ML040690940

02/2004 - Final Written Exam

ML040690940
Person / Time
Site:	Grand Gulf
Issue date:	09/16/2003
From:	NRC Region 4
To:	Entergy Operations
References
50-416/04-301 50-416/04-301
Download: ML040690940 (100)

Text

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 1 The plant was operating at rated conditions with the electrical distribution system in the preferred lineup.

Conditions on the Entergy Grid resulted in grid voltage dropping to 200 KV for 10 seconds then returning to 500 KV.

Which one of the following describes the response of the plant electrical buses?

ASSUME NO OPERATOR ACTIONS AND ALL SYSTEMS FUNCTIONED AS DESIGNED.

BOP BUSES 15AA 16AB 17AC 11HD; 12HE; 13AD; 14AE A. de-energized energized from energized from energized from diesel w/ loads shed diesel w/ loads shed diesel w/ loads and re-sequenced and re-sequenced locked out B. energized w/ all energized from ESF energized from ESF energized from ESF loads restored transformer w/ no transformer w/ no transformer w/ no loss of loads loss of loads loss of loads C. energized w/ no energized from ESF energized from ESF energized from ESF loss of loads transformer w/ no transformer w/ no transformer w/ no loss of loads loss of loads loss of loads D. energized w/ energized from energized from energized from major pumps diesel w/ loads shed diesel w/ loads shed diesel w/ loads locked out and re-sequenced and re-sequenced restored QUESTION NRC RECORD # WRI 801 ANSWER: D. SYSTEM # R27; R21 K/A 295003 AK1.02: 3.1/3.4 P75; P81 AK1.03: 2.9/3.2 AK2.03: 3.7/3.9 AK3.01: 3.3/3.5 LP# GLP-OPS-R2700 AK3.02: 2.9/3.1 OBJ. 10; 23 AK3.03: 3.5/3.6 LP# GG-1-LP-OP-R2100 AA1.01: 3.7/3.8 OBJ. 12; 14; 16; 18; 20; 34 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

GGNS Tech Specs 3.3.8.1 NEW 05-1-02-I-4 sect 3.3.4e MODIFIED BANK DIFF 2; CA 04-1-01-R21-1 sect 5.1 & table 1 04-1-01-P81-1 sect 3.22 RO SRO BOTH CFR 41.7 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 2 The plant was operating at 100% power when a low primary water tank level activated the Electrical Generator Protection (EGP) system.

Which of the following best describes valve response to stabilize reactor pressure?

ASSUME NO OPERATOR ACTIONS AND ALL SYSTEMS FUNCTIONED AS DESIGNED.

Main Turbine Main Turbine Main Turbine Safety Relief Stop Valves Control Valves Bypass Valves Valves A. Open Open throttling Closed Closed, throughout controlling pressure transient B. Closed Closed Cycling open & Closed following an closed to control initial opening pressure C. Closed Closed Cycling open & Closed, throughout closed to control transient pressure D. Closed Closed Closed Cycling open &

closed to control pressure QUESTION NRC RECORD # WRI 802 ANSWER: B. SYSTEM # N11; N30; K/A 295006 AA1.03: 3.7/3.7 N32-2; B21 AK2.07: 4.0/4.1 LP# GLP-OPS-N4151 AA2.04: 4.1/4.1 OBJ. 11; 12.7 LP# GLP-OPS-N3202 OBJ. 3.1; 4 LP# GG-1-LP-OP-MCD7b OBJ. 2 (trans 15.2.2 & 15.2.3) SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

05-1-02-I-2 sect 3.2; 4.2.1d NEW 05-1-02-I-1 sect 2.3; 3.1 MODIFIED BANK DIFF 2; CA 03-1-01-4 sect 5.2.1 FSAR Tables 15.2-2 & 15.2-4 RO SRO BOTH CFR 41.5 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 3 A LOCA occurred 45 minutes ago in the Auxiliary Building steam tunnel.

The inboard MSIVs failed to isolate and the outboard MSIV on the damaged pipe remained open.

HPCS is tagged out for motor repairs.

The following conditions exist:

Reactor water level -155 inches wide range and dropping Reactor pressure 285 psig and stable Drywell pressure 1.0 psig and stable Aux Bldg Steam Tunnel Temperature reached 285°F 5 minutes into the transient and is presentlyF235°F and dropping Which one of the following identifies a pump capable of restoring and maintaining reactor water level?

ASSUME NO OPERATOR ACTIONS AND ALL SYSTEMS FUNCTIONED AS DESIGNED.

A. Condensate Pump A B. LPCS C. RCIC D. RHR A QUESTION NRC RECORD # WRI 803 ANSWER: B. SYSTEM # E12; E21; K/A 295007 AK1.01: 2.9/3.2 E51; N19 AK2.03: 3.1/3.2 LP# GLP-OPS-E2100 AK2.04: 3.2/3.3 OBJ. 8.3; 19 LP# GLP-OPS-E1200 OBJ. 4.2; 8.9; 23 LP# GLP-OPS-E5100 OBJ. 11; 22 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

04-1-02-H13-P601 21A-F7; G2 NEW 03-1-01-1 sect 6.2.4 MODIFIED BANK DIFF 2; CA 02-S-01-33 Att I pg 39 of 47 FSAR table 6.3-8 fig 6.3-6; 6.3-7 RO SRO BOTH CFR 41.6 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 4 The plant is operating at 100% power.

Feed flow transmitter C34-N002B has failed upscale.

Which one of the following describes the response of actual reactor water level and the Digital Feed Control System (DFCS)?

A. Reactor water level will immediately drop then return to normal level when the level dominance of the DFCS takes over and automatically substitutes the opposite line feed flow for the failed value.

B. Reactor water level will immediately drop then return to normal level when the DFCS system automatically de-selects and locks out three element control and selects single element control.

C. Reactor water level will remain stable when the level dominance of the DFCS takes over and automatically substitutes the opposite line feed flow for the failed value.

D. Reactor water level will remain stable when the DFCS system automatically de-selects three element control and selects single element control.

QUESTION NRC RECORD # WRI 804 ANSWER: D. SYSTEM # C34; K/A 295009 AA2.02: 3.6/3.7 B21 LP# GLP-OPS-C3400 OBJ. 6.3; 6.5; 10.7; SRO TIER 1 GROUP 1 / RO TIER GROUP 23

REFERENCE:

04-1-02-1H13-P680 NEW 2A-C9 MODIFIED BANK DIFF 2; CA NRC 8/2002 wri705 RO SRO BOTH CFR 41.4/41.5/41.7/

REFERENCE MATERIAL REQUIRED: NONE 43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 5 Plant Service Water is lost to the Auxiliary Building.

Which one of the following best describes the effects on drywell chillers and the drywell atmosphere?

ASSUME NO OPERATOR ACTION.

A. Drywell chillers will continue to operate but drywell temperature will rise. Drywell pressure will drop to the point that the Drywell Normal Vacuum relief valves will open equalizing pressure between the drywell and containment.

B. Drywell chillers will continue to operate, but drywell temperature will rise. Drywell pressure will rise to the point that the Drywell Normal Vacuum relief valves will open equalizing pressure between the drywell and containment.

C. Drywell chillers will trip causing drywell temperature to rise. Drywell pressure will remain constant due to the communication between the containment and drywell atmospheres via the drywell vents.

D. Drywell chillers will trip causing drywell temperature to rise. Drywell pressure will rise such that High Drywell Pressure alarms will actuate and an isolation of containment, drywell, and Auxiliary Building will occur due to a high drywell pressure.

QUESTION NRC RECORD # WRI 805 ANSWER: D. SYSTEM # M51; K/A 295010 AK2.05: 3.7/3.8 P72; P44 223001 K6.01: 3.6/3.8 LP# GLP-OPS-M5100 A4.12: 3.5/3.6 OBJ. 6.3; 10.4 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

04-1-02-H13-P870 NEW 3A-D4; F2 MODIFIED BANK DIFF 2; CA 04-1-02-H13-P680 5A-C2 NRC 3/1998 wri70 RO SRO BOTH CFR 41.4/41.5 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 6 The plant is performing a reactor startup from cold shutdown.

The reactor was critical with a 120 second period.

The At-The-Controls Operator felt power was slow to rise and withdrew the next control rod from position 08 to 10 as allowed by the Control Rod Movement Sequence Sheet.

This resulted in a sustained 15 second period.

Which one of the following describes the next action the At-The-Controls Operator should take?

A. Immediately range all IRMs to range 10 and monitor overlap data between IRMs and APRMs.

B. Monitor reactor power and range IRMs as necessary and verify the effects of the reactor reaching the point of adding heat and turning power ascension.

C. Insert the Control Rod back to position 08 and any other control rods as necessary to obtain a reactor period of > 50 seconds and inform supervision.

D. Inform the Reactor Engineer of the power rise, and insert the Control Rod as far as necessary to turn power then when the point of adding heat is reached control rod withdrawals may resume.

QUESTION NRC RECORD # WRI 806 ANSWER: C. SYSTEM # C11-2; C51 K/A 295014 AA1.03: 3.2/3.3 AA2.02: 3.9/3.9 AA2.03: 4.0/4.3 2.1.2: 3.0/4.0 LP# GLP-OPS-IOI01 2.1.30: 3.9/3.4 OBJ. 3.3 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

03-1-01-1 sect. 2.1.4 NEW Pilgrim Reactivity Event MODIFIED BANK DIFF 1; M 2/2003 RO SRO BOTH CFR 41.1/41.2/ 41.6/43.6 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 7 The plant was operating at 100% power when the Operator At-The-Controls observed indications that require a reactor scram to be inserted.

He placed the Reactor Mode Switch to Shutdown and noted no movement of the control rods.

Given the attached pictures of the H13-P680 indications, determine which one of the following identifies the primary reason control rods have failed to insert?

See attached Control Room Photos.

A. The Scram Discharge Volume has a hydraulic block.

B. The Reactor Protection System failed to de-energize.

C. ATWS ARI failed to de-energize and vent the Scram Air Header.

D. The Reactor Mode Switch failed to transfer from RUN to SHUTDOWN.

QUESTION NRC RECORD # WRI 807 ANSWER: B. SYSTEM # C11-2; C71; K/A 295015 AA1.04: 3.4/3.7 C11-1A AA1.01: 3.8/3.9 AA1.02: 4.0/4.2 LP# GLP-OPS-C7100 2.1.31: 4.2/3.9 OBJ. 6.4; 20; 23 LP# GLP-OPS-C1102 OBJ. 12; 14.8; 23; 26 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

05-1-02-I-1 sect 2.1; 4.1; 4.2; NEW 4.3; 5.1 04-1-01-C11-2 sect 4.7.2e, f, j MODIFIED BANK DIFF 2; CA 04-1-02-H13-P680 4A2-D4 5A-A1; A3; 7A-A3; B2 RO SRO BOTH CFR 41.6/43.5 REFERENCE MATERIAL REQUIRED: Simulator Photos

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 8 Power to the Safety Relief Valves (SRVs) via circuit breakers 72-11A23 and 72-11B34 has been lost.

Status lights on H13-P628 (Upper Control Room) and H13-P631 (Main Control Room) are de-energized.

Which one of the following describes the operation of the Low-Low Set SRVs from the Remote Shutdown Panels?

A. Low-Low Set SRVs will only operate on the Safety Function.

B. Low-Low Set SRVs will only operate on the Safety Function or by using the Remote Shutdown Panel handswitches.

C. Only the two ADS SRVs will operate on the Relief, ADS (as applicable), and Safety Functions as well as the Remote Shutdown Panel handswitches.

D. Low-Low Set SRVs will operate on the Relief, ADS (as applicable), and Safety Functions as well as the Remote Shutdown Panel handswitches.

QUESTION NRC RECORD # WRI 808 ANSWER: B. SYSTEM # C61; E22-2 K/A 295016 AA1.05: 2.8/2.9 LP# GLP-OPS-C61 OBJ. 6; 7.3; 14; 23 LP# GLP-OPS-E2202 OBJ. 9.1; 10.4; 12.5; SRO TIER 1 GROUP 1 / RO TIER GROUP 19.3; 25

REFERENCE:

E-1161-011; 012; 014; 017 NEW MODIFIED BANK DIFF 1; M RO SRO BOTH CFR 41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 9 The plant is operating at 100% power.

Control Room HVAC has isolated and Standby Fresh Air has initiated.

Which one of the following identifies the signal with present plant conditions that caused the isolation of the Control Room envelope?

A. High-High radiation levels in the outside air intake duct.

B. Smoke detected in the Control Room HVAC fan inlet.

C. High inlet air chlorine concentrations.

D. High inlet air Freon concentrations.

QUESTION NRC RECORD # WRI 809 ANSWER: A. SYSTEM # Z51 K/A 295017 AK3.01: 3.6/3.9 LP# GLP-OPS-Z5100 AK3.05: 3.3/3.6 OBJ. 8; 11; 13.9 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

04-S-02-H13-P855 NEW 1A-A5, 1A-B4 MODIFIED BANK DIFF 1; M 04-1-02-H13-P601 19A-A10; A11 RO SRO BOTH CFR 41.11/41.13/43.4 04-S-01-Z51-1 sect 3.2; 5.4.1a REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 10 You have been directed to line up Suppression Pool Cooling.

Which one of the following identifies who must be notified and the reason for this notification?

A. The Auxiliary Building Operator to verify the RHR system piping is vented prior to starting the RHR pump.

B. Radiation Protection to allow personnel to perform surveys of the Containment for elevated radiation levels.

C. Plant Services to ensure NO personnel are working in the Suppression Pool or on the Suppression Pool in the plant boat.

D. Plant Security to determine which personnel are in the Containment Building such that they can be notified of the RHR System start.

QUESTION NRC RECORD # WRI 810 ANSWER: B. SYSTEM # E12 K/A 295026 2.3.2: 2.5/2.9 LP# GLP-OPS-E1200 2.1.32: 3.4/3.8 OBJ. 14 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

04-1-01-E12-1 sect 3.1 & NEW Caution sect 5.2 MODIFIED BANK DIFF 1; M NRC 6/2001 wri 287 RO SRO BOTH CFR 41.10/41.12/

REFERENCE MATERIAL REQUIRED: None 43.4/43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 11 A LOCA has occurred.

Reactor water level is being maintained by Condensate and Feedwater with Startup Level Control in automatic.

Drywell pressure is stable at 2.23 psig.

The Entergy power grid is stable.

Which one of the following describes the operation of the Diesel Generators?

ASSUME THE DIESELS ARE OPERATING NORMALLY.

A. Division 1 & 2 DG operating and may be tied to the bus or shutdown.

Division 3 DG operating and may be tied to the bus or shutdown.

B. Division 1 & 2 DG operating and are unable to be tied to the bus or shutdown.

Division 3 DG operating and is unable to be tied to the bus or shutdown.

C. Division 1 & 2 DG operating may be tied to the bus, however are unable to be shutdown.

Division 3 DG operating and may be tied to the bus or shutdown.

D. Division 1 & 2 DG operating may be tied to the bus, however are unable to be shutdown.

Division 3 DG operating and is unable to be tied to the bus or shutdown.

QUESTION NRC RECORD # WRI 811 ANSWER: C. SYSTEM # P75; P81 K/A 295024 EK2.06: 3.9/4.0 LP# GLP-OPS-P8100 EA1.06: 3.7/3.7 OBJ. 8; 10; 11; 19; 26; 29 LP# GLP-OPS-E2201 OBJ. 11; 23 LP# GLP-OPS-P7500 OBJ. 8; 10; 11; 19; 26; 29 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

04-1-01-P75-1 sect 3.18 NEW 04-1-01-P81-1 sect 3.26 MODIFIED BANK DIFF 1; M RO SRO BOTH CFR 41.7/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 12 The plant was operating at 100% power when a transient caused reactor pressure to rise rapidly.

Reactor pressure rose to maximum pressure of 1130 psig and is currently at 956 psig.

Reactor water level dropped to a minimum of -20 inches wide range before recovery to

+18 inches.

Which one of the following describes the operation of the RPS, ATWS ARI and SRV valves?

ASSUME NO OPERATOR ACTIONS.

RPS Scram Pilot Valves ATWS ARI Valves SRVs A. Energized ready to reset Energized ready to reset 2 SRVs open B. De-energized ready to reset De-energized ready to reset 6 SRVs open C. Energized ready to reset De-energized ready to reset 2 SRVs open D. De-energized ready to reset Energized ready to reset 6 SRVs open QUESTION NRC RECORD # WRI 812 ANSWER: D. SYSTEM # C11-1A; K/A 295025 EK2.04: 3.9/4.1 C71; E22-2 EK2.01: 4.1/4.1 LP# GLP-OPS-C111A EK2.05: 4.1/4.2 OBJ. 7.4; 8.7 LP# GLP-OPS-E2202 OBJ. 11.2; 12.1; 13 LP# GLP-OPS-C7100 OBJ. 6.4; 9; 10 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

Tech Specs3.3.1.1; 3.3.4.2 NEW 3.3.6.5; 3.4.4 MODIFIED BANK DIFF 1; M RO SRO BOTH CFR 41.3/41.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 13 An ATWS has caused Containment conditions to degrade.

Containment Sprays are operating.

Which one of the following situations would require operators to Emergency Depressurize the reactor?

Drywell Drywell Containment Containment Suppression Temperature Pressure Temperature Pressure Pool Level A. 188°F rising 8.5 psig 135°F rising 3.0 psig 18.2 ft B. 250°F rising 8.5 psig 188°F rising 3.0 psig 18.2 ft C. 188°F lowering 3.0 psig 135°F lowering 8.5 psig 22.2 ft D. 250°F lowering 3.0 psig 188°F lowering 8.5 psig 22.2 ft QUESTION NRC RECORD # WRI 813 ANSWER: B. SYSTEM # M41-1 K/A 295027 EA1.03: 3.5/3.8 LP# GG-1-LP-RO-EP03 EK3.01: 3.7/3.8 OBJ. 3 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

05-1-01-EP-3 steps 23 - 30 NEW MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR 41.10/41.9/

REFERENCE MATERIAL REQUIRED: 05-1-01-EP-3 43.2/43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 14 Suppression Pool level has dropped to 14 ft 6 inches due to a leak.

Which one of the following describes the significance of this loss of level?

A. The Suppression Pool temperature monitors will be uncovered such that they will indicate Containment air temperature instead of pool temperature.

B. The Safety Relief Valve (SRV) spiders do not have sufficient coverage to prevent uncondensed steam from being introduced to the Containment.

C. The ECCS pumps are at the limit below which cavitation may result due to insufficient suction pressure because of Suppression Pool temperatures.

D. The Drywell to Containment Suppression Pool vents are partially uncovered causing a loss of the pressure suppression capability of the Suppression Pool.

QUESTION NRC RECORD # WRI 814 ANSWER: C. SYSTEM # E12; E21; K/A 295030 EK1.02: 3.5/3.8 E22-1; E51; M24 LP# GLP-OPS-E1200 OBJ. 14 LP# GLP-OPS-E2201 OBJ. 15 LP# GLP-OPS-E2100 OBJ. 11 LP# GLP-OPS-E5100 OBJ. 13 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

04-1-01-E12-1 sect 3.2.6 NEW 04-1-01-E21-1 sect 3.13 MODIFIED BANK DIFF 1; M 04-1-01-E22-1 sect 3.19 04-1-01-E51-1 sect 3.4 RO SRO BOTH CFR 41.8/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 15 The following conditions exist in the plant:

Reactor power 2% with 44 rods withdrawn.

Reactor pressure 300 psig and stable.

Reactor level - 230 inches Fuel Zone and lowering 6 Safety Relief Valves have been manually opened.

RCIC is injecting into the reactor vessel.

Which one of the following identifies the state of core cooling?

A. Adequate core cooling is NOT assured.

B. Adequate core cooling is assured by Minimum Alternate RPV Flooding Pressure (MARFP).

C. Adequate core cooling is assured by Minimum Zero RPV Water Level without RPV injection.

D. Adequate core cooling is assured by Minimum Steam Cooling Water Level.

QUESTION NRC RECORD # WRI 815 ANSWER: B. SYSTEM # B21; EPs; K/A 295031 EK1.01: 4.6/4.7 Cond of Ops 2.1.1: 3.7/3.8 2.4.6: 3.1/4.0 2.4.21: 3.7/4.3 LP# GLP-OPS-PROC 2.4.18: 2.7/3.6 OBJ. 11.2 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

01-S-06-2 sect 5.18 NEW 05-1-01-EP-2 (EP2A) Table 2 MODIFIED BANK DIFF 2; CA NRC 8/2002 wri309 RO SRO BOTH CFR 41.2/41.3/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-1-01-EP-2/2A 41.14

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 16 The plant is being controlled from the Remote Shutdown Panels.

RCIC is being used to control Reactor Water level.

Reactor pressure is 400 psig.

Reactor Water Level indication on H22-P150 is - 30 inches.

Which one of the following identifies actual Reactor Water Level?

05-1-02-II-1 Attachments I & II are provided.

A. - 30 inches B. - 35 inches C. - 40 inches D. -45 inches QUESTION NRC RECORD # WRI 816 ANSWER: C. SYSTEM # C61; B21-1 K/A 295031 EK2.01: 4.4/4.4 2.1.31: 4.2/3.9 2.1.25: 2.8/3.1 2.4.11: 3.4/3.6 LP# GLP-OPS-C6100 EA2.01: 4.6/4.6 OBJ. 19 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

05-1-02-II-1 Att II NEW MODIFIED BANK DIFF 2; CA NRC 8/2002 wri603 RO SRO BOTH CFR 41.7/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-1-02-II-1 Att. I & II

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 17 Which one of the following is a basis for controlling Hydrogen Concentrations inside the Mark III Containment?

A. Ignition of excessive Hydrogen when combined with sufficient quantities of Oxygen could result in excessive internal peak pressures challenging the structural integrity of the Containment.

B. Ignition of excessive Hydrogen when combined with Oxygen that could result in an excessive rate of change in the Containment pressure resulting in excessive force on the Drywell structure which could cause its failure.

C. Ignition of excessive Hydrogen and Oxygen combinations would result in excessive temperatures degrading the structural integrity of the Containment and safety related equipment inside the Containment.

D. Uncontrolled ignition of excessive Hydrogen and Oxygen inside the Containment and Drywell would result in excessive temperatures impinging on safety related components inside the Containment and Drywell causing their ultimate failure.

QUESTION 17 NRC RECORD # WRI 817 ANSWER: A. SYSTEM # EP Bases K/A 500000 EK1.01: 3.3/3.9 LP# GG-1-LP-RO-EP03 2.4.18: 2.7/3.6 OBJ. 6 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

GGNS PSTG Appendix A NEW GGNS PSTG Appendix B PC/G Combustible Gas Conc. MODIFIED BANK DIFF 1; M RO SRO BOTH CFR 41.9 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 18 An ATWS has occurred.

RCIC has automatically started to assist in control of Reactor Water level.

Reactor pressure is rising.

Which one of the following describes the response of RCIC?

ASSUME NO OPERATOR ACTION.

A. RCIC speed will drop with injection rate remaining relatively stable.

B. RCIC speed will rise with injection rate remaining relatively stable.

C. RCIC speed will drop causing the injection rate to drop.

D. RCIC speed will rise causing the injection rate to rise.

QUESTION NRC RECORD # WRI 818 ANSWER: B. SYSTEM # E51 K/A 295025 EA1.05: 3.7/3.7 LP# GLP-OPS-E5100 217000 A1.04: 3.6/3.6 OBJ. 4.11; 8.17; 22 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

Vendor Manual 460000182 NEW Simulator Response MODIFIED BANK DIFF 1; M RO SRO BOTH CFR 41.4/41.5/41.14 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 19 The plant was operating in the MEOD region of the Power to Flow Map.

A reduction in total core flow combined with a reduction in Feedwater temperature could result in which one of the following conditions?

ASSUME NO OPERATOR ACTIONS.

A. Thermal hydraulic instability, which left unimpeded, would result in violation of the MCPR thermal and safety limits.

B. Actual core thermal power being higher than indicated that could lead to violation of the APLHGR thermal limit.

C. Thermal hydraulic instability, which immediately reduces the margin of safety to the LHGR and APLHGR thermal limits.

D. Actual core thermal power being higher than indicated that could lead to violation of the MCPR thermal and safety limits.

QUESTION NRC RECORD # WRI 819 ANSWER: A. SYSTEM # B33 K/A 295001 2.2.34: 2.8/3.2 AK1.04: 2.5/3.3 LP# GLP-OPS-B3300 AK1.03: 3.6/4.1 OBJ. 30; 46; 50 LP# GLP-OPS-ONEP OBJ. 24 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

Tech Spec Bases B3.3.1.1-2d NEW B3.2.2; B2.1.1.2; B3.3.1.3 MODIFIED BANK DIFF 1; M 05-1-02-III-3 sect 4.9 (NOTE)

& Sect 5.3 RO SRO BOTH CFR 41.5/41.10/43.5/43.1 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 20 An ATWS has occurred.

Reactor Feed Pumps are unavailable.

Reactor power is at 35 %. Reactor water level is at - 90 inches and lowering.

The plant is at rated pressure.

The SRO directing actions per EP-2A has directed reactor pressure to be lowered to 500 psig to support the condensate system feeding the reactor vessel.

Which one of the following describes the Reactor Power response when reducing Reactor Pressure with present conditions?

A. reactor power will rise due to the collapsing of the voids resulting in more neutron thermalization which in turn heats the moderator.

B. Reactor power will rise due to boron blowing out of the core region which was absorbing thermal neutrons.

C. Reactor power will drop due to the voiding of the core and remain lower than the original power.

D. Reactor power will initially drop due to voiding followed by a rise in power due to the lowering the moderator temperature.

QUESTION NRC RECORD # WRI 820 ANSWER: D. SYSTEM # J11 K/A 295015 AK1.04: 3.8/3.8 LP# GG-1-LP-RO-EP02A AK1.02: 3.9/4.1 OBJ. 2 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

PSTG App B RC/P-2 NEW MODIFIED BANK DIFF 1; M NRC 5/2000 wri 180 RO SRO BOTH CFR 41.1/41.2/41.5/43.6 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 21 The plant is in cold shutdown with reactor level being maintained by CRD and RWCU.

A breach of primary containment has occurred.

Suppression Pool level has dropped and the Control Room Operator is unable to determine level.

RCIC suction has been aligned to the Suppression Pool.

Drywell pressure -0.01 psig Containment pressure -0.1 psig RCIC suction pressure on E51-R604 +3.5 psig What is Suppression Pool Level?

05-1-01-EP-2 Att 29 is provided.

A. 11.8 feet B. 13.8 feet C. 14.1 feet D. 14.3 feet QUESTION NRC RECORD # WRI 821 ANSWER: D. SYSTEM # E30; M24 K/A 295030 EA2.01: 4.1/4.2 2.1.25: 2.8/3.1 LP# GG-1-LP-OP-EP007 2.4.21: 3.7/4.3 OBJ. 3 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

05-1-01-EP-2 Att 29 NEW 3.5psig - (-0.1psig) = 3.6 psig MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR 41.7/41.9/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-1-01-EP-2 Att 29

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 22 Given the following conditions:

Reactor power 45%

Reactor level -100 inches Reactor pressure 850 psig Suppression pool temperature 150°F Suppression pool level 19 feet 6 inches 4 SRVs are open Which one of the following best describes the required actions to be taken given the above conditions?

A. Close two of the four SRVs and raise the reactor pressure band to a top end of 1000 psig, to reduce the amount of heat entering the Suppression Pool.

B. Immediately commence an Emergency Depressurization in accordance with EP-2A because limits in the Containment have been exceeded based on Suppression Pool temperature.

C. Lower reactor pressure using cooldown rates that may exceed 100 °F/Hr, to avoid jeopardizing Containment by exceeding the heat capacity temperature limit of the Suppression Pool.

D. Conditions at present are acceptable, however add water from the CST to the Suppression Pool to lower temperature and raise level to add capacity for the Suppression Pool.

QUESTION NRC RECORD # WRI 822 ANSWER: C. SYSTEM # E30; M24 K/A 295026 EK1.02: 3.5/3.8 EOPs HCTL 2.4.14: 3.0/3.9 2.4.6: 3.1/4.0 LP# GG-1-LP-RO-EP02A 2.4.18: 2.7/3.6 OBJ. 2; 3b; 5 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

05-1-01-EP-2 steps 14 & 15 NEW HCTL curve Figure 1 MODIFIED BANK DIFF 2; CA 05-1-01-EP-3 steps 14 & 15 NRC 3/1998 wri 035 RO SRO BOTH CFR 41.5/41.9/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-1-01-EP-2 & 3 43.2

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 23 The plant is at 900 psig with the MSIVs open.

Reactor power is 5%.

Reactor water level is +45 inches.

Damage occurs to the IP Condenser boot resulting in Main Condenser vacuum dropping to 8 inches Hg vac in all three condensers.

Which one of the following describes the immediate response of the plant?

ASSUME NO IMMEDIATE OPERATOR ACTIONS.

A. The reactor will remain critical and the MSIVs will remain open.

B. The reactor would scram due to high reactor water level with pressure control on the SRVs.

C. The reactor would scram due to high reactor pressure with pressure control on the SRVs.

D. The reactor will remain critical with pressure control on the SRVs.

QUESTION NRC RECORD # WRI 823 ANSWER: C. SYSTEM # C71; N62 K/A 295002 AA1.03: 3.4/3.5 LP# GLP-OPS-C7100 AA1.04: 3.3/3.4 OBJ. 10 AK2.01: 3.5/3.5 LP# GLP-OPS-M7101 AK2.03: 3.5/3.6 OBJ. 7.1; 8.1 LP# GG-1-LP-OP-N6200 OBJ. 14 LP# GLP-OPS-ONEP OBJ. 39 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

Tech Spec 3.3.1.1 NEW 05-1-02-I-1 sect 4.5.2 MODIFIED BANK DIFF 2; CA 05-1-02-V-8 sect 5.4 03-1-01-1 sect 6.1.10 RO SRO BOTH CFR 41.5/41.4/41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 24 The plant is operating at 100% power.

125 VDC breakers 72-11D15 and 72-11D16, DC control power breakers for 6.9 KV bus 11HD, have tripped.

Reactor Recirculation Pump A motor amps are indicating off-scale high on H13-P680.

Which one of the following correctly describes the response of Reactor Recirculation Pump A breaker?

Reactor Recirculation Pump A breaker:

A. will trip on overcurrent.

B. will NOT trip on overcurrent, but can be tripped manually from the H13-P680.

C. will NOT trip on overcurrent, but can be tripped locally using the local control switch (pistol grip) on breaker 252-1103 (CB-5A).

D. will NOT trip on overcurrent, but can be tripped locally, using the local manual trip pushbutton on breaker 252-1103 (CB-5A)

QUESTION NRC RECORD # WRI 824 ANSWER: D. SYSTEM # L11/R27 K/A 295004 AA1.03: 3.4/3.6 LP# GLP-OPS-R2700 GLP-OP-L1100 OBJ. 14/10a SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

04-1-01-L11-1 Att. ID NEW E-1163-003 MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR41.5/41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 25 The plant was at 22% power.

An EGP trip of the Main Generator occurred.

Which one of the following describes the affect on the temperature of Feedwater entering the reactor?

A. Feedwater temperature will rise due to higher steam pressure applied to the Feedwater heaters.

B. Feedwater temperature will drop due to removal of heating steam to the Feedwater heaters.

C. Feedwater temperature will remain stable because the turbine stop and control valves will remain open.

D. Feedwater temperature will remain stable because the rise in steam pressure will offset the reduced steam flow to the Feedwater heaters.

QUESTION NRC RECORD # WRI 825 ANSWER: B. SYSTEM # N23; N30; K/A 295005 AK3.03: 2.8/3.0 N41; N11/36 LP# GLP-OPS-N2335 OBJ. 14.2; 24 LP# GLP-OPS-N1136 OBJ. 2 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

SFD-1054 conditions A & G NEW 05-1-02-I-2 sect 4.2.1 MODIFIED BANK DIFF 1; M RO SRO BOTH CFR 41.5/41.6 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 26 The plant is in Mode 2 following a refueling outage.

Reactor temperature is 140°F.

A leak in the RWCU Regen Heat Exchanger has caused the RWCU Heat Exchanger Room temperature to rise to 125°F.

Which one of the following describes the effect this will have on reactor water level?

A. Reactor level will remain stable.

B. Reactor water level will rise; the reactor will scram on high reactor water level.

C. Reactor water level will rise; the reactor will NOT scram on high reactor water level.

D. Reactor water level will lower; the reactor will scram on low reactor water level.

QUESTION NRC RECORD # WRI 826 ANSWER: C. SYSTEM # G33 K/A 295008 AA2.03: 2.9/3.0 AA2.04: 3.1/3.3 AA2.05: 2.9/3.1 LP# GLP-OPS-G3336 OBJ. 8.6 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

05-1-02-III-5 NEW 03-1-01-1 MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR41.4/41.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 27 A flange rupture on the RWCU Regenerative Heat Exchanger inlet has caused an isolation of RWCU.

RPV water level dropped to -20 inches on wide range before recovering.

The following conditions exist in the plant:

Reactor level +36 inches and stable Reactor pressure 1000 psig and stable Drywell pressure + 1.5 psig and slowly rising Drywell temperature 110 °F Containment pressure + 1.5 psig Containment temperature 175 °F Suppression Pool temperature 86 °F Suppression Pool level 18.6 feet Which one of the following describes the available methods to remove heat from the Containment under the present conditions?

A. Containment Coolers with chilled water Containment Steam Tunnel Coolers with chilled water B. Containment Coolers without chilled water Containment Steam Tunnel Coolers without chilled water C. Containment Coolers without chilled water Containment Steam Tunnel Coolers with chilled water D. Containment Coolers with chilled water Containment Steam Tunnel Coolers without chilled water QUESTION NRC RECORD # WRI 827 ANSWER: SYSTEM # M41; P72; M71 K/A 295011 AK3.01: 3.6/3.9 B & C.

LP# GLP-OPS-P7100 AK2.01: 3.7/4.0 OBJ. 1; 7 LP# GLP-OPS-ONEP OBJ. 28; 29 LP# GLP-OPS-M7101 OBJ. 7.6; 8.1; 20 LP# GLP-OPS-M5100 OBJ. 1; 6.5 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

05-1-02-III-5 sect 3.5.3d NEW 05-S-01-EP-3 steps 23 - 26 MODIFIED BANK DIFF 3; CA Steps 17-20 & Att 10 NRC 12/2000 wri323 M-1079; 1119 RO SRO BOTH CFR 41.5/41.9/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 28 A complete loss of offsite power has occurred.

Instrument air is unavailable to the Containment and Auxiliary Buildings.

All SRV accumulators have bled down to 0 psig.

The Control Room Supervisor has ordered nitrogen bottles be installed per the loss of Instrument Air ONEP to allow use of SRVs for reactor pressure control.

Which one of the following identifies the SRVs that may be used for reactor pressure control?

A. All ADS valves and the lead Low-Low Set valve B. All ADS valves and all Low-Low Set valves C. Only Low-Low Set valves D. Only ADS valves QUESTION NRC RECORD # WRI 828 ANSWER: A. SYSTEM # P53; E22-2 K/A 295019 AK3.01: 3.3/3.4 LP# GLP-OPS-P5300 OBJ. 5.5; 19 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

05-1-02-V-9 sect 3.12 NEW M-1067A & M MODIFIED BANK DIFF 1; M M-1077C RO SRO BOTH CFR 41.4/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 29 The plant is in Mode 2 following a refueling outage.

A disturbance on the grid caused a loss of offsite power to the plant.

The emergency diesel generators automatically started and restored power to their associated buses.

Unit I Instrument Air Compressor has been started.

HPCS and RCIC were manually started to control reactor water level.

The current conditions exist:

Reactor level +35 inches Drywell Pressure 0.23 psig Which one of the following describes the required actions to restore the Auxiliary Building Instrument Air Isolation valves (P53-F026A and P53-F026B)?

A. NO action is required. The valves did NOT automatically close.

B. The valves will automatically open when power and air are restored.

C. The valves can be manually opened using their associated handswitches as soon as power and air are restored D. The valves can be manually opened using their associated handswitches ONLY after the Auxiliary Building Bypass switches have been placed in the BYPASS position.

QUESTION NRC RECORD # WRI 829 ANSWER: C. SYSTEM # M71/P53 K/A 295020 AA2.06: 304/3.8 LP# GLP-OPS-P5300 OBJ. 15.2 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

E-1229-006 NEW E-1229-009 04-1-01-P53-1 Att. III MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR41.4/41.7/41.9/

REFERENCE MATERIAL REQUIRED: None 41.10/43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 30 The plant is in Mode 5.

ADHR is in service in Spent Fuel Pool to Reactor mode.

Under vessel work resulted in a leak causing RPV level to lower to -50 inches.

Which one of the following describes the affects on ADHR shutdown cooling operation?

A. ADHR pumps will trip due to a loss of suction pressure when E12-F008 and F009, RHR SDC Isolation valves isolate on low reactor level.

B. ADHR will continue to operate as long as level in the Spent Fuel Pool remains sufficient to provide makeup to the suction of the ADHR pumps.

C. ADHR suction valves from the Spent Fuel Pool will isolate causing a loss of suction pressure to the ADHR pumps which will subsequently trip.

D. ADHR will continue to operate indefinitely, however cooling to the ADHR heat exchangers is lost until Plant Service Water is restored.

DELETE QUESTION NO ANSWER CORRECT QUESTION NRC RECORD # WRI 830 ANSWER: SYSTEM # E12-1 K/A 295021 AA2.02: 3.4/.4 LP# GG-1-LP-OP-E1201 OBJ. 8a; 16 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

04-1-01-E12-1 sect 3.15.5 & NEW 5.13 MODIFIED BANK DIFF 1; M M-1085A, C, D M-1088E RO SRO BOTH CFR 41.5/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 31 The plant is critical in Mode 2 at 500 psig.

CRD pump B is tagged out for repairs.

CRD pump A has tripped.

Control rod 20-05 is at position 24; the Auxiliary Building Operator has reported accumulator pressure is at 1400 psig and there is a nitrogen leak on the accumulator.

Which one of the following describes the action to take for these conditions?

A. Declare control rod 20-05 slow, monitor conditions and attempt to restore the CRD system status to normal.

B. Declare control rod 20-05 INOP, monitor conditions and attempt to restore the CRD system status to normal.

C. Declare control rod 20-05 accumulator INOP and place the Reactor Mode Switch to SHUTDOWN.

D. Insert an individual control rod scram on control rod 20-05 by taking both SRI Test switches to TEST then return the switches to NORM, monitor conditions and attempt to restore the CRD system status to normal.

QUESTION NRC RECORD # WRI 831 ANSWER: C. SYSTEM # C11-1A K/A 295022 AK1.01: 3.3/3.4 2.1.7: 3.7/4.4 2.4.4: 4.0/4.3 LP# GLP-OPS-ONEP 2.4.49: 4.0/4.0 OBJ. 1 LP# GLP-OPS-C111A OBJ. 15 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

05-1-02-IV-1 sect 2.1.1 NEW Tech Specs & Bases 3.1.5 MODIFIED BANK DIFF 1; M Conditions C & D RO SRO BOTH CFR 41.5/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 32 The following conditions are observed after a Loss of Coolant Accident:

Reactor Pressure 50 psig 166 elev. temperature in the Drywell 320°F Drywell Pressure 5.8 psig 139 elev. temperature in the Containment 192 °F 119 elev. temperature in the Containment 181 °F Containment Pressure 2.0 psig Shutdown Range Level Indication +154 inches Upset Range Level Indication +154 inches Wide Range Level Indication + 60 inches All level indicators appear to be tracking.

Which one of the following is true regarding Reactor Water Level?

A. Assume all level indicators are indicating properly until spiking or notching is observed.

B. Shutdown and Upset Range level indications are the only reliable level indication.

C. Wide Range level indication is the only reliable level indication.

D. There are NO reliable level indications at this time.

QUESTION NRC RECORD # WRI 832 ANSWER: A. SYSTEM # B21-1 K/A 295028 EK1.01: 3.5/3.7 LP# GG-1-LP-RO-EP02 EA2.03: 3.7/3.9 OBJ. 11 LP# GLP-OPS-PROC OBJ. 56.3 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

05-S-01-EP-2 Caution 1 NEW 02-S-01-27 sect 6.1.8; 6.2.1 MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR 41.5/41.7/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-S-01-EP-2

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 33 The plant is operating at 100% power.

The Riley temperature switches for RCIC Room are in alarm with readings of 260°F and continuing to rise.

All other temperatures are indicating normal.

Reactor steam flow and feed flow indications are matched.

Which one of the following describes the actions to be taken?

A. Isolate all systems discharging into Secondary Containment except as required to suppress a fire and enter EP - 2.

B. Isolate all systems discharging into Secondary Containment and perform a plant shutdown per IOI 03-1-01-2.

C. Monitor plant conditions and determine the cause of the high temperature and restore Secondary Containment parameters to within limits per EP-4.

D. Isolate all systems connected to the RPV which are discharging into Secondary Containment and continue to monitor plant conditions per EP-4.

QUESTION NRC RECORD # WRI 833 ANSWER: A. SYSTEM # P64; E31 K/A 295032 EA1.04: 3.3/3.4 EK2.03: 3.3/3.4 LP# GG-1-LP-RO-EP04 EA2.03: 3.8/4.0 OBJ. 7 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

05-S-01-EP-4 steps 7 - 13 NEW PSTG Appendix B MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR 41.4/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-S-01-EP-4

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 34 The plant was operating at 100% power when Main Steam Tunnel temperature and radiation levels started rising.

The reactor was manually scrammed.

Reactor level and pressure are stable.

All equipment functioned as designed.

The following conditions exist:

Main Steam Line Rad. Monitors A 8.5 x 104 mR/HR B 7.8 x 104 mR/HR C 7.9 x 104 mR/HR D 7.5 x 104 mR/HR Main Steam Tunnel Temp 176°F RCIC Room Temp. 218°F RCIC Room Rad. Monitor 8.4 x 104 mR/HR RHR A Room Temp. 215°F RHR A Room Rad. Monitor 7.3 x 104 mR/HR RHR B Room Temp. 195°F RHR B Room Rad. Monitor 6.9 x 104 mR/HR Which one of the following lists the appropriate actions/responses for this situation?

A. MSIVs automatically isolated; Standby Gas Treatment System manually initiated; Emergency Depressurization required; EP-2 entered.

B. MSIVs manually isolated; Standby Gas Treatment System in standby; IOI-3 entered.

C. MSIVs automatically isolated; Standby Gas Treatment System automatically initiated; IOI-3 entered.

D. MSIVs open; Standby Gas Treatment System NOT initiated; EP-2 entered.

QUESTION NRC RECORD # WRI 834 ANSWER: B. SYSTEM # EP K/A 295033 2.1.7: 3.9/4.2 LP# GG-1-LP-RO-EP04 OBJ. 7 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

05-S-01-EP-4 Steps 1-18 NEW 05-1-02-III-5 MODIFIED BANK PSTG Appendix B DIFF 2; CA RO SRO BOTH CFR41.12/43.4 REFERENCE MATERIAL REQUIRED: 05-S-01-EP-4

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 35 The plant is operating at 100% power.

High radiation in the Fuel Handling Area caused initiation of Standby Gas Treatment System A.

Which one of the following describes the effect this will have on the fan coil units in the Auxiliary Building?

A. Auxiliary Building Zone 1, 2, 3, 4, and CRD Repair Rm. Fan Coil Units will trip.

Fuel Handling Area Fan Coil Units B002 (208 el.) and B005 (185 el.) will trip.

B. Auxiliary Building Zone 1, 3, and CRD Repair Rm. Fan Coil Units will trip.

Fuel Handling Area Fan Coil Units B005 (185 el.) and B007 (139 el.) will trip.

C. Auxiliary Building Zone 2 and 4 Fan Coil Units will trip.

Fuel Handling Area Fan Coil Units B002 (208 el.) and B005 (185 el.) will trip.

D. Auxiliary Building Zone 1, 2, 3, 4, and CRD Repair Rm. Fan Coil Units will trip.

All Fuel Handling Area Fan Coil Units will trip.

QUESTION NRC RECORD # WRI 835 ANSWER: A. SYSTEM # T41/T42 K/A 295034 2.1.7: 3.7/4.4 LP# GG-1-LP-OP-T4100 GG-1-LP-OP-T4200 OBJ. 9a, 10, 11d SRO TIER 1 GROUP 2 / RO TIER GROUP 8a, 11, 12d

REFERENCE:

04-1-01-T48-1 NEW E-1254-032 MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR41.4/41.10/41.13/

REFERENCE MATERIAL REQUIRED: None 43.4

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 36 The plant is operating at 100% power.

Fuel Handling Area differential pressure rose to 0 inches wc.

Which one of the following describes the possible cause and remedy for this situation?

A. Standby Gas Treatment is operating and requires securing.

B. Both Fuel Handling Area Supply fans are operating requiring one of the fans to be secured.

C. Both Fuel Handling Area Exhaust fans are operating requiring one of the fans to be secured.

D. Controller T42-PDK-R600 for T42-F021 Fuel Handling Area Pressure Cont has malfunctioned and closed the damper requiring manual control to open the damper.

QUESTION NRC RECORD # WRI 836 ANSWER: B. SYSTEM # T42; T48 K/A 295035 EK2.01: 3.6/3.6 LP# GG-1-LP-OP-T4200 OBJ. 3a; 10a SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

M-1104A NEW 04-1-02-H13-P842 1A-E3 MODIFIED BANK DIFF 1; M RO SRO BOTH CFR 41.4/41.7/41.13 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 37 A LOCA has occurred and the signals are still present.

When RHR A automatically started on high drywell pressure, a leak developed on the Suppression Pool side of E12-F004A, RHR A Suppression Pool Suction.

RHR A was secured and E12-F004A closed.

Water level in the RHR A pump room is 5 inches over the top of the floor on 93 ft Elevation.

Which one of the following describes the actions to be taken to control water level in the RHR A Pump Room?

A. Verify the RHR A Pump Room Sump Pumps are operating and pumping water to the Auxiliary Building Floor Drain Transfer tank and then on to Radwaste.

B. Verify the RHR A Pump Room Sump Pumps are operating and pumping water to the Auxiliary Building Floor Drain Transfer tank and then on to the Suppression Pool via P45-F273 and F274.

C. Declare RHR A INOP and dispatch an operator to open circuit breakers associated with RHR A equipment and allow water level in the RHR A Pump Room and Suppression Pool level to equalize.

D. Verify the RHR A Pump Room Sump Pumps are operating and pumping water to the Auxiliary Building Floor Drain Transfer tank and then transfer water on to Radwaste by overriding the Auxiliary Building isolation valves.

QUESTION NRC RECORD # WRI 837 ANSWER: B. SYSTEM # P45 K/A 295036 EA1.03: 2.8/3.0 LP# GLP-OPS-P4500 OBJ. 3b7 & 10; 7b; SRO TIER 1 GROUP 2 / RO TIER GROUP 9; 12

REFERENCE:

M-1094B & C NEW 05-1-02-VI-1 MODIFIED BANK DIFF 2; CA sect 1.1; 3.1; 3.3; 3.8 05-S-01-EP-4 step 7 RO SRO BOTH CFR 41.4/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 38 The plant is operating at rated conditions.

A smoke detector in the hallway area 8, 119 ft elevation of the Auxiliary Building is in alarm.

Which one of the following describes the actions that occur for this alarm?

A. The Control Room Security and Fire Protection Computer will activate the fire protection Wet Pipe Sprinkler system in area 8 119 ft elevation to suppress the fire.

B. The Control Room Security and Fire Protection Computer will activate the fire protection Carbon Dioxide (CO2) systems in area 8 119 ft elevation to suppress the fire.

C. Upon verification by an operator of a fire, the operator will manually activate the deluge valve for the Wet Pipe Sprinkler system in area 8 119 ft elevation.

D. The local fire panel and the Control Room Security and Fire Protection Computer will alarm requiring an operator to be dispatched to investigate. NO other automatic actions will occur from the detection system.

QUESTION NRC RECORD # WRI 838 ANSWER: D. SYSTEM # P65; P64; C83 K/A 600000 EA2.03: 2.8/3.2 LP# GLP-OPS-P6400 OBJ. 3.8; 3.13 SRO TIER 1 GROUP 2 / RO TIER GROUP

REFERENCE:

04-1-01-C83-1 NEW Sect 3.2; 4.2; Att V MODIFIED BANK DIFF 1; M 04-S-01-P64-1 Att VII M0035B & J RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 39 The plant is operating at 36% power.

A malfunction of the IPC results in the Main Steam Bypass Valves opening and passing 10% power to the Main Condenser.

The combined Main Stop and Control Valves responded lowering turbine output.

Which one of the following describes the movement of control rods with present plant conditions?

A. Rod movements are bound by the constraints of the Rod Pattern Controller. As long as the Rod Pattern is met control rods may be moved as desired.

B. Rod movements are bound by the Rod Withdrawal Limiter to limit control rod withdrawals to 2 notches. Control rod insertions are unlimited.

C. Rod movements are bound by the Rod Withdrawal Limiter to limit control rod withdrawals to 4 notches. Control rod insertions are unlimited.

D. Rod withdrawals are prohibited until the Main Steam Bypass valves are closed.

Control rod insertions are unlimited.

QUESTION NRC RECORD # WRI 839 ANSWER: D. SYSTEM # C11-2 K/A 201005 A1.01: 3.2/3.3 A2.04: 3.2/3.2 K6.01: 3.2/3.2 K5.10: 3.2/3.3 LP# GLP-OPS-C1102 K1.02: 3.3/3.5 OBJ. 6; 7; 13.3; 20; SRO TIER 2 GROUP 1 / RO TIER GROUP 26

REFERENCE:

Tech Spec 3.3.2.1 Cond A NEW Tech Spec 3.1.5 MODIFIED BANK DIFF 1; M 03-1-01-2 sect 2.15 RO SRO BOTH CFR 41.6/43.6 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 40 The plant is operating at rated conditions.

The Recirc System FCVs are 70% open when a Recirc Flow Control Valve Runback occurs.

Hydraulic Power Unit B Subloop 2 pump discharge pressure drops to 1600 psig causing FCV B HYD EQUIP REDUN SUBLP INOP annunciator to illuminate as valve movement begins.

All equipment operates as designed.

Which one of the following statements describes the response of the Recirculation System?

A. Both Recirc Flow Control Valves will fail to stroke, causing an automatic downshift of Recirc pumps to slow speed.

B. Recirc A Flow Control Valve will stroke to >>20% valve position.

Recirc B Flow Control Valve will stroke to >>20% valve position.

C. Recirc A Flow Control Valve will stroke to >>20% valve position.

Recirc B Flow Control Valve will stop in mid stroke.

D. Recirc Pump A Flow Control Valve will stroke to >>20% valve position.

Recirc Pump B Flow Control Valve will drift to >>20% valve position using residual HPU pressure.

QUESTION NRC RECORD # WRI 840 ANSWER: B. SYSTEM # B33 K/A 202002 K3.06: 3.7/3.7 LP# GLP-OPS-B3300 A1.08: 3.4/3.4 OBJ. 22; 23.2; 24 SRO TIER 2 GROUP 1 / RO TIER GROUP

REFERENCE:

04-1-02-H13-P680 NEW 3A-C8; 3A-C7; 3A-B8; 3A-D1 MODIFIED BANK DIFF 2; CA 4A1-C4 RO SRO BOTH CFR 41.6 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 41 A LOCA has occurred.

RHR A is injecting to the Reactor.

Standby Service Water A Pump trips on motor overcurrent.

Which one of the following statements describes the response of RHR A?

A. RHR A Pump will trip due to a loss of cooling water flow to the RHR pump seal cooler.

B. RHR A Pump will continue to run indefinitely since the LOCA signal has all of the trips bypassed.

C. RHR A Pump will continue to inject until the motor overheats or the shaft seals overheat and seize the pump.

D. RHR A Pump will continue to run but E12-F048A, RHR A BYP VLV will maintain itself open due to the loss of cooling water to the RHR A Heat Exchangers.

QUESTION NRC RECORD # WRI 841 ANSWER: C. SYSTEM # E12; P41 K/A 203000 K6.10: 3.0/3.1 LP# GLP-OPS-E1200 OBJ. 8.1; 13.3 SRO TIER 2 GROUP 1 / RO TIER GROUP

REFERENCE:

04-1-02-H13-P601 20A-A4 NEW 04-1-02-H13-P870 1A-A1 MODIFIED BANK DIFF 1; M E-1181-043 RO SRO BOTH CFR 41.8 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 42 The plant is in a refueling outage.

LPCS is in its normal standby lineup.

E21-F005, LPCS INJ VLV, has been repacked and requires stroke timing.

The CRO places the LPCS MOV Test Switch to TEST and the LPCS INJ VLV handswitch to the OPEN position.

The valve fails to stroke open.

NO abnormal alarms are noticed.

Which one of the following prevented E21-F005 LPCS INJ VLV from stroking open?

P&ID M-1087 is provided.

A. Pressure indicating switch E21-PSH-N655 has failed low B. Pressure indicating switch E21-PIS-N650 has failed high.

C. E21-F007, LPCS Injection HDR Manual Isolation, is closed.

D. Breaker 52-151114, LPCS INJ SHUTOFF VLV, thermal device has tripped.

QUESTION NRC RECORD # WRI 842 ANSWER: B. SYSTEM # E21 K/A 209001 K4.10: 2.8/2.9 K4.01: 3.2/3.4 LP# GLP-OPS-E2100 OBJ. 8.3 SRO TIER 2 GROUP 1 / RO TIER GROUP

REFERENCE:

04-1-01-E21-1 NEW M-1087 MODIFIED BANK E-1182-002 E-1182-026 DIFF 1; M RO SRO BOTH CFR41.7 REFERENCE MATERIAL REQUIRED: M-1087

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 43 An ATWS has occurred from rated conditions and actions are being directed from EP-2A.

Standby Liquid Control (SLC) was ordered started.

The Control Room Operator takes the handswitch to START for both SLC pumps.

The circuit breakers for SLC Pumps A & B magnetic overloads tripped immediately when the pump motors started.

Which one of the following identifies the present valve positions for SLC and RWCU?

C41-F001A/B C41-F004A/B G33-F001 G33-F004 G33-F251 STORAGE SLC SQUIB PMP SUCT PMP SUCT RWCU SPLY TANK OUTLET DRWL INBD CTMT OTBD TO RWCU HXS ISOL ISOL A. Open Open/Fired Closed Closed Closed B. Open Closed Closed Open Open C. Closed Open/Fired Open Open Closed D. Closed Closed Closed Closed Closed QUESTION NRC RECORD # WRI 843 ANSWER: A. SYSTEM # C41 K/A 211000 A1.04: 3.6/3.7 A2.06: 3.1/3.3 LP# GLP-OPS-C4100 A2.07: 2.9/3.2 OBJ. 9.1; 9.3; 10.1; SRO TIER 2 GROUP 1 / RO TIER GROUP 10.3; 10.4; 11

REFERENCE:

04-1-01-C41-1 sect 5.3 NEW E-1169-001; 005; 006; 012; 014 MODIFIED BANK DIFF 1; M E-1160-014; 015; 051 E-1203-005 RO SRO BOTH CFR41.6/41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 44 The plant is operating at 100 % power in a normal alignment.

RPS A Motor Generator has tripped.

Which one of the following identifies the status of the Main Steam Isolation Valves (MSIVs)?

ASSUME NO OPERATOR ACTIONS.

A. All MSIVs are isolated, causing a reactor scram.

B. The Inboard MSIVs are isolated and the Outboard MSIVs are open.

C. The Inboard MSIVs are open and the Outboard MSIVs are isolated.

D. All MSIVs are open with only one of the solenoids energized on each valve.

QUESTION NRC RECORD # WRI 844 ANSWER: D. SYSTEM # C71; B21 K/A 212000 K3.12: 3.2/3.3 LP# GLP-OPS-C7100 OBJ. 6.1; 23 LP# GLP-OPS-B1300 OBJ. 11.1 SRO TIER 2 GROUP 1 / RO TIER GROUP

REFERENCE:

05-1-02-III-2 sect 5.2.2; 5.1.2 NEW MODIFIED BANK DIFF 1; M RO SRO BOTH CFR41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 45 RF13 is in progress.

Control Rod Drive Mechanism change out is complete.

I & C has requested the Source Range Detectors be stroked in and out of the core for an I&C work order.

Which one of the following describes action that must be performed prior to moving the Source Range detectors?

A. An SRO with NO other concurrent assigned duties must be in the Control Room to supervise the activity.

B. Shorting Links must be verified installed in the Reactor Protection circuitry to prevent inadvertent scram signals from being received.

C. Personnel must be dispatched to the Drywell to ensure SRM cables and the under vessel service platform are placed to allow free travel.

D. Ensure NO other activities are being performed from the Refuel Floor during the activity to prevent multiple reactivity additions to occur at the same time.

QUESTION NRC RECORD # WRI 845 ANSWER: C. SYSTEM # C51; B13 K/A 215004 K1.06: 2.8/2.8 LP# GLP-OPS-C5101 OBJ. 11 SRO TIER 2 GROUP 1 / RO TIER GROUP

REFERENCE:

04-1-01-C51-1 NEW sect 3.5; 4.2 Caution MODIFIED BANK DIFF 1; M 01-S-06-2 sect 5.8; 6.7.1 RO SRO BOTH CFR41.5/41.6 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 46 The plant is in mode 2 at 12 % of rated power.

APRM H is bypassed due to failed power supply.

The following is the present status of the APRMs versus LPRM inputs and indicated power:

APRM A B C D E F G H LPRM LVL D 5 5 5 2 3 2 4 5 LPRM LVL C 5 4 3 5 4 4 3 4 LPRM LVL B 3 2 2 4 3 3 3 3 LPRM LVL A 2 4 4 4 4 4 5 3 INDICATED POWER 12% 13% 14% 10% 10% 11% 10% 14%

byp LPRM 26-27B power supply to the detector has failed.

With the present conditions, which one of the following describes current plant status?

The LPRM vs APRM assignments table is attached.

A. RPS and RCIS remain in a normal configuration.

B. Half Scram RPS A and Rod Withdrawal Block on RCIS.

C. Half Scram RPS B and Rod Withdrawal Block on RCIS D. A full actuation of RPS should have occurred, RCIS will have a Rod Withdrawal Block.

QUESTION NRC RECORD # WRI 846 ANSWER: C. SYSTEM # C51 K/A 215005 K2.01: 2.4/2.6 K1.01: 4.0/4.0 K4.02: 3.7/3.7 LP# GG-1-LP-OP-C5103 K5.06: 2.5/2.6 OBJ. 6; 7a; 9c; 10; 12; 18 SRO TIER 2 GROUP 1 / RO TIER GROUP

REFERENCE:

04-1-01-C51-1 sect 5.2 Caution NEW Tech Spec Bases 3.3.1.1 MODIFIED BANK DIFF 3; CA 17-S-02-40 Att V & VI; 5.1; 6.1; 6.3; 7.1 RO SRO BOTH CFR41.6/41.7 REFERENCE MATERIAL REQUIRED: 17-S-02-40 Att V & VI

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 47 The plant is operating at 100% power.

The reference leg for Condensing chamber B21-D004A has ruptured in Containment.

Which one of the following describes the actions that will occur due to this leak and the actions that need to be taken?

Partial P&ID M-1077B is provided.

A. Level indication will go low resulting in a half scram signal. The trip function must be declared INOP with the required actions already complete.

B. Level indication will go high resulting in a half scram signal and half isolation signal of RCIC. The trip functions must be declared INOP with the required actions already complete.

C. Level indication will go low resulting in a half scram signal, and a RCIC, RHR A, and LPCS initiation and half building isolation signal to the outboard valves. The trip functions must be declared INOP with the required actions already complete.

Required systems must be bypassed and restored.

D. Level indication will go high resulting in a half scram signal and half building isolation signal to the outboard valves. The trip functions must be declared INOP with the required actions already complete. Required systems must be bypassed and restored.

QUESTION NRC RECORD # WRI 847 ANSWER: B. SYSTEM # B21-1 K/A 216000 A2.03: 3.0/3.1 LP# GLP-OPS-B2101 K1.22: 3.6/3.8 OBJ. 4.2; 5; 7; 8; SRO TIER 2 GROUP 1 / RO TIER GROUP 12.2; 22

REFERENCE:

M-1077B NEW 17-S-06-5 MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR41.5/41.7 REFERENCE MATERIAL REQUIRED: Partial M-1077B

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 48 The plant is operating at 100% power.

RCIC has been started for a quarterly surveillance.

The operator at RCIC has informed the control room that there is a steam leak on E51-F045, RCIC Steam Supply.

The Control Room Operator depresses the RCIC Manual Isolation pushbuttons on H13-P601.

Which one of the following describes the results of the Control Room Operators actions?

A. RCIC will continue to run normally.

B. RCIC will isolate the Group 4 Outboard Isolation valves.

C. RCIC will completely isolate Group 4 Isolation valves and E51-F045 will remain open.

D. RCIC will completely isolate Group 4 Isolation valves and close E51-F045.

QUESTION NRC RECORD # WRI 848 ANSWER: A. SYSTEM # E51 K/A 217000 A4.04: 3.6/3.6 LP# GLP-OPS-E5100 A2.03: 3.4/3.3 OBJ. 11 SRO TIER 2 GROUP 1 / RO TIER GROUP

REFERENCE:

05-1-02-III-5 Group 4 NEW 17-S-06-5 Att II Group 4 MODIFIED BANK DIFF 1; M 04-1-01-E51-1 sect 3.11 RO SRO BOTH CFR41.5/41.7/41.10/

REFERENCE MATERIAL REQUIRED: None 43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 49 The plant transient resulted in the following plant conditions:

Reactor power 55%

Reactor water level + 20 inches narrow range and stable Reactor pressure 1045 psig and lowering Drywell pressure 0.93 psig and stable Drywell temperature 115 °F and stable Containment temperature 85 °F and stable Suppression Pool temperature 91 °F Suppression Pool level 18.1 ft Main Steam Tunnel temperature 155 °F and stable All other plant conditions are normal.

Which one of the following identifies the Emergency Procedures that should be entered?

A. EP-2 only B. EP-3 only C. EP-4 only D. EP-2 & 3 QUESTION NRC RECORD # WRI 849 ANSWER: B. SYSTEM # M41 K/A 223001 2.4.2: 3.9/4.1 LP# GG-1-LP-RO-EP03 OBJ. 5 SRO TIER 2 GROUP 1 / RO TIER GROUP LP# GG-1-LP-RO-EP02 OBJ. 10 LP# GG-1-LP-RO-EP04 OBJ. 6

REFERENCE:

05-S-01-EP-2 sect 2.1 NEW 05-S-01-EP-3 sect 2.1 MODIFIED BANK DIFF 1; M 05-S-01-EP-4 sect 2.1 Table 3 01-S-06-2 sect 6.3.6 RO SRO BOTH CFR41.9/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 50 The plant is operating at 100% power.

Radiography is in progress on 166 ft elevation inside Containment.

The following conditions exist:

Fuel Handling Area Exhaust Radiation Monitors 2.7 mR/hr Fuel Pool Sweep Exhaust Radiation Monitors 27 mR/hr Main Steam Line Radiation Monitors 1500mR/hr Containment Vent Exhaust Radiation Monitors 4.5 mR/hr Standby Gas Treatment Radiation Monitors 1.5 mR/hr All channels of the Radiation Monitors are reading the same.

Which one of the following describes the expected response of the plant?

FHA Vent FPS Vent CTMT Vent SBGT MSIVs A. Isolated Isolated Isolated Operating Isolated B. Operating Operating Operating Standby Open C. Isolated Isolated Isolated Operating Open D. Operating Operating Isolated Standby Open QUESTION NRC RECORD # WRI 850 ANSWER: D. SYSTEM # M41 K/A 223002 K6.03: 2.9/3.1 LP# GLP-OPS-M4100 272000 K1.09: 3.6/3.8 OBJ. 11.1; 12.1; 13.6, 7, 8 SRO TIER 2 GROUP 1 / RO TIER GROUP LP# GLP-OPS-D1721 OBJ. 15

REFERENCE:

05-1-02-III-5 Grp 7 & AB Vent NEW 04-1-02-H13-P601 MODIFIED BANK DIFF 1; M 18A-D5 & D6 04-1-01-T48-1 sect 5.2.1a RO SRO BOTH CFR41.7/41.9/

REFERENCE MATERIAL REQUIRED: None 41.11/43.4

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 51 A LOCA occurred 12 minutes ago actuating LSS.

Drywell conditions have caused all RPV level indications to fail upscale.

Drywell pressure is 5.6 psig.

Containment pressure is 8.2 psig.

Containment temperature is 120°F.

Which one of the following describes the response of Containment Spray?

A. Containment Spray should have automatically initiated.

B. Containment Spray is in standby and should NOT be manually initiated due to plant conditions being UNSAFE for initiation.

C. Containment Spray is in standby due to a lack of appropriate signals for automatic initiation but may be manually initiated.

D. Containment Spray is in standby due to a lack of the appropriate signals for automatic initiation and will NOT manually initiate due to the lack of appropriate signals.

QUESTION NRC RECORD # WRI 851 ANSWER: A. SYSTEM # E12; B21 K/A 226001 K6.08: 2.7/2.8 LP# GLP-OPS-E1200 A2.10: 3.0/3.1 OBJ. 9.1; 10.1; 10.3; 10.4; SRO TIER 2 GROUP 1 / RO TIER GROUP 13.8; 23 LP# GG-1-LP-RO-EP03 OBJ. 3; 6

REFERENCE:

05-S-01-EP-3 steps3, 6, 8 NEW 04-1-02-H13-P601 MODIFIED BANK DIFF 1; M 20A-A5; A6; B6 17-S-06-5 Att II 3.3.6.3-1 RO SRO BOTH CFR41.7/41.8/41.10/

REFERENCE MATERIAL REQUIRED: None 43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 52 The plant is operating at 100% power.

ADS/SRV B21-F041K opened at time 1000.

The following conditions exist:

H13-P601 section 19B Red light OFF 19C Red light ON H13-P628 Clear lights ON Red light ON H13-P631 Both sets of Clear lights ON Green light ON Annunciator H13-P601 18A-G2 ADS/SRV LEAK is in alarm.

Annunciator H13-P601 19A-A5 SRV/ADS VLV OPEN/DISCH LINE PRESS HI is in alarm.

Current time is 1003.

Given the above indications, which one of the following describes the status of B21-F041K?

A. The valve is closed.

B. Open via relief valve logic.

C. Open via ADS logic.

D. Open via safety actuation.

QUESTION NRC RECORD # WRI 852 ANSWER: A. SYSTEM # B21 K/A 239002 A3.08: 3.6/3.6 A3.03: 3.6/3.6 LP# GLP-OPS-E2202 A1.01: 3.3/3.4 OBJ. 18 SRO TIER 2 GROUP 1 / RO TIER GROUP

REFERENCE:

E-1161-008; 009; 011; 012 NEW 013; 016 MODIFIED BANK DIFF 2; CA 04-1-02-H13-P601 18A-D5 & D6 RO SRO BOTH CFR41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 53 The plant is operating at 20% power.

The Turbine IPC system gradually lowered its control to 800 psig.

Which one of the following describes the response of the plant?

ASSUME NO OPERATOR ACTIONS.

A. Turbine Control Valves will open to lower pressure followed by a MSIV closure and reactor scram, the Main Generator will trip on reverse power which will trip the Main Stop and Control Valves.

B. Turbine Control Valves will remain at present positions with the Main Steam Bypass Control Valves opening to lower pressure followed by an MSIV closure and reactor scram, the Main Generator will trip on reverse power which will trip the Main Stop and Control Valves.

C. Turbine Control Valves will close to lower pressure followed by a Main Turbine and Generator trip on Load reject followed by a reactor scram on Turbine Stop and Control valve closure, the MSIVs will close on low pressure.

D. Main Steam Bypass Control Valves will remain open to lower pressure followed by a MSIV closure and reactor scram, the Main Generator will trip on reverse power which will trip the Main Stop and Control Valves.

QUESTION NRC RECORD # WRI 853 ANSWER: A. SYSTEM # N32-2 K/A 241000 K4.06: 3.6/3.7 LP# GLP-OPS-N3202 OBJ. 4.0 SRO TIER 2 GROUP 1 / RO TIER GROUP

REFERENCE:

GGNS Simulator response NEW MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 54 The plant is operating at 100% power.

Narrow Range Level A indicates +36.2 inches Narrow Range Level B indicates +36.5 inches Narrow Range Level C indicates +36.1 inches Upset Range Level indicates +38.1 inches Level Transmitter C34-LT-N004C diaphragm ruptures.

Which one of the following describes the response of the Digital Feedwater Control System (DFCS)?

ASSUME NO OPERATOR ACTION.

A. Level indication on Narrow Range C will fail downscale and DFCS will shift from three element to single element and replace it with Upset Range.

B. Level indication on Narrow Range C will fail downscale and DFCS will automatically remove C from the level calculations and replace it with Upset Range.

C. Level indication on Narrow Range C will fail upscale and DFCS will automatically shift the controlling level signal to Narrow Range A from C.

D. Level indication on Narrow Range C will fail upscale and DFCS will automatically remove C from the level calculations and replace it with Upset Range.

QUESTION NRC RECORD # WRI 854 ANSWER: D. SYSTEM # C34; B21 K/A 259002 A4.02: 3.7/3.6 LP# GLP-OPS-C3400 OBJ. 3.4; 3.6; 23 SRO TIER 2 GROUP 1 / RO TIER GROUP LP# GLP-OPS-B2101 OBJ. 12.3

REFERENCE:

M-1077B NEW 04-1-02-H13-P680 2A-C9 MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR41.4/41.5/41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 55 Both trains of Standby Gas Treatment (SBGT) System were started for a surveillance using the Manual Initiation pushbuttons.

SBGT A was placed in standby per the SOI.

A valid Fuel Handling Area exhaust high-high radiation signal was received.

Which one of the following describes the response of the Standby Gas Treatment System and required operator response?

A. SBGT A will automatically re-initiate and SBGT B will remain operating, requiring NO additional operator action.

B. SBGT A will remain in standby and requires manual operator initiation to resume operation, SBGT B will remain operating.

C. SBGT A will remain in standby, SBGT B will perform all required actions as long as Enclosure Building pressures remain satisfactory NO further operator action is required.

D. SBGT A Enclosure Building and Exhaust fans will automatically restart without repositioning of building dampers and SBGT B will remain operating controlling building pressure. The operator must follow by depressing manual initiation pushbuttons to get the dampers operating.

DELETE QUESTION ANSWERS A, B, C CORRECT.

QUESTION NRC RECORD # WRI 855 ANSWER: B. SYSTEM # T48 K/A 261000 2.4.10: 3.0/3.1 LP# GLP-OPS-T4800 K4.01: 3.7/3.8 OBJ. 8.6; 8.7; 8.8; 8.9; 18 SRO TIER 2 GROUP 1 / RO TIER GROUP LP# GG-1-LP-RO-EP04 OBJ. 6

REFERENCE:

04-1-01-T48-1 sect 5.2 NEW 04-1-02-H13-P870 MODIFIED BANK DIFF 1; M 2A-D2 & D3 04-1-02-H13-P601 RO SRO BOTH CFR41.7/41.10/41.11/

19A-B9 & C9 43.4 05-S-01-EP-4 steps 3 - 6 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 56 The plant is operating at 100% power.

Bus 11HD experiences a bus over-current lockout.

Which one of the following identifies the final breaker lineup of the Reactor Recirculation Pumps?

RECIRC PUMP A CB1A CB2A CB3A CB4A CB5A A. OPEN OPEN CLOSED CLOSED OPEN B. OPEN OPEN CLOSED CLOSED CLOSED C. OPEN OPEN CLOSED OPEN OPEN D. CLOSED CLOSED CLOSED CLOSED OPEN QUESTION NRC RECORD # WRI 856 ANSWER: A. SYSTEM # R21; B33 K/A 262001 K3.01: 3.5/3.7 202001 K1.08: 3.1/3.2 LP# GLP-OPS-B3300 K6.03: 2.9/3.0 OBJ. 11.2; 12.1; 15.3; 19; SRO TIER 2 GROUP 1 / RO TIER GROUP 36.4; 47; 50 LP# GLP-OPS-R2700 OBJ. 6.3; 8; 24

REFERENCE:

E-1163-001; 003; 005; 008; 032 NEW 054; 058 MODIFIED BANK DIFF 2;CA 04-S-02-H13-P807 1A-C3 RO SRO BOTH CFR41.4/41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 57 The Division I Diesel Generator is being operated locally for Systems Engineering.

The STBY/EMER switch on H22-P113 is in the EMER position.

A rupture of a fuel oil pipe from the Division 1 Diesel Generator Day Tank causes a fire that required a complete evacuation of the room.

Which one of the following describes the required action(s) to shutdown the diesel generator?

A. Depress the REMOTE MANUAL DG 11 STOP pushbutton on 1H13-P864.

B. Depress the DG 11 EMERG SHUTDOWN pushbutton on 1H22-P113.

C. Open breakers 72-11A48 and 72-11A50, STBY DSL GEN 11 TRAIN A and B START/STOP CONTROL.

D. Open breaker 72-11A57, STBY DSL GEN 11 RLY PNL 1H22-P113 FIELD FLASHING.

QUESTION NRC RECORD # WRI 857 ANSWER: D. SYSTEM # P75 K/A 264000 K4.01: 3.5/3.7 K4.07: 3.3/3.4 K6.09: 3.3/3.5 LP# GLP-OPS-P7500 OBJ. 13, 14, 16, 18 SRO TIER 2 GROUP 1 / RO TIER GROUP

REFERENCE:

04-1-01-P75-1 sect 3.22 NEW E-1110-012, 013, 018, 019, 020, MODIFIED BANK 021, 035 DIFF 2; CA RO SRO BOTH CFR41.8/43.2 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 58 The plant is in Mode 1.

Which one of the following describes the normal alignment for ventilation systems in the Auxiliary Building to maintain proper building differential pressure?

AB General FHA FHA FPS Supply FPS T42-F021 Area Fan Supply Exhaust Fans Exhaust Pressure Coil Units Fans Fans Fans Control Valve A. Fans cycling 1 Fan 2 Fans 1 Fan 2 Fans Fully open on running running running running temperature B. Fans 2 Fans 2 Fans 2 Fans 2 Fans Full open running running running running running C. Fans 1 Fan 1 Fan 1 Fan 1 Fan Modulating running running running running running D. Fans cycling 1 Fan 1 Fan NO Fan 1 Fan Modulating on running running running running temperature QUESTION NRC RECORD # WRI 858 ANSWER: C. SYSTEM # T41; T42 K/A 290001 A4.03: 2.6/2.7 LP# GLP-OPS-T4100 OBJ. 3c SRO TIER 2 GROUP 1 / RO TIER GROUP LP# GLP-OPS-T4200 OBJ. 3a

REFERENCE:

04-1-01-T41-1 sect 4.1 NEW 04-1-01-T42-1 sect 3.1; 4.1; 4.3 MODIFIED BANK DIFF 1; M RO SRO BOTH CFR41.10; 43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 59 A station blackout has occurred.

It has been estimated offsite power will return in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Reactor water level is below the top of active fuel.

The Shift Manager has decided to cross tie Division III Diesel Generator to bus 16AB.

Which one of the following describes the return of loads to bus 16AB?

A. Division II Load Shedding and Sequencing panel will automatically sequence loads limiting load on Division III Diesel Generator.

B. Division II Load Shedding and Sequencing panel has initiated a failure signal locking out the panel requiring manual starting of loads.

C. Division II loads will only start as required for their LOCA signals with non-essential loads remaining de-energized.

D. Division II Load Shedding and Sequencing panel is manually turned off requiring loads to be manually started in a prescribed order.

QUESTION NRC RECORD # WRI 859 ANSWER: D. SYSTEM # R21 K/A 262001 A1.02: 3.1/3.4 LP# GG-1-LP-OP-R2100 OBJ. 18; 19; 25a; 37 SRO TIER 2 GROUP 1 / RO TIER GROUP LP# GLP-OPS-ONEP OBJ. 12; 13

REFERENCE:

05-1-02-I-4 sect 3.2.9 NEW MODIFIED BANK DIFF 1; M RO SRO BOTH CFR41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 60 The following annunciators have been received in the Control Room concerning the Control Rod Drive Hydraulic System:

CONT ROD DRIFT H13-P680-4A2-E4 CRD CLG WTR TO RX DP HI H13-P680-4A1-A6 All other annunciators are clear concerning CRD.

Which one of the following describes the status of the Control Rod Drive (CRD) Hydraulic System?

A. The CRD Pressure Control Valve C11-F003 is improperly positioned too far open.

B. CRD Hydraulic System is operating in response to a reactor scram.

C. The in-service CRD Flow Control Valve C11-F002 has failed closed.

D. The CRD Stabilizing Valves have failed closed.

QUESTION NRC RECORD # WRI 860 ANSWER: A. SYSTEM # C11-1A K/A 201001 A3.10: 3.0/2.9 LP# GLP-OPS-C111A OBJ. 7.3; 8.1; 8.2; 8.3; 9; SRO TIER 2 GROUP 2 / RO TIER GROUP 10.3; 13.1; 23

REFERENCE:

04-1-02-H13-P680 NEW 4A2-E4; 4A1-A6; 4A1-A7 MODIFIED BANK DIFF 1; M 04-1-02-H13-P601 22A-A3 04-1-01-C11-1 sect 3.10 RO SRO BOTH CFR41.5/41.6 M-1081B REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 61 The plant is starting up following a refueling outage.

Recirculation Pump A is running in fast speed.

Activities are in progress to shift Recirculation Pump B to fast speed.

As the operator depresses the START pushbutton on TRANS TO LFMG/START handswitch on 1H13-P680 for Recirculation Pump B, annunciator H13-P680-3A-E8, LFMG B OVERLD/TRIP alarms.

Which one of the following describes the affect this alarm will have on Recirculation Pump B?

A. Circuit breaker CB-1B will remain closed, CB-2B will trip and CB-5B will open.

Recirculation Pump B will coast down.

B. Circuit breaker CB-1B and CB-2B will trip and circuit breaker CB-5B will remain open. Recirculation Pump B will coast down.

C. Circuit breaker CB-1B will remain closed, CB-2B will trip, and CB-5B will close.

Recirculation Pump B will accelerate to rated speed.

D. Circuit breaker CB-1B and CB-2B will remain closed and circuit breaker CB-5B will remain open. Recirculation Pump B will remain in slow speed.

QUESTION NRC RECORD # WRI 861 ANSWER: B. SYSTEM # B33 K/A 202001 K6.06: 3.1/3.1 K5.10: 2.8/2.8 K4.17: 3.3/3.5 LP# GLP-OPS-B3300 OBJ. 11.3 SRO TIER 2 GROUP 2 / RO TIER GROUP

REFERENCE:

04-1-02-1G13-P680-3A-E8 NEW MODIFIED BANK DIFF 2:CA RO SRO BOTH CFR41.4 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 62 The plant is operating at 100% power.

All Containment Cooling is out of service.

A leak in the RWCU A Filter Demineralizer Room has caused room temperature to rise to 145°F.

Control Room annunciator RWCU DMIN RM 1 TEMP HI on H13-P680 is in alarm.

Which one of the following describes the affect this will have on the Reactor Water Cleanup (RWCU) System?

A. The RWCU system will continue to operate normally.

B. RWCU A Filter Demineralizer will isolate.

C. Both RWCU A and B will isolate.

D. RWCU A will isolate.

QUESTION NRC RECORD # WRI 862 ANSWER: A. SYSTEM # G33 K/A 204000 A2.09: 2.8/2.8 K1.15: 3.1/3.2 LP# GLP-OPS-G3336 OBJ. 8, 18 SRO TIER 2 GROUP 2 / RO TIER GROUP

REFERENCE:

04-1-02-1G13-P680-11A-B1 NEW MODIFIED BANK DIFF 1:M RO SRO BOTH CFR41.4 REFERENCE MATERIAL REQUIRED:

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 63 The plant is in Mode 4 at 170°F.

RHR A was in shutdown cooling when E12-F008 inadvertently isolated.

Which one of the following would be an indication of a change in plant Mode status?

A. RX TEMP LO annunciator has been received.

B. Source Range Nuclear Instrumentation count rate rising.

C. Steam is reported from B21-F001 and F002 Reactor Head vents.

D. Reactor Engineering reports available Shutdown Margin has been reduced.

QUESTION NRC RECORD # WRI 863 ANSWER: C. SYSTEM # E12 K/A 205000 K3.01: 3.3/3.3 LP# GLP-OPS-E1200 OBJ. 23 SRO TIER 2 GROUP 2 / RO TIER GROUP LP# GLP-OPS-TS001 OBJ. 4.16; 5

REFERENCE:

04-1-02-H13-P680 NEW 8A1-B7; 4A2-B2 MODIFIED BANK DIFF 1; M 05-1-02-III-1 sect 4.5 Tech Specs sect 1.1 Definitions RO SRO BOTH CFR41.2/41.3/41.4/41.5 Table 1.1-1 REFERENCE MATERIAL REQUIRED: None 43.2

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 64 The plant is operating at rated conditions.

RHR A was placed in Suppression Pool cooling following a RCIC run.

While RHR was operating, power was lost to bus 15AA for approximately 4 minutes.

Power has been restored from an offsite source.

Which one of the following describes the actions to be taken to re-establish Suppression Pool Cooling?

A. Start RHR A pump and verify E12-F024A Suppression Pool Test Return is open and E12-F048A RHR Heat Exchanger Bypass valve is closed.

B. With system as is, open E12-F073A and F074A RHR Heat Exchanger Vents for 30 seconds, then re-start RHR A pump in Suppression Pool Cooling.

C. Return RHR A system to LPCI standby then realign RHR A pump in Suppression Pool Cooling.

D. RHR A will automatically resume Suppression Pool Cooling when Load Sequencing occurs.

QUESTION NRC RECORD # WRI 864 ANSWER: C. SYSTEM # E12 K/A 219000 K5.01: 2.6/2.7 LP# GLP-OPS-E1200 OBJ. 14; 20; 23 SRO TIER 2 GROUP 2 / RO TIER GROUP LP# GLP-OPS-ONEP OBJ. 10

REFERENCE:

04-1-01-E12-1 sect. NEW 3.10; 4.1.1; 5.1; 5.2; 5.2.1 MODIFIED BANK DIFF 1; M 05-1-02-I-4 sect 3.1 Caution RO SRO BOTH CFR41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 65 The plant is in a refueling outage.

RHR A is operating in Shutdown Cooling mode.

Both loops of Spent Fuel Pool Cooling and Cleanup are operating.

Core alterations are in progress.

Due to a failed flow transmitter, E12-F064A, RHR A MIN FLO TO SUPP POOL opens.

Which one of the following would be a result of this event?

A. Containment and Spent Fuel Pool temperature rising.

B. Radiation levels on 208 el. Auxiliary Building and Containment rising.

C. Water clarity in the Containment and Spent Fuel Pool lowering.

D. Area temperature on 208 el. Auxiliary Building and Containment rising.

QUESTION NRC RECORD # WRI 865 ANSWER: B. SYSTEM # G41/E12 K/A 234000 K5.03: 2.9/3.4 GG-1-LP-RF-F1104 GLP-OPS-E1200 OBJ. 1b SRO TIER 2 GROUP 2 / RO TIER GROUP 21 23

REFERENCE:

04-1-01-E12-1 NEW sect 4.2.2e Caution MODIFIED BANK DIFF 1:M RO SRO BOTH CFR41.4/41.9/41.12/

REFERENCE MATERIAL REQUIRED: None 43.4/43.7

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 66 The river is at high water level when a barge strikes the Radial Well Switchgear house and causes all radial well pumps to trip.

Operators and Electricians estimate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore the switchgear at the river.

The Control Room has reduced power to 67 Mlbm/hr core flow.

The Turbine Building operator has reported TBCW temperature is 116°F and rising steadily.

Temperatures on all lube oil systems are rising rapidly.

Main Condenser vacuum is degrading and the plant was manually scrammed.

Which one of the following describes the actions to be taken for RPV level control?

A. Maintain reactor pressure stable and control RPV level with RCIC and secure the Condensate and Feedwater systems.

B. Emergency depressurize the reactor allowing low pressure ECCS to be used for RPV level control.

C. Maintain reactor pressure stable and alternate the Reactor Feed Pumps, Condensate Booster and Condensate pumps for level control.

D. Emergency depressurize the reactor to within the capabilities of the Condensate pumps and alternate operation of pumps while monitoring temperatures.

QUESTION NRC RECORD # WRI 866 ANSWER: B. SYSTEM # N21; N19; K/A 259001 K6.06: 2.7/2.7 P44; P43 K6.05: 2.7/2.7 LP# GLP-OPS-ONEP K1.10: 2.7/2.7 OBJ. 41; 44 SRO TIER 2 GROUP 2 / RO TIER GROUP LP# GLP-OPS-N1900 OBJ. 12; 22.12; 31 LP# GLP-OPS-N1900 OBJ. 12 LP# GLP-OPS-N2100 OBJ. 28.12; 39

REFERENCE:

05-1-02-V-2 sect 2.0 Note; 3.1.7 NEW 3.2.1 - 3.2.4; 3.2.8; 4.2 MODIFIED BANK DIFF 1; M 05-1-02-V-11 sect 2.1; 3.6 RO SRO BOTH CFR41.4/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 67 Annunciator H13-P807-3A-G3, STATIC INVRTR 1Y98 TROUBLE is in alarm.

Investigation reveals that the static inverter output frequency is oscillating.

The following indications are present on Inverter 1Y98:

INVERTER FAILURE light energized IN SYNC light de-energized INVERTER SUPPLYING LOAD energized ALT SOURCE SUPPLYING LOAD de-energized Which one of the following describes actions to be taken for this condition?

A. The static switch will transfer automatically when inverter output frequency lowers to the required setpoint.

B. Manually transfer Inverter 1Y98 to its alternate power source using the ALT SOURCE TO LOAD pushbutton.

C. Manually transfer Inverter 1Y98 to its alternate power source using the Manual Bypass Switch.

D. Open breaker 72-11K05, 1Y98 PRIMARY DC POWER to allow the inverter to automatically swap to its alternate power source.

QUESTION NRC RECORD # WRI 867 ANSWER: D. SYSTEM # L62 K/A 262002 A2.03: 2.4/2.6 A3.01: 2.8/3.1 K4.01: 3.1/3.4 GLP-OPS-L6200 OBJ. 4, 5, 10, 11 SRO TIER 2 GROUP 2 / RO TIER GROUP

REFERENCE:

04-1-01-L62-1 sect 3.2; 5.1 NEW 04-S-02-SH13-P807-3A-G3 MODIFIED BANK Vendor Manual DIFF 3;CA RO SRO BOTH CFR41.7/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 68 The plant is operating at 100% power.

The in-service Offgas train flow has steadily risen from 18 scfm to 50 scfm and remains stable.

A check of the Offgas system lineup reveals a normal valve lineup.

Which one of the following describes the possible cause of the elevated flow in Offgas?

A. Hydrogen Water Chemistry hydrogen injection has been secured resulting in more Oxygen passing out of the reactor with the steam.

B. Offgas loop seals have been blown resulting in leakage into the Offgas process from the drains.

C. A Hydrogen fire is ongoing in the Offgas Charcoal Adsorbers resulting in the release of gases in the Offgas system.

D. Leakage into the condenser has risen resulting in more flow of non-condensables into the Offgas system.

QUESTION NRC RECORD # WRI 868 ANSWER: D. SYSTEM # N64; N62 K/A 271000 A3.02: 2.9/2.8 A2.01: 3.1/3.3 LP# GLP-OPS-ONEP A2.10: 3.1/3.3 OBJ. 39 SRO TIER 2 GROUP 2 / RO TIER GROUP LP# GLP-OPS-N6465 OBJ. 11.2; 11.3; 11.4; 25 LP# GLP-OPS-N6200 OBJ. 24; 26

REFERENCE:

05-1-02-V-8 sect 3.7; 4.7 NEW 05-1-02-V-13 sect 3.1; 4.3 MODIFIED BANK DIFF 2; CA 04-1-02-H13-P845 1A-A4; B5; B4; RO SRO BOTH CFR41.4/41.10/41.13/

1B-C5; D5; D6 43.4/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 69 ESF Inverter 1Y88 Manual Bypass switch has failed resulting in a loss of power to UPS power panel 1Y84.

FH AREA EXH DIV 2,3 RAD HI-HI/INOP is in alarm.

FP EXH DIV 2,3 RAD HI-HI/INOP is in alarm.

Standby Gas Treatment System B has automatically started due to a loss of logic power.

A half scram on RPS B has occurred.

Which one of the following describes the affect on the Fuel Handling Area (FHA) Exhaust and Fuel Pool Sweep (FPS) Radiation Monitoring Systems?

A. Only FHA and FPS channel B has tripped due to an INOP signal.

B. Only FHA and FPS channel C has tripped due to an INOP signal.

C. FHA and FPS channels B and C have tripped due to INOP signals.

D. FHA and FPS channels B and D have tripped due to INOP signals.

QUESTION NRC RECORD # WRI 869 ANSWER: A. SYSTEM # D17; L62; K/A 272000 K2.05: 2.6/2.9 T42; T48 LP# GLP-OPS-D1721 OBJ. 9; 17.2; 25; 26; 28 SRO TIER 2 GROUP 2 / RO TIER GROUP LP# GLP-OPS-T4801 OBJ. 9.4

REFERENCE:

E-1026 NEW E-1160-046; 049 MODIFIED BANK DIFF 2; CA E-1177-003; 015 E-1257-01 04-1-02-H13-P601 19A-C9; C10; B9; B10 RO SRO BOTH CFR41.10/41.11/43.4 REFERENCE MATERIAL REQUIRED: None 43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 70 The plant is operating at 100 % power.

Unit I Instrument Air Compressor is tagged out of service for impeller replacement.

Service Air Compressor A is operating with Service Air Compressor B in standby.

Unit II Instrument Air Compressor trips on motor overload.

Instrument Air header pressure has dropped to 97 psig and lowering.

Which one of the following describes the operation of the Plant Air System in response to the current conditions?

A. Service Air will require manual cross-tying to the Instrument Air header upstream of the Plant Air Dryers.

B. Service Air will require manual cross-tying to the Instrument Air header downstream of the Plant Air Dryers.

C. Service Air will automatically cross tie to the Instrument Air header upstream of the Plant Air Dryers.

D. Service Air will automatically cross tie to the Instrument Air header downstream of the Plant Air Dryers.

QUESTION NRC RECORD # WRI 870 ANSWER: C. SYSTEM # P52; P53 K/A 300000 K1.02: 2.7/2.8 LP# GLP-OPS-P5300 OBJ. 3; 4 SRO TIER 2 GROUP 2 / RO TIER GROUP LP# GLP-OPS-P5200 OBJ. 3; 4

REFERENCE:

05-1-02-V-9 sect 3.2.1 NEW 04-1-02-H13-P870 MODIFIED BANK DIFF 1; M 7A-A3; B3; D3; F3 M-1067A; G RO SRO BOTH CFR41.4/41.10/43.5 M-1068D M-1126 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 71 Which one of the following describes the safety design basis for maintaining Control Room ambient air temperatures?

A. Ensure Control Room personnel remain comfortable during normal and accident conditions.

B. Ensure Control Room atmosphere remains at temperatures above that which would result in condensation in the Control Room.

C. Ensure optimal performance of the Control Room Standby Fresh Air charcoal adsorbers during accident conditions.

D. Ensure Control Room temperatures remain less than that required for equipment operability during normal and accident conditions.

QUESTION NRC RECORD # WRI 871 ANSWER: D. SYSTEM # Z51 K/A 290003 K5.03: 2.6/2.7 LP# GLP-OPS-Z5100 OBJ. 2 SRO TIER 2 GROUP 2 / RO TIER GROUP

REFERENCE:

FSAR 9.4.1.1.1f NEW Tech Spec Bases 3.7.4 MODIFIED BANK DIFF 1; M TRM 6.7.3 RO SRO BOTH CFR41.4 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 72 The plant is operating at 100% power.

Due to maintenance work, the Fuel Pool Cooling and Cleanup is lined up to supply only the Spent Fuel Pool.

Water level in the Upper Containment Pool was lowered approximately 1 foot to facilitate maintenance on temperature elements.

Maintenance work has been completed and Fuel Pool Cooling and Cleanup is being re-aligned to supply both the Spent Fuel Pool and the Upper Containment Pool.

Which one of the following is a potential consequence of opening the G41-F028, FPCC RTN VLV CTMT OTBD ISOL, too rapidly?

A. Possible overflow of the Fuel Pool Drain Tank due to the rapid return of water from the Upper Containment Pool.

B. Possible trip of the Fuel Pool Cooling and Cleanup pump(s) on low Fuel Pool Drain Tank level.

C. Possible overflow of the Spent Fuel Pool into the ventilation system.

D. Possible overflow of the Upper Containment Pools into the ventilation system.

QUESTION NRC RECORD # WRI 872 ANSWER: B. SYSTEM # L62 K/A 233000 A1.02: 2.9/3.1 K3.02: 3.1/3.2 GLP-OPS-G4100 OBJ. 23, 25 SRO TIER 2 GROUP 3 / RO TIER GROUP

REFERENCE:

04-1-01-G41-1 NEW MODIFIED BANK DIFF 1:M RO SRO BOTH CFR41.4/41.9 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 73 The plant is operating at rated pressure.

The Riley temperature switch E31-TS-N604C, MSL Temperature has failed upscale.

All other trip units are reading normal.

Which one of the following describes the status of the Main Steam Isolation Valves (MSIVs)?

A. The A MSIV solenoid on all eight MSIVs are de-energized with a half isolation signal.

B. The A MSIV solenoids on the Outboard MSIVs are de-energized with a half isolation signal.

C. The A MSIV solenoids on the Inboard MSIVs are de-energized with a half isolation signal.

D. Both MSIV solenoids on all eight MSIVs are energized requiring a second temperature signal to cause any isolation signals.

QUESTION NRC RECORD # WRI 873 ANSWER: A. SYSTEM # B21; E31; K/A 239001 K6.09: 3.9/4.1 M71 LP# GLP-OPS-B1300 OBJ. 10; 11.1 SRO TIER 2 GROUP 3 / RO TIER GROUP LP# GLP-OPS-M7101 OBJ. 7.1; 9; 26; 27

REFERENCE:

17-S-06-5 Att II 3.3.6.1-1 NEW MODIFIED BANK DIFF 1; M RO SRO BOTH CFR41.4/41.7/41.9 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 74 The plant was operating at 100% power.

A slight power rise has occurred.

The following indications are present in the Main Control Room:

CNDS PMP DISCH CNDCT HI NOT in alarm.

CNDS DMIN EFL CNDCT HI in alarm.

CNDS DMIN SYS TROUBLE in alarm.

RFP DISCH CNDCT HI in alarm.

FW TRBY HI/LO in alarm.

RWCU FLTR DMIN EFL CNDCT HI/LO NOT in alarm.

MSL RAD HI in alarm.

MSL A/MSL D RAD HI-HI/INOP in alarm.

MSL B/MSL C RAD HI-HI/INOP in alarm.

Which one of the following describes the cause of chemistry changes in the plant?

A. Condensate Pre-coat Filter isolation valve leaking by.

B. Condensate Deep Bed Demineralizer resin breakthrough.

C. Main Condenser tube break from Circulating Water System.

D. Reactor Water Cleanup Filter Demineralizer resin break through.

QUESTION NRC RECORD # WRI 874 ANSWER: B. SYSTEM # N19; N22 K/A 256000 A2.15: 2.8/3.1 LP# GLP-OPS-N1900 OBJ. 31 SRO TIER 2 GROUP 3 / RO TIER GROUP LP# GLP-OPS-N2200 OBJ. 3

REFERENCE:

05-1-02-V-12 sect 1.4; 4.0 NEW 04-1-02-H13-P680 1A-D2; D3 MODIFIED BANK DIFF 2; CA 2A-D1; D5; 4A1-D1; 11A-C3 04-1-02-H13-P601 18A-C4 RO SRO BOTH CFR41.4 19A-C4; D4 04-1-02-H22-P474 A4; C4; F8 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 75 The Drywell Floor Drain Sump A pump is tagged out of service.

The Drywell Floor Drain Sump B pump handswitch is selected for AUTO.

The handswitch on the control panel is in the ALTERNATE position.

The Floor Drain System Valve lineup is per SOI.

Which one of the following describes the operation of the Drywell Floor Drain Sump in the present configuration?

A. Sump Pump B will automatically start each time the sump High level switch is activated.

B. Sump Pump B will bypass the High level switch and automatically start on the High-High level switch.

C. Sump Pump B will automatically start when the leak detection timer times out on sump fill rate.

D. Sump Pump B will NOT cycle automatically on High or High-High level switches or any leak rate timers.

QUESTION NRC RECORD # WRI 875 ANSWER: D. SYSTEM # P45 K/A 268000 A4.01: 3.4/3.6 LP# GLP-OPS-P4500 OBJ. 7a; 7c; 11 SRO TIER 2 GROUP 3 / RO TIER GROUP

REFERENCE:

04-1-01-P45-2 sect 3.5 NEW E-1271-026; 027; 035 MODIFIED BANK DIFF 1; M RO SRO BOTH CFR41.13/43.4 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 76 The plant is operating at 100 % power following a return to power after a MSIV closure scram.

Multiple SRVs are weeping.

Which one of the following describes the limitations and monitoring of the Suppression Pool Temperature?

Tech Spec 3.6.2.1 and Surveillance 06-OP-1M24-V-001 are provided.

A. Suppression Pool Temperature is limited to 95ºF and must be monitored every sixty minutes when temperature is above 90ºF to verify average temperature is less than 105ºF. If Suppression Pool Temperature exceeds 95ºF, Suppression Pool Cooling must be placed in service.

B. Suppression Pool Temperature is limited to 95ºF and must be monitored every five minutes while testing is adding heat to the Suppression Pool. If Suppression Pool Temperature exceeds 100ºF Suppression Pool Cooling must be placed in service and reactor power reduced to less than 1%.

C. Suppression Pool Temperature is limited to 100ºF and must be monitored every five minutes while testing is adding heat to the Suppression Pool. If Suppression Pool Temperature exceeds 95ºF Suppression Pool Cooling must be placed in service. Testing must be secured if Suppression Pool Temperature exceeds 105ºF.

D. Suppression Pool Temperature is limited to 105ºF and must be monitored every sixty minutes while testing is in progress and Suppression Pool Temperature exceeds 90ºF. If Suppression Pool Temperature exceeds 105ºF Suppression Pool Cooling must be placed in service and testing secured.

QUESTION NRC RECORD # WRI 876 ANSWER: A. SYSTEM # M24 K/A 295013 AA2.01: 3.8/4.0 LP# GLP-OPS-M7101 AA1.02: 3.9/3.9 OBJ. 23 2.1.33: 3.4/4.0 LP# GG-1-LP-RO-EP03 2.4.4: 4.0/4.3 OBJ. 3; 5 LP# GLP-OPS-M4101 OBJ. 11 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

Tech Spec 3.6.2.1 NEW Tech Spec SR3.6.2.1.1 MODIFIED BANK DIFF 2; CA 05-S-01-EP-3 step 10 - 12 NRC 8/2002 wri605 06-OP-1M24-V-0001 sect 5.1.2a RO SRO BOTH CFR 41.10/43.2/43.5 REFERENCE MATERIAL REQUIRED: Tech Spec 3.6.2.1 &

06-OP-1M24-V-0001

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 77 The plant is in day 19 of RF13.

Fuel Pool Cooling and Cleanup pumps have tripped and are not expected back within the next 5 days.

Spent Fuel Pool temperature is 139°F and rising 1 degree every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Preparations are in progress for plant startup within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Which one of the following is the course of action that should be taken with regard to plant operations?

Procedure 05-1-02-III-1 and Tech Spec 6.7.4 are provided.

A. Proceed with plant startup and plant Mode change to MODE 2. Monitor Spent Fuel Pool temperatures and start additional area fan coil units on 208 ft elevation.

B. Install fire hoses on 208 ft elevation discharging into the Spent Fuel Pool and drain the Spent Fuel Pool to the Suppression Pool. Delay plant startup until Spent Fuel Pool temperatures are less than 140°F, then plant startup may proceed to MODE 2.

C. Align RHR A or B for Fuel Pool Cooling Assist. Delay plant startup until Fuel Pool Cooling can be restored to service or Spent Fuel Pool temperatures will remain less than 140°F without the assistance of RHR.

D. Proceed with plant startup and plant mode change to MODE 2. Enter LCO on Spent Fuel Pool temperature and take exception to 3.0.4. Align RHR A or B for Fuel Pool Cooling Assist to restore Spent Fuel Pool temperatures.

QUESTION NRC RECORD # WRI 877 ANSWER: C. SYSTEM # G41; E12 K/A 295023 AA1.02: 2.9/3.1 LP# GLP-OPS-G4146 OBJ. 18; 19; 25 LP# GLP-OPS-TS001 OBJ. 13; 17 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

05-1-02-III-1 sect 3.5.2 NEW TRM 6.7.4 A1 & A2; no 3.0.4 MODIFIED BANK DIFF 2; CA 04-1-01-G41-1 sect 3.5 04-1-01-E12-1 sect 6.1 NOTE 04-1-02-H13-P680 4A2-B7 RO SRO BOTH CFR 41.4/41.5/41.10 REFERENCE MATERIAL REQUIRED: 05-1-02-III-1 & T/S 6.7.4 43.5/43.7

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 78 Which one of the following identifies a condition that would require declaration of an ALERT?

10-S-01-1 is provided.

A. A scram occurred but 6 control rods remain withdrawn Reactor power is indicating stable on range 4 of the IRMs. All neutron monitoring detectors have been fully inserted into the core.

B. A scram should have occurred but the operator had to place the reactor mode switch to shutdown to activate RPS, once the mode switch was in SHUTDOWN all control rods inserted fully and reactor power has dropped to the source range.

C. A scram occurred with RPS actuation however 50 control rods failed to fully insert. Reactor power is 10% with Reactor water level being maintained at

-110 inches on the Fuel Zone indication.

D. A scram occurred when the Operator At-The-Controls placed the mode switch to SHUTDOWN but control rod 28-33 remains stuck fully withdrawn. Reactor power has dropped to the source range and is stable at 6000 counts.

QUESTION NRC RECORD # WRI 878 ANSWER: A. SYSTEM # K/A 295037 2.4.40: 2.3/4.0 Emergency Action Levels 2.4.41: 2.3/4.1 LP# GLP-EP-EPTS OBJ. 1 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

10-S-01-1 Att I 11.2 NEW MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: 10-S-01-1

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 79 A LOCA has occurred at GGNS.

Conditions have warranted the declaration of a General Emergency with a release in progress.

Dose projections on Highway 61 at the junction with Highway 18 in Port Gibson have been determined to be 100 mRem TEDE.

The wind is coming from 320 degrees at 15 mph.

Which one of the following identifies the protective action recommendations that should be made and the affected sectors?

10-S-01-1; 10-S-01-12 and the EPZ map are provided.

A. Evacuate 2 miles in all sectors, and 5 miles in sectors F, G, H. Shelter the remainder of the 10 mile emergency planning zone.

B. Evacuate 2 miles in all sectors, and 5 miles in sectors P, Q, R. Shelter the remainder of the 10 mile emergency planning zone.

C. Evacuate 2 miles in all sectors, and 10 miles in sectors F, G, H. Shelter the remainder of the 10 mile emergency planning zone.

D. Evacuate 2 miles in all sectors, and 10 miles in sectors P, Q, R. Shelter the remainder of the 10 mile emergency planning zone.

QUESTION NRC RECORD # WRI 879 ANSWER: A. SYSTEM # K/A 295038 EA1.02: 3.0/3.8 Protective Action Recom 2.4.41: 2.1/4.0 LP# GLP-EP-EPTS OBJ. 2 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

10-S-01-1 sect 6.1.4k NEW 10-S-01-12 sect 6.2.1 MODIFIED BANK DIFF 2; CA 5 mile EPZ map 140° is sectors F, G, H RO SRO BOTH CFR 41.10/41.12/43.5 REFERENCE MATERIAL REQUIRED: 10-S-01-1 & 10-S-01-12 43.4 5 mile EPZ Map

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 80 You are the Shift Manager.

Which one of the attached Batch Liquid Radwaste Discharge Permits should be approved to allow a liquid radwaste discharge?

01-S-08-11 and permits are provided.

A. Batch Liquid Discharge Radwaste Permit 04-02-06-1 B. Batch Liquid Discharge Radwaste Permit 04-02-06-2 C. Batch Liquid Discharge Radwaste Permit 04-02-06-3 D. Batch Liquid Discharge Radwaste Permit 04-02-06-4 QUESTION NRC RECORD # WRI 880 ANSWER: D. SYSTEM # Admin Proc K/A 295017 AA1.01: 3.1 LP# GLP-OPS-PROC 2.3.3: 2.9 OBJ. 34.3 SRO TIER 1 GROUP 1 / RO TIER GROUP

REFERENCE:

01-S-08-11 Att I NEW MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR41.10/ 41.13/43.2/

REFERENCE MATERIAL REQUIRED: Discharge Permits & 43.4/43.5 01-S-08-11

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 81 RF13 Mode 5 core alterations are in progress.

RHR A is operating in Shutdown Cooling with ADHR as the alternate decay heat removal subsystem.

LPCS Pump motor is disassembled.

11DB1 DC bus is de-energized for bus repairs.

Engineering just reported HPCS Pump and Motor bearing oil has metal filings in the oil.

Which one of the following statements describes the actions to be taken for plant conditions?

Tech Spec 3.5.2 is provided.

A. Restore required ECCS system to operable status within four hours.

B. Immediately suspend Core Alterations and OPDRVs, restore required ECCS system to operable status within four hours.

C. Verify Reactor water level is 22 ft 8 inches above the reactor pressure vessel flange and the upper containment reactor cavity and transfer canal gates are removed.

D. Restore required ECCS system to operable status within four hours. Immediately suspend OPDRVs. Initiate action to restore one SBGT Train and Secondary Containment to Operable status and isolate Secondary Containment penetrations unable to be isolated.

QUESTION NRC RECORD # WRI 881 ANSWER: C. SYSTEM # E22 K/A 209002 2.1.10: 2.7/3.9 LP# GLP-OPS-E2201 2.2.22: 3.4/4.1 OBJ. 17 2.2.25: 2.5/3.7 LP# GLP-OPS-TS001 OBJ. 34 SRO TIER 2 GROUP 1 / RO TIER GROUP

REFERENCE:

Tech Spec 3.5.2 applicability NEW Tech Spec Bases 3.5.2 MODIFIED BANK DIFF 3; CA RO SRO BOTH CFR 41.7/41.8/43.1/43.2 REFERENCE MATERIAL REQUIRED: Tech Spec 3.5.2

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 82 Both ADS Booster Compressors are broken.

ADS Air Receiver pressure dropped to 110 psig.

The Control Room Supervisor dispatched operators to install nitrogen bottles per the Loss of Instrument Air ONEP and establish 165 psig pressure on the ADS Air Receivers.

Which one of the following statements describes the actions to be taken for plant conditions?

Tech Spec 3.0; SR3.0; and 3.5.1 are provided.

A. Immediately enter LCO 3.0.3 until reactor pressure is 150 psig or ADS Air Receivers and ADS Booster Compressors have been restored to an operable status.

B. Be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor pressure to 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> or restore ADS Air Receivers and ADS Booster Compressors to an operable status.

C. Monitor ADS Air Receiver Pressure once per 7 days and restore ADS Booster Compressors to an operable status within 31 days.

D. Restore ADS Air Receivers and ADS Booster Compressors to an operable status within 31 days.

QUESTION NRC RECORD # WRI 882 ANSWER: B. SYSTEM # E22-2 K/A 218000 2.2.23: 2.6/3.8 LP# GLP-OPS-E2202 2.2.22: 3.4/4.1 OBJ. 22 2.2.25: 2.5/3.7 LP# GLP-OPS-TS001 OBJ. 13 SRO TIER 2 GROUP 1 / RO TIER GROUP

REFERENCE:

Tech Spec 3.5.1 Cond G NEW SR 3.5.1.3 MODIFIED BANK DIFF 2; CA Tech Spec Bases 3.5.1 SR3.5.1.3 SR3.0.1 RO SRO BOTH CFR 41.7/41.8/43.1/43.2 REFERENCE MATERIAL REQUIRED: Tech Spec 3.5.1; LCO 3.0 & SR3.0

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 83 The crew has had the shift for eight hours and a replacement Shift Manager is arriving to relieve the On-Duty Shift Manager.

Which one of the following is NOT required for the On-coming Shift Manager to assume the duties?

A. Log the relief in the Shift Managers Logbook B. The plant should be a stable condition before beginning turnover.

C. Complete a walkdown of the Control Room and understand plant conditions D. Conduct a shift briefing to discuss plant status and scheduled evolutions/tasks.

QUESTION NRC RECORD # WRI 883 ANSWER: D. SYSTEM # Shift K/A Generic 2.1.3: 3.0/3.4 Turnover LP# GLP-OPS-PROC OBJ. 46 SRO TIER 3 GROUP / RO TIER GROUP

REFERENCE:

02-S-01-4 sect 6.1; 6.2; 6.3.1 NEW MODIFIED BANK DIFF 1; M NRC 6/2001 wri481 RO SRO BOTH CFR41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 84 The Outside Operator is performing switching in the 500KV switchyard per Switching Orders from the dispatcher.

The Operator notes the switching orders from the dispatcher are ordering an operation that contradicts the 500KV System Operating Instruction (SOI).

The System Engineer has researched the procedure and the switching orders requested by the dispatcher and determined the switching orders would NOT violate any licensing basis documents.

Entergy - Mississippi personnel are awaiting the completion of the switching to begin work on a 500 KV equipment outage.

Which one of the following describes the method to prevent violating the switching orders and GGNS procedures?

A. The Shift Manager may approve deviation from the GGNS SOI since the work to be done is in the Entergy - Mississippi portion of the switchyard.

B. A revision of the SOI is required to be approved by the OSRC, Electrical Superintendent and Operations Manager prior to proceeding with work.

C. A Temporary Change Notice should be written and issued with a 50.59 evaluation for the 500 KV SOI.

D. A Component Position Control Form should be written and issued with a 50.59 evaluation.

QUESTION NRC RECORD # WRI 884 ANSWER: C. SYSTEM # Procedures K/A Generic 2.1.20: 4.3/4.2 2.1.2: 3.0/4.0 LP# GLP-OPS-PROC 2.1.23: 3.9/4.0 OBJ. 3.4; 3.5; 3.7; 3.8; 11.4 SRO TIER 3 GROUP / RO TIER GROUP

REFERENCE:

01-S-02-2 sect 6.1 NEW 01-S-02-3 sect 6.4 & 6.5 MODIFIED BANK DIFF 1; M 01-S-02-1 sect 2.5; 2.6; 5.2; 6.1 01-S-06-2 sect 6.2.1a2 RO SRO BOTH CFR41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 85 The Control Room Supervisor has dispatched operators with the N19 Condensate System Manual Valve Lineup and Electrical Lineup Check sheets for restoration following RF13.

Which one of the following describes the use of these checksheets?

A. An SRO verifies proper revisions and marks the checksheets as a controlled copy and signs and dates the checksheet, this is valid the entire time the checksheet is in use.

The operators may perform the checksheet in any sequence, unless otherwise noted.

B. An SRO verifies proper revisions and marks the checksheets as a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> controlled copy and signs, dates and times the checksheet, this validation must be renewed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the checksheet is complete. The operators may perform the checksheet in any sequence, unless otherwise noted.

C. An SRO verifies proper revisions and marks the checksheets as a controlled copy and signs, dates and times the checksheet, this is valid the entire time the checksheet is in use. The operators must perform the checksheet in sequence with the manual valve lineups completed before electrical lineups.

D. An SRO verifies proper revisions and marks the checksheets as a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> controlled copy and signs, dates and times the checksheet, this validation must be renewed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the checksheet is complete. The operators must perform the checksheet in sequence, unless otherwise specified by the SRO.

QUESTION NRC RECORD # WRI 885 ANSWER: A. SYSTEM # Procedures K/A Generic 2.1.21: 3.1/3.2 2.1.20: 4.3/4.2 2.2.14: 2.1/3.0 LP# GLP-OPS-PROC 2.1.29: 3.4/3.3 OBJ. 3.15; 4.5; 11.4; SRO TIER 3 GROUP / RO TIER GROUP 11.18; 45.2

REFERENCE:

02-S-01-2 sect 6.12 NEW 01-S-06-2 sect 6.2.1a.2 MODIFIED BANK DIFF 1; M RO SRO BOTH CFR41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 86 The plant is in a refueling outage on day 4 with core alterations in progress.

Shutdown cooling and forced circulation are lost due to a loss of electrical power.

The fuel handlers on 208 ft in Containment report an excessive amount of steam emanating from the Reactor Cavity area.

RWCU inlet temperature indications in the Main Control Room are indicating 215°F and rising.

Which one of the following identifies the Plant Mode of Operation and any actions that should be taken?

Tech Specs 3.6.1.1 and 3.6.4.1 are provided.

A. Mode 3 requiring Primary Containment to be operable.

B. Mode 4 requiring Secondary Containment to be operable.

C. Mode 4 with NO requirements for Primary or Secondary Containment.

D. Mode 5 with NO requirements for Primary or Secondary Containment.

QUESTION NRC RECORD # WRI 886 ANSWER: D. SYSTEM # Procedures K/A Generic 2.1.22: 2.8/3.3 LP# GLP-OPS-TS01 OBJ. 4.16; 5 SRO TIER 3 GROUP / RO TIER GROUP

REFERENCE:

Tech Specs NEW definitions Table 1.1-1 MODIFIED BANK DIFF 2; CA 3.6.4.1 & bases Applicability RO SRO BOTH CFR43.2 REFERENCE MATERIAL REQUIRED: Tech Specs 3.6.1.1 & 3.6.4.1

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 87 Refueling Outage 13 is in progress.

Surveillance 06-OP-1E12-C-0014, LPCI A Cold Shutdown Valve Test needs to be run according to the outage schedule.

Which one of the following individuals may directly order the Control Room Operator to perform this test?

A. Outage Director B. Operations Manager C. Control Room Supervisor D. Training Department SRO with an inactive license coordinating surveillance testing QUESTION NRC RECORD # WRI 901 ANSWER: C. SYSTEM # Procedures K/A Generic 2.1.9: 2.5/4.0 LP# GLP-OPS-PROC OBJ. 11.4; 11.20 SRO TIER 3 GROUP / RO TIER GROUP

REFERENCE:

01-S-06-2 sect 6.1.4; 6.2.1n; NEW 6.2.4a(1) MODIFIED BANK DIFF 1; M 01-S-06-4 sect 6.1.3 RO SRO BOTH CFR41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 88 The plant is operating at 100% power.

A fire has been reported on the 133 ft. elevation of the Turbine Building.

The Fire Brigade Leader has requested that both Smoke Exhaust Fans be started and the Turbine Building Roof Hatches be opened to facilitate the removal of smoke that has accumulated in the areas.

Which one of the following describes the correct actions to be taken prior to starting the fans and opening the hatches?

A. Verify Turbine Building Vent Stack Monitors are in operation and install Continuous Air Monitors (CAMs) at the roof hatches to be opened.

B. Ensure Turbine and Radwaste Building rollup doors are open to allow air flow to enhance smoke removal.

C. Operate both Turbine Building Exhaust fans through the Turbine Building Exhaust Filter Train for representative sampling of the Turbine Building atmosphere.

D. Notify Radiation Protection/Plant Chemistry to draw an air sample from the Smoke Exhaust Fans and setup monitoring equipment at the roof hatches to be opened.

QUESTION 88 SRO NRC RECORD # WRI 888 ANSWER: D. SYSTEM # U41 K/A 286000 2.3.8: 3.2 LP# GLP-OPS-U4100 2.3.11: 3.2 OBJ. 14 SRO TIER 2 GROUP 2 / RO TIER GROUP

REFERENCE:

Tech Specs 6.3.10 Condition A NEW 06-CH-1D17-V-0032 MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR 41.10/41.11/41.13 REFERENCE MATERIAL REQUIRED: None 43.4/43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 89 Which one of the following describes actions that would require the completion of a Component Position Control Tag (CPC)?

A. Construction Water is being connected to the Instrument and Service Air compressors to allow disconnecting TBCW piping for replacement, estimated job should be completed in 10 days.

B. I&C is lifting leads in H13-P625 to allow E22-F015, HPCS PMP SUCT FM SUPP POOL to be closed and E22-F001, HPCS PMP SUCT FM CST to be opened due to an inoperable Suppression Pool level transmitter.

C. Plant Services has a fire hose hooked up at the SSW basin to wash out mud per a WO and is estimated to take 2 days and hoses will be removed per the WO instructions.

D. An operator is opening a service air valve to allow mechanical maintenance to use air operated tools to support work, per a WO.

QUESTION NRC RECORD # WRI 889 ANSWER: C. SYSTEM # Equipment K/A Generic 2.2.15: 2.2/2.9 Control - Configuration 2.2.11: 2.5/3.4 Control LP# GLP-OPS-PROC OBJ. 45.6; 45.18 SRO TIER 3 GROUP / RO TIER GROUP

REFERENCE:

02-S-01-37 sect 5.1; 6.1.6; 6.1.7 NEW 6.2.1; 6.3.1d MODIFIED BANK DIFF 2; CA 01-S-06-3 sect 5.24 02-S-01-2 sect 6.10.1 RO SRO BOTH CFR41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 90 The plant is operating at full power.

Which one of the following activities could be worked WITHOUT initiating a Work Request?

A. Tighten the packing to stop a leak on P66-F029A, Domestic Water Supply to the Auxiliary Building.

B. Repair the open (green) light bulb on circuit breaker 152-1411 LFMG B MTR FDR (CB-1B).

C. Repair the latching mechanism for the Control Room Door OC506 that accesses the Operations Kitchen.

D. Install a pressure gauge on the discharge of RHR A Jockey pump at PP-N408A on E12-FX056.

QUESTION NRC RECORD # WRI 890 ANSWER: B. SYSTEM # Procedures K/A Generic 2.2.19: 2.1/3.1 LP# GLP-OPS-PROC OBJ. 26.3 SRO TIER 3 GROUP / RO TIER GROUP

REFERENCE:

WM-100 sect 5.1 & Att 9.1 NEW 05-1-02-III-5 AB isolation MODIFIED BANK valves TRM3.6.4.2-1 DIFF 2; CA 04-1-01-B33-1 Att III 04-1-01-E12-1 Att IA RO SRO BOTH CFR41.10/43.5 06-OP-SP64-D-0044 Att I Sect 5.2 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 91 A plant startup is in progress.

At 60 psig, RCIC was placed in Standby per the SOI.

At 150 psig, RCIC surveillance 06-OP-1E51-C-0005 was started.

During the surveillance, the Auxiliary Building Operator found pieces of the RCIC Gland Seal Compressor on the floor.

Soon after this discovery, the compressor tripped its circuit breaker.

The balance of the surveillance was completed satisfactorily.

All other systems are Operable.

Which one of the following describes the status of RCIC and the allowances for continuing power ascension to Mode 1?

Tech Spec 3.5.3, Surveillance 06-OP-1E51-C-0005, and 01-S-06-44 are provided.

A. RCIC is operable and power ascension and mode change may continue as planned.

B. RCIC is functional but inoperable; however, power ascension may continue up to 950 psig but the plant must remain in Mode 2.

C. RCIC is functional but inoperable, power ascension halted and reactor pressure maintained less than 150 psig. The plant may remain in Mode 2.

D. RCIC is functional but inoperable; however, power ascension may continue up to 950 psig and the mode change to Mode 1 is allowed provided HPCS is operable.

QUESTION NRC RECORD # WRI 891 ANSWER: C. SYSTEM # Procedures K/A Generic 2.2.24: 2.6/3.8 LP# GLP-OPS-TS01 OBJ. 13; 17 SRO TIER 3 GROUP / RO TIER GROUP

REFERENCE:

TS 3.5.3 & bases Applicability NEW 06-OP-1E51-C-0005 sect 5.9 MODIFIED BANK DIFF 3; CA 03-1-01-1 sect 6.2.11a & Note 01-S-06-44 sect 5.2; 5.12 RO SRO BOTH CFR41.10/43.5 04-1-01-E51-1 sect 3.12 REFERENCE MATERIAL REQUIRED: TS3.5.3; 01-S-06-44; 06-OP-1E51-C-0005

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 92 A Refueling Outage is in progress.

Which one of the following describes a core alteration requiring the presence of a Senior Reactor Operator dedicated to Refueling?

04-1-01-C11-2 Figure 1a, 17-S-02-40 Att. V and 17-S-02-300 Att. V are provided.

A. During core loading verification, LPRM string 26-35 has been found out of position.

A tool is to be used to reposition the string for proper alignment.

B. During core loading verification, fuel bundle 1-36 has been found oriented improperly requiring reorientation.

C. Control rod 16-45 has a full blade guide installed with the cell de-fueled and requires withdrawal to allow for control rod drive mechanism maintenance.

D. Replacement of IRM detector G is complete and the technicians want to stroke the detector in and out of the core with it bypassed.

QUESTION NRC RECORD # WRI 892 ANSWER: B. SYSTEM # Procedures K/A Generic 2.2.34: 2.8/3.2 LP# GLP-OPS-PROC 2.2.32: 2.3/3.3 OBJ. 11.2 SRO TIER 3 GROUP / RO TIER GROUP

REFERENCE:

01-S-06-2 sect 5.8 NEW 04-1-01-C11-2 Figure 1a MODIFIED BANK DIFF 2; CA 17-S-02-40 Att V 17-S-02-300 Att V RO SRO BOTH CFR43.6/43.7 REFERENCE MATERIAL REQUIRED: 17-S-02-40 Att. V; 17-S-02-300 Att. V; 04-1-01-C11-2 Figure 1a

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 93 A shipment of spent RWCU resin was being transported on Grand Gulf Road when the truck overturned.

The Claiborne County Sheriffs office notified Grand Gulf of the accident.

Who is responsible for making the initial notification of Entergy personnel and who is notified?

A. Shift Manager contacts the Duty Manager B. Shift Manager contacts the On-Duty Radiation Protection Supervisor C. On-Duty Radiation Protection Supervisor contacts the Duty Manager D. On-Duty Radiation Protection Supervisor contacts the Shift Manager QUESTION NRC RECORD # WRI 893 ANSWER: A. SYSTEM # Radiation K/A Generic 2.3.3: 1.8/2.9 Protection - SRO Responsibilities LP# GLP-EP-EPTS6 OBJ. 3 SRO TIER 3 GROUP / RO TIER GROUP

REFERENCE:

10-S-01-32 sect 6.1.1c NEW Att II MODIFIED BANK DIFF 1; M RO SRO BOTH CFR41.10/41.12/43.4 REFERENCE MATERIAL REQUIRED: None 43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 94 The plant is operating at rated conditions.

A leak has been identified on a valve in the Auxiliary Building Steam Tunnel.

Plans are to lower reactor power to 40% to allow a Steam Tunnel entry to attempt to back seat the valve to stop the leak allowing continued plant operation.

Moderate Hydrogen Water Chemistry has just been secured.

It is estimated the job will take 10 minutes to complete.

The worker selected to perform the job has a current exposure of 200 mRem.

Radiation levels are expected to be 2 - 2.5 R/Hr in the area of the work.

Determine whether the General Operations Radiation Work Permit is acceptable for this job?

NMM RP-105 and Operations RWP are provided.

A. The General Operations RWP is sufficient without further actions.

B. The General Operations RWP is sufficient however a special pre-job briefing is required.

C. A job specific RWP is required. Radiation Protection will require more in-depth planning and a pre-job briefings.

D. A job specific RWP is required. Radiation Protection will require approval of the ALARA Committee due to the excessive dose commitment.

QUESTION NRC RECORD # WRI 894 ANSWER: C. SYSTEM # Radiation K/A Generic 2.3.7: 2.0/3.3 Protection - RWP LP# ELP-GET-RWT-OBJ. 32; 68 SRO TIER 3 GROUP / RO TIER GROUP

REFERENCE:

NMM RP-105 sect 5.2.3; 5.2.5 NEW 5.2.6; 5.3.1 MODIFIED BANK DIFF 2; CA NRC 8/2002 wri710 RO SRO BOTH CFR41.10/41.12/43.4 REFERENCE MATERIAL REQUIRED: NMM RP-105; Ops RWP; 43.5 Calculator

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 95 The plant has experienced a transient.

Multiple control rods did NOT fully insert.

Plant conditions are as follows:

Reactor water level: -189 and stable Reactor pressure: 124 psig Drywell temperature: 245°F Drywell pressure: 4.8 psig Drywell hydrogen: 3.1%

Containment temperature: 125°F Containment pressure: 2.3 psig Containment hydrogen: 1.2%

Suppression Pool level: 23.9 ft.

Suppression Pool temperature: 113°F Which one of the following sets of procedures takes priority for controlling the current plant conditions?

A. System Operating Instructions (SOIs) and Off-Normal Event Procedures (ONEPs)

B. Off-Normal Event Procedures (ONEPs) and Emergency Procedures (EPs)

C. Emergency Procedures (EPs) and Severe Accident Procedures (SAPs)

D. Severe Accident Procedures (SAPs) only QUESTION NRC RECORD # WRI 895 ANSWER: D. SYSTEM # SAPs K/A Generics 2.4.8: 3.7 2.4.11: 3.6 GLP-EP-EPT19 OBJ. 5, 7 SRO TIER 3

REFERENCE:

05-S-01-SAP-1 Attachment I NEW MODIFIED BANK DIFF 2:CA RO SRO BOTH CFR41.10/43.5 REFERENCE MATERIAL REQUIRED: EPs and SAPs

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 96 The plant is operating at 100% power.

The shift has the normal complement of licensed operators: 3 SROs and 3 ROs.

Security has notified the Shift Manager that armed adversaries have entered the Protected Area and are presently occupying the Fire Water Pump House.

Which one of the following describes the actions to be taken by Operations personnel and the Emergency Response Organization (ERO)?

A. Operations personnel are to immediately return to the Control Room using routes inside the power block. The ERO should report to their Emergency Response Facilities.

B. One SRO and one RO will man the Remote Shutdown Panels with the balance of operations personnel taking cover. The ERO should report to any Emergency Response Facilities outside of the Protected Area.

C. One SRO and one RO will man the Remote Shutdown Panels with the balance of operations personnel taking cover. The ERO should take cover until the all clear is given by Security.

D. Control Room personnel should remain in the Control Room. All other personnel on site should take cover in their locations and remain out of the power block until the all clear is given by Security.

QUESTION NRC RECORD # WRI 896 ANSWER: C. SYSTEM # Procedures K/A Generic 2.4.12: 3.4/3.9 LP# GLP-OPS-ONEP 2.4.28: 2.3/3.3 OBJ. 1; 45 SRO TIER 3 GROUP / RO TIER GROUP LP# GLP-EP-EPTS6 OBJ. 7

REFERENCE:

05-1-02-VI-4 NEW sect 1.5; 1.6; 1.7; 2.0 MODIFIED BANK DIFF 2; CA 10-S-01-1 sect 6.1.5 Caution RO SRO BOTH CFR41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 97 A LOCA has occurred and reactor water level has dropped to -220 inches and is slowly lowering.

ECCS Systems are in a degraded condition.

The Emergency Response Organization has manned all Emergency Facilities.

The Shift Manager determines that entry into the Severe Accident Procedures is warranted.

Which one of the following describes whose concurrence is required and the basis for this concurrence?

A. Offsite Emergency Coordinator concurrence is required to ensure the proper Emergency Declaration and protective action recommendations are made.

B. Emergency Director and Offsite Emergency Coordinator concurrence is required to ensure proper Emergency Declaration and protective action recommendations are made.

C. Radiation Emergency Manager concurrence is required due to the significant radiological consequences associated with actions in the SAPs.

D. Emergency Director concurrence is required due to the significant radiological consequences associated with actions in the SAPs.

QUESTION NRC RECORD # WRI 897 ANSWER: D. SYSTEM # Procedures K/A Generic 2.4.18: 2.7/3.6 LP# GLP-EP-EPT19 OBJ. 3 SRO TIER 3 GROUP / RO TIER GROUP

REFERENCE:

05-S-01-SAP-1 NEW General Comments MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 98 The plant is operating at rated conditions.

A fire in the Turbine Building resulted in a loss of power to the D and E DC buses.

The Operator-at-the Controls notices his H13-P680 annunciators are NOT functioning.

Further investigation concludes NONE of the Control Room annunciators are working.

Which one of the following describes the actions to be taken for these conditions?

A. Dispatch operators in the Control Room and plant to continuously monitor applicable in-plant instrumentation and declare an Alert to activate the ERO.

B. Setup the PDS computers to sound alarms on key parameters and call in additional operators to begin a controlled shutdown of the plant.

C. Declare an Alert to activate the ERO and manually scram the reactor then proceed to place the plant in Cold Shutdown.

D. Dispatch operators in the Control Room and plant to continuously monitor applicable in-plant instrumentation, call in additional operators to begin a controlled shutdown of the plant.

QUESTION NRC RECORD # WRI 898 ANSWER: A. SYSTEM # Procedures K/A Generic 2.4.32: 3.3/3.5 LP# GLP-EP-EPTS6 OBJ. 1; 3; 7; 8 SRO TIER 3 GROUP / RO TIER GROUP

REFERENCE:

10-S-01-1 sect 6.2.1a NEW Att I 17.2.1 MODIFIED BANK DIFF 2; CA 04-1-01-C82-1 Att III & X RO SRO BOTH CFR41.10/43.5 REFERENCE MATERIAL REQUIRED: 10-S-01-1

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 99 The plant has experienced an ATWS.

Due to rising Drywell pressure, the CRS decided to dispatch two operators from the Control Room to shutdown the Drywell Purge Compressors, per EP-2, Attachment 15.

Which one of the following describes the radiological monitoring required for this task?

EP-2 Attachment 15 is provided.

A. Operators can perform this task under current RWPs with NO additional monitoring required.

B. The Control Room RP technician can provide the required radiological monitoring for this task.

C. The operators must report to the Operational Support Center (OSC) to be briefed prior to performing this task, with NO additional monitoring required.

D. The operators must report to the Operational Support Center (OSC) to be briefed and have an RP technician assigned to perform radiological monitoring for this task.

QUESTION NRC RECORD # WRI 899 ANSWER: B. SYSTEM # Emergency K/A Generics 2.4.36: 2.8 Procedures GLP-EP-EPTS6 OBJ. 5 SRO TIER 3

REFERENCE:

10-S-01-29 sect 6.4.2a NEW MODIFIED BANK DIFF 2:CA RO SRO BOTH CFR41.10/43.5 REFERENCE MATERIAL REQUIRED: EP-2 Attachment 15

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION FEBRUARY 2004 SENIOR REACTOR OPERATOR QUESTION 100 The plant has experienced an ATWS.

Due to severe weather in the area, ALL telephone lines for the plant are damaged.

Which one of the following describes the backup method for notification of state and local agencies?

A. Satellite phone B. Cellular phones C. Shortwave radios D. Emergency Response Data System (ERDS) computer system QUESTION NRC RECORD # WRI 900 ANSWER: A. SYSTEM # Emergency K/A Generics 2.4.43: 3.5 Procedures GLP-EP-EPTS6 OBJ. 3 SRO TIER 3

REFERENCE:

10-S-01-6 Section 6.3.1 NEW MODIFIED BANK DIFF 1;M RO SRO BOTH CFR41.10/43.5 REFERENCE MATERIAL REQUIRED: