b) Diesel Fire Pump to RHR X-connect must be used to maintain level.
40153 c) All injection sources will be lost. Steam Cooling is required when RPV level drops to -19.
d) RClC will continue to inject but must be controlled in manual from the Control Room.
Proposed Answer: a ) HPCl system will have to be used to control RPV level.
Explanation (Optional):
Technical Reference(s): AOP-45, AOP-49 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP 13, EO 1.09.A, 1.10.8 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RClC Group # 1 1 Ability to monitor automatic operations of the W A # 217000 A3.06 A3.06 REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) including: (CFR: 41.7 I45.7)
Lights and alarms Importance Rating 3.5 3.4 Proposed Question: RClC has received an Initiation Signal and is operating as expected when the following alarm is received.
09-4-1 RClC PMP SUCT PRESS LO Which of the below indications is consistent with this alarm?
a) 13MOV-21, RClC Pump Discharge to RPV, closes.
RO/SRO b) 13HOV-1, RClC Turbine Trip/lhrottle Valve, closes then reopens.
41/54 c) 13MOV-131, RClC Turbine Steam Inlet Isolation, closes.
d) RClC Pump suction has realigned to the Torus.
Proposed Answer: a ) 13MOV-21, RClC Pump Discharge to RPV, closes.
Explanation (Optional): Low suction is a trip that closes 13HOV-1. 13MOV-21 closes on interlock.
13HOV-1 must be manually reset to reopen. HPCI will cycle on a low suction pressure trip signal.
13MOV-131 will only close on RPV high water level.
RClC suction swap will occur without causing a RClC Turbine Trip.
v Technical Reference(s) OP-19, ARP 0941-23 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-13, EO 1.05.b,1.12, (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level SRO Tier # 2 APRM / LPRM Group # 1 Abilityt~~ain-primaly-aRd-seconcC;Hy K/A # 215005 2 - 4 34 ckertli&~wit.hin-a+lowable4mk~
- q.....
7k RANDOMLY RESELECTED Ability to supervise and assume a management role during plant transients and upset conditions.
(CFR: 43.5/45.12/45.13)
Importance Rating Proposed Question: A reactor startup is in progress at 8 O h power, Mode 2, when the Feedwater Startup Valve, 34FCV-137, fails full open with the following results:
FullSCRAM.
0 SNO-1 reports 4 rods at positions between 04 and 48.
0 All four of the rods are widely separated across the core.
0 APRMs are Downscale.
v 0 RPV level peaked low at 190 inches and is slowly rising.
0 SNO-2 is able to control Feedwater Startup Valve, 34FCV-137, in manual.
0 RPV pressure is 900 psig and slowly trending down.
Which of the following is the cause of the SCRAM and the appropriate direction to the SNO-I?
a) RPV High Water Level, Determine an EP-3 success path and insert control rods.
RO/SRO b) APRM Upscale, Determine an EP-3 success path and insert control rods.
s55 c) RPV High Water Level, insert control rods per AOP-I, Subsequent Operator Actions.
d) APRM Upscale, Insert control rods per AOP-1, Subsequent Operator Actions.
Proposed Answer: d) APRM Upscale, Insert control rods per AOP-1, Subsequent Operator Actions.
Explanation (Optional): APRM set-down SCRAM of 15% when RMS is out of RUN.
Even if RPV HIGH LEVEL is reached and the Turbine Trips, the TSV Closure SCRAM is bypassed when the RMS is out of RUN.
Stem provides no EOP-2 entry conditions, therefore EOP-3 and thus EP-3 are not applicable.
AOP-1 provides Operator guidance for these conditions.
Technical Reference(s): AOP-1, SDLP-O-IC, SDLP-05 (Attach if not previously provided)
~~ ~~ ~
Proposed references to be provided to applicants during examination: None
Learning Objective: SDLP-O7C, SDLP-05 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the infomation will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 5 Comments:
Examination Outline Cross-reference: Level SRO Tier # 2 RCIC Group # 1 Knowledge of electrical power supplies to the W A # 217000 K2.04 following: (CFR: 41 -7) cuum pump)
Importance Rating 2.6 Proposed Question: The Plant is operating at 100% power steady state with HPCl tagged out for maintenance. You are currently in day three (3) of the HPCl LCO. The SNO is performing a surveillance test to demonstrate RClC is operable and reports that the Barometric Compressor Vacuum Pump, (13P-3), will not start.
Which of the below describes appropriate action and the impact on Technical Specifications?
a) Dispatch NPO to check breaker status on BMCC-113. LCO is more restrictive.
RO/SRO b) Dispatch NPO to check breaker status on BMCC-1/3. No change in LCO.
556 c) Dispatch NPO to check breaker status on BMCC-2/4. LCO is more restrictive.
d) Dispatch NPO to check breaker status on BMCC-24. No change in LCO.
Proposed Answer: a) Dispatch NPO to check breaker status on BMCC-1/3. LCO is more restrictive.
Explanation (Optional):
Technical Reference(s): OP-43, OP-19 (Attach if not previously provided)
W Proposed references to be provided to applicants during examination: NONE Learning Objective: SDLP-13, EO-I .04.A (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 2 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 ADS Group # 1 1 Knowledge of the physical connections andlor WA # 218000 K1.06 K1.06 cause effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following:
(CFR: 41.2 to 41.9 145.7 to 45.8)
Safetykelief valves Importance Rating 3.9 3.9 Proposed Question: The Plant is at 70% power and normal operating pressure. The Control Room receives annunciator, 09-4-2-37, "SRV Electric Lift Initiated or Bypassed".
All SRV green lights are on.
Which of the following describes how this impacts the operation of the ADS Valve(s)?
a) Will operate on hydraulic overpressure.
ROISRO b) Will operate pneumatically on High RPV pressure.
42157 c) Only operate manually from Panel 02ADS-071.
d) Only operate manually from 09-4Panel.
Proposed Answer: a) Will operate on hydraulic overpressure.
Explanation (Optional):
Technical Reference(s): ARP-09-4-2-37,OP-68 (Attach if not previously provided)
GE DWG 791E453 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-29, EO-1.05.A.4 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO L-Tier # 2 2 PCIS/Nuclear Steam Supply Shutoff Group # 1 I Ability to (a) predict the impacts of the following KIA # 223002 Au4 A2424 on the PRIMARY CONTAINMENT ISOLATION A2.05 A2.05 SYSTEMlNUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
(RANDOMLY RESELECTED)
Nuclear boiler instrumentation failures Importance Rating M K!
3.3 3.6 Proposed Question: During full power operation, l&C is conducting PClS Group 1 Testing per Instrument Surveillance Procedures on transmitters, 02-3LT-57A through 57D. The technician commences testing RPV Level 1 response on level transmitter 02-3LT-57A At the same moment the technician imposes a level signal of c 59.5on LT-57A, level transmitter 02-3LT-57C fails downscale.
Which of the following describes the expected results of this combination of events (1) and Operator response (Z)?
a) (1) inboard MSIVs closed causing Reactor SCRAM. (2) Operators respond per AOP-1, Reactor SCRAM, AOP-15, ISOLATION VERIFICATION AND RECOVERY and EOPs.
b) (1) All MSIVs closed causing Reactor SCRAM. (2) Operators respond per ROlSRO AOP-1, Reactor SCRAM, AOP-15, ISOLATION VERIFICATION AND RECOVERY and EOPs.
43/58 c) (1) Half Isolation signal on PClS A. (2) Operators direct I&C Technician to restore 02-3LT-57A per AP-02.06, PROCEDURE USE AND ADHERANCE and reset the isolation signal per AOP-15, ISOLATION VERIFICATION AND RECOVERY.
d) (1) Half Isolation signal on PClS A,(2) Operators direct 1&C Technician to restore 02-3LT-57A per AP-02.06, PROCEDURE USE AND ADHERANCE and stop further testing.
Proposed Answer: d) (1) Half Isolation signal on PClS A. (2) Operators direct I&C Technician to restore 02-3LT-57A per AP-02.06, PROCEDURE USE AND ADHERANCE and stop further testing.
Explanation (Optional): Group 1 MSlV Logic includes RPV Low Level. To permit testing and guard against spurious instrument failure, the de-energized- to-function, fail-safe logic, initiates a LJ Iout of 2 taken twice arrangement, which is also divided between two divisions.
The failure of any one transmitter (either LT-57A or IC, B or ID) will not cause MSlV Closure, just indication of partial logic actuation; the consequences will depend upon which ones are involved. Any combination of one on A logic (LT-57A or 57C) and one on 6 logic (LT-57B or 57D) will fulfill complete logic and MSIVs (all) will automatically close. If both transmitters on the same logic train fail, it will result in no more than the half isolation logic actuation.
a) Incorrect because MSIVs will not close with signals from 2 PClS Logic A Train signals.
b) Incorrect because MSIVs will not close with signals from 2 PClS Logic A Train signals.
c) Incorrect because even with the LT-57A returned to service, the % isolation signal is still present from LT-57C and cant be reset. Also- AOP-15 assumes isolation occurred.
d) Correct- 12 isolation occurs and AP requires backing out of procedure by placing component manipulated into a safe condition.
Technical Reference(s): AP-12.03, AP-02.06 (Attach if not previously provided)
~
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-02B, EO-1.09.C (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 5 Comments:
Examination Outline Cross-reference: Level SRO Tier # 2 PClSlNuclear Steam Supply Shutoff Group # 1 Knowledge of electrical power supplies to the W A # 223002 K2.01 following: (CFR: 41.7)
Logic power supplies Importance Rating 2.7 Proposed Question: In which of the following complete system loss events, would you expe!Ct to find at least one MSlV in each Main Steam Line closed?
a) AOP-59, Loss of RPS Bus A Power
-OR AOP-45, Loss of DC Power System A b) AOP-18, LOSSOf 10500 BUS ROfSRO -
AND AOP-45, Loss of DC Power System A s59 C) AOP-21, LOSSOf UPS AND AOP-46, Loss of DC Power System B d) AOP-19, LOSS Of 10600 BUS OR AOP-46, Loss of DC Power System B Proposed Answer: b) AOP-18, LOSSOf 10500 BUS AND AOP-45, Loss of DC Power System A Explanation (Optional):
Technical Reference(s): AOP-18, AOP-19, AOP-21 (Attach if not previously provided)
AOP-45, AOP-46 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-29, EO 1.04.a, EO 1.05.a. 1.c (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 PCIS/Nuclear Steam Supply Shutoff Group # 1 1 Knowledge of conditions and limitations in the WA # 223002 2.1.10 2.1.10 facility license.(CFR: 43.1 / 45.13)
Importance Rating 2.7 3.9 Proposed Question: The following data was collected on a Post Trip Review of a reactor scram from 100%
power:
Time + 0, reactor scrammed manually, all immediate operator actions of AOP-1 completed, all rods inserted fully 0 Time + 1 minute - Main Condenser Vacuum 22.5" HG - Main Turbine tripped, all Turbine Stop Valves shut Time + 1.5 minutes - Main Condenser Vacuum 20" HG - Both Reactor Feed pumps tripped Time + 2.5 minutes, HPCl started and injected into the RPV on RPV level of 126.5" 0 Time + 3 minutes, Main Condenser Vacuum 8" HG - Bypass Valves and MSIV's shut 0 Time + 4 minutes, RPV pressure stabilized 800-1000 psig by manually operating SRV's 0 Time + 8 minutes, RPV level restored to Normal band of 196.5 to 222.5" 0 Time + 30 minutes, plant stabilized in Mode 3 Which system did not respond as required?
a) Main Turbine.
RO/SRO b) Feedwater Pumps.
44/60 c) HPCI.
d) MSIV's.
Proposed Answer: d ) MSIV's.
Explanation (Optional): a) Main Turbine responded as expected to loss of vacuum.
b) FW pumps responded as expected to loss of vacuum.
c) HPCl responded as expected to low RPV level.
d) MSIV's did not respond, as expected- should have not closed on loss of vacuum, bypassed low vacuum trip at 8" with Main Stop Valves Closed and Mode Switch not in Run. Immediate Operator Actions per AOP-1 to place Rx Mode switch in Shutdown provided in stem at Time 0 minute.
Technical Reference(s): AOP-1, AOP-31 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAOP EO-I .02 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam
(Optional - Questionsvalidated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 1 Comments:
Examination Outline Cross-reference: Level SRO Tier # 2 Main and Reheat Steam Group # 2 Ability to (a) predict the impacts of the following K/A +y 239001 A2.01 on the MAIN AND REHEAT STEAM SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
Malfunction of reactor turbine pressure regulating Importance Rating 3.9 Proposed Question: The reactor is operating at 25% power with recirculation flow at minimum.
If a turbine trip occurs and the bypass valves fail to open, which of the following would be the appropriate procedure(s) to respond to the event?
a) AOP-1, Reactor Scram, AND AOP-6, Malfunction of EHC Pressure Regulator.
RO/SRO b) AOP-2, Main Turbine Trip Without Scram, AND AOPS, Malfunction of EHC Pressure Regulator.
S6 1 c) EOP-2, RPV Control, AND AOP-2, Main Turbine Trip Without Scram.
d) AOP-1, Reactor Scram, AND EOP-2, RPV Control.
Proposed Answer: d) AOP-1, Reactor Scram, AND EOP-2, RPV Control.
Explanation (Optional): High pressure entry into EOP-2 and AOP-1 Technical Reference(s): AOP-1, EOP-2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-05, EO 1.07.a.6, 7, & 10 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 SRVs Group # 1 1 Knowledge of the physical connections andlor KIA # 239002 K1.O1 K1.O1 cause effect relationships betweem RELlEFlSAFETY VALVES and the following:
(CFR: 41.2 to 41.9 145.7 to 45.8)
Nuclear boiler Importance Rating 3.8 3.9 Proposed Question: After steady state conditions are achieved, which of the below is confirmation of an inadvertent SRV full opening while in normal full power operation?
a) Reactor Power at 1OOoh and Main Generator Output at 850 MWe.
ROISRO b) RPV Water level at 207 inches and Level Set at 202 inches.
45162 c) Feed flow at 11 x IOs I b d h r and Steam flow at 10 x 1Os Ibm/hr.
d) Torus water temperature trending down slowly with Torus cooling in service.
Proposed Answer: c) Feed flow at 11 x 1O6 lbmlhr and Steam flow at 10 x 1Os Ibm/hr.
Explanation (Optional):
Technical Reference(s): AOP-36 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAOP, EO-I .02,2.27 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Reactor Water Level Control Group # 1 1 Knowledge of the physical connections andlor WA # 259002 K1.13 K1.13 cause effect relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following: (CFR: 41.2 to 41.9 145.7 to 45.8) condensate system Importance Rating 3.2 3.2 Proposed Question: The plant is operating at 100% power with a normal Feed and Condensate alignment.
There are no systems or components inoperable. The A Condensate Pump trips due to an electrical fault.
Which one of the following is the expected result of this trip?
a) The operating pumps assume the additional load and the RFPs are not affected. A normal power reduction is required.
RO/SRO b) The A Condensate Booster Pump trips on interlock, but the RFPs are not affected. A normal power reduction is required.
46/63 c) The A Condensate Booster Pump trips on interlock causing RFPs to trip on low suction pressure. A manual SCRAM is required.
d) Condensate Booster Pump suction pressure decreases causing RFPs to trip on low suction pressure. A manual SCRAM is required.
Proposed Answer: a) The operating pumps assume the additional load and the RFPs are not affected. A normal power reduction is required.
Explanation (Optional):
Technical Reference(s): AOP-42, OP-3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-33, EO 1.05.b.2 8 1 . 1 4 . ~ (As available)
Question Source: Bank # JAF LOR# 25601012B02C Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO L--
Tier # 2 2 SGTS Group # 1 1 Knowledge of the effect that a loss or W A # 261000 K3.05 K3.05 malfunction of the STANDBY GAS TREATMENT SYSTEM will have on following: (CFR: 41.7 145.6)
Secondary containment radiation/ contamination levels Importance Rating 3.2 3.5 Proposed Question: A Station Blackout has occurred resulting in a full Reactor SCRAM with all rods in. HPCl is operating to maintain RPV water level and pressure control.
As a result of HPCl operation:
a) The A Station Battery is expected to rapidly deplete.
RO/SRO b) The Crescent Area contamination levels are expected to rise.
47/64 c) The HPCl Turbine MUST be manually tripped on RPV high water level.
d) RPV water level is expected to slowly drop until injection overcomes decay heat losses.
Proposed Answer. b) The Crescent Area contamination levels are expected to rise.
Explanation (Optional):
Technical Reference(s): AOP-49, AOP-45, AOP-46 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-01B, EO-1.09.A, F, LPAOP-49, EO-1.04 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 AC Electrical Distribution Group # 1 1 Knowledge of electrical power supplies to the K/A# 262001 K2.01 K2.01 following: (CFR: 41.7)
Off-site sources of power Importance Rating 3.3 3.6 Proposed Question: The Plant is in day 8 of a refuel outage. A full core off load has just been completed.
Niagara Mohawk called to report that the Lake Road 13.2 KV line is being taken out of service immediately.
Which of the below is a priority Control Room action?
a) Dispatch an NPO to transfer DHR power to the Diesel Generator.
RO/SRO b) Transfer in house electrical distribution from Normal to Reserve Station Service.
4 0165 c) Dispatch an NPO to align 115 KV control power to the alternate source.
d) Implement alternate temperature monitoring of Interim Spent Fuel Storage.
Proposed Answer. a) Dispatch an NPO to transfer DHR power to the Diesel Generator.
Explanation (Optional):
Technical Reference(s): AOP-71 (Attach if not previously provided)
~ ~~
Proposed references to be provided to applicants during examination: None 4
Learning Objective: LPAOP, EO-1.03, SDLP-71S, EO-1.09, SDLP-32, EO-1.04, l.lO.A (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 UPS (AC/DC) Group # 1 1 Knowledgeof tagging and clearance procedures. K/A # 262002 2.2.13 2.2.13 (CFR: 41.10/45.13)
Importance Rating 3.6 3.8 Proposed Question: Operators are tagging out the UPS M/G set for bearing replacement.
Worker protection is assured by:
a) A Maintenance PTR.
ROlSRO b) A Special Condition PTR.
49/66 c) A Guarantee PTR.
d) A Hold PTR.
Proposed Answer: d) A Hold PTR.
Explanation (Optional): PTR Purposes:
a)Maintenance PTR- used outside the power block.
b)Special Condition PTR- Not personnel protection-cautionary statement.
c) Guarantee PTR- a formal agreement from Ops controller to another controller-does not mean equipment is deenergized. Hold tags are used for a Guarantee PTR.
d)Hold PTR- Personnel protection in power block to support maintenance.
Technical Reference(s): OP-46B, AP-12.01 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-71A, EO I. 13.e (As available)
Question Source: Bank #
~~
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam
~~
(Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Examination Outline Cross-reference: Level SRO Tier # 2 Offgas Group # 2 Ability to predict andlor monitor changes in WA # 271000 A1.06 parameters associated with operating the OFFGAS SYSTEM controls including:
(CFR: 41.5 / 45.5)
Filter differential pressure Ej& @ - ; f f ~ f +
~ , f ~
Importance Rating 2.5 Proposed Question: The plant is operating at rated power when the following indications simultaneously OCCUT:
0 09-6-1-23, OFF GAS LINE FILTER DlFF PRESS HI in and clear.
0 09-6 Off Gas Flow Recorder (38FR-101) drops from 120 to 70 SCFM.
0 EPIC OFFGASRAD alarm.
09-10 Off Gas Radiation Monitors both reading 150-175 mr/hr.
Which of the following describes the plant event and the appropriate mitigating procedure?
a) Off Gas line blockage will cause a loss of condenser vacuum. AOP-31, LOSS OF CONDENSER VACUUM.
RO/SRO b) A hydrogen fire has ignited in piping downstream of the SJAEs. AOP-5, COMBUSTION IN SJAE AFTERCONDENSER.
S67 c) Fuel failure has resulted in a large radioactive gas release from the RPV. AOP-3, HIGH ACTIVITY IN REACTOR COOLANT OR OFFGAS.
d) An explosion has breached the SJAE discharge piping. AOP4, EXPLOSION IN AIR EJECTOR DISCHARGE PIPING.
Proposed Answer: b) A hydrogen fire has ignited in piping downstream of the SJAE's. AOP-5, COMBUSTION IN SJAE AFTERCONDENSER.
Explanation (Optional):
Technical Reference(s): ARP-09-6-1-23/AOP-4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAOP-EO 1.01 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5,lO
55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 DC Electrical Distribution Group # 1 1 Ability to (a) predict the impacts of the following WA # 263000 A2.01 A2.01 on the D.C. ELECTRICAL DISTRIBUTION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 I45.6)
Grounds Importance Rating 2.8 3.2 Proposed Question: The plant is operating at 100% with operators attempting to locate a ground on the A station battery. The next breaker to be opened is the supply for 10700 BKR Control Power (71DCA3 Ckt 24).
How will the opening of this DC breaker affect the 4KV breakers on the 10700 bus AND which procedures will provide guidance?
a) The breakers will immediately trip, AOP-20, Loss of 10700 Bus, and AOP-22, DC Power System A Ground Isolation.
b) The breakers can be tripped mechanically only, OP-46A, 4160 V & 600 V ROISRO Normal AC Power Distribution, and AOP-22, DC Power System A Ground Isolation.
50168 c) The breakers will immediately trip, OP-464 4160 V & 600 V Normal AC Power Distribution, and AOP-22, DC Power System A Ground Isolation.
d) The breakers can be tripped mechanically, AOP-20, Loss of 10700 Bus, and AOP-22, DC Power System A Ground Isolation.
Proposed Answer: b) The breakers can be tripped mechanically only, OP-46A, 4160 V & 600 V Normal AC Power Distribution, and AOP-22, DC Power System A Ground Isolation.
Explanation (Optional):
Technical Reference(s): AOP-22 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-7 1B, EO-I .09.C. 18 (As available)
Question Source: Bank # JAF LOR # 20004211BO1C Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 EDGs Group #
Knowledge of the effect that a loss or W A # 264000 K6.02 K6.02 malfunction of the following will have on the EMERGENCY GENERATORS (DIESEUJET) :
(CFR: 41.7 / 45.7)
Fuel oil pumps Importance Rating 3.6 3.6 Proposed Question: A LOCA and LOOP exists. Off-site power is not expected to be returned to service for two days. All EDG equipment is operable with the exception that Fuel Oil Transfer Pumps 93Pl-A1 8 2 have just tripped and cannot be started.
Based upon these events, select the expected plant response assuming NO Operator action.
a) EDG's 'A", 'B", "c", 8 'D" will continue to run until off- site power is restored.
ROlSRO b) EDG "A"will trip immediately, EDG's 'B","C", 8 "D" will continue to run until off-site power is restored.
51/69 c) EDG "A" will continue to run for up to three (3) hours then trip, EDGs "B",'C", 8 "D" will continue to run until off- site power is restored.
d) EDG "A"and "c" will continue to run for up to three (3) hours then trip. EDG's "B" 8 "D" will continue to run until off- site power is restored.
Proposed Answer. c) EDG 'A" will continue to run for up to three (3) hours then trip,. EDG's "B","C", 8 "D" will continue to run until off- site power is restored.
Explanation (Optional):
Technical Reference(s): OP-22 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-93, EO-I .05.A.4, 1.10.G (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of the process for making changes in Group #
procedures as described in the safety analysis report. (CFR: $3.3 I 45.13)
KIA # 2.2.6 Importance Rating 3.3 Proposed Question: Who must approve a temporary change to a Technical Specification related procedure?
a) The procedure RPO and an Operations QTR.
ROISRO b) A plant management QTR and a SRO license QTR.
S70 c) The General Manager-Plant Operations and a plant management QTR.
d) A plant management QTR and any operations licensed QTR.
Proposed Answer: b) A plant management QTR and a SRO license QTR.
Explanation (Optional):
Technical Reference(s): AP-2.04 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAP, EO-4.05 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
L.'
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 3 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Instrument Air Group # 1 1 Knowledge of the connections and I or cause UJA# 300000 K1.05 K1.05 effect relationships between INSTRUMENT AIR SYSTEM and the following:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
Main Steam Isolation Valve air Importance Rating 3.1 3.2 Proposed Question: The plant is operating at 75% reactor power.
The SNO-1 depresses the TEST pushbutton for 29AOV-86B, B OUTBOARD MSIV.
Which one of the following describes the response of 29AOV-86B?
a) Instrument Nitrogen bleeds off the bottom portion of the MSlV air cylinder and the top portion of the MSlV air cylinder is pressurized to stroke the MSlV closed in 3-5 seconds.
ROlSRO ) Instrument Air bleeds off the bottom portion of the MSlV air cylinder causing the MSlV to slowly close.
52/71 I Instrument Nitrogen bleeds off the bottom portion of the MSlV air cylinder zausing the MSlV to slowly close.
d nstrument Air bleeds off the bottom portion of the MSlV air cylinder and the top portion of the MSlV air cylinder is pressurized to stroke the MSlV closed in 3-5 seconds.
Proposed Answer: b) Instrument Air bleeds off the bottom portion of the MSlV air cylinder causing the MSlV to slowly close.
W Explanation (Optional):
Technical Reference(s): ST-1I, OP-I (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank # Perry 1 INPO # 21861 (Modified to JAF)
Modified Bank ## (Note changes or attach parent)
New Question History: Last NRC Exam 1/1n001 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of the effects of alterations on core Group #
configuration. (CFR: $$l$$)
WA # 2.2.32 Importance Rating 3.3 Proposed Question: The purpose of core spiral fuel un-loading is which one of the following?
a) it minimizes the possibility of flow induced vibration of nuclear instrumentation b) it precludes the formation of moderator filled cavities surrounded on all sides by RO/SRO fuel s72 c) it prevents SRM count rates from spiking when fuel is being off-loaded d) it enables the completion of a full core off-load in less time Proposed Answer: b) it precludes the formation of moderator filled cavities surrounded on all sides by fuel Explanation (Optional)'
Technical Reference(s1: ITS - Bases, RAP- 7.1.24 (Attach if not previously provided)
RAP-7.1.04B Section 5.10.3 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O7B, EO 1.13.E, 1.17.G (As available)
Question Source: Bank # JAF LOR # 1332 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 6 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Component Cooling Water Group # 1 1 Knowledge of electrical power supplies to the KIA# 400000 K2.01 K2.01 following: (CFR: 41.7)
CCW pumps Importance Rating 2.9 3.0 Proposed Question: An electrical transient has occurred and Switchgear L-14 is de-energized.
Which of the following equipment would be lost due to the degraded electrical source?
a) 12P-IA, RWCU Pump A RO/SRO b) 15P-2B, RBCLC Pump B 53ff 3 c) 11P-2B, SLC Pump 8 d) 46P-1 B, Service Water Pump B Proposed Answer b) 15P-2B, RBCLC Pump B Explanation (Optional):
Technical Reference(s): OP-46 (Attach if not previously provided)
OP40 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-15, EO-1.038 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowiedge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of radiation exposure limits and Group #
contamination control I including permissible levels in excess of those authorized.
KIA # 2.3.4 Importance Rating 3.1 Proposed Question: Authorization to receive radiological exposures in excess of 10CFR20 limits is the responsibility of the a) Radiation Protection Manager RO/SRO b) Emergency Director s74 c) TSC Manager d) General Manager- Plant Operations Proposed Answer: b) Emergency Director Explanation {Optional):
Technical Reference(s): EAP-15, AP-07.05 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: EP-12.5.3, EO-1.20.6 (As available)
W Question Source: Bank ## LaSalle 1INPO # 19298 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 11/20/2000 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis I O CFR Part 55 Content: 55.41 55.43 4 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 CRD Hydraulic Group ## 2 2 Knowledge of electrical power supplies to the WA# 201001 K2.03 K2.03 following: (CFR: 41.7)
Backup SCRAM valve solenoids Importance Rating 3.5 3.6 Proposed Question: WHICH ONE of the following supplies power to the Backup Scram Valves?
a) 120VACUPS ROISRO b) 125 VDC 54/75 c) 24VDC d) 120VACRPS Proposed Answer b) 125 VDC Explanation (Optional):
Technical Reference(s): OP-18 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-05, EO-I .04.A (As available)
Question Source: Bank # Oyster Creek 1 INPO # 13001 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 412911996 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level SRO W Tier # 3 Knowledge of the requiremen ewing and Group #
approving release permits. (C 45.10)
WA # 2.3.6 Importance Rating 3.1 Proposed Question: You are the Control Room Supervisor when approached by a Shift Chemistry Technician about the need to discharge a Waste Sample Tank to the Lake. The Shift Chemistry Technician has all of the necessary paperwork.
Prior to starting the discharge, it must be approved by (1)
During the discharge, the loss of a circulating water pump will result in I'll
~
a) (1) Shift Manager (2) higher concentration radioactive releases ROlSRO b) (1) Chemistry Superintendent (2)higher concentration radioactive releases S76 c) (1) Shift Manager (2) lower concentration radioactive releases d) (1 ) Chemistry Superintendent (2) lower concentration radioactive releases Proposed Answer: a) (1) Shift Manager (2) higher concentration radioactive releases Explanation (Optional):
Technical Reference(s): SP-1.05 Attachment # 2 (Attach if not previously provided)
~ ~~~
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-20, EO-1.13.B (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 4 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Control Rod and Drive Mechanism Group # 2 2 Ability to (a) predict the impacts of the following WA # 201003 A2.01 A2.01 on the CONTROL ROD AND DRIVE MECHANISM ;
and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 145.6)
Stuck rod Importance Rating 3.4 3.6 Proposed Question: A normal reactor startup is in progress at 7% reactor power and 970 psig RPV pressure.
Control Rod 26-35 did not move when given a withdraw signal from it's current notch position 12. Drive water differential pressure has been adjusted to 450 psid. All previous attempts to move this rod have been unsuccessful.
The operator's next action should be t o . . .
a) Individually SCRAM Control Rod 26-35, then disarm it electrically and hydraulically.
RO/SRO b) Attempt to move Control Rod 26-35 by performing "Double Clutching."
55/77 c) Declare Control Rod 26-35 INOPERABLE, then disarm it electrically and hydraulically.
d) Raise drive water differential pressure an additional 50 psig and attempt to withdraw Control Rod 26-35.
Proposed Answer: d) Raise drive water differential pressure an additional 50 psig and attempt to withdraw Control Rod 26-35.
W Explanation (Optional):
Technical Reference(s): AOP-24 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-OBC, EO-1.15.C (As available)
Question Source: Bank # Quad Cities 1 INPO # 19545 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 811312001 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RWM Group # 2 2 Ability to manually operate andlor monitor i n the WA # 201006 A4.06 A4.06 control room: (CFR: 41.7 / 45.5 to 45.8)
Selected rod position indication:P-Spec(Not-BWR6)
Importance Rating 3.2 3.2 Proposed Question: A manual SCRAM was inserted based on lowering RPV water level. The condition has been corrected and RPV level has been returned to the Green Band. During the SCRAM, two (2) control rods failed to fully insert. The SNO has attempted to insert control rods using the CRD System per EP-3, Backup Control Rod Insertion.
Which of the following conditions could prevent manual control rod insertion?
a) SDlV High Level Over-ride Switch in Normal.
RO/SRO b) Rod Worth Minimizer Bypass Switch in Normal.
56/78 c) Alternate Rod Insertion (ARI) U T reset.
d) Reactor Protection System SCRAM NOT reset.
Proposed Answer: b) Rod Worth Minimizer Bypass Switch in Normal.
Explanation (Optional):
Technical Reference(s): EP-3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: EOPBLP, EO 1.07 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6, 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RWCU Group # 2 2 Knowledge of REACTOR WATER CLEANUP W A # 204000 K4.03 K4.03 SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)
Over temperature protection for system components Importance Rating 2.9 2.9 Proposed Question: The plant is performing a reactor startup and heatup, currently at 200 psig.
Reactor water level control is via Reactor Water Cleanup (RWCU) rejecting to the main condenser hotwell 0 Main condenser vacuum has been established with the vacuum pump The operator is cautioned to carefully monitor system parameters while rejecting.
Which of the following RWCU system trips/isolations provide RWCU Demineralizer protection while in this lineup?
a) Cleanup Blowdown Flow Control Valve (12FCV-55) closure on low upstream pressure.
RO/SRO b) RWCU system isolation on non-regenerative heat exchanger high outlet temperature.
57/79 c) Cleanup Blowdown Flow Control Valve (12FCV-55) closure high downstream pressure.
d) RWCU system Containment Isolation Valve closure on high system flowrate.
Proposed Answer: b) RWCU system isolation on non-regenerative heat exchanger high outlet temperature.
W Explanation (Optional): Per DBD, 12FCV-55 closures are to protect system piping from under (vacuum) and over pressure.
System high temperature isolation protects resin.
System high flow isolation protects public from excessive releases.
Technical Reference(s): OP-28, ARP 09-4-2-35, DBD System 12 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-12, EO 1.05.c.1 (As available)
Question Source: Bank # Peach Bottom 2 INPO # 18536 (Modified to JAF)
~ ~ _ _ _ _ _ _
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 911911997 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7
55.43 d Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RPlS Group # 2 2 Ability to monitor automatic operations of the W A # 214000 A3.02 A3.02 ROD POSITION INFORMATION SYSTEM including: (CFR: 41.7 I45.7)
Alarm and indicating lights Importance Rating 3.2 3.1 Proposed Question: During a plant startup, with reactor power at 12%, control rod 18-11 was selected and the following indications occur:
Annunciator 09-5-2-2, ROD WITHDRAWAL BLOCK Annunciator 09-5-2-1, RWM ROD BLOCK A loss of ALL rod position indications on the Four Rod Display occurred A loss of ALL red Full-Out and green Full-In indications of Full Core Display Which of the following may be the cause for these indications?
a) Loss of 120 VAC Panel 71RBACB5 ROlSRO b) Loss of Panel 71AC10 58/80 c) Loss of Reactor Protection System (RPS) Distribution Panel A d) Loss of Uninterruptible Power Supply (UPS)
Proposed Answer: d) Loss of Uninterruptible Power Supply (UPS)
Explanation (Optional):
v Technical Reference(s): AOP-21 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O3G, EO-1.04, LPAOP, EO-1.01 (As available)
Question Source: Bank # Fermi 2 2 INPO # 7322 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 13.11/ I 995 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RBM Group # 2 2 Ability to predict andlor monitor changes in K/A# 215002 A1 . O l A I .01 parameters associated with operating the ROD BLOCK MONITOR SYSTEM controls including:
(CFR: 41.5 / 45.5)
Trip reference: BWR-3,4,5 Importance Rating 2.7 2.8 Proposed Question: While withdrawing control rod 26-27 at 4Oohpower, which of the below is the probable indication of a withdraw rod block?
a) SDlV High Level Alarm reading 20 gallons.
RO/SRO b) Rod Block Monitor green Push To Setup lamp is lit.
59/81 c) Control Rod 26-27 has withdrawn more than one (1) notch beyond the other rods in that group.
d) All Detector A Bypass lamps are lit on the Four Rod Display.
Proposed Answer: b) Rod Block Monitor green Push To Setup lamp is lit.
Explanation (Optional):
Technical Reference(s): OP-16 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: RAP-7.3.16, Attachment 3 Learning Objective: SDLP-7C, EO-1.05.B.4.F (As available)
Question Source: Bank ##
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Nuclear Boiler Inst. Group #
Knowledge of the effect that a loss or K/A # 216000 K6.01 K6.01 malfunction of the following will have on the NUCLEAR BOILER INSTRUMENTATION :
(CFR: 41.7 145.7)
A.C. electrical distribution Importance Rating 3.1 3.3 Proposed Question: Given the following plant conditions immediately after a SCRAM from full power:
Drywell Instrument Run temperature - 120 O F 0 Reactor Building temperature - 94 "F 0 Reactor pressure - 880 psig Immediately following a loss of all AC power, WHAT is the MINIMUM reactor water level that can be monitored from the control room?
a) +44" RO/SRO b) -150" 60182 C) -145" d) +164.5" Proposed Answer: c) -1 45" Explanation (Optional): Requires evaluation of what indication remains after a loss of all AC power. As a result of automatic backup swap on DC power at least one instrument remains for each response. Must recognize that EOP entry is applicable. Must apply RPV Sat curve to determine correct answer.
Technical Reference(s): EOP- Caution #1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOP's Learning Objective: EOP2LP, EO-1.01, SDLP-O2B, EO- (As available) 1.10.A, 1.04 Question Source: Bank # LaSalle 1 INPO # 11671 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam lot611995 (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RHRILPCI: CTMT Spray Mode Group # 2 2 Knowledge of the physical connections andlor WA # 226001 K1.12 K1.12 cause effect relationships between RHWLPCI:
CONTAINMENT SPRAY SYSTEM MODE and the following: (CFR: 41.2 to 41.9 /45.7 to 45.8)
Suppression pool (spray penetration): Plant-Specific Importance Rating 3.0 3.0 Proposed Question: Why does the Torus Spray flowpath of EOP-4, Primary Containment Control, prohibit initiation of Torus Spray if Torus Level is greater than 26 feet?
a) Less than 95% of non-condensable gasses exist in the Torus air space.
ROlSRO b) The spray header is covered by Torus water level.
61/83 c) The DW to Torus Vent flowpath has been lost.
d) Initiation of Sprays could bring the Torus to subatmospheric conditions.
Proposed Answer. b) The spray header is covered by Torus water level.
Explanation (Optional):
Technical Reference(s): BWROG EPGs (Attach if not previously provided)
EOP4LP
~~ ~
Proposed references to be provided to applicants during examination: None Learning 0bjective: EOP4LP, EO-1.05 (As available) u' Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RHRILPCI: TorusJPool Spray Mode Group # 2 2 Ability to manually operate and/or monitor in the WA # 230000 A4.02 A4.02 control room: (CFR: 41.7 I45.5 to 45.8)
Spray valves Importance Rating 3.8 3.6 Proposed Question: During a LOCA, the A and C RHR Pumps are injecting in the LPCl Mode. An Operator attempts to place the A loop of RHR in Torus Spray as directed by the Control Room Supervisor.
Without further Operator action, design interlocks will result in which of the following when valve operation is initiated?
a) The valves may be opened but will immediately close due to a LPCl Initiation signal present.
b) 10MOV39A and 34A will open, but 10MOV-38A will NOT open allowing Torus RO/SRO cooling mode of operation.
62/84 c) 10MOV-39A and 38A will open but 10MOV-34A will NOT open allowing Torus spray mode of operation.
d) The valves will NOT open due to a LPCl initiation signal being present unless the initiation signal is first overridden.
Proposed Answer. d) The valves will NOT open due to a LPCI initiation signal being present unless the initiation signal is first overridden.
Explanation (Optional):
Technical Reference(s): OP-13 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-10, EO-I .05.A2 (As available)
Question Source: Bank # Fermi 2 2 INPO # 8890 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 4/6/1998 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Fuel Pool CoolinglCleanup Group # 2 2 Knowledge of FUEL POOL COOLING AND W A # 233000 K4.03 K4.03 CLEAN-UP design feature@) andlor interlocks which provide for the following: (CFR: 41.7)
Maintenance of adequate pool temperature Importance Rating 2.8 3.1 Proposed Question: A design basis of the Fuel Pool Cooling and Cleanup System is to maintain the Spent Fuel Pool outlet temperature below for a peak annual refueling heat load of 10 X lo6 BTU/Hr.
a) 155°F ROlSRO b) 145°F 63/85 c) 135 "F d) 125'F Proposed Answer: c) 135 "F Explanation (Optional):
Technical Reference(s): OP-30, AOP-53, FSAR (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-19, EO-1.02 (As available)
W Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis.
10 CFR Part 55 Content: 55.41 7 55.43 7 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Reactornurbine Pressure Regulator Group # 2 2 Knowledge of the operational Implications of the WA # 241000 K5.05 K5.05 following concepts as they apply to REACTOWWRBINE PRESSURE REGULATING SYSTEM : (CFR: 41.5 145.3)
Turbine inlet pressure vs. turbine load Importance Rating 2.8 ' 2.9 Proposed Question: Reactor Power is reduced from 100% to 95% by lowering recirculation flow.
Turbine Control Valves are repositioned by EHC sensing as compared to ?
a) RPV Pressure, Pressure Setpoint.
RO/SRO b) RPV Pressure, Turbine 1"Stage Pressure 64/86 c) Turbine Inlet Pressure, Turbine 1"Stage Pressure d) Turbine Inlet Pressure, Pressure Setpoint.
Proposed Answer: d) Turbine Inlet Pressure, Pressure Setpoint.
Explanation (Optional):
Technical Reference(s): SLP-74c (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-74C, EO-1.05.A.4 (As available)
Question Source: Bank # Dresden 2 INPO # 6524 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New
~~
Question History: Last NRC Exam 3/11/1996 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Secondary CTMT Group # 2 2 Knowledge of the effect that a loss or KIA# 290001 K3.01 K3.01 malfunction of the SECONDARY CONTAINMENT will have on following: (CFR: 41.7 /45.4) tOff-site radioactive release rates Importance Rating 4.0 4.4 Proposed Question: The plant was operating at 100% power when a large steam leak occurred inside the Reactor Building. SGT Train 'A" and "B" are operating at rated flows. Secondary Containment pressure is +1.5" WG.
Off-Site radioactivity release rates are expected to be.. . .. .
a) Ground releases via SGT only RO/SRO b) Ground releases via SGT and Reactor Building Ventilation 65/87 c) Elevated releases via SGT and Ground releases via Reactor Building leakage d) Elevated releases via SGT and Ground releases via Reactor Building Ventilation Proposed Answer: c) Elevated releases via SGT and Ground releases via Reactor Building leakage Explanation (Optional):
Technical Reference(s): OP-51A (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-16A EO-I.OQb,SDLP-664 EO-1.05.C (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Ability to explain and apply system limits and Group #
precautions. (CFR: 41.10 / 43.2 I 45.12)
KIA # 2.1.32 2.1.32 Importance Rating 3.4 3.8 Proposed Question: Prior to returning to two loop operation from one loop operation, which of the following LIMITS must be met and what is the REASON for that limit?
a) LIMIT - The temperature difference between the bottom head coolant and the recirc loop coolant in the loop to be started is 5 145 deg F.
REASON - To prevent a violation of the RPV pressure and temperature limitationthat minimize the chances of brittle fracture from occurring.
b) LIMIT - The temperature difference between the recirc loop coolant in the loop to be started and the reactor vessel coolant is 5 50 deg F.
ROISRO REASON - To prevent a violation of the RPV pressure and temperature limitation that minimize the chances of brittle fracture from occurring.
c) LIMIT - The temperature difference between the bottom head coolant and the recirc loop coolant in the loop to be started is 5 145 deg F.
66/88 REASON - To prevent damage to the fuel cladding that.would result from the sudden increase in power due to the injection of cold water.
d) LIMIT - The temperature difference between the recirc loop coolant in the loop to be started and the reactor vessel coolant is 5 50 deg F.
REASON - To prevent damage to the fuel cladding that would result from the sudden increase in power due to the injection of cold water.
Proposed Answer: b) LIMIT - The temperature difference between the recirc loop coolant in the loop,to be started and the reactor vessel coolant is 5 50 deg F.
REASON - To prevent a violation of the RPV pressure and temperature limitation that minimize the chances of brittle fracture from occurring.
Explanation (Optional): UFSAR - Section 14.5.7.2 STARTUP of IDLE Recirculation Pump analysis starts from following conditions- idle loop is filled with 100 F water, active loop is at 67%
O flow, core flow is at 50 %, Core power is 65%. When the idle loop is started at these conditions there is a resultant short duration peak of 90% neutron power with no SCRAM, the analysis predicts that Damage To Fuel Barrier Occurs.
Based upon the ITS and FSAR analysis; a) Distracter has incorrect limit- it states the difference is between bottom head and idle loop temperature- the SR verifies difference between bottom head coolant temperature 8, RPV coolant temperature is 5 145°F. The reason it gives is correct for the purpose of the LCO.
b) Correct response has correct limit- The reason it gives is correct for the purpose of the LCO.
c) Distracter has incorrect limit- see distracter A. The reason it gives is incorrect based upon UFSAR - Section 14.5.7.2- analysis predicts that NO Damage To Fuel Barrier Occurs.
d) Distracter has correct limit- The reason it gives is incorrect based upon UFSAR - Section 14.5.7.2- analysis predicts that Damage To Fuel Barrier Occurs.
Technical Reference(s): ST-26K, ITS SR 3.4.9.3, SR 3.4.9.5 ITS Bases 3.4.9, SR (Attach if not B3.4.9.3, SR 83.4.9.5, FSAR- 14.5.7.2 previously provided)
Proposed references to be provided to applicants during examination: NONE Learning Objective: SDLP-2H, EO-1.139 (As available)
Question Source: Bank # Dresden 2 INPO # 21373 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 6/14/2002 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of how to conduct and Verify valve Group ##
lineups. (CFR: 41.10145.1 145.12)
WA ## 2.1.29 2. I.29 Importance Rating 3.4 3.3 Proposed Question: LCO maintenance to replace the A EDG Jacket Water Cooler has just been completed and restoration is in progress. The valve lineup calls for the Cooler ESW Outlet Isolation Valve, 46ESW-5A, located in A EDG Room, to be locked 4 turns closed.
To complete this valve lineup, you must:
a) Perform initial positioning, then call for another qualified operator to verify your actions.
RO1SRO b) Perform initial positioning in the presence of another qualified operator verifying your actions.
67/89 c) Perform initial positioning and NIA the verification requirement.
d) Perform initial positioning and obtain Shift Manager concurrence to waive the verification.
Proposed Answer: b) Perform initial positioning in the presence of another qualified operator verifying your actions.
Explanation (Optional): Cannot independently verify position of throttled valves.
Correct response. Required because of A .
Component verification required on all safety related components unless waived.
SM can waive verification for excessive exposure or hazardous environment.
The A EDG Room is neither.
Technical Reference(s): AP-12.06 Section 7.6 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAP-48.03, 48.02 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Ability to perform Pre-StartUp procedures for the Group #
facility I including operating those controls associated with plant equipment that could affect reactivity. (CFR: 45.1)
WA # 2.2.1 2.2.1 Importance Rating 3.7 3.6 Proposed Question: A plant startup is in progress with the Mode Selector Switch in Startup. Control rods are being withdrawn.
0 The Rod Worth Minimizer (RWM) has just failed with 25% of the control rods withdrawn.
What action is required?
a) Bypass the RWM, verify all further control rod movements are in compliance using a qualified person, and continue the reactor startup.
RO/SRO b) Suspend withdrawal of the control rods, manually SCRAM the reactor, and verify operability of the RWM before commencing a reactor startup.
68/90 c) Suspend withdrawal of the control rods, verify operability of the Rod Block Monitor, and continue the reactor startup.
d) Bypass the RWM, fully insert all control rods, and verify operability of the RWM before commencing a reactor startup.
Proposed Answer: a) Bypass the RWM, verify all further control rod movements are in compliance using a qualified person, and continue the reactor startup Explanation (Optional):
Technical Reference(s): OP-64 Section E . l (Attach if not previously provided)
OP-65 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP3D, EO-1.15.A, LPAP, EO-46.04 (As available)
Question Source: Bank # Quad Cities 1 INPO # 20445 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 811312001 (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of surveillance procedures. Group #
(CFR: 41.10 145.13)
WA # 2.2.12 2.2.12 Importance Rating 3.0 3.4 Proposed Question: The Plant is operating at 98% Reactor power. The Control Room Supervisor has ordered you to perform ST-24J, RClC Flow Rate and Inservice Test, following maintenance.
During RClC pump operations:
a) RClC must be shutdown if Torus water temperature exceeds 95 "F.
ROISRO b) EHC Pressure Set must be adjusted to maintain RPV pressure .c 1050 psig.
69/91 c) Recirculation flow must be reduced to maintain Reactor power 400%.
d) Torus cooling must be in service to prevent Torus water temperature from exceeding 105 "F.
Proposed Answer: d) Torus cooling must be in service to prevent Torus water temperature from exceeding 105 "F.
Explanation (Optional):
Technical Reference(s): ST-24J, ITS-3.6.2.1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None
'd Learning Objective: SDLP-13, EO-1.13.d (As available)
Question Source: Bank # LaSalle 1 INPO # 19132 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1112012000 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of the process for determining the Group #
internal and external effects on core reactivity.
(CFR: 43.6)
KIA # 2.2.34 2.2.34 Importance Rating 2.8 3.2 Proposed Question: A Reactor startup from Cold Shutdown is in progress. The ECP was calculated based upon the following:
Reactor Coolant temperature at 140 "F Total Core Flow at 15 X 1Os Ibmlhr At time of criticality, Reactor has been shutdown for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> Feedwater temperature 80 "F Which of the below will result in criticality later in the rod pull sequence than the Predicted ECP?
a) Criticality occurs 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after shutdown.
ROISRO b) Feedwater temperature drops to 75 "F.
70192 c) Total Core Flow is reduced to 10 X lo6Ibmlhr d) Reactor Coolant temperature drops to 125 "F.
Proposed Answer: a) Criticality occurs 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after shutdown.
Explanation (Optional): a) 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> vs. 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> results in criticality occurring at a higher Xeon concentration requiring more total rod worth to overcome, therefore later than predicted.
b) A 5 ' F drop in FW Temperature is a net positive reactivity effect if FW is injecting. If not, there will be no effect. Criticality will occur earlier or as predicted.
c) Change in core flow is of no effect until core voiding. Criticality as predicted Ref. 7.3.25- Single Loop Operation. Attachment #I For two loop operation, 18 mlbmlhr core flow results from 5 mlbm/hr drive flow per loop.
Ref. ST-23C, Jet Pump Operability Test, Attachment # 3 30% Pump Speed is approximately 6 mlbm/hr drive flow per loop 25% Pump Speed is approximately 5 mlbmlhr drive flow per loop Ref. FSAR Section 4.3 Pump Speed Control Range 20-100 d) A reactor coolant drop in temperature is a net positive reactivity effect Criticality earlier than predicted.
Technical Reference(s): RAP-7.3.13 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPOP-65A, EO-1.10 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW
Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 1 55.43 6 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of the process for performing a Group #
containment purge.
(CFR: 43.4 145.10)
KIA # 2.3.9 2.3.9 Importance Rating 2.5 3.4 Proposed Question: The plant is conducting a shutdown, power is currently 30°! and lowering. It is desired to de-inert the Primary Containment (both the Drywell and Torus) as soon as possible to permit containment access for maintenance for a forced outage.
Which procedurally allowed flowpath would be used for de-inerting the Primary Containment?
a) Through the Standby Gas Treatment System with the Drywell and Torus de-inerted simultaneously.
ROlSRO b) Through the Reactor Building Ventilation System with the Drywell de-inerted first and then the Torus de-inerted.
71/93 c) Through the Reactor Building Ventilation System with the Drywell and Torus de-inerted simultaneously.
d) Through the Standby Gas Treatment System with the Drywell de-inerted first and then the Torus de-inerted.
Proposed Answer: d) Through the Standby Gas Treatment System with the Drywell de-inerted first and then the Torus de-inerted.
Explanation (Optional):
Technical Reference(s): OP-37 (Attach if not previously provided)
~~ ~ ~ ~~
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-I.OGC, EO-1.13.C (As available)
Question Source: Bank # Quad Cities 1 INPO # 20444 (Modified to JAF)
~~ ______ ~~ ~
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 8/13/2001 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of EOP implementation hierarchy and G~~~~#
coordination with other support procedures.
(CFR: 41.10 / / 45.13)
KIA # 2.4.16 Importance Rating 4.0 Proposed Question: In an emergency event the reactor scrammed due to high drywell pressure.
The following plant conditions exist:
Drywell temperature SPDS display DVVT "VERTICAL RUN TEMP' indicates 300 deg F.
e RPV pressure is 40 psig and equalized with the drywell.
e RPV water level indications are very erratic and do not correlate well with one another.
Under these circumstances, the operating crew would be required to execute the following Emergency Operating Procedure(s):
a) EOP-4, Primary Containment Control, ONLY.
ROlSRO b) EOP-2, RPV Control, AND EOP-4, Primary Containment Control, concurrently ONLY.
s94 c) Initially, EOP-2, RPV Control, EOP-4. Primary Containment Control, concurrently, Then EOP-2, RPV Control, AND EOP-7, RPV Flooding, concurrently.
d) Initially EOP-2, RPV Control, AND EOP-4, Primary Containment Control, concurrently, THEN EOP-7, RPV Flooding, &Q EOP-4, Primary Containment Control, concurrently.
Proposed Answer: d) Initially EOP-2, RPV Control, AND EOP-4, Primary Containment Control, concurrently, THEN EOP-7, RPV Flooding, AND EOP-4, Primary Containment Control, concurrently.
Explanation (Optional):
Technical Reference(s): EOP-2, 4, 7 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOP's Learning Objective: EOP2LP, EO-1.02, 1.03, EOP4LP, EO-4.02 (As available)
Question Source: Bank # JAF LOR # 20005204B04C Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10
55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of 10 CFR: 20 and related facility Group #
radiation control requirements.
(CFR: 41.12 143.4. 45.9 / 45.10)
WA # 2.3.1 2.3.1 Importance Rating 2.6 3.0 Proposed Question: As a result of degrading emergency conditions, the Shift Manager has directed you to immediately investigate an equipment problem inside a locked high radiation area. The duty RP technician is outside the Protected Area fence performing surveys. The ERO is not yet staffed.
What action should be taken to expedite your entry?
a) Using the key on any NPO Duty key ring.
ROlSRO b) Go to the RP office and sign out a key yourself.
72/95 c) Contact and meet the RP tech to obtain a key.
d) Obtain a radiological master key from the Shift Manager.
Proposed Answer: d) Obtain a radiological master key from the Shift Manager.
Explanation (Optional):
Technical Reference(s): AP-07.06 (Attach if not previously provided)
L- Proposed references to be provided to applicants during examination: None Learning Objective: LPAP, EO-31.03.H (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 4 Comments:
Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of which events related to system Group #
operationdstatus should be outside agencies.
45.11)
WA # 2.4.30 Importance Rating 3.6 Proposed Question: As a result of an error during IIC Surveillance testing at 100% power, the MSIV's inadvertently closed resulting in the following:
0 A Full SCRAM on MSlV closure.
HPCl initiation and injection.
HPCI tripped by the Operators.
Operator control of level using Feed and Condensate.
0 Operator control of RPV pressure using SRV's.
Which of the below is the earliest required NRC report?
a) Immediate Notification due to an Emergency Plan event declaration RO/SRO b) One (1) Hour Notification due to violation of the Reactor Coolant System Pressure Safety Limit.
S96 6) Four (4) Hour Notification due to ECCS discharge to Reactor Coolant System resulting from a valid signal.
d) Eight (8) Hour Notification due to a valid Containment Isolation signal affecting Containment Isolation Valves.
Proposed Answer: c) Four (4) Hour Notification due to ECCS discharge to Reactor Coolant System resulting from a valid signal.
Explanation (Optional): a) Distracter requires immediate notification of E-Plan declaration. There were no E-Plan declarations required in the stem.
b) Distracter requires 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notification upon safety limit violation. Per the stem- no information was given that the RPV pressure safety limit was violated. If the pressure safety limit had been violated, notifications would be required to be made per (50.36 ( c)(l) and requirements of (50.72(b)(3) (ii)
(b) which requires a 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notificationfor any event that results in the NPP being in a unanalyzed condition that significantly degrades plant safety. The plant is analyzed to not exceed the pressure safety limit on a closure of MSIV's at power.- see AP-03. I 1 Attachment # 7.
c) Correct response- 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification is required due to ECCS injection (50.72(b)(2) (iv) (a) - see AP-03.11 Attachment # 7 d) Distracter requires 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification due to valid Containment Isolation signal. Per the stem- Group Isolations did occur on valid low levels- question asked for earliest required notification which at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per answer IC".
Technical Reference(s): ENN-LI-102, AP-03.11 (Attach if not previously provided) 10CFR 50.36, 50.72 & NUREG-1022 Rev 2.
Proposed references to be provided to applicants during examination: AP-03.11 Learning Objective: LPAP-10.06 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of communications procedures Group #
associated with EOP implementation.
(CFR: 41.10/45.13)
KIA # 2.4.15 2.4.15 Importance Rating 3.0 3.5 Proposed Question: The Shift Manager has implemented the Emergency Plan based on high.Drywell pressure and assigned an Operator as the NRC Communicator. .
WHICH ONE of the following describes when communications with the NRC may be secured?
a) Technical Support Center is activated ROISRO b) NRC disconnects or authorizes securing line 73/97 c) Transient is over and the plant is recovering d) Once initial classification notice is provided to the NRC Proposed Answer: b) NRC disconnects or authorizes securing line Explanation (Optional):
Technical Reference(s): EAP-1.1 attachment 14 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: EP-I 2.5.5.1 , EO-2.07 (As available)
Question Source: Bank # Limerick 1 INPO # 12345 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1/20/1998 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of facility protection requirements Group #
including fire brigade and portable fire fighting equipment usage.
(CFR: 43.5 / 45.12)
WA # 2.4.26 2.4.26 Importance Rating 2.9 3.3 Proposed Question: Given the following conditions:
0 You are responding to an electrical fire as a member of the plant's fire brigade team.
0 You have brought a Class B/C fire extinguisher to the scene.
0 Other members have rigged a fire hose with a solid-stream nozzle.
Which one of the following actions should be taken?
a) Do not use the fire hose. Put the fire out with the Class B/C fire extinguisher.
b) Use the fire hose first. If it does not put out the fire, use the Class BIC fire RO/SRO extinguisher.
74/98 c) Wait for the fire brigade member assigned to bring a Class D fire extinguisher, then use the Class D fire extinguisher.
d) Do not use the Class B/C fire extinguisher. Put the fire out with the fire hose.
Proposed Answer: a) Do not use the fire hose. Put the fire out with the Class B/C fire extinguisher.
Explanation (Optional):
Technical Reference(s): (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDL-76, EO-1.05.A (As available)
Question Source: Bank # LaSalle 1 INPO # 11156 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1019/1995 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of the RO's responsibilities in Group #
emergency plan implementation.
(CFR: 45.11)
WA # 2.4.39 2.4.39 Importance Rating 3.3 3.1 Proposed Question: You are a licensed Reactor Operator on dayshift, working on the FIN Team. You do not have assigned responsibilities in the Emergency Response Organization (ERO). A transient occurs that results in the declaration of an ALERT Emergency and Protected Area Evacuation. To which of the following locations do you report?
a) The Operations Support Center (OSC).
ROISRO b) The Training Building assembly area.
75/99 c) The Technical Support Center (TSC).
d) The Offsite Assembly Area (Airport).
Proposed Answer: b) The Training Building assembly area.
Explanation (Optional):
Technical Reference(s): EAP-I 0 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: EP-12.5.3, EO-1.18 (As available)
Question Source: Bank # Duane Arnold 1 INPO # 8781 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 9120l1999 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowiedge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Examination Outline Cross-reference: Level SRO
\-- Tier # 3 Knowledge of the bases for prioritizing safety Group #
functions during ergency operations.
5.12)
WA # 2.4.22 Importance Rating 4.0 Proposed Question: If Torus level cannot be maintained above 10.75 feet, EOP-4, Primary Containment Control, directs the operator to ensure the HPCl turbine is tripped.
Which of the following describes the bases for RClC and HPCl operation under the same EOP circumstances (ie, Torus water level cannot be maintained above 10.75 feet)?
a) RClC operation may continue ONLY if it is the last operable high pressure injection source available to provide adequate core cooling.
RO/SRO b) RClC must be secured at the same time as HPCl to minimize the containment pressure rise.
s100 c) RClC operation may continue because the turbine exhaust energy does not exceed the vent capability of the containment.
d) RClC must be secured prior to HPCl to prevent erratic turbine operation due to exhaust back pressure fluctuation.
Proposed Answer: c) RClC operation may continue because the turbine exhaust energy does not exceed the vent capability of the containment.
Explanation (Optional):
Technical Reference(s): EPGI REV 2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: EOP4LP, EO-1.05 (As available)
Question Source: Bank # Fermi 2 2 INPO# 19714 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 6/14/2001 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments: