Text

Init Exam -08/2002- Draft RO-SRO Written Exam
	ML030420309
Person / Time
Site:	Grand Gulf
Issue date:	02/23/2002
From:	Gody A; NRC Region 4
To:	Eaton W; Entergy Operations
References
	50-416/02-301 50-416/02-301
	Download: ML030420309 (200)
	v • d • e

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 1 The plant is operating at 20 % power following RF11.

A ground fault on the Entergy Grid resulted in the Main Generator output circuit breakers J5228 and J5232 automatically opening on a Generator Lockout.

Which one of the following describes the reaction of the plant to this trip?

ASSUME NO OPERATOR ACTIONS.

A. Main Steam Bypass Valves will automatically open maintaining reactor pressure. The reactor will scram following the closure of the Main Turbine Stop and Control Valves.

B. Main Steam Bypass Valves will automatically open maintaining reactor pressure. The reactor will remain at power with the Main Turbine remaining in operation.

C. Main Steam Bypass Valves will automatically open maintaining reactor pressure. The reactor will remain at power with the Main Turbine Stop and Control Valves closing.

D. The reactor will scram due to the Main Turbine Control Valve fast closure that will result in a subsequent normal closure of the Main Turbine Control Valves to maintain reactor pressure.

QUESTION 1 NRC RECORD # WRI 553 ANSWER: C SYSTEM # N41; K/A 295005 AA2.05: 3.8/3.9 N32-2; C71 AA2.04: 3.7/3.8 LP# GLP-OPS-C7100 AA2.03: 3.1/3.1 OBJ. 9, 10 LP# GG-1-LP-OP-N3202 OBJ. 2 LP# GG-1-LP-OP-N4151 OBJ. 11 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 1

REFERENCE:

Tech. Spec 3.3.1.1 NEW 05-1-02-I-2 sect 5.1 - 5.4 MODIFIED BANK DIFF 2; CA NRC 6/2001 RO SRO BOTH CFR 41.5/41.6 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 1 The plant is operating at 20 % power following RF11.

A ground fault on the Entergy Grid resulted in the Main Generator output circuit breakers J5228 and J5232 automatically opening on a Generator Lockout.

Which one of the following describes the reaction of the plant to this trip?

ASSUME NO OPERATOR ACTIONS.

A. Main Steam Bypass Valves will automatically open maintaining reactor pressure. The reactor will scram following the closure of the Main Turbine Stop and Control Valves.

B. Main Steam Bypass Valves will automatically open maintaining reactor pressure. The reactor will remain at power with the Main Turbine remaining in operation.

C. Main Steam Bypass Valves will automatically open maintaining reactor pressure. The reactor will remain at power with the Main Turbine Stop and Control Valves closing.

D. The reactor will scram due to the Main Turbine Control Valve fast closure that will result in a subsequent normal closure of the Main Turbine Control Valves to maintain reactor pressure.

QUESTION 1 NRC RECORD # WRI 553 ANSWER: C SYSTEM # N41; K/A 295005 AA2.05: 3.8/3.9 N32-2; C71 AA2.04: 3.7/3.8 LP# GLP-OPS-C7100 AA2.03: 3.1/3.1 OBJ. 9, 10 LP# GG-1-LP-OP-N3202 OBJ. 2 LP# GG-1-LP-OP-N4151 OBJ. 11 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 1

REFERENCE:

Tech. Spec 3.3.1.1 NEW 05-1-02-I-2 sect 5.1 - 5.4 MODIFIED BANK DIFF 2; CA NRC 6/2001 RO SRO BOTH CFR 41.5/41.6 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 2 A reactor scram has occurred.

Which one of the following is a correct method of verifying the position of the control rods?

(The scram has NOT been reset.)

A. Using the full core display on H13-P680, depress ALL RODS with RCIS in Raw Data and observe a blank display with only green LEDs for all control rods.

B. Using the full core display on H13-P680, depress ALL RODS with RCIS in Raw Data and observe all control rods indicate 00 with a green LED for all control rods.

C. Using the full core display on H13-P680, depress ALL RODS with RCIS out of Raw Data and observe a blank display with only red LEDs for all control rods.

D. Using the full core display on H13-P680, depress ALL RODS with RCIS out of Raw Data and observe all control rods indicate 00 with a red LED for all control rods.

QUESTION 2 NRC RECORD # WRI 10 ANSWER: A. SYSTEM # C11-2; K/A 295006 AA2.02: 4.3/4.4 C11-1B 201005 A3.02: 3.5/3.5 LP# GG-1-LP-OP-C111B A4.02: 3.7/3.7 OBJ. 3c, 3f 201003 K4.05: 3.2/3.3 A3.01: 3.7/3.6 LP# GG-1-LP-OP-C1102 A4.02: 3.5/3.5 OBJ. 12 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

04-1-01-C11-2 NEW sect. 4.7.2p & 4.8.2i MODIFIED BANK DIFF: 2; CA 05-1-02-I-1 NRC 6/2001 sect. 2.1; 3.7; & 3.7.4 RO SRO BOTH CFR 41.6/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 3 Plant conditions are as follows:

MODE: Mode 1 Rx power: 28 %

T-G Load: 365 MWE Load Demand 390 MWE Bypass position: 0 %

All other parameters are per plant design.

The operator withdraws a control rod that raises Reactor power to 29 %.

How will the Turbine EHC Control System respond?

A. Bypass Control Valves will throttle open as required to maintain Rx pressure.

B. HP Turbine Control Valves will throttle open as required to maintain Rx pressure.

C. LP Turbine Control Valves will throttle open as required to maintain Rx pressure.

D. HP Turbine Control and Bypass Control Valves will throttle open as required to maintain Rx pressure.

QUESTION 3 NRC RECORD # WRI 69 ANSWER: B. SYSTEM # N32-2 K/A 295007 AK2.01: 3.5/3.7 241000 A2.02: 3.7/3.7 LP# GG-1-LP-OP-N3202 K4.01: 3.8/3.8 OBJ 4b, 6b, 7b SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

03-1-01-2 sect. 5.2 NEW MODIFIED BANK DIFF: 2; CA NRC 6/2001 RO SRO BOTH CFR 41.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 4 The plant was operating at 20 % power when a Feedwater rupture in the Turbine Building caused Reactor water level to drop.

The Control Room Operator manually initiated HPCS and RCIC.

Level in the Reactor dropped to ? 32 inches before HPCS and RCIC turned level, and level is now rising.

Which one of the following best describes the status of the Recirculation System?

A. Recirc Pumps are in Slow Speed with the Flow Control Valves Locked up (motion inhibit).

B. Recirc Pumps are tripped with the Flow Control Valves Locked up (motion inhibit).

C. Recirc Pumps are in Slow Speed with the Flow Control Valves in their pre-transient positions.

D. Recirc Pumps are tripped with the Flow Control Valves in their pre-transient positions.

QUESTION 4 NRC RECORD # WRI 121 ANSWER: C. SYSTEM # B33 K/A 295009 AK1.02: 3.0/3.1 AK2.03: 3.1/3.2 LP# GLP-OPS-B3300 AA1.03: 3.0/3.1 OBJ. 23, 24, 25, 47 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

ARI 04-1-02-H13-P680 NEW 3A-D4; 3A-D10 MODIFIED BANK DIFF 1; M NRC 3/1998 RO SRO BOTH CFR 41.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 5 Given the following plant conditions:

Reactor Power 100%

Reactor Level +36 inches Reactor Pressure 1025 psig Containment Temperature 85 °F steady Containment Pressure 0.03 psig steady Suppression Pool Temperature 81 °F steady Drywell Pressure 1.1 psig rising slowly Drywell Temperature 110 °F steady Drywell Area Sumps show no unusual changes in level, flow, or temperature.

Drywell Atmosphere radiation monitor show no changes.

The Roving NOA has noted that Drywell Pressure is rising slowly.

Drywell atmosphere radiation levels are steady.

Which one of the following describes a possible cause of the conditions as noted above?

A. Small leak on the Main Steam Line Flow Elbows Instrument Line.

B. Small leak on the RWCU suction from the Reactor Bottom Head.

C. Small leak on the Instrument Air header inside the Drywell.

D. Small leak on Recirc Pump Seals.

QUESTION 5 NRC RECORD # WRI 286 ANSWER:. C. SYSTEM # M41; P53 K/A 295010 AK3.04: 3.5/3.8 AK3.05: 3.5/3.4 LP# GLP-OPS-P5300 2.4.21: 3.7/4.3 OBJ. 26.1 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

GGNS 1998 event NEW CR 1998-0952 MODIFIED BANK DIFF 3, CA NRC 4/2000 RO SRO BOTH CFR 41.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 6 The plant is performing a reactor startup from cold shutdown.

The reactor was at the point of adding heat.

The Control Room Supervisor instructed the operators to stop the startup for a short duration to perform a surveillance.

During this time, the reactor went subcritical and power dropped to range 3 of the IRMs.

The At-The-Controls Operator, noting that reactor power had dropped selected the next control rod and withdrew the control rod from 20 to 48 with continuous motion as allowed by the Control Rod Movement Sequence Sheet.

This resulted in a sustained 20 second period.

The following are the plant parameters at present:

Reactor Pressure 80 psig Reactor Level + 40 inches 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. Perform the coupling checks for the Control Rod, and inform the Reactor Engineer of the power rise.

C. Withdraw the next in sequence Control Rod to maintain the power rise to reach the point of adding heat.

D. Insert the Control Rod to a position which causes reactor period to be > 50 seconds.

QUESTION 6 NRC RECORD # WRI 204 ANSWER: D. SYSTEM # C11-2; C51 K/A 295014 AK3.01: 4.1/4.1 LP# GG-1-LP-OP-IOI01 OBJ. 3c & d SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

03-1-01-1 sect. 2.1.4; 2.1.16 NEW Susquehanna reactivity MODIFIED BANK DIFF 1; M Event 7/98 NRC 6/2001 04-1-01-C51-1 sect 4.3.2 NOTE RO SRO BOTH CFR 41.1/41.2/ 41.6/43.6 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 7 Scram conditions exist. All control rods did NOT fully insert.

Reactor water level is being maintained at -60 inches wide range.

Reactor pressure is being maintained at 910 psig.

Reactor power is 20 %.

The following initial indications exist:

RPS white lights on H13-P680 are extinguished.

Scram Air Header Pressure low annunciator is illuminated.

RX SCRAM TRIP annunciator is illuminated.

The following actions have been taken:

Defeat the RPS scram signal and reset RPS Unisolate the Instrument Air header Defeat Alternate Rod Insertion A CRD pump is confirmed operating and the CRD FCV is open to achieve 250 psid Drive pressure.

Which one of the following contains the minimum actions required to drive the control rods to position 00 using Rod Control and Information System?

A. Bypass Control Rod Drive withdrawal blocks, select control rods and insert.

B. Bypass Control Rod Drive withdrawal blocks, select control rods in sequence and insert.

C. Bypass Control Rod Drive insert and withdrawal blocks, select control rods and insert.

D. Select control rods in sequence and insert.

QUESTION 7 NRC RECORD # WRI 203 ANSWER: C. SYSTEM # C11-2; C71; C11- K/A 295015 AK3.01: 3.4/3.7 1A LP# GG-1-LP-RO-EP02A OBJ. 5 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

EP 05-S-01-EP-2A NEW Step 48 Att. 18, 19 & 20 MODIFIED BANK DIFF 3; CA NRC 6/2001 RO SRO BOTH CFR 41.6/43.6 REFERENCE MATERIAL REQUIRED: 05-S-01-EP-2 EP-2A

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 8 A loss of coolant accident has occurred.

Determine which one of the following describes conditions that would direct / allow manual initiation of Containment Spray per the Emergency Procedures?

Supp Pool Lvl CTMT Temp CTMT Press Drywell Temp Drywell Press A. 18 FT 186ºF 1 psig 200ºF 3 psig B. 17 FT 120ºF 7 psig 215ºF 15 psig C. 18 FT 120ºF 6.5 psig 200ºF 15 psig D. 25 FT 140ºF 4 psig 215ºF 7 psig QUESTION 8 NRC RECORD # WRI 601 ANSWER: B. SYSTEM # M41-1; E12 K/A 295024 AK2.14: 3.9/3.9 LP# GG-1-LP-RO-EP03 OBJ. 3 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

EP 05-S-01-EP-3 NEW CTMT Temp & Press Legs MODIFIED BANK DIFF 2; CA Figures 3 & 4 RO SRO BOTH CFR 41.9/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-S-01-EP-3 & Figures 3 & 4

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 9 The following conditions exist in the plant:

Reactor power 0% all rods inserted.

Reactor pressure 230 psig and rising.

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

RHR C is injecting into the reactor vessel.

Which one of the following identifies adequate core cooling?

A. Adequate core cooling is NOT assured.

B. Adequate core cooling is assured by Minimum Alternate RPV Flooding Pressure (MARFP) with SRVs and Reactor pressure.

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 9 NRC RECORD # WRI 309 ANSWER: A. SYSTEM # Eps & K/A 295031 EK1.01: 4.6/4.7 Conduct of Ops. 2.1.1: 3.7/3.8 LP# GG-1-LP-RO-EP02 2.4.21: 3.7/4.3 OBJ. 15, 16 LP# GG-1-LP-RO-EP01 OBJ. 4a LP# GG-1-LP-OP- PROC OBJ. 10b3 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

01-S-06-2 section 5.18 NEW 05-S-01-EP-2 steps 69 - 74 MODIFIED BANK DIFF 3, CA PSTG App. B EPG Cont 3 NRC 12/2000 Steam Cooling RO SRO BOTH CFR 41.2/41.3/41.10/

REFERENCE MATERIAL REQUIRED: 05-S-01-EP-2 43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 10 An ATWS has occurred. Actions of EP-2A are being taken.

Which one of the following describes an allowance to terminate injection of Standby Liquid Control?

A. Control rods have been inserted to the equivalent of the first banked position with RPV temperature at < 200°F making the reactor subcritical.

B. All control rods are inserted to the Maximum Subcritical Banked Withdrawal Position, which assures the reactor will remain subcritical under all conditions.

C. RPV temperature has been reduced to < 200 °F and indicated reactor power on all IRMs is downscale on range 1, which indicates a subcritical reactor.

D. Standby Liquid Control has been injected such that Hot Shutdown Boron Weight (HSBW) has been injected and confirmed by chemical analysis.

QUESTION 10 NRC RECORD # WRI 216 ANSWER:. B. SYSTEM # C41; K/A 295037 EA1.04: 3.4/4.5 C11; C71 EK1.04: 3.4/3.6 EK1.05: 3.4/3.6 LP# GG-1-LP-RO-EP02A EA2.03: 4.3/4.4 OBJ. 2, 3, 5 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

05-S-01-EP-2A NEW step 2 & 4 MODIFIED BANK DIFF 3, CA GGNS PSTG App B NRC 4/2000 RC/Q-1 RO SRO BOTH CFR 41.1/41.2/

REFERENCE MATERIAL REQUIRED: 05-S-01-EP-2A 41.6/43.6

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 11 A LOCA has occurred. The Plant Supervisor has ordered the Hydrogen Recombiners started for Hydrogen removal in Containment.

Determine the final Hydrogen Recombiner Power Setting and the time to final Recombiner power.

Pre-LOCA Containment Temperature was 85 °F.

Post LOCA Containment Pressure +1.0 psig.

A. 47.73 kw after 20 minutes B. 47.73 kw after 25 minutes C. 49.02 kw after 20 minutes D. 49.02 kw after 25 minutes QUESTION 11 NRC RECORD # WRI 219 ANSWER: B. SYSTEM # E61 K/A 500000 EA1.03: 3.4/3.2 2.1.20: 4.3/4.2 LP# GQC-RO-CRO01 2.1.25: 2.8/3.1 OBJ. E61 task 5 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

04-1-01-E61-1 sec. 5.4.2 NEW Figure 1 MODIFIED BANK DIFF 2, CA NRC 4/2000 RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: 04-1-01-E61-1 & Calculator

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION: 12 Concerning the Fast Opening of One Recirculation Flow Control Valve transient, which one of the following CONDITIONS would result in the more severe transient on the reactor?

A. Reactor Power is at 30 % with Recirc in Slow Speed with Maximum Valve position.

B. Reactor Power is at 36 % with Recirc in Fast Speed with Minimum Valve position.

C. Reactor Power is at 60 % with Recirc in Fast Speed with Minimum Valve position.

D. Reactor Power is at 100 % with Recirc in Fast Speed with 68% Valve position.

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 13 The plant was operating at power.

A transient caused the Recirculation Pump B trip to OFF.

The following parameters are indicated:

Reactor power 65 %

Core Flow 54 Mlbm/hr Recirc A Flow 39,000 gpm Recirc B Flow 0 gpm Reactor Engineering is calculating FCBB.

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

A. Immediately Scram the Reactor.

B. Monitor Neutron Monitoring for thermal hydraulic instability. Immediately reduce core thermal power by only inserting control rods to exit the region.

C. Monitor Neutron Monitoring for thermal hydraulic instability. Immediately exit the region by reducing core thermal power by inserting control rods, or raising core flow by opening Recirc FCV A.

D. Monitor Neutron Monitoring for thermal hydraulic instability and scram the reactor if any is noted, close B33-F067B and re-open after five minutes, restart the Recirc Pump as soon as possible.

QUESTION 13 NRC RECORD # WRI 602 ANSWER: C. SYSTEM # B33 K/A 295001 AA2.01: 3.5/3.8 2.4.1: 4.3/4.6 LP# GLP- OPS-B3300 2.4.11: 3.4/3.6 OBJ. 41.2; 42; 43 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

05-1-02-III-3 sect NEW Figure 1 MODIFIED BANK DIFF: 2; CA 03-1-01-2 sect 2.24 NRC 4/2000 RO SRO BOTH CFR 41.10/41.5/43.5 REFERENCE MATERIAL REQUIRED: 05-1-02-III-3 w/o Immediate actions

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 14 The plant is in a normal electrical line-up with all busses fed from their preferred power source. If a lockout of BOP Transformer 12B were to occur, Which of the following indicates the correct status of BOP busses?

A. 11HD ENERGIZED 12HE DE-ENERGIZED 13AD DE-ENERGIZED 14AE ENERGIZED 18AG ENERGIZED 28AG DE-ENERGIZED B. 11HD DE-ENERGIZED 12HE ENERGIZED 13AD ENERGIZED 14AE DE-ENERGIZED 18AG ENERGIZED 28AG ENERGIZED C. 11HD ENERGIZED 12HE DE-ENERGIZED 13AD DE-ENERGIZED 14AE ENERGIZED 18AG DE-ENERGIZED 28AG DE-ENERGIZED D. 11HD DE-ENERGIZED 12HE ENERGIZED 13AD ENERGIZED 14AE DE-ENERGIZED 18AG DE-ENERGIZED 28AG ENERGIZED QUESTION 14 NRC RECORD # WRI 507 ANSWER: B SYSTEM # R21 K/A 295003 A1.01: 3.7/3.8 LP# GLP-OPS-R2700 OBJ. 8 &15.1. SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

04-1-01-R21-11 sect 3.2 NEW 04-1-01-R21-12 sect 3.2 MODIFIED BANK DIFF 1; M 04-1-01-R21-13 sect 3.2 NRC 6/2001 04-1-01-R21-14 sect 3.2 RO SRO BOTH CFR 41.7 04-1-01-R21-18 sect 3.2 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 15 Which one of the following describes the reason for isolating the Main Steam Isolation Valves on a Low Main Condenser Vacuum?

A. Prevent erosion damage to the Main Steam Isolation Valve and Main Steam Bypass Valve seats due to steam condensation in the Main Steam Lines that would prevent their complete isolation in an emergency.

B. Prevent erosion damage to turbine blading in the Low Pressure Turbine due to steam condensation in the Main Steam Lines.

C. Prevent over-pressurization of low pressure piping on the suction of the Condensate pumps that could result in a rupture introducing steam outside Secondary Containment.

D. Prevent rupture of the turbine rupture diaphragms or damage to the turbine exhaust hood that could lead to leakage of radiation to the environment.

QUESTION 15 NRC RECORD # WRI 220 ANSWER: D. SYSTEM # B21; K/A 295002 AK3.05: 3.4/3.4 N11; N62 AA1.04: 3.3/3.4 LP# GG-1-LP-OP-M7101 OBJ. 6 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

GGNS Tech Spec Bases NEW 3.3.6.1-1d MODIFIED BANK DIFF 1; M NRC 4/2000 RO SRO BOTH CFR 41.4/43.4 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 16 The plant was operating at 80 % power.

Reactor Narrow Range Water Level transmitter C34-N004B has failed downscale and brought in annunciator RX WTR LVL SIG FAIL HI/LO.

The Operator at the Controls notices the Reactor Narrow Range Level indicator C34-LI-R606A indicates offscale HIGH and annunciator RFPT/MN TURB LVL 9 TRIP is in.

Reactor Narrow Range Water Level indicator R606C is reading + 36 inches.

Reactor Upset Range Water Level indicator is reading + 38 inches.

Reactor Wide Range Water Level indicator on P680 is reading + 40 inches.

Reactor Wide Range Water Level indicators A & B on P601 are reading + 40 inches.

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

(NO OTHER ALARMS ARE PRESENT.)

A. Immediately initiate a Reactor Scram and trip the Main Turbine and the Reactor Feed Pump Turbines because they failed to trip.

B. Manually select Reactor Water Level Control to Single Element control and verify Reactor level returns to normal.

C. Select the Master Level Controller to MANUAL to lock the level signals at the present setting to prevent any level perturbations and establish stable level control.

D. Continue monitoring Reactor Water Level on P680 and compare with other indications on P601 and the PDS computer and contact I&C.

QUESTION 16 NRC RECORD # WRI 275 ANSWER: D. SYSTEM # C34; N21; K/A 295008 AK1.01: 3.0/3.2 N30 245000 A3.01: 3.6/3.6 LP# GG-1-LP-OP-C3401 259001 K6.07: 3.8/3.8 OBJ. 3f; 3g2; 3i; 22; 23 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

04-1-02-H13-P680 NEW 4A2-A2 & D1 MODIFIED BANK DIFF 2; CA NRC 6/2001 RO SRO BOTH CFR 41.4/41.5 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 17 The Control Room has been abandoned and control has been established at the Remote Shutdown Panels.

Reactor pressure 600 psig Indicated Reactor level at the Remote Shutdown Panel 66 inches With present plant conditions, which one of the following describes Narrow Range Level, Actual Level and the availability of RCIC for level control?

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

NARROW RANGE ACTUAL LEVEL RCIC LEVEL A. 55 inches 52 inches Not available B. 45 inches 43 inches Available C. 48 inches 43 inches Available D. 60 inches 60 inches Not available QUESTION 17 NRC RECORD # WRI 603 ANSWER: A. SYSTEM # C61; B21 K/A 295016 AA2.02: 4.2/4.3 2.1.25: 2.8/3.1 LP# GLP-OPS-C6100 2.4.11: 3.4/3.6 OBJ 19 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

05-1-02-I-1 Att I & II NEW MODIFIED BANK DIFF 2; CA NRC 6/2001 RO SRO BOTH CFR 41.5/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-1-02-II-1 Att. I & II

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 18 The plant is operating at 100 % power.

RCIC is to be operated for surveillance testing to return the system to operable.

Standby Service Water A is operating.

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

A. Suppression Pool Temperature is limited to 95ºF and must be monitored every five minutes while RCIC is operating. If Suppression Pool Temperature exceeds 100ºF Suppression Pool Cooling must be placed in service.

B. Suppression Pool Temperature is limited to 95ºF and must be monitored every five minutes while RCIC is operating. If Suppression Pool Temperature exceeds 95ºF Suppression Pool Cooling must be placed in service and RCIC secured.

C. Suppression Pool Temperature is limited to 100ºF and must be monitored every five minutes while RCIC is operating. If Suppression Pool Temperature exceeds 95ºF Suppression Pool Cooling must be placed in service. RCIC must be secured if Suppression Pool Temperature exceeds 100ºF.

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

QUESTION 18 NRC RECORD # WRI 605 ANSWER: C. SYSTEM # E51; M24 K/A 295013 AA2.01: 3.8/4.0 AA1.02: 3.9/3.9 LP# GG-1-LP-OP-E5100- 2.1.33: 3.4/4.0 OBJ. 12a; 22 2.4.4: 4.0/4.3 LP# GG-1-LP-OP-M7101 OBJ. 23 LP# GG-1-LP-RO-EP03 OBJ. 3; 5 LP# GG-1-LP-OP-M4101 OBJ. 11 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

Tech Spec 3.6.2.1 NEW Tech Spec SR3.6.2.1.1 MODIFIED BANK DIFF 2; CA 04-1-01-E51-1 sect 5.2.1e 05-S-01-EP-3 step 10 - 12 RO SRO BOTH CFR 06-OP-1M24-V-0001 sect 41.5/41.10/43.3/43.5 5.1.2a REFERENCE MATERIAL REQUIRED: Tech Spec 3.6.2.1 & 06-OP-1M24-V-0001

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 19 A LOCA has occurred.

The reactor is depressurized.

The Drywell remains inaccessible due to pressure of 2.5 psig and temperatures of 200ºF.

MSIV Leakage Control Outboard System has been initiated.

Which one of the following identifies the method to monitor radiation released outside Secondary Containment at this time with MSIV Leakage Control in operation?

A. The operating Standby Gas Treatment Train Radiation Monitors will give the indication of radiation release.

B. Auxiliary Building Fuel Handling Area Exhaust Radiation Monitors will give indication of radiation release.

C. Auxiliary Building Fuel Handling Area Exhaust and Fuel Pool Sweep Exhaust Radiation Monitors will give indication of radiation release.

D. Auxiliary Building Fuel Handling Area Exhaust and Standby Gas Treatment Radiation Monitors will give indication of radiation release.

QUESTION 19 NRC RECORD # WRI 606 ANSWER: A. SYSTEM # T48; K/A 295017 AA1.08: 3.1/3.4 E32; D17 AA1.09: 3.6/3.8 LP# GG-1-LP-OP-E3200 239003 K1.02: 2.9/3.0 OBJ. 12c; 13 LP# GG-1-LP-OP-D1721 OBJ. 2 LP# GG-1-LP-OP-T4801 OBJ. 9f SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

04-1-01-E32-1 sect 3.2 NEW sect 3.2; 5.1.1c; 5.2.1c MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR 41.11/41.13 REFERENCE MATERIAL REQUIRED: NONE 43.4

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 20 The plant is operating at 100 % power.

The Component Cooling Water temperature control valve closes to 25% in response to a temperature controller malfunction and the valve is unable to be reopened.

CCW temperatures have risen and continue to rise slowly.

Recirculation Pumps A and B pump bearing temperatures are in alarm on H13-P614.

Which one of the following describes the foremost actions to be taken for these conditions?

Loss of Component Cooling Water ONEP is provided.

A. Immediately scram the reactor and manually trip both Recirculation Pumps.

B. Start the standby CCW pump. Trip the RWCU pumps then close CCW to the Non-Regenerative Heat Exchangers, P42-F103. Reduce core flow to 60% only.

C. Trip the RWCU pumps then close CCW to the Non-Regenerative Heat Exchangers, P42-F103. Isolate CCW to Fuel Pool Heat Exchangers by closing P42-F105 and F205 only.

D. Start the standby CCW pump. Trip the RWCU pumps then close CCW to the Non-Regenerative Heat Exchangers, P42-F103. Isolate CCW to Fuel Pool Heat Exchangers by closing P42-F105 and F205. Reduce core flow to 60% only.

QUESTION 20 NRC RECORD # WRI 314 ANSWER: A. SYSTEM # P42; K/A 295018 AK3.03: 3.1/3.3 ONEPs LP# GLP-OPS-ONEP OBJ. 1; 2; 34 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

05-1-02-V-1 sect 3.1 NEW Section 3.2.2 Note for 2.1 MODIFIED BANK DIFF 1; M NRC 12/2000 RO SRO BOTH CFR 41.4/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-1-02-V-1 w/o Imm.

Actions

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 21 The plant is operating at rated conditions.

Plant Air Dryer A is in service and Dryer B is tagged out for maintenance.

Plant Air Dryer A has undergone an Executed Stop.

Annunciator PLANT AIR DRYR TROUBLE on H13-P870 is in alarm.

Instrument Air Header pressure is 110 psig and stable.

Which one of the following identifies actions to be taken?

A. Immediately scram the reactor in preparation for a complete loss of Instrument Air.

B. Immediately dispatch an operator to isolate Plant Air Dryer A by closing isolation valves P51-F207A and P51-F208A.

C. Dispatch an operator to manually crosstie Service Air via SP52-F010 or SP52-F300 and monitor Instrument Air header pressure.

D. Monitor Instrument Air header pressure and determine the cause of the Executed Stop using the PDS computer and a local operators observations.

QUESTION 21 NRC RECORD # WRI 607 ANSWER: D. SYSTEM # P51; P53 K/A 295019 AK2.14: 3.2/3.2 2.4.31: 3.3/3.4 LP# GLP-OP-P5101 300000 A2.01: 2.9/2.8 OBJ. 2d 2.1.32: 3.4/3.8 LP# GLP-OPS-P5300 OBJ. 20.1; 21 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

04-1-02-H13-P870 7A-E3 NEW 04-1-01-P51-1 MODIFIED BANK DIFF 1; M sect 3.5.4 & 3.7 05-1-02V-9 sect 3.4 RO SRO BOTH CFR 41.4/41.10/43.5 REFERENCE MATERIAL REQUIRED: 04-1-02-H13-P870-7A-E3

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 22 The plant is operating at rated conditions.

A rupture in the Instrument Air header has resulted in Instrument Air header pressure dropping to 40 psig.

Which one of the following describes the response of the Containment Cooling System and Containment temperature?

A. Containment Coolers will operate recirculating the Containment atmosphere maintaining temperature stable.

B. Containment Coolers will operate isolated without cooling water allowing Containment temperatures to rise.

C. Containment Coolers will trip due to a loss of cooling water resulting in Containment temperatures rising.

D. Containment Coolers will trip due to the isolation of the inlet and outlet dampers resulting in Containment temperatures rising.

QUESTION 22 NRC RECORD # WRI 608 ANSWER: B. SYSTEM # P71; K/A 295020 AK2.03: 3.1/3.3 P53; M41 AK3.03: 3.2/3.2 LP# GG-1-LP-OP-M4100 300000 K1.03: 2.8/2.9 OBJ. 10a; 13a LP# GLP-OPS-P7100 OBJ. 16.2 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

E-1213-009 & 012 NEW M-1100B MODIFIED BANK DIFF 1; M M-1109D 05-1-02V-9 sect 5.5 RO SRO BOTH CFR 41.4/41.7/41.9 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 23 The plant is in a startup following a 32 day outage.

MSIVs are closed.

Recirc loop temperatures are at 180 ?F.

Control rods are being withdrawn to achieve criticality. (Minimal decay heat)

Feedwater is operating in long cycle cleanup.

The operating CRD Pump tripped.

What will be the response of the plant?

(ASSUME NO FURTHER OPERATOR ACTIONS)

A. The reactor water level will remain stable at its present level.

B. The reactor water level will rise to the point that a reactor scram is received on High water level.

C. The reactor water level will drop to the point that a reactor scram is received on Low water level.

D. The plant will scram due to a loss of charging water pressure to the Hydraulic Control Units.

QUESTION 23 NRC RECORD # WRI 55 ANSWER: C. SYSTEM # C11-1A; K/A 295022 AK2.04: 2.5/2.7 G33/36; IOI- 1 AK2.05: 2.4/2.5 LP# GLP-OPS-G3336 AA1.04: 2.5/2.6 OBJ 3.3, 8.6, 21 LP# GG-1-LP-OP-C111A OBJ 23 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

03-1-01-1 NEW sect. 2.2.5; 3.3.1d; 3.3.3a MODIFIED BANK DIFF 2; CA NRC 6/2001 RO SRO BOTH CFR 41.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 24 An ATWS has occurred. Reactor pressure is being controlled with SRVs.

Standby Liquid Control has been initiated.

Reactor power has just dropped below 4%.

The following conditions exist:

Reactor Power 2 % and stable Reactor Pressure 1000 psig and stable Reactor Level - 100 inches Fuel Zone and stable Suppression Pool Level 16.5 feet and rising Suppression Pool Temperature 150 °F and rising Drywell Pressure + 1.0 psig and rising Which one of the following describes actions to be taken?

A. Maintain RPV water level between -192 and + 53.5 inches and stabilize RPV pressure < 1064.7 psig.

B. Maintain RPV water level between -192 and + 53.5 inches. Lower RPV pressure to 700 - 900 psig, and initiate SPMU.

C. Terminate and prevent all injection into the RPV except for Boron, CRD and RCIC, and lower RPV water level to the top of active fuel.

D. Terminate and prevent all injection into the RPV except for Boron, CRD and RCIC and emergency depressurize the RPV.

QUESTION 24 NRC RECORD # WRI 301 ANSWER: D. SYSTEM # M41; K/A 295026 EK2.01: 3.9/4.0 B21 LP# GG-1-LP-RO-EP03 OBJ. 2, 3 LP# GG-1-LP-RO-EP02A OBJ. 7, 10h SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

05-S-01-EP2 EP2A NEW Step 33, 14, & 15 MODIFIED BANK DIFF 2, CA 05-S-01-EP3 Step 15 and NRC 12/2000 HCTL RO SRO BOTH CFR 41.7/41.9/

REFERENCE MATERIAL REQUIRED: 05-S-01-EP-2A and EP-3 41.10/41.14/43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION JUNE 2001 REACTOR OPERATOR QUESTION 25 The plant was operating at 100 % Power.

A steam leak has developed in the Containment steam tunnel.

Containment temperature has gone up to 85°F and still rising.

A power reduction has commenced but Containment temperature continues to rise.

Tech. Specs states if Containment temperature exceeds 95°F to restore to < 95°F within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

If Containment temperature is unable to be restored to < 95°F within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; then be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and be in MODE 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Which of the following is the reason for this action?

Tech Specs 3.0 & 3.6.1.5 are provided.

A. Shut down of the Reactor is done to prevent having to initiate Containment Spray to maintain Containment temperature below 185°F.

B. Shut down of the Reactor is done to place the plant in a MODE that the LCO does NOT apply.

C. Shut down of the Reactor is done to prevent having to Emergency Depressurize to maintain Containment temperature below 185°F.

D. Shut down of the Reactor is done to prevent damaging operating equipment inside Containment due to current high temperature.

QUESTION 25 NRC RECORD # WRI 512 ANSWER: B. SYSTEM # M41-1 K/A 295027 K3.03: 3.7/3.7 LP# GG-1-LP-OP-TS001 OBJ. 13 LP# GG-1-LP-OP-M4101 OBJ. 11, 12 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

TECH. SPEC. 3.6.1.5 NEW TECH. SPEC. BASES MODIFIED BANK DIFF 1; M 3.6.1.5 NRC 6/2001 TECH SPEC 3.0.2 ** RO SRO BOTH CFR 41.9/41.10/43.2 REFERENCE MATERIAL REQUIRED: Tech Spec 3.6.1.5 & 3.0.2

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 26 The plant is operating at rated conditions.

Temperatures in the lower CRD Cavity are rising.

Which one of the following describes the expected response of the Drywell Cooling System?

A. At 145ºF, M51-F009 and M51-F016 open and M51-F006 and M51-F014 close to divert Drywell Cooling flow from the lower Drywell area to the CRD Cavity.

B. At 145ºF, the Standby fans on M51-B001 and B002 will auto start and M51-F009 and M51-F016 open raising the flow of air to the CRD Cavity.

C. At 145ºF, M51-F009 and M51-F016 open raising the flow of air to the CRD Cavity and at 155ºF M51-F006 and M51-F014 close to divert more air flow.

D. At 155ºF, the Standby fans on M51-B001 and B002 will auto start and M51-F009 and M51-F016 open raising the flow of air to the CRD Cavity.

QUESTION 26 NRC RECORD # WRI 609 ANSWER: A. SYSTEM # M51 K/A 295028 EA1.03: 3.9/3.9 EK2.04: 3.6/3.6 LP# GG-1-LP-OP-M5100 EK3.04: 3.6/3.8 OBJ. 6a SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

04-1-02-H13-P870 NEW 4A-H2; 10A-H2 MODIFIED BANK DIFF 2; CA M-1101 E-1214-001; 004; 009 RO SRO BOTH CFR 41.4/41.7 REFERENCE MATERIAL REQUIRED: Drywell Cooling System Drawing

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 27 Which of the following is the basis for Emergency Reactor Pressure Vessel (RPV)

Depressurization when Suppression Pool Level CANNOT be maintained below 24.4 feet?

A. 24.4 feet is the highest Suppression Pool level at which the pressure suppression capability of Containment can be maintained.

B. 24.4 feet is the highest Suppression Pool level at which the Suppression Pool will NOT overflow the weir wall resulting in flooding the Drywell.

C. 24.4 feet is the highest Suppression Pool level at which Suppression Pool level instrumentation taps will become covered resulting in loss of ability to monitor Suppression Pool level.

D. 24.4 feet is the highest Suppression Pool level at which opening Safety Relief Valves (SRVs) will NOT result in exceeding the design blowdown rate for the SRV discharge piping.

QUESTION 27 NRC RECORD # WRI 513 ANSWER: A. SYSTEM # M41-1 K/A 295029 K1.01: 3.4/3.7 LP# GG-1-LP-OP-EP03 OBJ. 6 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

GGNS PSTG APP B SP/L-3 NEW GGNS PSTG APP A SP/L-3 MODIFIED BANK DIFF 1; M NRC 6/2001 RO SRO BOTH CFR 41.9/41.10 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 28 The plant is in a LOCA with ECCS systems injecting to the reactor.

Suppression Pool level has lowered to 13.5 feet.

Which one of the following is a condition that exists due to this level?

A. The SRV tailpipe exhaust spiders have been uncovered.

B. The RCIC Turbine Exhaust pipe has been uncovered.

C. Suppression Pool temperature is unable to be determined.

D. Containment Pressure is unable to be determined QUESTION 28 NRC RECORD # WRI 8 ANSWER: C. SYSTEM # E30 K/A 295030 EA2.02: 3.9 LP# GG-1-LP-OP-EP01 OBJ. 5 LP# GG-1-LP-RO-EP03 OBJ. 6 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

05-S-01-EP-3 Caution 2 NEW GGNS EOP PSTG App B MODIFIED BANK DIFF 1; M Caution 2 NRC 6/2001 RO SRO BOTH CFR 41.9/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 29 Preparations are being made to startup following a Refueling Outage.

During the Refueling outage several Control Rod Drive Mechanisms were replaced and the old mechanisms taken to the CRD Maintenance Room for rebuilding.

The Auxiliary Building Operator has reported elevated radiation levels indicated on his dositec in Area 9 185 ft. of 150 mr/hr.

Which one of the following is the possible cause of the elevated radiation levels?

A. This is normal radiation levels for general area Auxiliary Building.

B. Used CRD Drive Mechanisms in the Auxiliary Building C. Waterborne contamination in the Fuel Pool Cooling System D. Normal radiation levels from the Fuel Pool Filter Demineralizers.

QUESTION 29 NRC RECORD # WRI 610 ANSWER: C. SYSTEM # G41/46 K/A 295033 EA2.03: 3.7/3.7 LP# GLP-OPS-G4146 OBJ. 2; 4.8; 5.4; SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 5.5; 5.6; 15.1

REFERENCE:

M-1051A; M-1016 NEW UFSAR 9.1.3.1.2; MODIFIED BANK DIFF 1; M 12.2.1.3.3 & Table 12.2.-12 FPCCU System Figure 4 RO SRO BOTH CFR 41.12/43.4 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 30 A GENERAL EMERGENCY has been declared due to a DBA LOCA.

Plant conditions as follows:

Reactor water level 195

Approximately 15% fuel damage

Large fission product inventory in Containment

All Containment parameters are within design limits

Radiation levels at the site boundary are at 15 mRem/hr TEDE What protective actions, if any, should be recommended?

10-S-01-1 Activation of the Emergency Plan is provided.

A. No evacuation of surrounding areas is required.

B. Evacuate a 2-mile radius and shelter five miles downwind.

C. Evacuate a 2-mile radius and five miles downwind. Shelter the rest of the 10-mile EPZ.

D. Evacuate a 2-mile radius and 10 miles downwind. Shelter the rest of the 10-mile EPZ.

QUESTION 30 NRC RECORD # WRI 611 ANSWER: C. SYSTEM # EPP PARs K/A 295038 EK1.02: 4.2/4.4 LP# GG-1-LP-EP-EPTS6 OBJ 2 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

10-S-01-1 sect. 6.1.4.j(1) NEW MODIFIED BANK DIFF 2; CA EPTS 6 EXAM 2 RO SRO BOTH CFR 41.10/41.12 REFERENCE MATERIAL REQUIRED: 10-S-01-1 43.4/43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 31 A fire has been reported by security at the Hydrogen Bulk Storage Facility.

The Outside Operator has identified the leak as coming from the vent stack on the discharge of the cryogenic compressors.

Which one of the following describes the actions to be taken by operators at the scene?

A. Establish a perimeter of at least 160 feet, and allow the fire to burn itself out.

B. Establish a perimeter of at least 160 feet, and establish a monitor nozzle spraying water on the point where the fire is originating.

C. Evacuate a 2 mile radius of the fire, and establish a monitor nozzle spraying water on the point where the fire is originating.

D. Evacuate a 2 mile radius of the fire, and establish a fire team to enter the area with a fire hose cooling the team and isolate the source of the fire.

QUESTION 31 NRC RECORD # WRI 612 ANSWER: A. SYSTEM # P73 K/A 600000 AK3.04: 2.8/3.4 2.1.32: 3.4/3.8 LP# GG-1-LP-OP-P7300 2.4.25: 2.9/3.4 OBJ 12; 14 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

04-1-01-P73-1 NEW sect 3.1; 3.2; 3.3; 3.7 MODIFIED BANK DIFF 1; M RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 32 The plant is in a Refueling Outage with the reactor disassembled five (5) days after the plant was shutdown in Refueling Outage 11.

Reactor Coolant Temperature is 140 ?F The Fuel shuffle has just begun.

The in-service shutdown cooling pump has just tripped off.

Assume no further operator action.

Determine for this condition:

1. The time to reach 200?F for the Reactor Vessel.

2. The time for level to reach the Top of Active Fuel.

05-1-02-III-1 Attachment I is provided.

A. 1) 0.75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months

2) 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months B. 1) 1.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

2) 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months C. 1) 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

2) 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months D. 1) 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

2) 75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months QUESTION 32 NRC RECORD # WRI 604 ANSWER: C. SYSTEM # G41/46; K/A 295021 AK1.01: 3.6/3.8 ONEP LP# GLP-OPS-ONEP AA2.01: 3.5/3.6 OBJ 18; 19 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 3

REFERENCE:

05-1-02-III-1 Att. I NEW Figures 2 & 5 MODIFIED BANK DIFF 2; CA Pre-shuffle 150ºF NRC 3/1998 WRI37 RO SRO BOTH CFR 41.5/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-1-02-III-1 Att I

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 33 RHR A Pump Room temperature rises to 170 ?F.

Which one of the following identifies the systems or components in addition to RHR A that will be affected by this temperature?

A. HPCS B. MSIVs and RCIC C. RCIC D. RWCU QUESTION 33 NRC RECORD # WRI 048 ANSWER: C. SYSTEM # E12; E51; K/A 219000 A1.08: 3.7/3.6 E31 A2.14: 4.1/4.3 A3.01: 3.3/3.3 LP# GLP-OPS-E3100 A4.06: 3.9/3.7 OBJ 7.4 295032 EK3.03: 3.8/3.9 EK3.07: 3.6/3.8 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 3

REFERENCE:

ARI 04-1-02-H13-P601 NEW 20A-B1 MODIFIED BANK DIFF 1; M 05-1-02-III-5 Group 2, 3, 4 NRC 3/1998 RO SRO BOTH CFR 41.4/41.10/43.5 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 34 Following a Recirc line rupture, reactor level has dropped to - 80 inches.

Both trains of the Standby Gas Treatment System have initiated.

Which of the following best describes the operation of the Standby Gas Treatment System flow control dampers?

A. When 0.2 inches water column is obtained in the Enclosure Building, the steam tunnel cooler dampers throttle to their intermediate position. 90 seconds later the remaining flow control dampers throttle to their intermediate position.

B. When 0.25 inches water column is obtained in the Enclosure Building, the steam tunnel cooler dampers throttle to their intermediate position. 120 seconds later the remaining flow control dampers throttle to their intermediate position.

C. After 90 seconds, the flow control dampers will go to their intermediate positions to maintain ? 0.75 inches water column in the Auxiliary Building and 0.25 inches water column in the Enclosure Building.

D. After 90 seconds the flow control dampers throttle to maintain 0.25 inches water column in the Auxiliary Building. If the Enclosure Building pressure reaches 0.75 inches water column, the flow control dampers go to their intermediate positions.

QUESTION 34 NRC RECORD # WRI 007 ANSWER: A. SYSTEM # T48 K/A 295035 EA1.02: 3.8/3.8 LP# GG-1-LP-OP-T4801 261000 A1.04: 3.0/3.3 OBJ. 8e SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 3

REFERENCE:

E- 1257- 08, 11, 23 NEW MODIFIED BANK DIFF 2; CA NRC 3/1998 RO SRO BOTH CFR 41.4/41.7/41.13 REFERENCE MATERIAL REQUIRED:

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 35 The plant is operating normally at 100 % power.

The Suppression Pool Hi/Lo Level and a LPCS Room Sump Level Hi-Hi annunciators have been received on the H13-P870 panel.

The Control Room Operator has noted that Suppression Pool Level is at 18.4 feet. An operator dispatched to the room reports that water is spraying from the LPCS Suction piping, but he was unable to tell the exact location.

Which one of the following are the appropriate actions for this event?

A. Immediately scram the reactor, initiate Suppression Pool Makeup, and emergency depressurize the plant, and isolate the LPCS Suction from the Suppression Pool.

B. Ensure the LPCS Room sump pumps are operating, isolate LPCS Suction from the Suppression Pool and observe the status of the leak and makeup to the Suppression Pool via normal means, if required open the LPCS Room Door.

C. Monitor and control LPCS Room sump levels, rack out the LPCS Pump Breaker and isolate LPCS Suction from the Suppression Pool, scram the reactor since the Max Safe Level has been reached.

D. Verify the LPCS Room sump pumps are operating, isolate LPCS Suction from the Suppression Pool and rack out the LPCS Pump Breaker, and observe the status of the leak and makeup to the Suppression Pool via normal means.

QUESTION 35 NRC RECORD # WRI 058 ANSWER: D. SYSTEM # P45; E12; K/A 295036 EA2.03: 3.4/3.8 EOP- 4 EK3.03: 3.5/3.6 LP# GG-1-LP-RO-EP03 EA2.02: 3.1/3.1 OBJ 3 LP# GG-1-LP-RO-EP04 OBJ 4 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 3

REFERENCE:

05-S-01-EP-3 step 41 NEW 05-S-01-EP-4 step 9 - 13 MODIFIED BANK DIFF 2; CA ARI 04-1-02-H13-P680 NRC 3/1998 8A1-A4 ARI 04-1-02-H13-P870 RO SRO BOTH CFR 41.4/41.10/43.5 4A-A3; 2A-F1; 4A-C3 REFERENCE MATERIAL REQUIRED: 05-S-01-EP-3 & 4

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 36 A reactor scram has occurred.

All control rods have fully inserted.

Reactor level is stable at + 4 inches narrow range with reactor pressure at rated conditions.

Which one of the following describes the flows observed with present conditions for the Control Rod Hydraulic System?

CRD Pump Min Charging Water Cooling Water Recirc Pump Seal Flow Header Header Purge A. 20 gpm 0 gpm 60 gpm 2 gpm / pump B. 20 gpm 165 gpm 5 gpm 2 gpm / pump C. 20 gpm 165 gpm 5 gpm 0 gpm D. 0 gpm 165 gpm 60 gpm 0 gpm QUESTION 36 NRC RECORD # WRI 613 ANSWER: B. SYSTEM # C11-1A K/A 295036 K5.02: 2.6/2.6 LP# GG-1-LP-OP-C111A OBJ 4 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 1

REFERENCE:

SFD-1081 NEW 04-1-01-B33-1 MODIFIED BANK DIFF 2; CA sect 4.1.2b(2)(e) 04-1-01-C11-1 sect 4.1.2v(2) RO SRO BOTH CFR 41.5/41.6 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 37 The plant is operating at 45% power while returning to full power conditions.

A fatigue failure of the High Pressure Turbine First Stage Pressure connection resulted in a loss of all inputs of pressure to Rod Control and Information System.

Which one of the following describes the ability to move control rods with present plant conditions?

Assume the only actions to move control rods are from H13-P680.

A. Control rod movement is restricted to the Rod Pattern Controller.

B. Control rod movements are unrestricted for both withdrawals and insertions.

C. Control rod movement is restricted to 2 notch withdrawals and unlimited insertions.

D. Control rod movement is restricted to 4 notch withdrawals and unlimited insertions.

QUESTION 37 NRC RECORD # WRI 614 ANSWER: A. SYSTEM # C11-2; K/A 201005 K6.01: 3.2/3.2 N11 LP# GG-1-LP-OP-C1102 OBJ 5, 6, 7, 13b, 23, 26 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

04-1-02-1H13-P680-4A2-D5 NEW 03-1-01-2 MODIFIED BANK DIFF 2; CA sect 2.15; 5.3; 5.7; 6.4 Tech Spec 3.3.2.1 & bases RO SRO BOTH CFR 41.6/43.6 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 38 The plant is operating at 45% power raising power to full power conditions.

A disturbance on the Entergy Grid resulted in a trip of Circuit Breakers J5228 and J5232 in the GGNS Switchyard.

Reactor water level control responded to control reactor water level above -20 inches wide range.

Reactor pressure control responded to an actuation of B21-F051D momentarily with subsequent pressure control using Main Steam Bypass Valves.

All other systems functioned as designed.

Which one of the following identifies the current configuration of the Recirculation Pump circuit breakers?

CB-1 CB-2 CB-3 CB-4 CB-5 A. CLOSED CLOSED CLOSED CLOSED OPEN B. OPEN OPEN CLOSED OPEN OPEN C. CLOSED CLOSED OPEN OPEN OPEN D. OPEN OPEN OPEN OPEN OPEN QUESTION 38 NRC RECORD # WRI 615 ANSWER: C. SYSTEM # B33; N32 K/A 202002 A2.01: 3.4/3.4 LP# GLP-OPS-B3300 202001 A2.15: 3.7/3.9 OBJ 27; 50 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

04-1-02-1H13-P680 NEW 3A-D4 & D10 MODIFIED BANK DIFF 2; CA Tech Spec 3.3.4.1 & bases Tech Spec 3.3.1.1; 3.3.6.5; RO SRO BOTH CFR 41.6 3.3.4.2 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 39 Which one of the following is the reason the LPCI Injection Valves, E12-F042A, B, and C, are designed to remain closed at normal reactor vessel pressure following a LOCA initiation signal?

A. This allows the pump time to pressurize the header, thus minimizing the differential pressure across the injection valve.

B. This ensures reactor pressure has dropped sufficiently to prevent the possibility of over pressurizing low pressure piping.

C. This allows the pump to develop enough discharge head to overcome reactor pressure for injection preventing back flow of hot reactor water into LPCI piping.

D. This ensures reactor pressure has equalized with LPCI pressure to prevent the injection check valves E12-F041A, B, C from slamming the injection piping causing damage.

QUESTION 39 NRC RECORD # WRI 060 ANSWER: B. SYSTEM # E12 K/A 203000 K1.17: 4.0/4.0 K4.01: 4.2/4.2 K4.02: 3.3/3.4 A3.01: 3.8/3.7 A3.08: 4.1/4.1 LP# GLP-OPS-E1200 A4.08: 4.3/4.3 OBJ 8.9; 14.2 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

04-1-01-E12-1 sect. 3.4 NEW Tech Spec Bases B3.3.5.1 MODIFIED BANK DIFF 1; M NRC 6/2001 RO SRO BOTH CFR 41.8 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 40 A LOCA has occurred.

Power to bus 15AA has been lost.

All other systems are functioning as required.

Reactor water level is - 170 inches.

Reactor pressure is 840 psig.

Drywell pressure is 8.3 psig.

Which one of the following identifies the response of the Automatic Depressurization System?

A. ADS valves are unable to open due to a loss of air needed to open the valves.

B. ADS valves are unable to automatically open due to a loss of power needed to actuate.

C. ADS valves will automatically open using both solenoids associated with the air actuator.

D. ADS valves will automatically open when the appropriate time delays have been met.

QUESTION 40 NRC RECORD # WRI 616 ANSWER: D. SYSTEM # E21; E12; K/A 209001 K3.02: 3.8/3.9 E22-2; R21 218000 K6.01: 3.9/4.1 LP# GG-1-LP-OP-E2202 K6.02: 4.1/4.1 OBJ 5c,f; 9a,b; 12b,c; SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1 15; 17; 25

REFERENCE:

04-1-02-1H13-P601 NEW 18A-B3 & E2 MODIFIED BANK DIFF 2; CA 19A-B3 & E2 04-1-01-B21-1 sect 5.1.1 RO SRO BOTH CFR 41.7 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 41 The plant is starting up and is currently operating at 80% power.

All systems are operating properly.

There is a spurious High Pressure Core Spray (HPCS) initiation.

All other systems respond properly.

NO operator action is taken.

Which of the following identifies the effect on Reactor Water Level the spurious HPCS initiation will have?

A. Reactor Water Level will RISE, Feedwater Level Control will respond, and Reactor Water level will stabilize at a HIGHER than normal condition.

B. Reactor Water level will RISE, Feedwater Level Control will respond, and Reactor Water level will be returned to NORMAL level.

C. Reactor Water level will RISE, Feedwater Level Control will respond, and Reactor Water level will stabilize at a LOWER than normal condition.

D. Reactor Water level will not be affected due to Feedwater Level Control will respond and maintain Reactor Water level at NORMAL level.

QUESTION 41 NRC RECORD # WRI 519 ANSWER: A. SYSTEM # C34; E22 K/A 209002 K3.01: 3.9/3.9 LP# GG-1-LP-OP-MCD7b 259002 A2.08: 4.5/4.5 OBJ. 2 A SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

UFSAR 15.5.1.2.1 NEW UFSAR FIG. 15.5-1 MODIFIED BANK DIFF 2; CA NRC 6/2001 RO SRO BOTH CFR 41.7/41.8 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 42 An ATWS has occurred.

Standby Liquid Control has just been initiated for injection into the reactor vessel.

Reactor power is 10 %.

Reactor pressure is 850 psig.

Reactor water level - 110 inches Fuel Zone.

Which one of the following identifies indications of Standby Liquid Control injection into the Reactor?

SLC Tank SLC Pump Reactor Power Squib Valve SLC SQUIB Level Pressure Ready lights OOSVC LOSCONT Alarms OR PWR LOSS lights A. 4400 gal 1050 psig 9% OFF ON ON B. 4600 gal 860 psig 10% ON OFF OFF C. 4600 gal 1050 psig 9% ON OFF OFF D. 4400 gal 860 psig 10% OFF OFF ON QUESTION 42 NRC RECORD # WRI 617 ANSWER: A. SYSTEM # C41 K/A 211000 A4.04: 4.5/4.6 A4.01: 3.9/3.9 A4.03: 4.1/4.1 LP# GLP-OPS-C4100 A4.07: 3.6/3.6 OBJ 12 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

04-1-01-C41-1 sect 5.3.2b NEW MODIFIED BANK DIFF 1; M RO SRO BOTH CFR 41.1/41.6/41.7 REFERENCE MATERIAL REQUIRED: NONE 43.6

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 43 The plant was operating at 60% power when Division I UPS Power Panel 1Y89 trips its incoming circuit breaker.

Which one of the following is the response of the Reactor Protection System?

A. RPS A system tripped resulting in a half scram due to a loss of power to the RPS A logic relays.

B. RPS A system logic energized causing a half scram due to a loss of power to the RPS A Scram Pilot Valve solenoids.

C. RPS A system tripped with no half scram because the RPS A solenoids still have power available.

D. RPS A system logic energized causing alarms indicating the loss of power with no half scram due to RPS Bus A still being energized.

QUESTION 43 NRC RECORD # WRI 102 ANSWER: A. SYSTEM # C71; L62 K/A 212000 K6.01: 3.6 K1.04: 3.4 LP# GLP-OPS-C7100 OBJ. 11.10; 23 LP# GG-1-LP-OP-L6200 OBJ. 6b SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

E- 1026 NEW E- 1173 - 14, 15, & 19 MODIFIED BANK DIFF 1; M NRC 3/1998 RO SRO BOTH CFR 41.2/41.6 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 44 Which one of the following identifies the power sources to the Intermediate Range Neutron Monitors (IRMs) and the impact on the monitors upon a loss of power?

A. IRMs are powered from Division I and II DC buses and fail upscale on a loss of power.

B. IRMs are powered from Division I and II RPS buses and fail upscale on a loss of power.

C. IRMs are powered from Division I and II UPS buses and fail downscale on a loss of power.

D. IRMs are powered from the BOP UPS buses and fail downscale on a loss of power.

QUESTION 44 NRC RECORD # WRI 618 ANSWER: C. SYSTEM # C51-2 K/A 215003 K2.01: 2.5/2.7 LP# GG-1-LP-OP-C5102 OBJ 6b,c; 11a; 18 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 1

REFERENCE:

04-1-01-C51-1 Att III NEW 04-1-01-L62-1 Att I MODIFIED BANK DIFF 1; M 04-1-02-1H13-P680 5A-A8, B8; 7A-A8, B8, B9 RO SRO BOTH CFR 41.6 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 45 GGNS is operating at 10% power with the mode switch in the STARTUP position.

APRM H has four (4) LPRMs currently bypassed.

What would APRM H indicate when the Meter Function switch is taken to COUNT after a fifth (5th)

LPRM is taken to BYPASS?

A. 75 B. 80 C. 85 D. 90 QUESTION 45 NRC RECORD # WRI 326 ANSWER:. C. SYSTEM # C51 K/A 215005 K6.03: 3.1/3.3 LP# GG-1-LP-OP-C5104 OBJ. 3b; 9b SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

06-OP-1C51-V-0003 NEW Att I sect 5.2.1g- i MODIFIED BANK DIFF 2; CA NRC 12/2000 RO SRO BOTH CFR 41.6/41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 46 Which one of the following describes conditions that would prohibit withdrawing Source Range Neutron (SRM) detectors from the core?

A. IRMs are on Range 1 reading between 10 and 25 and all SRMs are reading between 6000 and 9000 cps.

B. Source Range A detector has lost power and has been declared INOP and is bypassed at H13-P680.

C. Under vessel work has been performed during the ongoing refueling outage and work is still in progress.

D. Drywell entry has been made to inspect and identify leaks at rated pressure during a reactor startup.

QUESTION 46 NRC RECORD # WRI 619 ANSWER: C. SYSTEM # C51-1 K/A 215004 2.1.32: 3.4/3.8 LP# GG-1-LP-OP-C5101 A4.04: 3.2/3.2 OBJ 7a; 11a SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

04-1-01-C51-1 sect 3.5 NEW 04-1-02-1H13-P680-7A-C11 MODIFIED BANK DIFF 1; M RO SRO BOTH CFR 41.5/41.6 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 47 Which one of the following conditions will allow a start of a Recirculation Pump A?

A. Steam Dome Pressure 950 psig Recirc Loop A Temperature 485 ?F Recirc Loop B Temperature 529 ?F Bottom Head Temperature 500 ?F Reactor Power 30 %

B. Steam Dome Pressure 1014 psig Recirc Loop A Temperature 495 ?F Recirc Loop B Temperature 529 ?F Bottom Head Temperature 495 ?F Reactor Power 90 %

C. Steam Dome Pressure 981 psig Recirc Loop A Temperature 499 ?F Recirc Loop B Temperature 529 ?F Bottom Head Temperature 500 ?F Reactor Power 60 %

D. Steam Dome Pressure 960 psig Recirc Loop A Temperature 481 ?F Recirc Loop B Temperature 525 ?F Bottom Head Temperature 495 ?F Reactor Power 60 %

QUESTION 47 NRC RECORD # WRI 122 ANSWER: C. SYSTEM # B33; B21 K/A 216000 K1.23: 3.3/3.4 202001 A4.01: 3.7/3.7 LP# GLP-OPS-B3300 OBJ. 26.3 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

06-OP-1B33-V-0005 NEW Data Sheet IV sect. 5.4 MODIFIED BANK DIFF 2; CA 04-1-01-B33-1 sect 3.3 NRC 3/1998 RO SRO BOTH CFR 41.5 REFERENCE MATERIAL REQUIRED: STEAM TABLES

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 48 RCIC was manually initiated for level control following a loss of feedwater.

Reactor level has risen to + 55 inches.

Which one of the following describes the operation of RCIC?

A. RCIC Injection Shutoff valve, E51-F013, will close and RCIC will operate on minimum flow. If Reactor water level drops to < - 41.6 inches, the RCIC Injection Shutoff valve will re-open.

B. RCIC Steam Supply to RCIC Turbine valve, E51-F045, will close securing RCIC.

If Reactor water level drops to < - 41.6 inches the RCIC Steam Supply to RCIC Turbine valve will open and RCIC will re-inject into the Reactor.

C. RCIC Turbine Trip/Throttle valve will close securing RCIC. RCIC will require a manual restart if further operation becomes necessary.

D. RCIC Steam Supply to RCIC Turbine valve, E51-F045, will close securing RCIC.

RCIC Injection Shutoff valve, E51-F013, will remain open. If Reactor water level drops to < - 41.6 inches the RCIC Steam Supply to RCIC Turbine valve will open and RCIC will re-inject into the Reactor.

QUESTION 48 NRC RECORD # WRI 328 ANSWER:. B. SYSTEM # E51 K/A 217000 A1.01: 3.7/3.7 LP# GG-1-LP-OP-E5100 OBJ. 8c, i, k, 19 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

E-1185-02, 06, 15, 34, 35 NEW MODIFIED BANK DIFF 1; M NRC 12/2000 RO SRO BOTH CFR 41.5/41.7/41.10 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 49 A LOCA has occurred. High Pressure Core Spray is inoperable.

ADS Inhibit switches are in INHIBIT.

Drywell pressure is 1.05 psig and rising.

Reactor pressure is 890 psig and falling.

Reactor water level is - 155 inches and stable on wide range indication.

ALL systems are functioning as designed and CRD is maximized.

Which one of the following describes the operation of the Automatic Depressurization System (ADS) valves?

A. ADS valves can ONLY be opened using their handswitches.

B. ADS will automatically initiate after the ADS 105 second timer has timed out.

C. ADS can be manually initiated using the ADS Manual Initiation pushbuttons.

D. ADS will automatically initiate after both the 9.2 minute and 105 second timers have timed out.

QUESTION 49 NRC RECORD # WRI 620 ANSWER: C. SYSTEM # E22-2 K/A 218000 K4.03: 3.8/4.0 LP# GG-1-LP-OP-E2202 K5.01: 3.8/3.8 OBJ. 15 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

04--1-02-1H13-P601 NEW 18A-B3, B4, C2 MODIFIED BANK DIFF 1, M 19A-B3, B4, C2 NRC 12/00 RO SRO BOTH CFR 41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 50 The plant is operating at 100% power steady state.

All power from offsite is lost.

All systems respond and function properly.

All plant parameters remain in their normal band.

Division 1 and 2 Load Shedding and Sequencing (LSS) functions properly.

Which of the following components is without power at this time?

A. Drywell Chillers A.

B. Division 1 Drywell Cooler Fans.

C. Drywell Chillers B.

D. Division 2 Drywell Cooler Fans.

QUESTION 50 NRC RECORD # WRI 528 ANSWER: A. SYSTEM # M51 K/A 223001 K2.09: 2.7/2.9 K2.10: 2.7/2.9 LP# GG-1-LP-OP-M5100 K2.08: 2.7/3.0 OBJ. 7a&c; 9a SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

04-1-01-R21-1 Table 1 NEW 04-1-01-M51-1 Att III MODIFIED BANK DIFF 2; CA 04-1-01-P72-1 Att II NRC 6/2001 RO SRO BOTH CFR 41.7/41.8 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 51 An LOCA has occurred.

Reactor scram immediate actions are complete.

The following conditions exist:

Reactor water level + 20 inches after dropping to - 32 inches Reactor pressure 40 psig Drywell pressure 1.5 psig Which one of the following should be isolated?

A. Reactor Water Cleanup system B. Reactor Sample Isolation valves C. Reactor Core Isolation Cooling Vacuum Breaker valves D. Main Steam Line Drain Isolation valves B21-F016 & F019 QUESTION 51 NRC RECORD # WRI 385 ANSWER: C. SYSTEM # E51 K/A 223002 A2.09: 3.6/3.7 2.4.4: 4.0/4.3 LP# GG-1-LP-OP-E5100 2.4.21: 3.7/4.3 OBJ. 8o SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

05-1-02-III-5 Group 9 NEW MODIFIED BANK DIFF 1; M NRC 12/2000 RO SRO BOTH CFR 41.7/41.9 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 52 The reactor has scrammed.

A loss of offsite power (LOP) and loss of coolant accident (LOCA) signal were received by LSS eight (8) hours ago.

The ESF buses were restored by their respective diesel generators.

When the handswitch for SRV B21-F051F was taken to OPEN, the valve did NOT change position.

Instrument air system header pressure and ADS receiver pressure indicates 0 psig.

Which one of the following correctly describes a method to allow further operation of this SRV?

A. The SRV can be opened after installing nitrogen bottles in area 9, 139 ft elevation and pressurizing the ADS air header.

B. The SRV can be opened by placing the A solenoid (H13-P601) and B solenoid (H13-P631) handswitches to OPEN.

C. The SRV can be opened by placing the handswitch on Division I or II Remote Shutdown Panels to OPEN.

D. Due to the loss of instrument air, this SRV will open only in the Safety function.

QUESTION 52 NRC RECORD # WRI 337 ANSWER: D. SYSTEM # E22-2 K/A 239002 K1.05: 3.1/3.3 LP# GG-1-LP-OP E2202 OBJ. 7; 11; 13; 19d; 28 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

05-1-02-V-9 sect 3.15 NEW 04-1-01-B21-1 MODIFIED BANK DIFF 2; CA Sect 4.2.2 NRC 12/2000 M-1077C & E RO SRO BOTH CFR 41.3 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 53 The plant is operating at 45% power.

An incident at the Front Standard of the Main Turbine resulted in a local manual trip of the Main Turbine.

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

A. The reactor will scram on Turbine Valve fluid pressure and the Turbine Bypass Valves will open.

B. The reactor will scram on high reactor flux and the Turbine Bypass Valves will open.

C. The reactor will scram on Turbine Valve fluid pressure, the Turbine Bypass Valves will open, and 9 Safety Relief Valves will open.

D. The reactor will scram on high reactor flux, the Turbine Bypass Valves will open, and Safety Relief Valves will open.

QUESTION 53 NRC RECORD # WRI 244 ANSWER: A. SYSTEM # N32; C71 K/A 241000 K6.11: 3.4/3.4 A1.01: 3.9/3.8 A1.02: 4.1/3.9 LP# GLP-OP-C7100 A1.07: 3.8/3.7 OBJ. 9, 10; 23 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

FSAR Table 15.2-4 NEW Tech Spec 3.3.1.1 & bases MODIFIED BANK DIFF 2; CA FSAR Table 15.2-2 NRC 4/2000 RO SRO BOTH CFR 41.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 54 Which one of the following signals will cause a SBGT system 'A' initiation?

A. Fuel handling area exhaust rad monitor 'A' indicating 4.2 mRem/hr.

Fuel handling area exhaust rad monitor 'D' mode switch NOT in operate.

B. Fuel pool sweep exhaust rad monitor 'A' indicating 32 mRem/hr.

Fuel pool sweep exhaust rad monitor 'C' indicating 34 mRem/hr.

C. Fuel handling area exhaust rad monitor 'B' indicating 5.3 mRem/hr.

Fuel handling area exhaust rad monitor 'C' inop trip.

D. Fuel pool sweep exhaust rad monitor 'A' indicating 31 mRem/hr.

Fuel pool sweep exhaust rad monitor 'C' inop trip.

QUESTION 54 NRC RECORD # WRI 621 ANSWER: A. SYSTEM # T48; K/A 261000 K4.01: 3.7/3.8 D17; T42 LP# GG-1-LP-OP-T4801 OBJ. 8f SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

04-1-02-1H13-P870 2A-A3 NEW 04-1-02-1H13-P601 MODIFIED BANK DIFF 1, M 19A-B9 LOT 9/99 ESF TRM table 3.3.6.2-1 RO SRO BOTH CFR 41.7/41.11 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 55 A plant startup is in progress.

The Operator at the Controls has just shifted the A Recirculation Pump to fast speed. The B Recirculation pump is running in slow speed with its flow control valve at 100% open.

Immediately after the pump up-shift, the following indications were received in the Main Control Room:

Reactor Power 34 % and stable.

Reactor level dropped to + 32 inches.

Annunciator RECIRC FCV A PARTIAL CLOSE/ RFP TRIP (P680-3A-D1) is illuminated.

Which one of the following would be the expected response of the Recirculation System?

(No other alarms or indicating lights have been received.)

A. The A Recirc Flow Control Valve will remain at present position and will require resetting via the RECIRC PUMP A CAV INTLK RESET pushbutton.

B. The A Recirc Flow Control Valve Hydraulic Power Unit will require resetting from the Control Room Back Panels and then the valve opened to 15 - 20 % valve position.

C. The A Recirc Flow Control Valve runback to 0 % valve position and B Recirc Flow Control valve will runback to 15 - 20 % valve position and then both valves will be reset via the RECIRC PUMP A CAV INTLK RESET pushbutton.

D. The A Recirc Flow Control Valve will remain at present position and B Recirc Flow Control Valve will runback to 15 - 20 % valve position and then both valves will be reset via the RECIRC PUMP A CAV INTLK RESET pushbutton QUESTION 55 NRC RECORD # WRI 235 ANSWER: A. SYSTEM # B33 K/A 202002 A3.01: 3.6/3.4 LP# GLP-OPS-B3300 OBJ. 20, 24, 50 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

04-1-02-H13-P680 NEW 3A-D1 MODIFIED BANK DIFF 1; M 04-1-01-B33-1 section 6.6 NRC 4/2000 RO SRO BOTH CFR 41.6 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 56 The plant was operating at full power when a leak on the RFPT A lube oil reservoir resulted in a trip of both AC RFPT Lube Oil Pumps.

The ACRO noted bearing oil pressure at 2 psig.

Which one of the following describes the response of the Feedwater System and the ability to inject water to the Reactor?

A. RFPT A will continue to operate using the Emergency RFPT Lube Oil Pump.

Reactor water level will continue to be maintained by Feedwater.

B. RFPT A will trip and isolate the RFPT and power will be reduced with Recirc to within the capabilities of RFPT B allowing reactor level to be recovered to close to normal levels.

C. RFPT A will trip and isolate feedwater to the reactor vessel resulting in a lowering reactor water level and eventual reactor scram. Water level must be recovered using ECCS and RCIC.

D. RFPT A will trip and power will be reduced with Recirc to within the capabilities of RFPT B with level being maintained at 18 inches until the Digital Feedwater Control System can be restored to three element control.

QUESTION 56 NRC RECORD # WRI 622 ANSWER: B. SYSTEM # N21; B33 K/A 259001 A2.01: 3.7/3.7 K1.11: 2.7/2.7 K4.06: 2.5/2.6 LP# GG-1-LP-OP-N2100 K6.09: 2.8/2.9 OBJ. 12; 14; 16; 37 LP# GLP-OPS-B3300 OBJ. 24.1; 24.2; 47 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 1

REFERENCE:

04-1-02-1H13-P680 NEW 2A-A2, B1, B2; 3A-A3 MODIFIED BANK DIFF 2; CA 04-1-01-N21-1 sect 3.7 RO SRO BOTH CFR 41.4 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 57 The plant was operating at full power.

The variable leg of the transmitters off of D004C Condensing pot has broken and drained the water from the associated level transmitters.

The Digital Feed Control System was aligned for normal operations.

Which one of the following describes the response of the Digital Feed Control System (DFCS) and actual Reactor Water Level?

A. Reactor water level will drop initially then is restored when DFCS automatically transfers from Three Element Control to Single Element Control.

B. Reactor water level will drop initially then is restored when the DFCS automatically deselects C Narrow Range and is replaced with Upset Range for level control.

C. Reactor water level will rise initially then is restored when the DFCS automatically deselects C Narrow Range and is replaced with Upset Range for level control.

D. Reactor water level will remain stable and the DFCS automatically deselects C Narrow Range and is replaced with Upset Range for level control.

QUESTION 57 NRC RECORD # WRI 623 ANSWER: D. SYSTEM # N21; K/A 259002 K1.03: 3.8/3.9 C34; B21 LP# GG-1-LP-OP-C3401 OBJ. 3f; 22; 25 LP# GLP-OPS-B2101 OBJ. 8.1; 8.4; 20 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

04-1-02-1H13-P680 NEW 2A-C9 MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR 41.4/41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 58 The Electrical line up is normal.

A LOCA condition has caused Drywell Pressure to rise to 1.6 psig.

A switching error causes 500 kV voltage to drop.

The voltage to ALL ESF busses drops to 3290 volts.

The voltage transient duration is 10 seconds and then voltage returns to normal.

Which one of the following statements is the condition of the ESF busses after this voltage transient?

A. 15AA is being supplied from ESF 11 16AB is being supplied from ESF 21 17AC is being supplied from ESF 21 B. 15AA is being supplied from Div I D/G 16AB is being supplied from Div II D/G 17AC is being supplied from Div III D/G C. 15AA is being supplied from ESF 11 16AB is being supplied from ESF 21 17AC is being supplied from Div III D/G D. 15AA is being supplied from Div I D/G 16AB is being supplied from Div II D/G 17AC is being supplied from ESF 21 QUESTION 58 NRC RECORD # WRI 11 ANSWER: B. SYSTEM # R21 K/A 264000 2.4.4: 4.0/4.3 LP# GG-1-LP-OP-R2100 K4.05: 3.2/3.2 A3.05: 3.4/3.4 OBJ. 12; 20; 22; 37 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1

REFERENCE:

04-1-01-R21-1 sect 5.1.1a NEW 04-1-01-P81-1 sect 3.22 MODIFIED BANK DIFF 2; CA NRC 6/2001 RO SRO BOTH CFR 41.8 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 59 Select the statement that describes the MOST probable cause of the following plant conditions:

Annunciator RECIRC PMP B SEAL STG FLO HI/LO alarms.

Annunciator RECIRC PMP B OUTR SEAL LEAK HI alarms.

Recirc pump B # 1 seal cavity pressure: 1020 psig.

Recirc pump B # 2 seal cavity pressure: 100 psig A. Failure of the # 1 seal.

B. Failure of the # 2 seal.

C. Failure of the CRD seal purge regulator.

D. Plugging of the orifice between # 1 and # 2 seals.

QUESTION 59 NRC RECORD # WRI 540 ANSWER: B. SYSTEM # B33 K/A 202001 A2.10: 3.5/3.9 LP# GLP-OPS-B3300 A1.09: 3.3/3.3 A1.10: 2.6/2.7 OBJ. 29.4 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2

REFERENCE:

04-1-02-1H13-P680 NEW 3A-A12 & 3A-B11 MODIFIED BANK DIFF 2; CA NRC 6/2001 RO SRO BOTH CFR 41.3/41.5 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 60 The plant is in a refuel outage.

Reactor Water Clean-Up (RWCU) is operating.

Residual Heat Removal (RHR) B is in Shutdown Cooling.

E12-F048B RHR B Heat Exchanger Bypass valve is FULL OPEN.

E12-F003B RHR B Heat Exchanger Outlet valve is FULL CLOSED.

Which of the following would be a valid indication of Reactor Coolant Temperature under present plant conditions?

P & IDs M-1079 and M-1085A are provided.

A. RHR B heat exchanger B001B inlet temperature E12 TE-N004B B. RHR B heat exchanger B002B inlet temperature E12 TE-N002B.

C. RHR B heat exchanger discharge temperature E12 TE-N027B.

D. RWCU Non-Regen heat exchanger inlet temperature G33 TE-N006.

QUESTION 60 NRC RECORD # WRI 541 ANSWER: C. SYSTEM # E12 K/A 205000 K1.03: 3.4/3.5 LP# GLP-OPS-E1200 OBJ. 14 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2

REFERENCE:

04-1-01-E12-1 NEW sect 4.2.2.e.13 Caution MODIFIED BANK DIFF 2; CA P&ID M1085A NRC 6/2001 M-1079 RO SRO BOTH CFR 41.2/41.3/41.4 REFERENCE MATERIAL REQUIRED: M-1079 & M-1085A 41.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 61 Suppression Pool Cooling is in service to support a RCIC surveillance.

Which of the following statements accurately depicts the required operation of E12-F003A(B), RHR Heat Exchanger Outlet Valve, and E12-F048A(B), RHR Heat Exchanger Bypass Valve, while in Suppression Pool cooling?

A. Maintain flow greater than 4000 gpm. If the F048A(B) is not full OPEN, avoid using F003A(B) to throttle flow for extended periods (30 minutes) in the 0% to 15% open range.

B. Maintain flow greater than 4000 gpm. If the F048A(B) is not full CLOSED, avoid using F003A(B) to throttle flow for extended periods (60 minutes) in the 0% to 15% open range.

C. Maintain heat exchanger flow less than 8600 gpm. If the F048A(B) is not full OPEN, avoid using F003A(B) to throttle flow for extended periods (30 minutes) in the 0% to 15% open range.

D. Maintain heat exchanger flow less than 8600 gpm. If the F048A(B) is not full CLOSED, avoid using F003A(B) to throttle flow for extended periods (60 minutes) in the 0% to 15% open range.

QUESTION 61 NRC RECORD # WRI 624 ANSWER: A. SYSTEM # E12 K/A 219000 K5.04: 2.9/2.9 LP# GLP-OPS-E1200 2.1.32: 3.4/3.8 OBJ. 14.1 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2

REFERENCE:

04-1-01-E12-1 NEW sect 3.2.3; 3.6.3b; MODIFIED BANK DIFF 1; M 5.2.2a(5) Caution NRC 12/00 id 282 RO SRO BOTH CFR 41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 62 In which one of the following situations would containment spray be in service?

A. 5 minutes since LOCA signal received Reactor water level - 192 inches RHR A and B pumps operating on minimum flow Drywell pressure 2 psig CTMT pressure 9 psig B. 12 minutes since LOCA signal received Reactor water level - 192 inches RHR A and B pumps operating on minimum flow Drywell pressure 0.7 psig CTMT pressure 8 psig C. 13 minutes since LOCA signal received Reactor water level - 92 inches RHR A and B pumps were overridden off 2 min. after LOCA Drywell pressure 5 psig CTMT pressure 4 psig D. 15 minutes since LOCA signal received Reactor water level - 92 inches RHR A and B pumps were overridden off 2 min. after LOCA Drywell pressure 3 psig CTMT pressure 8 psig QUESTION 62 NRC RECORD # WRI 625 ANSWER: D. SYSTEM # E12 K/A 226001 A3.07: 3.5/3.5 LP# GLP-OPS-E1200 A3.01: 3.0/3.0 OBJ. 10.1; 11.2 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 2

REFERENCE:

04-1-01-E12-1 sect 3.3 NEW 04-1-02-1H13-P601 MODIFIED BANK DIFF 1; M 17A-F3; 20A-B6 LOT 7/95 RO SRO BOTH CFR 41.7/41.8 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 63 GGNS Main Generator has a limit to carry no more than ? 250 MVARs.

What is the basis for this limitation?

A. This is the Maximum reactive load allowed by the manufacturer due to the heat build up in the stator windings at full power.

B. GGNS is a base load station such that Entergy dispatchers are required to minimize the reactive load carried on the Main Generator.

C. GGNS Main Generator reverse power relays will not recognize a reverse power condition at high reactive load and will not provide the required trip.

D. The Generator V-Curves supplied by the manufacturer limit the power factor on the generator to reduce hysteresis losses.

QUESTION 63 NRC RECORD # WRI 44 ANSWER: C. SYSTEM # N41 K/A 245000 A4.14: 2.5/2.5 A3.10: 2.5/2.6 LP# GG-1-LP-OP-N4151 K4.06: 2.7/2.8 OBJ 7; 13 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2

REFERENCE:

04-1-01-N40-1 sect. 3.8 NEW MODIFIED BANK DIFF 1; M NRC 3/1998 RO SRO BOTH CFR 41.4/41.10/43.5 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 64 The plant is operating normally at full power rated conditions.

A rupture of tubes in the 3A Low Pressure Feedwater Heater resulted in an automatic isolation of the Feedwater Heater String.

Which one of the following describes the limitations on plant operations?

A. Power is limited to a maximum of 50% rated thermal power utilizing Condensate and Heater Drain pumps to a limit of 250 psid differential pressure across the LP Feedwater Heaters due to 1/3 Condensate System capacity.

B. Power is limited to a maximum of 75% rated thermal power without restrictions on the use of Condensate and Heater Drain pumps.

C. Power is limited to a maximum of 100% rated thermal power without restrictions on the use of Condensate and Heater Drain pumps opening the LP Feedwater Heater Bypass valve as necessary to reduce differential pressure.

D. Power is limited as necessary to maintain the turbine parameters within limits without restrictions on the operation of the Condensate and Heater Drain Pumps.

QUESTION 64 NRC RECORD # WRI 294 ANSWER: D. SYSTEM # N23; N19 K/A 256000 A3.07: 2.9/2.9 A3.04: 3.0/3.0 A3.01: 2.7/2.7 LP# GLP-OPS-N2335 A2.08: 3.1/3.1 OBJ. 15 SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 2

REFERENCE:

04-1-01-N23-1 sect 3.10 NEW 03-1-01-2 sect 2.10.2 MODIFIED BANK DIFF 1; M NRC 4/2000 RO SRO BOTH CFR 41.4 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 65 Bus 16AB has experienced an undervoltage condition.

Diesel Generator 12 failed to start.

Power to bus 16AB has been restored by the Control Room Operator from transformer ESF 11.

Reactor water level and pressure are stable.

Which one of the following is the type of load sequencing that will take place?

A. Bus Undervoltage Sequence B. Loss of Offsite Power Sequence C. Loss of Coolant Accident Sequence D. Bus Undervoltage Sequence following automatic source selection QUESTION 65 NRC RECORD # WRI 626 ANSWER: A. SYSTEM # R21 K/A 262001 A3.04: 3.4/3.6 LP# GG-1-LP-OP-R2100 OBJ. 15; 16; 17; 18 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 2

REFERENCE:

E-1039 NEW MODIFIED BANK DIFF 1; M LOT 8/02 RO SRO BOTH CFR 41.4/41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 66 Static inverter 1Y95 has automatically transferred to its alternate power source because of a fault on its normal power source.

Two hours later, the electricians have repaired the fault and the normal power source for 1Y95 is re-energized.

Which one of the following statements describes the restoration of the inverter to its NORMAL source?

A. The inverter static switch can be manually transferred back to the normal power source, only if the power sources are IN SYNC.

B. The inverter static switch will automatically transfer back to the normal power source, only if the power sources are IN SYNC.

C. The inverter static switch will automatically transfer back to the normal power source, regardless of whether the power sources are IN SYNC.

D. The inverter static switch can be manually transferred back to the normal power source, regardless of whether the power sources are IN SYNC.

QUESTION 66 NRC RECORD # WRI 544 ANSWER: A. SYSTEM # L62 K/A 262002 A3.01: 2.8/3.1 LP# GG-1-LP-OP-L6200 OBJ. 4b; 10a SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2

REFERENCE:

04-1-01-L62-1 NEW sect 3.2 & 3.5 MODIFIED BANK DIFF 1; M NRC 6/2001 RO SRO BOTH CFR 41.7/41.10/43.5 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 67 The plant was operating at 100 % power when Control Room Operators rapidly reduced power due to changes in feedwater heating.

Annunciator MSL RAD HI was received.

No other radiation alarms were received and indications on the Main Steam Line Radiation Monitors are elevated.

All other radiation levels have dropped from previous readings.

Which one of the following is the most likely cause of these conditions?

A. Fuel cladding failure B. Resin intrusion into the reactor C. Hydrogen water chemistry is in service D. Release of crud containing Co-60 into the reactor QUESTION 67 NRC RECORD # WRI 627 ANSWER: C. SYSTEM # P73; D17 K/A 272000 A1.01: 3.2/3.2 LP# GG-1-LP-OP-P7300 OBJ. 8 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2

REFERENCE:

04-1-02-1H13-P601 NEW 19A-D4; 18A-D4 MODIFIED BANK DIFF 1; M RO SRO BOTH CFR 41.10/41.11 REFERENCE MATERIAL REQUIRED: None 43.4/43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 68 The Plant is operating at 100 % power.

The Motor Driven Fire pump is out of service.

A fire in Transformer ESF 12 has initiated the Deluge system for the transformer.

The A Diesel Driven Fire Pump received a signal to start.

Which one of the following describes the starting limitations of the Diesel Driven Fire Pump?

A. The diesel engine will attempt to start for 15 minutes. If the diesel does not start it alarms in the Control Room, and it must be reset from the Control Room before it will attempt to start again.

B. The diesel engine will attempt to start for 15 seconds, then wait for 15 seconds. It will attempt this start sequence for 6 attempts. After that, it must manually be reset before any further start attempts occur.

C. The diesel engine will attempt to start for 15 seconds then wait for 15 minutes to allow the battery to recharge, then it will attempt this cycle again. After that it must manually be reset before any further start attempts occur.

D. The diesel engine will attempt to start as long as air pressure is > 60 psig. After that, the air bank must recharge before additional start attempts can occur.

QUESTION 68 NRC RECORD # WRI 3 ANSWER: B. SYSTEM # P64 K/A 286000 A2.08: 3.2/3.3 K5.05: 3.0/3.1 K4.07: 3.3/3.3 A3.01: 3.4/3.4 LP# GG-1-LP-OP-P6400 A4.06: 3.4/3.4 OBJ. 6d SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2

REFERENCE:

ARI 04-S-02-SH13-P862 NEW 1A-B3; 1A-B5 MODIFIED BANK DIFF 1; M sect. 1.2; 2.1; & 4.5 NRC 3/1998 RO SRO BOTH CFR 41.4 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 69 A station blackout has occurred.

A fire has broken out in the Division II ESF Switchgear Room on 119 ft elevation area 10.

Which one of the following describes the ability to combat the fire?

A. Fire fighting will be limited to the use of portable fire extinguishers.

B. The CO2 fire suppression system can be overridden open and the Auxiliary Building Isolation Valves opened using the AUX BLDG ISO BYPASS Switch.

C. The Fire Water System Auxiliary Building Isolation Valves can be opened using the AUX BLDG ISO BYPASS Switch to provide fire water to hoses.

D. The Fire Water System Auxiliary Building Isolation Valves can be bypassed by manually opening the motor operated bypass valves.

QUESTION 69 NRC RECORD # WRI 264 ANSWER: D. SYSTEM # T10; K/A 290001 K6.09: 3.4/3.6 P64; M71; R21 A2.06: 3.7/4.0 LP# GG-1-LP-OP-M7101 286000 A2.09: 2.7/2.8 OBJ. 8a; 10; 16d,e,f; 20; 29 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 2

REFERENCE:

05-1-02-V-9 NEW Sect 3.19 & 5.46 MODIFIED BANK DIFF 2; CA 05-1-02-III-5 NRC 4/2000 sect 3.4.4 M-0035E RO SRO BOTH CFR 41.9 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 70 An accident has occurred and a radiological release is in progress.

RCIC was manually initiated to maintain reactor water level.

The Main Steam failed to completely isolate upon a manual isolation when a steam leak was reported in the Turbine Building.

Annunciator CONT RM VENT RAD HI has been received.

There are various Turbine Building radiation alarms.

No other radiation alarms were received.

Which one of the following describes the response of the Control Room HVAC System (Z51)?

(ASSUME NO OPERATOR ACTIONS)

A. Control Room HVAC is operating in a normal configuration.

B. Control Room HVAC has isolated and the Control Room Standby Fresh Air Unit associated with the in service Control Room Air Conditioning Unit will automatically start.

C. Both Control Room Standby Fresh Air Units will automatically start to filter Control Room air with the remaining Control Room HVAC components continuing to operate as normal.

D. Control Room HVAC will isolate and both Control Room Standby Fresh Air Units will initiate to filter Control Room air with the in service Control Room Air Conditioning Unit remaining in operation.

QUESTION 70 NRC RECORD # WRI 628 ANSWER: A. SYSTEM # Z51; D17 K/A 290003 A1.05: 3.2/3.3 LP# GG-1-LP-OP-Z5100 OBJ. 8; 9; 11 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2

REFERENCE:

04-1-02-1H13-P601 NEW 19A-A9; A10; A11 MODIFIED BANK DIFF 2; CA 04-S-02-SH13-P855 1A-A5; 2A-A5 RO SRO BOTH CFR 41.7/41.11/43.4 TRM Table 6.3.1-1 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 71 The operating Instrument Air Compressor has tripped.

The Standby Instrument Air Compressor fails to start.

Which one of the following describes the response of the Instrument/Plant/Service Air Systems?

A. Instrument Air pressure will continue to drop until an operator manually cross ties Service Air to the Instrument Air header down stream of the Instrument Air Dryers.

B. Instrument Air pressure will continue to drop until an operator manually cross ties Service Air to the Instrument Air header up stream of the Instrument Air Dryers.

C. Instrument Air pressure will continue to drop until the Service to Instrument Air cross tie automatically opens to connect to the Instrument Air header down stream of the Instrument Air Dryers.

D. Instrument Air pressure will continue to drop until the Service to Instrument Air cross tie automatically opens to connect to the Instrument Air header up stream of the Instrument Air Dryers.

QUESTION 71 NRC RECORD # WRI 629 ANSWER: D. SYSTEM # P51; K/A 300000 K4.02: 3.0/3.0 P52; P53 LP# GG-1-LP-OP-P5200 OBJ. 3; 4 LP# GLP-OPS-P5300 OBJ. 3; 4 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2

REFERENCE:

05-1-02-V-9 sect 3.1.1 NEW M-1067A MODIFIED BANK DIFF 1; M M-1068D M-1126 RO SRO BOTH CFR 41.4/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 72 The plant is operating in the normal electrical lineup.

CCW pumps A and C are operating with B selected for STANDBY.

A Loss of Coolant Accident occurs resulting in a shedding of loads.

The C CCW pump trips on overcurrent.

Which of the following describes the resulting status of the CCW system?

(Assume NO operator action.)

A. Pump A operating; Pump B operating; Pump C tripped B. Pump A operating; Pump B not operating; Pump C tripped C. Pump A not operating; Pump B operating; Pump C tripped D. Pump A not operating; Pump B not operating; Pump C tripped QUESTION 72 NRC RECORD # WRI 105 ANSWER: B. SYSTEM # P42 K/A 400000 K4.01: 3.4/3.9 LP# GG-1-LP-OP-P4200 OBJ 5; 7a; 11h; 23 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2

REFERENCE:

04-1-01-R21-1 Table 1 NEW MODIFIED BANK DIFF 1; M NRC 3/1998 RO SRO BOTH CFR 41.4 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 73 Radwaste is discharging the Floor Drain Sample Tank to the River.

Which one of the following would result in an isolation of the G17-F355 Liquid Radwaste Discharge Isolation Valve?

A. The Floor Drain Sample Pump discharge pressure is too low.

B. The blow down flow rate is too low.

C. The discharge flow rate is too low.

D. The effluent radiation monitor high radiation setpoint is reached.

QUESTION 73 NRC RECORD # WRI 110 ANSWER: B. SYSTEM # G17 K/A 268000 A1.02: 2.6/3.6 LP# GG-1-LP-OP-G1718 OBJ 6a, 7h, l SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 3

REFERENCE:

ARI 04-1-02-H13-P601 NEW 19A-H7; 19A-H8 MODIFIED BANK DIFF 1; M ARI 04-1-02-H13-P870 NRC 3/1998 6A-F3 RO SRO BOTH CFR 41.13/43.4 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 74 A rise in Fuel Handling Area Exhaust Radiation levels have resulted in an actuation of Standby Gas Treatment. Standby Gas Treatment A and B initiated, but Standby Gas Treatment B Exhaust and Enclosure Building fans failed to start.

Which of the following identifies the condition of the Auxiliary Building Ventilation System?

A. All fan coil units operating, Div I and Div II Secondary Containment Isolation dampers open.

B. All fan coil units shutdown, the Division I Secondary Containment Isolation dampers closed.

C. All fan coil units shutdown, Div I and Div II Secondary Containment Isolation dampers closed.

D. Div I fan coil units shutdown, the Division I Secondary Containment Isolation dampers closed.

QUESTION 74 NRC RECORD # WRI 630 ANSWER: C. SYSTEM # T41; K/A 288000 K1.05: 3.3/3.6 D17; T48 K1.02: 3.4/3.4 LP# GG-1-LP-OP-T4100 OBJ. 7; 9; 10; 11 SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 3

REFERENCE:

05-1-02-III-5 Check list NEW E-1257-01 MODIFIED BANK DIFF 2; CA E-1253-01, 02, 06, 07, 12 LOT 9/99 Vent RO SRO BOTH CFR 41.7/41.11 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 75 The plant is operating at rated conditions.

The following are the parameters taken from the recent Mon Edit.

MFLCPR 0.94 MAPRAT 1.02 MFLPD 0.96 FDLRX 0.92 FCBB 2.42 Which of the following identifies the consequences of continued operation with thermal limits at the present values?

A. Fuel cladding could exhibit in excess of 1% plastic strain during a LOCA.

B. Possible peak cladding temperatures in excess of 2200ºF during a DBA LOCA.

C. There is a possibility of the onset of transition boiling in greater than 0.1 % of the fuel rods.

D. The core may become unstable during operation in the Restricted Region of the Power to Flow Map.

QUESTION 75 NRC RECORD # WRI 631 ANSWER: B. SYSTEM # B21; K/A 290002 A2.05: 3.7/4.2 Thermal Limits K5.01: 3.5/3.9 LP# General Physics HTFF Chapter 9 Thermal Limits OBJ. 9; 10 SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 3

REFERENCE:

Tech Spec Bases NEW 3.2.1; 3.2.2; 3.2.3; 3.2.4 MODIFIED BANK DIFF 2; CA 10CFR50.46(b)(1)

RO SRO BOTH CFR 41.3/41.14/43.2 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 76 The plant is performing the Reactor Vessel In-Service Leak Test (03-1-01-6) following refueling operations. A miscommunication results in a significant reactor pressure rise.

Pressure as read on the Control Room Wide Range Pressure indication on P680 is pegged upscale.

The Post Accident Pressure recorders indicate pressure has reached 1350 psig.

Which one of the following is a correct statement with regard to the GGNS Safety Limit for Reactor Pressure?

A. Reactor Pressure is above the Safety Limit of 1250 psig, because Tech Specs specifically references the P680 Reactor Wide Range Instrument.

B. Reactor Pressure is above the Safety Limit of 1325 psig, because the Post Accident indication is sensed from the Reactor Water Level instruments reference leg tubing.

C. Reactor Pressure is below the Safety Limit of 1375 psig, because the Post Accident indication is sensed from the Reactor Bottom Head instrument tap.

D. Reactor Pressure is below the Safety Limit of 1550 psig.

QUESTION 76 RO NRC RECORD # WRI 30 ANSWER: B. SYSTEM # K/A 295025 EK1.05: 4.4 Tech Specs EK1.02: 4.1 LP# GG-1-LP-OP-TS001 Generic 2.2.22: 3.4 OBJ 24; 28 2.2.25: 2.5 LP# GG-1-LP-OP-B2102 OBJ 3 SRO TIER GROUP / RO TIER 1 GROUP 1

REFERENCE:

Tech Specs 2.1.2 NEW Bases B2.1.2 MODIFIED BANK DIFF 1; M NRC 12/2000 RO SRO BOTH CFR 41.3/43.2 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 77 Concerning the operation of the Reactor Feed Pump (RFP) Turbines governor control in MANUAL and SPEED AUTO, which of the following correctly identifies the limitations imposed in MANUAL and SPEED AUTO if the raise pushbutton is depressed and held from 0 to 100%?

A. In MANUAL, the governor will stroke 0-100% in 15 seconds.

In SPEED AUTO, the speed setpoint will change at a rate of 10 rpm/sec for one second and 120 rpm/ second thereafter.

B. In MANUAL, the governor will stroke 0-100% in 10 seconds.

In SPEED AUTO, the speed setpoint will change at a rate of 15 rpm/sec for one second and 120 rpm/ second thereafter.

C. In MANUAL, the governor will stroke 0-100% in 15 seconds.

In SPEED AUTO, the speed setpoint will change at a rate of 10 rpm/sec for one second and 150 rpm/ second thereafter.

D. In MANUAL, the governor will stroke 0-100% in 10 seconds.

In SPEED AUTO, the speed setpoint will change at a rate of 15 rpm/sec for one second and 150 rpm/ second thereafter.

QUESTION 77 RO NRC RECORD # WRI 548 ANSWER: A. SYSTEM # N21 K/A 259001 K5.03: 2.8 LP# GG-1-LP-OP-N2100 OBJ. 19; 29 SRO TIER GROUP / RO TIER 2 GROUP 1

REFERENCE:

04-1-01-N21-1 sect 3.14 NEW MODIFIED BANK DIFF 1; M NRC 6/2001 RO SRO BOTH CFR 41.5/41.10/43.5 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 78 The plant is operating at 80 % power.

Feedwater Level Control is selected for Three Element Control.

Feedwater Flow A indicates 6.8 mlbm/hr Feedwater Flow B indicates 6.5 mlbm/hr The sensing line for the A Feedwater Flow Transmitter has broken loose.

Which one of the following describes the reaction of the Feedwater Level Control System?

A. A hard failure would be registered de-selecting 3-element control.

B. A soft failure would be registered de-selecting 3-element control.

C. A hard failure would be registered causing the Feedwater Level Control System to automatically input an Estimated Flow maintaining 3-element control D. A soft failure would be registered de-selecting 3-element control and disabling the use of 3-element control.

QUESTION 78 RO NRC RECORD # WRI 273 ANSWER: A. SYSTEM # C34 K/A 259002 K4.10: 3.4/3.4 LP# GG-1-LP-OP-C3401 K6.04: 3.1/3.1 OBJ. 6c SRO TIER GROUP / RO TIER 2 GROUP 1

REFERENCE:

04-1-02-H13-P680-2A-C9 NEW MODIFIED BANK DIFF 1; M NRC 4/2000 RO SRO BOTH CFR 41.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 79 Which one of the following would result in an automatic alignment of Standby Service Water A to Division I Diesel Generator?

A. RCIC has been manually started using the controls on H22-P150.

B. RHR A Pump has been manually started using the pump controls on H13-P601.

C. SSW A Pump has been manually started using the pump controls on H13-P870.

D. LPCS/LPCI A has been manually initiated using the pushbuttons on H13-P601.

QUESTION 79 RO WRI 677 ANSWER: D. SYSTEM # P75; P41 K/A 264000 K1.04: 3.2 LP# GLP-OPS-P4100 SRO TIER GROUP / RO TIER 2 GROUP 1 OBJ. 6.8; 9.5; 10.1

REFERENCE:

E-1225-001; 008 NEW DIFF 1; M MODIFIED BANK RO SRO BOTH CFR 41.4/41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 80 RCIC was operating for a surveillance when RCIC tripped on overspeed.

The overspeed trip has NOT been reset.

Which one of the following describes the indications if the operator placed the Trip/Throttle valve handswitch to close and then back to open?

See attached drawing of handswitch and indications.

A. The TRIP/THROTTLE SUPV will indicate full open The TRIP/THROTTLE VLV will indicate full closed B. The TRIP/THROTTLE SUPV will indicate full closed The TRIP/THROTTLE VLV will indicate full open C. The TRIP/THROTTLE SUPV will indicate full open The TRIP/THROTTLE VLV will indicate full open D. The TRIP/THROTTLE SUPV will indicate full closed The TRIP/THROTTLE VLV will indicate full closed QUESTION 80 RO WRI 676 ANSWER: D. SYSTEM # E51 K/A 217000 A4.02: 3.9 A4.03: 3.4 2.1.31: 4.2 LP# GG-1-LP-OP-E5100 SRO TIER GROUP / RO TIER 2 GROUP 1 OBJ. 9e

REFERENCE:

E-1185-015; 035; 039 NEW DIFF 2; CA MODIFIED BANK AUDIT 6/2001 Q 59 RO SRO BOTH CFR 41.7 REFERENCE MATERIAL REQUIRED: Drawing of handswitch

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 81 Which one of the following describes the function of the Safety Relief Valve Tailpipe Vacuum breakers?

A. Equalize pressure conditions in the SRV tailpipes to prevent Suppression Pool water from drawing into the piping resulting in excessive hydraulic loads.

B. Relieve steam from the SRV tailpipes following SRV actuation to the Drywell to prevent excessive back pressure on the SRV discs.

C. Relieve differential pressures built up in the SRV tailpipes following actuation that would result in lift pressures for SRVs in excess of design.

D. Equalize pressure established in the SRV tailpipes to assure positive actuation of the SRV tailpipe pressure switches to give accurate SRV position indication.

QUESTION 81 RO WRI 678 ANSWER: A. SYSTEM # E22-2 K/A 239002 K5.01: 2.7 2.1.28: 3.2 LP# GG-1-LP-OP-E2202 SRO TIER GROUP / RO TIER 2 GROUP 1 OBJ. 5b

REFERENCE:

UFSAR sect 5.2.2.4.1 NEW DIFF 1; M MODIFIED BANK RO SRO BOTH CFR 41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 82 The plant is operating at rated conditions.

The following indications are illuminated on the H13-P680 panel.

Pushbutton HCU FAULT Pushbutton ACKN HCU FAULT Annunciator HCU TROUBLE The operator identifies Control Rod 28-05 has a flashing RED led when HCU FAULT pushbutton is depressed.

ASSUME ALL OTHER ANNUNCIATORS AND STATUS LIGHTS ARE NORMAL.

Which one of the following is the probable cause for these indications?

A. HCU Accumulator pressure at 1620 psig B. Loss of power to RPS bus A C. A single control rod scram D. Detectable water in an HCU Instrument Block QUESTION 82 NRC RECORD # WRI 341 ANSWER: D. SYSTEM # C11-1B K/A 201003 A2.08: 3.8 LP# GG-1-LP-OP-C111A OBJ. 8d SRO TIER GROUP / RO TIER 2 GROUP 2 LP# GG-1-LP-OP-C1102 OBJ. 12; 26

REFERENCE:

04-1-02-1H13-P680 NEW 4A2-D4, E4; 7A-A2 MODIFIED BANK DIFF 2; CA 04-1-01-C11-2 NRC 12/2000 Sect 4.7.2.e,f; 4.8.2.a,d RO SRO BOTH CFR 41.6/41.10 05-1-02-III-2 sect 5.2.3 REFERENCE MATERIAL REQUIRED: None 43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 83 A plant startup is in progress.

The reactor is at the point of adding heat with reactor temperature at 165ºF.

Reactor water level is at +35 inches.

Which one of the following describes the response of the plant to an isolation of Reactor Water Cleanup?

ASSUME NO OPERATOR ACTION.

A. Reactor water level will rise to the point that a RPS scram signal is received and a Feedwater isolation occurs.

B. Reactor water level will rise to the point that water level has reached the main steam lines.

C. Reactor water level will rise until steam flow offsets the expansion of water in the reactor at which point level will stabilize.

D. Reactor water level will remain stable with steam flow offsetting the expansion of water in the reactor.

QUESTION 83 RO WRI 679 ANSWER: B. SYSTEM # G33; K/A 204000 K3.02: 3.1 C11-1A; C71 LP# GLP-OP-C7100 OBJ. 9; 10 LP# GLP-OPS-G3336 SRO TIER GROUP / RO TIER 2 GROUP 2 OBJ. 1; 2; 19; 21

REFERENCE:

03-1-01-1 NEW DIFF 2; CA Sect 2.2.5; 3.3.3a MODIFIED BANK RO SRO BOTH CFR 41.4 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 84 The Main Steam Isolation Valves just isolated with the plant operating at rated conditions.

Plant systems were operating in a normal configuration at the time of the isolation.

Which one of the following describes the response of the Condensate and Feedwater Systems?

ASSUME NO OPERATOR ACTION.

A. Condensate and Condensate Booster Pumps will continue to operate on minimum flow. Feedwater will continue to operate for a period of time until steam from the MSRs is exhausted. Reactor pressure will require reduction to allow feeding with Condensate and Condensate Booster Pumps.

B. Condensate and Condensate Booster Pumps will trip on low hotwell level causing a cascading effect tripping the Feedwater Pumps on low suction pressure. This results in a loss of all feed to the reactor until hotwell level is restored. Once the hotwell is restored reactor pressure will require reduction to allow feeding with Condensate and Condensate Booster Pumps.

C. Condensate and Condensate Booster Pumps will continue to operate supplying flow. Feedwater will trip on low reactor water level, but can be restored using steam from the MSRs until exhausted. Reactor pressure will require reduction to allow feeding with Condensate and Condensate Booster Pumps.

D. Condensate and Condensate Booster Pumps will trip on minimum flow. Feedwater will trip on low suction pressure, but Condensate and Condensate Booster Pumps can be restored at a lower pump combination and return Feedwater to service.

Reactor pressure will require reduction following a loss of steam from the MSRs to allow feeding with Condensate and Condensate Booster Pumps.

QUESTION 84 RO WRI 680 ANSWER: A. SYSTEM # N21; N19; K/A 239001 K3.03: 3.2 N11; B21 K1.22: 3.1 LP# GG-1-LP-OP-MCD7b SRO TIER GROUP / RO TIER 2 GROUP 2 OBJ. 2

REFERENCE:

UFSAR Figure 15.2-6 NEW DIFF 2; CA Simulator performance MODIFIED BANK RO SRO BOTH CFR 41.4/41.14 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 85 Which one of the following identifies the start signal and power supply for the TBCW DC Emergency Water Cooling Pump?

A. Seal Oil Pump C running will start the DC TBCW pump powered from 11DA.

B. Seal Oil Pump C running will start the DC TBCW pump powered from 11DF.

C. Low H2 Side Seal Oil Pressure will start the DC TBCW pump powered from 11DA.

D. Low H2 Side Seal Oil Pressure will start the DC TBCW pump powered from 11DF.

QUESTION 85 RO WRI 681 ANSWER: B. SYSTEM # L11; P43 K/A 263000 K2.01: 3.1 LP# GG-1-LP-OP-P4300 SRO TIER GROUP / RO TIER 2 GROUP 2 OBJ. 6; 8a

REFERENCE:

04-1-01-P43-1 Att III NEW DIFF 1; M E-1227-018 MODIFIED BANK E-1147-003 RO SRO BOTH CFR 41.4 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 86 Preparations are being made for a plant startup.

Which one of the following describes the process for startup of Offgas to prevent excessive levels of explosive gas buildup?

A. Preheat the Offgas system using Oxygen from the Hydrogen Water Chemistry System to scavenge any residual Hydrogen and recombining the two in the Hydrogen Recombiners prior to startup of the Steam Jet Air Ejectors.

B. Establish a purge of the Offgas system with Oxygen from the Hydrogen Water Chemistry System to heat up the Hydrogen Recombiner prior to startup of the Steam Jet Air Ejectors, then maintain constant flow through Offgas.

C. Preheat the Offgas system using air from the Instrument Air System to scavenge any residual Hydrogen and recombining the two in the Hydrogen Recombiners prior to startup of the Steam Jet Air Ejectors.

D. Establish a purge of the Offgas system with air from the Instrument Air System to heat up the Hydrogen Recombiner prior to startup of the Steam Jet Air Ejectors, then maintain minimum flow through Offgas.

QUESTION 86 RO WRI 682 ANSWER: D. SYSTEM # N64 K/A 271000 A4.09: 3.3 2.1.32: 3.4 LP# GLP-OPS-N64565 OBJ. 14.2; 15; 21.2 LP# GG-1-LP-OP-N6200 SRO TIER GROUP / RO TIER 2 GROUP 2 OBJ. 15

REFERENCE:

04-1-01-N64-1 NEW DIFF 1; M sect 3.7; 3.9; 4.1.2c NOTE MODIFIED BANK 04-1-01-N62-1 sect 3.8; 3.10; 4.3.2i RO SRO BOTH CFR 41.7/41.13 Caution; 4.4.1a; 4.4.2c 03-1-01-1 sect 3.3.8g 04-1-01-P73-1 sect 3.9; 3.10 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 87 Which one of the following identifies the maximum allowed Spent Fuel Pool Temperature per the Technical Requirements Manual?

A. 70ºF B. 125ºF C. 140ºF D. 200ºF QUESTION 87 RO WRI 683 ANSWER: C. SYSTEM # G41 K/A 233000 A2.07: 3.0 LP# GG-1-LP-OP-G4146 SRO TIER GROUP / RO TIER 2 GROUP 3 OBJ. 14a; 19

REFERENCE:

04-1-01-G41-1 sect 3.5 NEW DIFF 1; M TRM 6.7.4 MODIFIED BANK RO SRO BOTH CFR 41.7 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 88 The plant is at 135?F.

All ECCS systems are in standby.

The Reactor Mode Switch is in SHUTDOWN.

Refuel Floor Surveillances are being performed on the Refuel Bridge.

The Reactor Head is installed with the first head closure bolt de-tensioned.

Secondary Containment is in effect.

With the above conditions, which one of the following is the Plant Operational Mode?

A. Mode 2 - Startup B. Mode 3 - Hot Shutdown C. Mode 4 - Cold Shutdown D. Mode 5 - Refueling QUESTION 88 RO NRC RECORD # WRI 585 ANSWER: D. SYSTEM # ADMIN K/A Generic 2.1.22: 2.8 Conduct of Ops.

LP# GG-1-LP-OP-TS001 OBJ. 5 SRO TIER GROUP / RO TIER 3 GROUP

REFERENCE:

Tech Specs sect. 1.1 NEW table 1.1-1 MODIFIED BANK DIFF 1; M NRC 6/2001 RO SRO BOTH CFR 41.10/43.1 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 89 Which one of the following evolutions would be the manipulation of Controls as defined in Licensed Operator duties?

A. Shifting Control Rod Drive Flow Control Valves in Containment.

B. Lowering Control Rod Drive Pressure by opening C11-F003 on H13-P601.

C. Inserting local manual scrams of individual control rods in Containment at the SRI switches.

D. Controlling Reactor water level during a plant operation at 30% power using Feedwater.

QUESTION 89 RO WRI 684 ANSWER: C. SYSTEM # Conduct K/A Generics 2.1.1: 3.7 of Operations LP# GG-1-LP-OP-PROC SRO TIER GROUP / RO TIER 3 GROUP OBJ. 11.b.1

REFERENCE:

01-S-06-2 sect 5.1 NEW DIFF 2; CA 10 CFR 50.2 definitions MODIFIED BANK RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 90 You are the Operator-at-the-Controls.

Which one of the following Operations personnel is allowed to relieve you of Operator-at-the-Controls duties with a simple verbal exchange of current plant status?

ASSUME ONLY LEGAL AND PROCEDURAL REQUIREMENTS; DISREGARD UNION CONTRACTS.

A. A Tagging Group Reactor Operator B. Operations Office Shift Manager qualified active Senior Reactor Operator C. Training shift Nuclear Operator A visiting the Control Room during Training D. Nuclear Operator A assigned as the Auxiliary Building Roving Operator QUESTION 90 RO WRI 685 ANSWER: D. SYSTEM # Conduct K/A Generics 2.1.3: 3.0 of Operations LP# GG-1-LP-OP-PROC SRO TIER GROUP / RO TIER 3 GROUP OBJ. 45d

REFERENCE:

02-S-01-4 sect 2.7; 6.1.6 NEW DIFF 2; CA MODIFIED BANK RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 91 The basis for having Hydrogen Water Chemistry is to remove Oxygen from the reactor to reduce the possibility of Intergranular Stress Corrosion Cracking.

Which one of the following is the reason for injecting Oxygen into the Condensate System?

A. Oxygen is used in the reactor to scavenge any excess Hydrogen remaining in the Reactor environment to prevent explosive concentrations from remaining.

B. Oxygen raises concentrations in the Condensate System piping to prevent the flaking off of oxide layers inside the Condensate System components.

C. Oxygen concentrations inside the Reactor are insufficient to support recombination requiring additional Oxygen to allow complete removal.

D. Oxygen raises concentrations in the Reactor to prevent the removal of oxide layers from components in the Reactor during the recombination with Hydrogen.

QUESTION 91 RO WRI 686 ANSWER: B. SYSTEM # P73 K/A Generics 2.1.28: 3.2 LP# GG-1-LP-OP-P7300 SRO TIER GROUP / RO TIER 3 GROUP OBJ. 1

REFERENCE:

ER 96/0936 Supp 0 NEW DIFF 1; M sect 1.2.1 MODIFIED BANK RO SRO BOTH CFR 41.4 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 92 The Access Control Area Fan Coil Unit Heater, Z17-B003, has malfunctioned resulting in extremely high temperatures in the HP Lab and the heater control panel.

An MAI has been generated to investigate the problem.

Which one of the following describes the most appropriate configuration control method for this equipment?

A. Secure the fan coil unit. Log the handswitch position in the Control Room Operators Logbook and add the MAI number for reference.

B. Open the heater circuit breaker. Hang an information tag on the circuit breaker stating the problem and reference the MAI number.

C. Secure the fan coil unit. Document the handswitch position on a Component Position Control Form and reference the MAI number.

D. Open the heater circuit breaker. Hang a red tag on the circuit breaker and reference the MAI number on the clearance.

QUESTION 92 RO NRC RECORD # WRI 387 ANSWER: D. SYSTEM # K/A Generic 2.2.11: 2.5 Procedures LP# GG-1-LP-OP-PROC OBJ. 10b1,2; SRO TIER GROUP / RO TIER 3 GROUP

REFERENCE:

02-S-01-2 NEW Sect 6.12.6b MODIFIED BANK DIFF 2; CA 01-S-06-1 sect 5.4 & 5.9 NRC 12/2000 04-S-01-Z17-1 Att III RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 93 The Component Cooling Water Supply to the Non-Regenerative Heat Exchangers P42-F103 is to be used as a boundary valve for a tagout for Mechanical Maintenance.

Which one of the following describes the items to be tagged to utilize this valve as a boundary valve?

P&IDs M-1063B and M-1067H, and Electrical E-1226-05 are provided as reference.

A. Handswitch tagged in closed position.

Air supply P53-FY087 tagged in closed position.

Air supply to actuator vented and tagged.

B. Handswitch tagged in closed position.

P42-F103 jacked closed with an installed jacking device.

Air supply to actuator vented and tagged.

C. Air supply P53-FY087 tagged in closed position.

P42-F103 jacked closed with an installed jacking device.

Air supply to actuator vented and tagged.

D. Handswitch tagged in closed position.

Air supply P53-FY087 tagged in closed position.

P42-F103 jacked closed with an installed jacking device.

QUESTION 93 RO NRC RECORD # WRI 580 ANSWER: A. SYSTEM # Protective K/A Generics 2.2.13: 3.6 Tagging LP# GG-1-LP-OP-PROC OBJ. 10j, k, l SRO TIER GROUP / RO TIER 3 GROUP

REFERENCE:

01-S-06-1 sect 6.2.1h - i NEW M-1067H MODIFIED BANK DIFF 2; CA M-1063B NRC 6/2001 E-1226-05 RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: M-1063B , M-1067H, &

E-1226-05

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 94 Mechanical Maintenance is preparing to repack a valve in the Radwaste Equipment Drain Filter Demin Room. The job is estimated to take 30 - 45 minutes to complete.

Which one of the following is an acceptable method of reducing the radiation levels the Mechanics will be exposed to?

A. Install fans and HEPA filters to ventilate the room.

B. Plant Services builds a room of lead blankets around the valve.

C. Flush the Equipment Drain Filter and refill with demineralized water.

D. Cut the valve from the system and remove the valve to outside the room.

QUESTION 94 RO WRI 687 ANSWER: C. SYSTEM # RAD K/A Generics 2.3.10: 2.9 CON LP# ELP-GET-RWT01 SRO TIER GROUP / RO TIER 3 GROUP OBJ. RWT41

REFERENCE:

10CFR 20.1101(b) NEW DIFF 2; CA 01-S-08-2 sect 6.1.1 MODIFIED BANK 08-S-01-28 sect 5.9.2; 6.1 RO SRO BOTH CFR 41.12/43.4 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 95 Residual Heat Removal A is being lined up to operate in Suppression Pool Cooling.

The Plant Supervisor has requested you contact Health Physics.

Which one of the following describes the purpose of this phone notification?

A. Allows Health Physics personnel to evacuate any personnel from the Containment.

B. Allows Health Physics personnel to perform surveys of the RHR rooms and Containment for elevated radiation levels.

C. Informs Health Physics of elevated heat and noise levels in the vicinity of the RHR Rooms such that personnel entering the areas may be informed.

D. Informs Health Physics that the transient High Radiation areas for the RHR loop are now in effect.

QUESTION 95 RO NRC RECORD # WRI 287 ANSWER: B. SYSTEM # Rad Con - K/A Generic 2.3.2: 2.5 ALARA 2.1.32: 3.4 LP# GLP-OPS-E1200 OBJ. 14 SRO TIER GROUP / RO TIER 3 GROUP

REFERENCE:

04-1-01-E12-1 sect 3.1 NEW MODIFIED BANK DIFF 1, M NRC 6/2001 RO SRO BOTH CFR 41.10/41.12/43.4 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 96 An ATWS has occurred.

The MSIVs have closed.

RPV level is unable to be determined.

The following parameters exist:

Reactor Pressure 70 psig and lowering.

Suppression Pool Level 16.5 feet Suppression Pool Temperature 150 °F Drywell Pressure + 1.0 psig 8 Safety Relief Valves are OPEN.

LPCI A is injecting through Shutdown Cooling at 5000 GPM.

HPCS, LPCS, and RHR B & C are unavailable for injection.

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

A. Continue injection as long as water is available from any source to restore pressure above Minimum Alternative RPV Flooding Pressure.

B. Exit the Emergency Procedures and enter Severe Accident Procedure 4.

C. Exit the Emergency Procedures and enter Severe Accident Procedure 5.

D. Exit the Emergency Procedures and enter Severe Accident Procedure 6.

QUESTION RO 96 NRC RECORD # WRI 298 ANSWER:. B. SYSTEM # Conduct K/A Generics 2.4.4: 4.0 of Ops - EOP and SAP LP# GLP-EP-EPT19 OBJ. 7 SRO TIER GROUP / RO TIER 3 GROUP

REFERENCE:

05-S-01-EP2A NEW Steps 94 & 96 MODIFIED BANK DIFF 2, CA SAP - 4 step 4 MDRIR NRC 4/2000 RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-1-01-EP-2A and SAPs

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 97 An emergency condition has resulted in an Alert being declared.

The Emergency Response Organization is in route for manning.

How many and what are the responsibilities of Non-Licensed Operators dispatched to the Control Room during the initial phase of the emergency?

A. One Non-Licensed Operator is to perform the duties of safe shutdown operator, communications will be handled by the TSC when manned. All other operators report to the OSC.

B. Two Non-Licensed Operators are to perform the duties of communicators. All other operators report to the OSC.

C. Two Non-Licensed Operators are to perform the duties of communicators and one operator as the safe shutdown operator. All other operators report to the OSC.

D. Two Non-Licensed Operators are to perform the duties of communicators and two operators to perform equipment operations required outside the Control Room. All other operators report to the OSC.

QUESTION 97 RO NRC RECORD # WRI 577 ANSWER: D. SYSTEM # E-Plan K/A Generics 2.4.35: 3.3 LP# GG-1-LP-OP-PROC OBJ. 11d SRO TIER GROUP / RO TIER 3 GROUP

REFERENCE:

01-S-10-6 Att II & III NEW 01-S-06-2 sect 6.2.1d MODIFIED BANK DIFF 1; M NRC 6/2001 RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 98 The plant is operating at 100% power.

A short circuit results in N71-F041A, CLG TWR BYP VLV inadvertently opening.

Main Generator load immediately began to drop and is continuing to drop.

Which one of the following describes the Immediate Operator Actions to be taken for plant conditions?

A. Reduce reactor power as necessary to remain above turbine trip setpoint, maintaining within power to flow limitations.

B. Reduce recirculation flow immediately until thermal power decreases by 20% or 60% core flow is reached, and determine core stability actions.

C. Reduce reactor power as necessary to maintain offgas system below isolation setpoint and condenser vacuum above turbine trip setpoint remaining within power to flow limitations.

D. Reduce core flow to 60% (67Mlbm/hr) but not less than 55% (62Mlbm/hr) core flow, followed by control rod insertion, to reduce reactor power to less than 60%

of rated thermal power.

QUESTION 98 RO WRI 688 ANSWER: A. SYSTEM # ONEPs K/A Generics 2.4.49: 4.0 LP# GLP-OPS-ONEP1 SRO TIER GROUP / RO TIER 3 GROUP OBJ. 1

REFERENCE:

05-1-02-V-8 sect 2.0 ** NEW DIFF 1; M 05-1-02-V-5 sect 2.0 MODIFIED BANK 05-1-02-V-11 sect 2.0 05-1-02-V-13 sect 2.0 RO SRO BOTH CFR 41.10/43.5 04-1-01-N71-1 Att II REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 99 An ALERT has been declared, and notifications to offsite agencies are being made on schedule.

A Followup Notification was made at 0917.

At 0930 another Followup Notification was approved for transmission after a new development in the situation. This notification was begun at 0932 and completed at 0936.

The Operational Hotline has become inoperable.

When does the next Followup Notification become due and how should the notification be made?

A. Time 1017; using the Standard Telephone System B. Time 1030; using the Emergency Notification System (ENS)

C. Time 1032; using the Standard Telephone System D. Time 1036; using the Emergency Notification System (ENS)

QUESTION 99 RO WRI 689 ANSWER: C. SYSTEM # EPPs K/A Generics 2.4.43: 2.8 LP# GLP-EP-EPTS6 SRO TIER GROUP / RO TIER 3 GROUP OBJ. 3

REFERENCE:

10-S-01-6 NEW DIFF 1; M sect 6.1.1b(1); 6.3.1; MODIFIED BANK 6.6.3.a(1) EPTS 6 Exm 1 Form EPP 06-01 RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 REACTOR OPERATOR QUESTION 100 A fire has been reported in the Division I Diesel Generator Room.

The Fire Brigade is in route to the fire.

Which one of the following describes responses of the Control Room Operator?

A. Manually start all three fire pumps and the outside air fans for Division II and III Diesel Generator Rooms.

B. Manually start the motor driven fire pump and the outside air fans for Division II and III Diesel Generator Rooms.

C. Manually start all three fire pumps and the outside air fans for all three Diesel Generator Rooms.

D. Manually start the motor driven fire pump and the outside air fans for all three Diesel Generator Rooms.

QUESTION 100 RO NRC RECORD # WRI 578 ANSWER: B. SYSTEM # Fire K/A Generics 2.4.25: 2.9 Protection LP# GG-1-LP-OP-PROC OBJ. 61c(1) SRO TIER GROUP / RO TIER 3 GROUP

REFERENCE:

10-S-03-2 NEW Sect 6.2.2c NOTE & 6.2.2d MODIFIED BANK DIFF 1; M NRC 6/2001 RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 2 A reactor scram has occurred.

Which one of the following is a correct method of verifying the position of the control rods?

(The scram has NOT been reset.)

A. Using the full core display on H13-P680, depress ALL RODS with RCIS in Raw Data and observe a blank display with only green LEDs for all control rods.

B. Using the full core display on H13-P680, depress ALL RODS with RCIS in Raw Data and observe all control rods indicate 00 with a green LED for all control rods.

C. Using the full core display on H13-P680, depress ALL RODS with RCIS out of Raw Data and observe a blank display with only red LEDs for all control rods.

D. Using the full core display on H13-P680, depress ALL RODS with RCIS out of Raw Data and observe all control rods indicate 00 with a red LED for all control rods.

QUESTION 2 NRC RECORD # WRI 10 ANSWER: A. SYSTEM # C11-2; K/A 295006 AA2.02: 4.3/4.4 C11-1B 201005 A3.02: 3.5/3.5 LP# GG-1-LP-OP-C111B A4.02: 3.7/3.7 OBJ. 3c, 3f 201003 K4.05: 3.2/3.3 A3.01: 3.7/3.6 LP# GG-1-LP-OP-C1102 A4.02: 3.5/3.5 OBJ. 12 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

04-1-01-C11-2 NEW sect. 4.7.2p & 4.8.2i MODIFIED BANK DIFF: 2; CA 05-1-02-I-1 NRC 6/2001 sect. 2.1; 3.7; & 3.7.4 RO SRO BOTH CFR 41.6/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 3 Plant conditions are as follows:

MODE: Mode 1 Rx power: 28 %

T-G Load: 365 MWE Load Demand 390 MWE Bypass position: 0 %

All other parameters are per plant design.

The operator withdraws a control rod that raises Reactor power to 29 %.

How will the Turbine EHC Control System respond?

A. Bypass Control Valves will throttle open as required to maintain Rx pressure.

B. HP Turbine Control Valves will throttle open as required to maintain Rx pressure.

C. LP Turbine Control Valves will throttle open as required to maintain Rx pressure.

D. HP Turbine Control and Bypass Control Valves will throttle open as required to maintain Rx pressure.

QUESTION 3 NRC RECORD # WRI 69 ANSWER: B. SYSTEM # N32-2 K/A 295007 AK2.01: 3.5/3.7 241000 A2.02: 3.7/3.7 LP# GG-1-LP-OP-N3202 K4.01: 3.8/3.8 OBJ 4b, 6b, 7b SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

03-1-01-2 sect. 5.2 NEW MODIFIED BANK DIFF: 2; CA NRC 6/2001 RO SRO BOTH CFR 41.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 4 The plant was operating at 20 % power when a Feedwater rupture in the Turbine Building caused Reactor water level to drop.

The Control Room Operator manually initiated HPCS and RCIC.

Level in the Reactor dropped to ? 32 inches before HPCS and RCIC turned level, and level is now rising.

Which one of the following best describes the status of the Recirculation System?

A. Recirc Pumps are in Slow Speed with the Flow Control Valves Locked up (motion inhibit).

B. Recirc Pumps are tripped with the Flow Control Valves Locked up (motion inhibit).

C. Recirc Pumps are in Slow Speed with the Flow Control Valves in their pre-transient positions.

D. Recirc Pumps are tripped with the Flow Control Valves in their pre-transient positions.

QUESTION 4 NRC RECORD # WRI 121 ANSWER: C. SYSTEM # B33 K/A 295009 AK1.02: 3.0/3.1 AK2.03: 3.1/3.2 LP# GLP-OPS-B3300 AA1.03: 3.0/3.1 OBJ. 23, 24, 25, 47 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

ARI 04-1-02-H13-P680 NEW 3A-D4; 3A-D10 MODIFIED BANK DIFF 1; M NRC 3/1998 RO SRO BOTH CFR 41.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 5 Given the following plant conditions:

Reactor Power 100%

Reactor Level +36 inches Reactor Pressure 1025 psig Containment Temperature 85 °F steady Containment Pressure 0.03 psig steady Suppression Pool Temperature 81 °F steady Drywell Pressure 1.1 psig rising slowly Drywell Temperature 110 °F steady Drywell Area Sumps show no unusual changes in level, flow, or temperature.

Drywell Atmosphere radiation monitor show no changes.

The Roving NOA has noted that Drywell Pressure is rising slowly.

Drywell atmosphere radiation levels are steady.

Which one of the following describes a possible cause of the conditions as noted above?

A. Small leak on the Main Steam Line Flow Elbows Instrument Line.

B. Small leak on the RWCU suction from the Reactor Bottom Head.

C. Small leak on the Instrument Air header inside the Drywell.

D. Small leak on Recirc Pump Seals.

QUESTION 5 NRC RECORD # WRI 286 ANSWER:. C. SYSTEM # M41; P53 K/A 295010 AK3.04: 3.5/3.8 AK3.05: 3.5/3.4 LP# GLP-OPS-P5300 2.4.21: 3.7/4.3 OBJ. 26.1 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

GGNS 1998 event NEW CR 1998-0952 MODIFIED BANK DIFF 3, CA NRC 4/2000 RO SRO BOTH CFR 41.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 6 The plant is performing a reactor startup from cold shutdown.

The reactor was at the point of adding heat.

The Control Room Supervisor instructed the operators to stop the startup for a short duration to perform a surveillance.

During this time, the reactor went subcritical and power dropped to range 3 of the IRMs.

The At-The-Controls Operator, noting that reactor power had dropped selected the next control rod and withdrew the control rod from 20 to 48 with continuous motion as allowed by the Control Rod Movement Sequence Sheet.

This resulted in a sustained 20 second period.

The following are the plant parameters at present:

Reactor Pressure 80 psig Reactor Level + 40 inches 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. Perform the coupling checks for the Control Rod, and inform the Reactor Engineer of the power rise.

C. Withdraw the next in sequence Control Rod to maintain the power rise to reach the point of adding heat.

D. Insert the Control Rod to a position which causes reactor period to be > 50 seconds.

QUESTION 6 NRC RECORD # WRI 204 ANSWER: D. SYSTEM # C11-2; C51 K/A 295014 AK3.01: 4.1/4.1 LP# GG-1-LP-OP-IOI01 OBJ. 3c & d SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

03-1-01-1 sect. 2.1.4; 2.1.16 NEW Susquehanna reactivity MODIFIED BANK DIFF 1; M Event 7/98 NRC 6/2001 04-1-01-C51-1 sect 4.3.2 NOTE RO SRO BOTH CFR 41.1/41.2/ 41.6/43.6 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 7 Scram conditions exist. All control rods did NOT fully insert.

Reactor water level is being maintained at -60 inches wide range.

Reactor pressure is being maintained at 910 psig.

Reactor power is 20 %.

The following initial indications exist:

RPS white lights on H13-P680 are extinguished.

Scram Air Header Pressure low annunciator is illuminated.

RX SCRAM TRIP annunciator is illuminated.

The following actions have been taken:

Defeat the RPS scram signal and reset RPS Unisolate the Instrument Air header Defeat Alternate Rod Insertion A CRD pump is confirmed operating and the CRD FCV is open to achieve 250 psid Drive pressure.

Which one of the following contains the minimum actions required to drive the control rods to position 00 using Rod Control and Information System?

A. Bypass Control Rod Drive withdrawal blocks, select control rods and insert.

B. Bypass Control Rod Drive withdrawal blocks, select control rods in sequence and insert.

C. Bypass Control Rod Drive insert and withdrawal blocks, select control rods and insert.

D. Select control rods in sequence and insert.

QUESTION 7 NRC RECORD # WRI 203 ANSWER: C. SYSTEM # C11-2; C71; C11- K/A 295015 AK3.01: 3.4/3.7 1A LP# GG-1-LP-RO-EP02A OBJ. 5 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

EP 05-S-01-EP-2A NEW Step 48 Att. 18, 19 & 20 MODIFIED BANK DIFF 3; CA NRC 6/2001 RO SRO BOTH CFR 41.6/43.6 REFERENCE MATERIAL REQUIRED: 05-S-01-EP-2 EP-2A

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 8 A loss of coolant accident has occurred.

Determine which one of the following describes conditions that would direct / allow manual initiation of Containment Spray per the Emergency Procedures?

Supp Pool Lvl CTMT Temp CTMT Press Drywell Temp Drywell Press A. 18 FT 186ºF 1 psig 200ºF 3 psig B. 17 FT 120ºF 7 psig 215ºF 15 psig C. 18 FT 120ºF 6.5 psig 200ºF 15 psig D. 25 FT 140ºF 4 psig 215ºF 7 psig QUESTION 8 NRC RECORD # WRI 601 ANSWER: B. SYSTEM # M41-1; E12 K/A 295024 AK2.14: 3.9/3.9 LP# GG-1-LP-RO-EP03 OBJ. 3 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

EP 05-S-01-EP-3 NEW CTMT Temp & Press Legs MODIFIED BANK DIFF 2; CA Figures 3 & 4 RO SRO BOTH CFR 41.9/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-S-01-EP-3 & Figures 3 & 4

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 9 The following conditions exist in the plant:

Reactor power 0% all rods inserted.

Reactor pressure 230 psig and rising.

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

RHR C is injecting into the reactor vessel.

Which one of the following identifies adequate core cooling?

A. Adequate core cooling is NOT assured.

B. Adequate core cooling is assured by Minimum Alternate RPV Flooding Pressure (MARFP) with SRVs and Reactor pressure.

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 9 NRC RECORD # WRI 309 ANSWER: A. SYSTEM # Eps & K/A 295031 EK1.01: 4.6/4.7 Conduct of Ops. 2.1.1: 3.7/3.8 LP# GG-1-LP-RO-EP02 2.4.21: 3.7/4.3 OBJ. 15, 16 LP# GG-1-LP-RO-EP01 OBJ. 4a LP# GG-1-LP-OP- PROC OBJ. 10b3 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

01-S-06-2 section 5.18 NEW 05-S-01-EP-2 steps 69 - 74 MODIFIED BANK DIFF 3, CA PSTG App. B EPG Cont 3 NRC 12/2000 Steam Cooling RO SRO BOTH CFR 41.2/41.3/41.10/

REFERENCE MATERIAL REQUIRED: 05-S-01-EP-2 43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 10 An ATWS has occurred. Actions of EP-2A are being taken.

Which one of the following describes an allowance to terminate injection of Standby Liquid Control?

A. Control rods have been inserted to the equivalent of the first banked position with RPV temperature at < 200°F making the reactor subcritical.

B. All control rods are inserted to the Maximum Subcritical Banked Withdrawal Position, which assures the reactor will remain subcritical under all conditions.

C. RPV temperature has been reduced to < 200 °F and indicated reactor power on all IRMs is downscale on range 1, which indicates a subcritical reactor.

D. Standby Liquid Control has been injected such that Hot Shutdown Boron Weight (HSBW) has been injected and confirmed by chemical analysis.

QUESTION 10 NRC RECORD # WRI 216 ANSWER:. B. SYSTEM # C41; K/A 295037 EA1.04: 3.4/4.5 C11; C71 EK1.04: 3.4/3.6 EK1.05: 3.4/3.6 LP# GG-1-LP-RO-EP02A EA2.03: 4.3/4.4 OBJ. 2, 3, 5 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

05-S-01-EP-2A NEW step 2 & 4 MODIFIED BANK DIFF 3, CA GGNS PSTG App B NRC 4/2000 RC/Q-1 RO SRO BOTH CFR 41.1/41.2/

REFERENCE MATERIAL REQUIRED: 05-S-01-EP-2A 41.6/43.6

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 11 A LOCA has occurred. The Plant Supervisor has ordered the Hydrogen Recombiners started for Hydrogen removal in Containment.

Determine the final Hydrogen Recombiner Power Setting and the time to final Recombiner power.

Pre-LOCA Containment Temperature was 85 °F.

Post LOCA Containment Pressure +1.0 psig.

A. 47.73 kw after 20 minutes B. 47.73 kw after 25 minutes C. 49.02 kw after 20 minutes D. 49.02 kw after 25 minutes QUESTION 11 NRC RECORD # WRI 219 ANSWER: B. SYSTEM # E61 K/A 500000 EA1.03: 3.4/3.2 2.1.20: 4.3/4.2 LP# GQC-RO-CRO01 2.1.25: 2.8/3.1 OBJ. E61 task 5 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

04-1-01-E61-1 sec. 5.4.2 NEW Figure 1 MODIFIED BANK DIFF 2, CA NRC 4/2000 RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: 04-1-01-E61-1 & Calculator

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION: 12 Concerning the Fast Opening of One Recirculation Flow Control Valve transient, which one of the following CONDITIONS would result in the more severe transient on the reactor?

A. Reactor Power is at 30 % with Recirc in Slow Speed with Maximum Valve position.

B. Reactor Power is at 36 % with Recirc in Fast Speed with Minimum Valve position.

C. Reactor Power is at 60 % with Recirc in Fast Speed with Minimum Valve position.

D. Reactor Power is at 100 % with Recirc in Fast Speed with 68% Valve position.

QUESTION 12 NRC RECORD # WRI 116 ANSWER: C. SYSTEM # B33; K/A 295014 AA1.02: 3.6/3.8 FSAR CHPT 15 202002 K1.02: 4.2/4.2 LP# GG-1-LP-OP-MCD7b K3.02: 4.0/4.0 OBJ 2 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 1

REFERENCE:

UFSAR 15.4.5.3.2.1 NEW MODIFIED BANK DIFF: 1; M NRC 3/1998 RO SRO BOTH CFR 41.6/43.6 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 13 The plant was operating at power.

A transient caused the Recirculation Pump B trip to OFF.

The following parameters are indicated:

Reactor power 65 %

Core Flow 54 Mlbm/hr Recirc A Flow 39,000 gpm Recirc B Flow 0 gpm Reactor Engineering is calculating FCBB.

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

A. Immediately Scram the Reactor.

B. Monitor Neutron Monitoring for thermal hydraulic instability. Immediately reduce core thermal power by only inserting control rods to exit the region.

C. Monitor Neutron Monitoring for thermal hydraulic instability. Immediately exit the region by reducing core thermal power by inserting control rods, or raising core flow by opening Recirc FCV A.

D. Monitor Neutron Monitoring for thermal hydraulic instability and scram the reactor if any is noted, close B33-F067B and re-open after five minutes, restart the Recirc Pump as soon as possible.

QUESTION 13 NRC RECORD # WRI 602 ANSWER: C. SYSTEM # B33 K/A 295001 AA2.01: 3.5/3.8 2.4.1: 4.3/4.6 LP# GLP- OPS-B3300 2.4.11: 3.4/3.6 OBJ. 41.2; 42; 43 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

05-1-02-III-3 sect NEW Figure 1 MODIFIED BANK DIFF: 2; CA 03-1-01-2 sect 2.24 NRC 4/2000 RO SRO BOTH CFR 41.10/41.5/43.5 REFERENCE MATERIAL REQUIRED: 05-1-02-III-3 w/o Immediate actions

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 14 The plant is in a normal electrical line-up with all busses fed from their preferred power source. If a lockout of BOP Transformer 12B were to occur, Which of the following indicates the correct status of BOP busses?

A. 11HD ENERGIZED 12HE DE-ENERGIZED 13AD DE-ENERGIZED 14AE ENERGIZED 18AG ENERGIZED 28AG DE-ENERGIZED B. 11HD DE-ENERGIZED 12HE ENERGIZED 13AD ENERGIZED 14AE DE-ENERGIZED 18AG ENERGIZED 28AG ENERGIZED C. 11HD ENERGIZED 12HE DE-ENERGIZED 13AD DE-ENERGIZED 14AE ENERGIZED 18AG DE-ENERGIZED 28AG DE-ENERGIZED D. 11HD DE-ENERGIZED 12HE ENERGIZED 13AD ENERGIZED 14AE DE-ENERGIZED 18AG DE-ENERGIZED 28AG ENERGIZED QUESTION 14 NRC RECORD # WRI 507 ANSWER: B SYSTEM # R21 K/A 295003 A1.01: 3.7/3.8 LP# GLP-OPS-R2700 OBJ. 8 &15.1. SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

04-1-01-R21-11 sect 3.2 NEW 04-1-01-R21-12 sect 3.2 MODIFIED BANK DIFF 1; M 04-1-01-R21-13 sect 3.2 NRC 6/2001 04-1-01-R21-14 sect 3.2 RO SRO BOTH CFR 41.7 04-1-01-R21-18 sect 3.2 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 15 Which one of the following describes the reason for isolating the Main Steam Isolation Valves on a Low Main Condenser Vacuum?

A. Prevent erosion damage to the Main Steam Isolation Valve and Main Steam Bypass Valve seats due to steam condensation in the Main Steam Lines that would prevent their complete isolation in an emergency.

B. Prevent erosion damage to turbine blading in the Low Pressure Turbine due to steam condensation in the Main Steam Lines.

C. Prevent over-pressurization of low pressure piping on the suction of the Condensate pumps that could result in a rupture introducing steam outside Secondary Containment.

D. Prevent rupture of the turbine rupture diaphragms or damage to the turbine exhaust hood that could lead to leakage of radiation to the environment.

QUESTION 15 NRC RECORD # WRI 220 ANSWER: D. SYSTEM # B21; K/A 295002 AK3.05: 3.4/3.4 N11; N62 AA1.04: 3.3/3.4 LP# GG-1-LP-OP-M7101 OBJ. 6 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

GGNS Tech Spec Bases NEW 3.3.6.1-1d MODIFIED BANK DIFF 1; M NRC 4/2000 RO SRO BOTH CFR 41.4/43.4 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 16 The plant was operating at 80 % power.

Reactor Narrow Range Water Level transmitter C34-N004B has failed downscale and brought in annunciator RX WTR LVL SIG FAIL HI/LO.

The Operator at the Controls notices the Reactor Narrow Range Level indicator C34-LI-R606A indicates offscale HIGH and annunciator RFPT/MN TURB LVL 9 TRIP is in.

Reactor Narrow Range Water Level indicator R606C is reading + 36 inches.

Reactor Upset Range Water Level indicator is reading + 38 inches.

Reactor Wide Range Water Level indicator on P680 is reading + 40 inches.

Reactor Wide Range Water Level indicators A & B on P601 are reading + 40 inches.

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

(NO OTHER ALARMS ARE PRESENT.)

A. Immediately initiate a Reactor Scram and trip the Main Turbine and the Reactor Feed Pump Turbines because they failed to trip.

B. Manually select Reactor Water Level Control to Single Element control and verify Reactor level returns to normal.

C. Select the Master Level Controller to MANUAL to lock the level signals at the present setting to prevent any level perturbations and establish stable level control.

D. Continue monitoring Reactor Water Level on P680 and compare with other indications on P601 and the PDS computer and contact I&C.

QUESTION 16 NRC RECORD # WRI 275 ANSWER: D. SYSTEM # C34; N21; K/A 295008 AK1.01: 3.0/3.2 N30 245000 A3.01: 3.6/3.6 LP# GG-1-LP-OP-C3401 259001 K6.07: 3.8/3.8 OBJ. 3f; 3g2; 3i; 22; 23 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

04-1-02-H13-P680 NEW 4A2-A2 & D1 MODIFIED BANK DIFF 2; CA NRC 6/2001 RO SRO BOTH CFR 41.4/41.5 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 17 The Control Room has been abandoned and control has been established at the Remote Shutdown Panels.

Reactor pressure 600 psig Indicated Reactor level at the Remote Shutdown Panel 66 inches With present plant conditions, which one of the following describes Narrow Range Level, Actual Level and the availability of RCIC for level control?

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

NARROW RANGE ACTUAL LEVEL RCIC LEVEL A. 55 inches 52 inches Not available B. 45 inches 43 inches Available C. 48 inches 43 inches Available D. 60 inches 60 inches Not available QUESTION 17 NRC RECORD # WRI 603 ANSWER: A. SYSTEM # C61; B21 K/A 295016 AA2.02: 4.2/4.3 2.1.25: 2.8/3.1 LP# GLP-OPS-C6100 2.4.11: 3.4/3.6 OBJ 19 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

05-1-02-I-1 Att I & II NEW MODIFIED BANK DIFF 2; CA NRC 6/2001 RO SRO BOTH CFR 41.5/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-1-02-II-1 Att. I & II

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 18 The plant is operating at 100 % power.

RCIC is to be operated for surveillance testing to return the system to operable.

Standby Service Water A is operating.

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

A. Suppression Pool Temperature is limited to 95ºF and must be monitored every five minutes while RCIC is operating. If Suppression Pool Temperature exceeds 100ºF Suppression Pool Cooling must be placed in service.

B. Suppression Pool Temperature is limited to 95ºF and must be monitored every five minutes while RCIC is operating. If Suppression Pool Temperature exceeds 95ºF Suppression Pool Cooling must be placed in service and RCIC secured.

C. Suppression Pool Temperature is limited to 100ºF and must be monitored every five minutes while RCIC is operating. If Suppression Pool Temperature exceeds 95ºF Suppression Pool Cooling must be placed in service. RCIC must be secured if Suppression Pool Temperature exceeds 100ºF.

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

QUESTION 18 NRC RECORD # WRI 605 ANSWER: C. SYSTEM # E51; M24 K/A 295013 AA2.01: 3.8/4.0 AA1.02: 3.9/3.9 LP# GG-1-LP-OP-E5100- 2.1.33: 3.4/4.0 OBJ. 12a; 22 2.4.4: 4.0/4.3 LP# GG-1-LP-OP-M7101 OBJ. 23 LP# GG-1-LP-RO-EP03 OBJ. 3; 5 LP# GG-1-LP-OP-M4101 OBJ. 11 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

Tech Spec 3.6.2.1 NEW Tech Spec SR3.6.2.1.1 MODIFIED BANK DIFF 2; CA 04-1-01-E51-1 sect 5.2.1e 05-S-01-EP-3 step 10 - 12 RO SRO BOTH CFR 06-OP-1M24-V-0001 sect 41.5/41.10/43.3/43.5 5.1.2a REFERENCE MATERIAL REQUIRED: Tech Spec 3.6.2.1 & 06-OP-1M24-V-0001

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 19 A LOCA has occurred.

The reactor is depressurized.

The Drywell remains inaccessible due to pressure of 2.5 psig and temperatures of 200ºF.

MSIV Leakage Control Outboard System has been initiated.

Which one of the following identifies the method to monitor radiation released outside Secondary Containment at this time with MSIV Leakage Control in operation?

A. The operating Standby Gas Treatment Train Radiation Monitors will give the indication of radiation release.

B. Auxiliary Building Fuel Handling Area Exhaust Radiation Monitors will give indication of radiation release.

C. Auxiliary Building Fuel Handling Area Exhaust and Fuel Pool Sweep Exhaust Radiation Monitors will give indication of radiation release.

D. Auxiliary Building Fuel Handling Area Exhaust and Standby Gas Treatment Radiation Monitors will give indication of radiation release.

QUESTION 19 NRC RECORD # WRI 606 ANSWER: A. SYSTEM # T48; K/A 295017 AA1.08: 3.1/3.4 E32; D17 AA1.09: 3.6/3.8 LP# GG-1-LP-OP-E3200 239003 K1.02: 2.9/3.0 OBJ. 12c; 13 LP# GG-1-LP-OP-D1721 OBJ. 2 LP# GG-1-LP-OP-T4801 OBJ. 9f SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

04-1-01-E32-1 sect 3.2 NEW sect 3.2; 5.1.1c; 5.2.1c MODIFIED BANK DIFF 2; CA RO SRO BOTH CFR 41.11/41.13 REFERENCE MATERIAL REQUIRED: NONE 43.4

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 20 The plant is operating at 100 % power.

The Component Cooling Water temperature control valve closes to 25% in response to a temperature controller malfunction and the valve is unable to be reopened.

CCW temperatures have risen and continue to rise slowly.

Recirculation Pumps A and B pump bearing temperatures are in alarm on H13-P614.

Which one of the following describes the foremost actions to be taken for these conditions?

Loss of Component Cooling Water ONEP is provided.

A. Immediately scram the reactor and manually trip both Recirculation Pumps.

B. Start the standby CCW pump. Trip the RWCU pumps then close CCW to the Non-Regenerative Heat Exchangers, P42-F103. Reduce core flow to 60% only.

C. Trip the RWCU pumps then close CCW to the Non-Regenerative Heat Exchangers, P42-F103. Isolate CCW to Fuel Pool Heat Exchangers by closing P42-F105 and F205 only.

D. Start the standby CCW pump. Trip the RWCU pumps then close CCW to the Non-Regenerative Heat Exchangers, P42-F103. Isolate CCW to Fuel Pool Heat Exchangers by closing P42-F105 and F205. Reduce core flow to 60% only.

QUESTION 20 NRC RECORD # WRI 314 ANSWER: A. SYSTEM # P42; K/A 295018 AK3.03: 3.1/3.3 ONEPs LP# GLP-OPS-ONEP OBJ. 1; 2; 34 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

05-1-02-V-1 sect 3.1 NEW Section 3.2.2 Note for 2.1 MODIFIED BANK DIFF 1; M NRC 12/2000 RO SRO BOTH CFR 41.4/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-1-02-V-1 w/o Imm.

Actions

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 21 The plant is operating at rated conditions.

Plant Air Dryer A is in service and Dryer B is tagged out for maintenance.

Plant Air Dryer A has undergone an Executed Stop.

Annunciator PLANT AIR DRYR TROUBLE on H13-P870 is in alarm.

Instrument Air Header pressure is 110 psig and stable.

Which one of the following identifies actions to be taken?

A. Immediately scram the reactor in preparation for a complete loss of Instrument Air.

B. Immediately dispatch an operator to isolate Plant Air Dryer A by closing isolation valves P51-F207A and P51-F208A.

C. Dispatch an operator to manually crosstie Service Air via SP52-F010 or SP52-F300 and monitor Instrument Air header pressure.

D. Monitor Instrument Air header pressure and determine the cause of the Executed Stop using the PDS computer and a local operators observations.

QUESTION 21 NRC RECORD # WRI 607 ANSWER: D. SYSTEM # P51; P53 K/A 295019 AK2.14: 3.2/3.2 2.4.31: 3.3/3.4 LP# GLP-OP-P5101 300000 A2.01: 2.9/2.8 OBJ. 2d 2.1.32: 3.4/3.8 LP# GLP-OPS-P5300 OBJ. 20.1; 21 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

04-1-02-H13-P870 7A-E3 NEW 04-1-01-P51-1 MODIFIED BANK DIFF 1; M sect 3.5.4 & 3.7 05-1-02V-9 sect 3.4 RO SRO BOTH CFR 41.4/41.10/43.5 REFERENCE MATERIAL REQUIRED: 04-1-02-H13-P870-7A-E3

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 22 The plant is operating at rated conditions.

A rupture in the Instrument Air header has resulted in Instrument Air header pressure dropping to 40 psig.

Which one of the following describes the response of the Containment Cooling System and Containment temperature?

A. Containment Coolers will operate recirculating the Containment atmosphere maintaining temperature stable.

B. Containment Coolers will operate isolated without cooling water allowing Containment temperatures to rise.

C. Containment Coolers will trip due to a loss of cooling water resulting in Containment temperatures rising.

D. Containment Coolers will trip due to the isolation of the inlet and outlet dampers resulting in Containment temperatures rising.

QUESTION 22 NRC RECORD # WRI 608 ANSWER: B. SYSTEM # P71; K/A 295020 AK2.03: 3.1/3.3 P53; M41 AK3.03: 3.2/3.2 LP# GG-1-LP-OP-M4100 300000 K1.03: 2.8/2.9 OBJ. 10a; 13a LP# GLP-OPS-P7100 OBJ. 16.2 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

E-1213-009 & 012 NEW M-1100B MODIFIED BANK DIFF 1; M M-1109D 05-1-02V-9 sect 5.5 RO SRO BOTH CFR 41.4/41.7/41.9 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 23 The plant is in a startup following a 32 day outage.

MSIVs are closed.

Recirc loop temperatures are at 180 ?F.

Control rods are being withdrawn to achieve criticality. (Minimal decay heat)

Feedwater is operating in long cycle cleanup.

The operating CRD Pump tripped.

What will be the response of the plant?

(ASSUME NO FURTHER OPERATOR ACTIONS)

A. The reactor water level will remain stable at its present level.

B. The reactor water level will rise to the point that a reactor scram is received on High water level.

C. The reactor water level will drop to the point that a reactor scram is received on Low water level.

D. The plant will scram due to a loss of charging water pressure to the Hydraulic Control Units.

QUESTION 23 NRC RECORD # WRI 55 ANSWER: C. SYSTEM # C11-1A; K/A 295022 AK2.04: 2.5/2.7 G33/36; IOI- 1 AK2.05: 2.4/2.5 LP# GLP-OPS-G3336 AA1.04: 2.5/2.6 OBJ 3.3, 8.6, 21 LP# GG-1-LP-OP-C111A OBJ 23 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

03-1-01-1 NEW sect. 2.2.5; 3.3.1d; 3.3.3a MODIFIED BANK DIFF 2; CA NRC 6/2001 RO SRO BOTH CFR 41.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 24 An ATWS has occurred. Reactor pressure is being controlled with SRVs.

Standby Liquid Control has been initiated.

Reactor power has just dropped below 4%.

The following conditions exist:

Reactor Power 2 % and stable Reactor Pressure 1000 psig and stable Reactor Level - 100 inches Fuel Zone and stable Suppression Pool Level 16.5 feet and rising Suppression Pool Temperature 150 °F and rising Drywell Pressure + 1.0 psig and rising Which one of the following describes actions to be taken?

A. Maintain RPV water level between -192 and + 53.5 inches and stabilize RPV pressure < 1064.7 psig.

B. Maintain RPV water level between -192 and + 53.5 inches. Lower RPV pressure to 700 - 900 psig, and initiate SPMU.

C. Terminate and prevent all injection into the RPV except for Boron, CRD and RCIC, and lower RPV water level to the top of active fuel.

D. Terminate and prevent all injection into the RPV except for Boron, CRD and RCIC and emergency depressurize the RPV.

QUESTION 24 NRC RECORD # WRI 301 ANSWER: D. SYSTEM # M41; K/A 295026 EK2.01: 3.9/4.0 B21 LP# GG-1-LP-RO-EP03 OBJ. 2, 3 LP# GG-1-LP-RO-EP02A OBJ. 7, 10h SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

05-S-01-EP2 EP2A NEW Step 33, 14, & 15 MODIFIED BANK DIFF 2, CA 05-S-01-EP3 Step 15 and NRC 12/2000 HCTL RO SRO BOTH CFR 41.7/41.9/

REFERENCE MATERIAL REQUIRED: 05-S-01-EP-2A and EP-3 41.10/41.14/43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION JUNE 2001 SENIOR REACTOR OPERATOR QUESTION 25 The plant was operating at 100 % Power.

A steam leak has developed in the Containment steam tunnel.

Containment temperature has gone up to 85°F and still rising.

A power reduction has commenced but Containment temperature continues to rise.

Tech. Specs states if Containment temperature exceeds 95°F to restore to < 95°F within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

If Containment temperature is unable to be restored to < 95°F within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; then be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and be in MODE 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Which of the following is the reason for this action?

Tech Specs 3.0 & 3.6.1.5 are provided.

A. Shut down of the Reactor is done to prevent having to initiate Containment Spray to maintain Containment temperature below 185°F.

B. Shut down of the Reactor is done to place the plant in a MODE that the LCO does NOT apply.

C. Shut down of the Reactor is done to prevent having to Emergency Depressurize to maintain Containment temperature below 185°F.

D. Shut down of the Reactor is done to prevent damaging operating equipment inside Containment due to current high temperature.

QUESTION 25 NRC RECORD # WRI 512 ANSWER: B. SYSTEM # M41-1 K/A 295027 K3.03: 3.7/3.7 LP# GG-1-LP-OP-TS001 OBJ. 13 LP# GG-1-LP-OP-M4101 OBJ. 11, 12 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

TECH. SPEC. 3.6.1.5 NEW TECH. SPEC. BASES MODIFIED BANK DIFF 1; M 3.6.1.5 NRC 6/2001 TECH SPEC 3.0.2 ** RO SRO BOTH CFR 41.9/41.10/43.2 REFERENCE MATERIAL REQUIRED: Tech Spec 3.6.1.5 & 3.0.2

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 26 The plant is operating at rated conditions.

Temperatures in the lower CRD Cavity are rising.

Which one of the following describes the expected response of the Drywell Cooling System?

A. At 145ºF, M51-F009 and M51-F016 open and M51-F006 and M51-F014 close to divert Drywell Cooling flow from the lower Drywell area to the CRD Cavity.

B. At 145ºF, the Standby fans on M51-B001 and B002 will auto start and M51-F009 and M51-F016 open raising the flow of air to the CRD Cavity.

C. At 145ºF, M51-F009 and M51-F016 open raising the flow of air to the CRD Cavity and at 155ºF M51-F006 and M51-F014 close to divert more air flow.

D. At 155ºF, the Standby fans on M51-B001 and B002 will auto start and M51-F009 and M51-F016 open raising the flow of air to the CRD Cavity.

QUESTION 26 NRC RECORD # WRI 609 ANSWER: A. SYSTEM # M51 K/A 295028 EA1.03: 3.9/3.9 EK2.04: 3.6/3.6 LP# GG-1-LP-OP-M5100 EK3.04: 3.6/3.8 OBJ. 6a SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

04-1-02-H13-P870 NEW 4A-H2; 10A-H2 MODIFIED BANK DIFF 2; CA M-1101 E-1214-001; 004; 009 RO SRO BOTH CFR 41.4/41.7 REFERENCE MATERIAL REQUIRED: Drywell Cooling System Drawing

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 27 Which of the following is the basis for Emergency Reactor Pressure Vessel (RPV)

Depressurization when Suppression Pool Level CANNOT be maintained below 24.4 feet?

A. 24.4 feet is the highest Suppression Pool level at which the pressure suppression capability of Containment can be maintained.

B. 24.4 feet is the highest Suppression Pool level at which the Suppression Pool will NOT overflow the weir wall resulting in flooding the Drywell.

C. 24.4 feet is the highest Suppression Pool level at which Suppression Pool level instrumentation taps will become covered resulting in loss of ability to monitor Suppression Pool level.

D. 24.4 feet is the highest Suppression Pool level at which opening Safety Relief Valves (SRVs) will NOT result in exceeding the design blowdown rate for the SRV discharge piping.

QUESTION 27 NRC RECORD # WRI 513 ANSWER: A. SYSTEM # M41-1 K/A 295029 K1.01: 3.4/3.7 LP# GG-1-LP-OP-EP03 OBJ. 6 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

GGNS PSTG APP B SP/L-3 NEW GGNS PSTG APP A SP/L-3 MODIFIED BANK DIFF 1; M NRC 6/2001 RO SRO BOTH CFR 41.9/41.10 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 28 The plant is in a LOCA with ECCS systems injecting to the reactor.

Suppression Pool level has lowered to 13.5 feet.

Which one of the following is a condition that exists due to this level?

A. The SRV tailpipe exhaust spiders have been uncovered.

B. The RCIC Turbine Exhaust pipe has been uncovered.

C. Suppression Pool temperature is unable to be determined.

D. Containment Pressure is unable to be determined QUESTION 28 NRC RECORD # WRI 8 ANSWER: C. SYSTEM # E30 K/A 295030 EA2.02: 3.9 LP# GG-1-LP-OP-EP01 OBJ. 5 LP# GG-1-LP-RO-EP03 OBJ. 6 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

05-S-01-EP-3 Caution 2 NEW GGNS EOP PSTG App B MODIFIED BANK DIFF 1; M Caution 2 NRC 6/2001 RO SRO BOTH CFR 41.9/41.10/43.5 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 29 Preparations are being made to startup following a Refueling Outage.

During the Refueling outage several Control Rod Drive Mechanisms were replaced and the old mechanisms taken to the CRD Maintenance Room for rebuilding.

The Auxiliary Building Operator has reported elevated radiation levels indicated on his dositec in Area 9 185 ft. of 150 mr/hr.

Which one of the following is the possible cause of the elevated radiation levels?

A. This is normal radiation levels for general area Auxiliary Building.

B. Used CRD Drive Mechanisms in the Auxiliary Building C. Waterborne contamination in the Fuel Pool Cooling System D. Normal radiation levels from the Fuel Pool Filter Demineralizers.

QUESTION 29 NRC RECORD # WRI 610 ANSWER: C. SYSTEM # G41/46 K/A 295033 EA2.03: 3.7/3.7 LP# GLP-OPS-G4146 OBJ. 2; 4.8; 5.4; SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 5.5; 5.6; 15.1

REFERENCE:

M-1051A; M-1016 NEW UFSAR 9.1.3.1.2; MODIFIED BANK DIFF 1; M 12.2.1.3.3 & Table 12.2.-12 FPCCU System Figure 4 RO SRO BOTH CFR 41.12/43.4 REFERENCE MATERIAL REQUIRED: None

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 30 A GENERAL EMERGENCY has been declared due to a DBA LOCA.

Plant conditions as follows:

Reactor water level 195

Approximately 15% fuel damage

Large fission product inventory in Containment

All Containment parameters are within design limits

Radiation levels at the site boundary are at 15 mRem/hr TEDE What protective actions, if any, should be recommended?

10-S-01-1 Activation of the Emergency Plan is provided.

A. No evacuation of surrounding areas is required.

B. Evacuate a 2-mile radius and shelter five miles downwind.

C. Evacuate a 2-mile radius and five miles downwind. Shelter the rest of the 10-mile EPZ.

D. Evacuate a 2-mile radius and 10 miles downwind. Shelter the rest of the 10-mile EPZ.

QUESTION 30 NRC RECORD # WRI 611 ANSWER: C. SYSTEM # EPP PARs K/A 295038 EK1.02: 4.2/4.4 LP# GG-1-LP-EP-EPTS6 OBJ 2 SRO TIER 1 GROUP 1 / RO TIER 1 GROUP 2

REFERENCE:

10-S-01-1 sect. 6.1.4.j(1) NEW MODIFIED BANK DIFF 2; CA EPTS 6 EXAM 2 RO SRO BOTH CFR 41.10/41.12 REFERENCE MATERIAL REQUIRED: 10-S-01-1 43.4/43.5

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 31 A fire has been reported by security at the Hydrogen Bulk Storage Facility.

The Outside Operator has identified the leak as coming from the vent stack on the discharge of the cryogenic compressors.

Which one of the following describes the actions to be taken by operators at the scene?

A. Establish a perimeter of at least 160 feet, and allow the fire to burn itself out.

B. Establish a perimeter of at least 160 feet, and establish a monitor nozzle spraying water on the point where the fire is originating.

C. Evacuate a 2 mile radius of the fire, and establish a monitor nozzle spraying water on the point where the fire is originating.

D. Evacuate a 2 mile radius of the fire, and establish a fire team to enter the area with a fire hose cooling the team and isolate the source of the fire.

QUESTION 31 NRC RECORD # WRI 612 ANSWER: A. SYSTEM # P73 K/A 600000 AK3.04: 2.8/3.4 2.1.32: 3.4/3.8 LP# GG-1-LP-OP-P7300 2.4.25: 2.9/3.4 OBJ 12; 14 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2

REFERENCE:

04-1-01-P73-1 NEW sect 3.1; 3.2; 3.3; 3.7 MODIFIED BANK DIFF 1; M RO SRO BOTH CFR 41.10/43.5 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 32 The plant is in a Refueling Outage with the reactor disassembled five (5) days after the plant was shutdown in Refueling Outage 11.

Reactor Coolant Temperature is 140 ?F The Fuel shuffle has just begun.

The in-service shutdown cooling pump has just tripped off.

Assume no further operator action.

Determine for this condition:

1. The time to reach 200?F for the Reactor Vessel.

2. The time for level to reach the Top of Active Fuel.

05-1-02-III-1 Attachment I is provided.

A. 1) 0.75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months

2) 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months B. 1) 1.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

2) 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months C. 1) 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

2) 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months D. 1) 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

2) 75 hours8.680556e-4 days 0.0208 hours 1.240079e-4 weeks 2.85375e-5 months QUESTION 32 NRC RECORD # WRI 604 ANSWER: C. SYSTEM # G41/46; K/A 295021 AK1.01: 3.6/3.8 ONEP LP# GLP-OPS-ONEP AA2.01: 3.5/3.6 OBJ 18; 19 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 3

REFERENCE:

05-1-02-III-1 Att. I NEW Figures 2 & 5 MODIFIED BANK DIFF 2; CA Pre-shuffle 150ºF NRC 3/1998 WRI37 RO SRO BOTH CFR 41.5/41.10/43.5 REFERENCE MATERIAL REQUIRED: 05-1-02-III-1 Att I

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 33 RHR A Pump Room temperature rises to 170 ?F.

Which one of the following identifies the systems or components in addition to RHR A that will be affected by this temperature?

A. HPCS B. MSIVs and RCIC C. RCIC D. RWCU QUESTION 33 NRC RECORD # WRI 048 ANSWER: C. SYSTEM # E12; E51; K/A 219000 A1.08: 3.7/3.6 E31 A2.14: 4.1/4.3 A3.01: 3.3/3.3 LP# GLP-OPS-E3100 A4.06: 3.9/3.7 OBJ 7.4 295032 EK3.03: 3.8/3.9 EK3.07: 3.6/3.8 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 3

REFERENCE:

ARI 04-1-02-H13-P601 NEW 20A-B1 MODIFIED BANK DIFF 1; M 05-1-02-III-5 Group 2, 3, 4 NRC 3/1998 RO SRO BOTH CFR 41.4/41.10/43.5 REFERENCE MATERIAL REQUIRED: NONE

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 34 Following a Recirc line rupture, reactor level has dropped to - 80 inches.

Both trains of the Standby Gas Treatment System have initiated.

Which of the following best describes the operation of the Standby Gas Treatment System flow control dampers?

A. When 0.2 inches water column is obtained in the Enclosure Building, the steam tunnel cooler dampers throttle to their intermediate position. 90 seconds later the remaining flow control dampers throttle to their intermediate position.

B. When 0.25 inches water column is obtained in the Enclosure Building, the steam tunnel cooler dampers throttle to their intermediate position. 120 seconds later the remaining flow control dampers throttle to their intermediate position.

C. After 90 seconds, the flow control dampers will go to their intermediate positions to maintain ? 0.75 inches water column in the Auxiliary Building and 0.25 inches water column in the Enclosure Building.

D. After 90 seconds the flow control dampers throttle to maintain 0.25 inches water column in the Auxiliary Building. If the Enclosure Building pressure reaches 0.75 inches water column, the flow control dampers go to their intermediate positions.

QUESTION 34 NRC RECORD # WRI 007 ANSWER: A. SYSTEM # T48 K/A 295035 EA1.02: 3.8/3.8 LP# GG-1-LP-OP-T4801 261000 A1.04: 3.0/3.3 OBJ. 8e SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 3

REFERENCE:

E- 1257- 08, 11, 23 NEW MODIFIED BANK DIFF 2; CA NRC 3/1998 RO SRO BOTH CFR 41.4/41.7/41.13 REFERENCE MATERIAL REQUIRED:

U.S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION AUGUST 2002 SENIOR REACTOR OPERATOR QUESTION 35 The plant is operating normally at 100 % power.