ML030580747

Initial Submittal of the SRO Written Examination for the Duane Arnold Examination - November 2002

ML030580747
Person / Time
Site:	Duane Arnold
Issue date:	11/18/2002
From:	Nuclear Management Co
To:	Office of Nuclear Reactor Regulation
References
50-331/02-301
Download: ML030580747 (186)

Text

INITIAL SUBMITTAL OF THE SRO WRITTEN EXAMINATION FOR THE DUANE ARNOLD EXAMINATION - NOVEMBER 2002

ES-401 Site-Specific Written Examination Form ES-401-8 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information Name: Region:

Date: Facility/Unit:

License Level: RO / Reactor Type: W CE / BW GE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent Examination papers will be collected five hours after the examination starts.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicant's Signature Results Examination Value Points Applicant's Score Points Applicant's Grade Percent 43 of 46 NUREG-1021, Revision 8, Supplement I

QF-1030-03 Rev. 0 (FP-T-SAT-30)

WRITTEN EXAMINATION KEY to Nudd to. .up' COVERSHEET EXAMINATION REVIEW AN]) APPROVAL:

Submitted by: Date: /20 4 Reviewed by: , 2 >L -. jDate: /go/too .z Approved by-: Cer Date: ?1acto /e;W Written Examination key Attach answer key to this page.

Key should contain the following:

"o Enabling Objective Number "o Test Item o Question or Statement o All possible answers o Correct Answer Indicated o Point Value "o References Indicate in the following table if any changes are made to the exam after approval:

1 1 4 4 t 1 4 4 i I I U ____________________________ i ______________________________ I ___________ I ______

Retention: Life of plant insurance policy + 10 yr. Disposition: Instructor, Training Supervisor, and TAC Retain in: Training Records 57_2002-ILC-SRO-W_xm.doc Rev. 0

QF-1030-02 Rev. 0 (FP-T-SAT-30)

WRITTEN EXAMINATION COVERSHEET Committed to NuclearExcelec TRAINEE INFORMATION:

Job Name:

Title:

Employee Number: Site: DAEC Training program: Senior Reactor Operator Course/lesson plan Number(s):

50007 Examination Number/Title: 2002 ILC SRO Written Exam GRADE:

Total Points Possible: 100 Grade: I =_  %

Graded by: Date:

Co-graded by (not required if Scantron graded): Date:

EXAMINATION RULES

1. References may not be used during this examination, unless otherwise stated.

2. Read each question carefully before answering. If you have any questions or need clarification during the examination, contact the examination proctor.

3. Conversation with other trainees during the examination is prohibited.

4. Partial credit will not be considered.

5. Rest room trips are limited and only one examinee at a time may leave.

6. Ensure you have answered each item prior to turning in your exam.

7. For exams with time limits:

a. You have _ minutes to complete the examination.

b. The proctor will collect all examinations after this time expires.

EXAMINATION INTEGRITY STATEMENT Dishonesty Policy: Cheating or compromising the exam will result in disciplinary actions up to and including termination.

"I have not given, received, or observed any aid or information regarding this examination prior to or during its administration that could compromise this examination's integrity."

EXR minPe'g ien2ture Date:

REMEDIATION ACKNOWLEDGEMENT "I acknowledge that the correct answers to the exam questions were indicated to me following the completion of the exam. I have had the opportunity to review the examination questions with the instructor to ensure my understanding.

Examninee's Signature: Date:

Retention: 6 years Disposition: Instructor, Training Supervisor, and TAC Retain in: Training Records 57_2002-ILC-SRO-Wxm.doc Rev. 0

QF-1030-02 Rev. 0 (FP-T-SAT-30)

NC M-

-,WRITTEN EXAMINATION COVERSHEET Committed to Nuclear Excellence i Training program: Senior Reactor Operator Examination Number/Title: 2002 ILC SRO Written ExamI Course/lesson plan Number(s):

50007 i Retention: 6 years Disposition: Instructor, Training Supervisor, and TAC Retain in: Training Records 57_2002-ILC-SRO-Wxm.doc Rev. 0

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

1. The plant is at 90% power when a Turbine trip occurs.

Which of the following plant responses is correct?

a. A pressure switch in the air header to the Extraction Steam Valves down stream of the Turbine Extraction Relay Dump Valve senses a low pressure and sends a CLOSED signal to MO-9147 and MO-9148 "MSR 1E-18A and B 2nd Stage Reheat Steam High Load Isolation Valves".

b. A pressure switch in the air header to the Extraction Steam Valves down stream of the Turbine Extraction Relay Dump Valve senses a low pressure and sends a CLOSED signal to CV-1106 and CV-1363 "HP Heater 1E6 A and B Drain Bypass Valves".

c. Air pressure in the air header to the Extraction Steam Valves down stream of the Turbine Extraction Relay Dump Valve is bled off causing CV-1106 and CV-1363 "HP Heater 1E6 A and B Drain Bypass Valves" to go CLOSED.

d. Air pressure in the air header to the Extraction Steam Valves down stream of the Turbine Extraction Relay Dump Valve is bled off causing CV-1237 "High Pressure Extraction Drain to Condenser Valve" to go CLOSED.

ANSWER: a Answer: The reason the MSRs isolate on a Turbine Trip is because PS1097 in the air header to the Extraction Steam Valves down stream of the Turbine Extraction Relay Dump Valve senses the turbine trip and sends the closed signal to MO-9147 and MO-9148. IAW IPOI 5 verification is required by the operator.

Reference:

SD 646 Rev 5 page 20, IPOI 5 Rev 33 step 3.3 (4) (b) page 6 Distracter 1: Plausible because these valves require closing after a scram. However, there is no auto closed signal. These valves require manual action to close. IPOI 5 Rev. 33 step 3.3 (6) (a and b) page 7.

Distracter 2: Plausible because these valves require closing after a scram. However, there is no auto closed signal. These valves require manual action to close. IPOI 5 Rev 33 step 3.3 (6) (a and b) page 7.

Distracter 3: Plausible because this valve is in the airline from the Extraction Relay Dump Valve. However, this is a fail open valve.

K/A System: 295005 Main Turbine Generator Trip K/A Number: AK3.05 Knowledge of the reason for the following responses as they apply to MAIN TURBINE GENERATOR TRIP: Extraction steam/moisture separate isolations K/A Value: 2.5/2.6 DAEC Objective 46.00.00.07 Number:

DAEC Objective Evaluate plant conditions and control room indications to determine if the Statement: Extraction Steam and Feedwater Heating system is operating as expected, and identify any actions that may be necessary to place the Extraction Steam and Feedwater Heating system or the plant in the correct condition Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -I 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Cognitive Level: 2RI - The candidate has to know the extraction dump valve will vent the air header to the pressure switch. This occurs due to EHC system actuation due to a turbine trip signal. They then need to understand this will generate a closed signal to the MSR high load valves and if the valves do not close the operator is directed in IPOI 5 to close them.

Source: New Operationally Required actions after a scram and turbine trip Validity:

OE: MSR failed to fully isolate on recent scram.

Estimated Completion Time: EB#

Time Validation: N/A m- (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -2 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

2. A Reactor SCRAM has occurred and has NOT been reset.

The OSS directs you to perform IPOI 5 "Reactor Scram" and verify all control rods fully inserted.

The Full Core Display shows 4 control rods with their green FULL IN lights OFF.

lAW IPOI 5 "Reactor Scram", what other method is allowed to verify all control rods are fully inserted?

a. Requesting a Rod Log

b. Check SPDS for ALL RODS IN

c. Use of Refuel One Rod Selected Permissive

d. Check the rods are at position 00 on the Four Rod Display ANSWER: c Answer: Refuel One-Rod Selected Permissive is the only other authorized method to verify all control rods have fully inserted IAW IPOI 5. This condition occurred at DAEC due to burned out light bulbs. At that time the only method to verify all rods in was to see all Green Full In lights on. The second method was subsequently added to IPOI 5.

REFERENCE:

IPOI 5 Rev 33 step 3.2 (6) (a)

Distracter 1: This is plausible because normally if you want to know rod position you are directed to request a Rod Log in several procedures. However, the Rod Log will indicate -99 for all rods because they are beyond 00 and not authorized in IPOI 5.

Distracter 2: This is plausible because SPDS does indicate ALL RODS IN. However, this is not an approved method to determine that all rod are inserted Distracter 3: This is plausible because the Four Rod Display shows rod position. However, with the scram not reset the display will be blank because the rods are beyond position 00.

K/A System: 295006 Scram K/A Number: AA2.02 Ability to determine and/or interpret the following as they apply to scram:

Control Rod Position.

K/A Value: 4.3/4.6 DAEC Objective 93.22.01.09 Number:

DAEC Objective Explain how to verify all rods in with the Refuel One Rod Select Permissive switch Statement:

Cognitive Level: 1P - This is a procedure step in IPOI 5 Source: New Operationally This is a required operator action for a scram Validity:

OE: Yes ATWS was entered at DAEC in the past before we could use the Refuel One Rod Selected Permissive due to failed light bulbs.

Estimated Completion Time: EB#

Time Validation: N/A Vt (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -3 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops

'> TMARs:

Question Usage (exams): 57_2002 ILC-RO-W xm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -4 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

3. A plant transient has occurred and ALC has been entered.

The following conditions currently exist:

"* All rods are in

"* Reactor Pressure has been stabilized at 550 psig

"* RPV level is +70 inches and lowering at 10 inches/min.

"* All available high pressure injection sources are injecting at maximum flow

"* RHR is being used to spray the Torus and Drywell

"* Drywell pressure is 3 psig and holding steady The OSS orders the panel operator to line up and maximize injection with RHR and Core Spray.

The 1C03 operator verifies the "A" Core Spray pump is running and MO-2115 "Outboard Inject Valve" is OPEN.

The operator then attempts to OPEN MO-2117 "Inboard Inject Valve" to maximize flow but the valve will NOT OPEN.

Which of the following would explain these conditions?

a. The Core Spray Automatic Initiation signal does NOT exist with these conditions. When RPV level drops to the Core Spray Initiation setpoint the operator will be able to OPEN the valve.

b. Valve logic prevents simultaneously OPENING both Inboard and Outboard Inject Valve at this time to prevent over pressurizing the low pressure Core Spray piping.

c. The differential pressure across the valve is too high. The Outboard Inject valve will need to be CLOSED and the Inboard Inject valve cracked off its closed seat. Then both valves can be OPENED.

d. The "A" SBDG logic has failed to provide an OPEN permissive to the Core Spray Inject valve logic. The operator should verify the "A" SBDG is running.

ANSWER: b Answer: There is a valid initiation signal at this point (>2 psig DW). The valve is still closed because the 450 psig permissive has not been reached. With RPV pressure

>450 psig only one inject valve at a time can be opened to protect low pressure Core Spray piping.

REFERENCE:

SD 151 Rev 5, ARP 1C 03B B-5 Rev 2 Distracter 1: This is a plausible choice if because the 64 inch initiation setpoint has not been reached. There is a valid initiation signal at this point (>2 psig DW). However, the valve is still closed because the 450 psig permissive has not been reached.

Distracter 2: This is a plausible choice because there is a higher than normal DP across the valve. However, the valve is still closed because the 450 psig permissive has not been reached.

Distracter 3: This is a plausible choice because the Core Spray and SBDG logic do have a connection. However, the Core Spray logic sends a signal to the SBDG logic to start. The valve is still closed because the 450 psig permissive has not been reached.

K/A System: 295007 High Reactor Pressure K/A Number: AK 2.04 Knowledge of the interrelationships between HIGH REACTOR PRESSURE and the following: LPCS Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W-xm. doc -5 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

K/A Value: 3.2/3.3 DAEC Objective 4.02.01.10 a

~ Number:

DAEC Objective Describe the Core Spray System interlocks, including purpose, setpoints, logic, Statement: and when/how they are bypassed.

a. Pump and Valve Interlocks Cognitive Level: 3SPK The problem is that the injection valves will not open and the operator must determine what is the cause of the system conditions. System interlocks for overpressure protection is preventing the valve operation.

Source: New Operationally System k nowledge and proper system response for ECCS systems.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A [m (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -6 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

4. The plant was operating at 92% power.

A plant transient occurred requiring a Manual SCRAM.

An electrical ATWS was entered due to failure to SCRAM conditions.

Defeat 15 "MSIV and MSL Drain LO-LO-LO Level Isolation Defeat" is installed.

Reactor power has stabilized at 4%.

RPV water level is 135 inches and steady.

The Main Turbine is still on line.

At 0910 a steam leak is detected in the Steam Tunnel reading 180°F and increasing at 10OF per minute.

Which of the following will occur if no operator action is taken and the above conditions continue?

a. At 0912 power will increase due to a rapid rise in RPV pressure.

b. At 0912 a Group 1 signal will occur. However, power will not be affected due to installation of Defeat 15.

c. At 0922 the MSIVs will close causing power to increase.

d. At 0922 the Main Turbine will trip. However, the Bypass Valves will maintain power stable at 4%.

ANSWER: a Answer: A Group 1 will occur at 200OF and the MSIVs will close. This will induce a large pressure increase in the RPV and add positive reactivity to the core. The candidate will have to predict the pressure increase based on MSIV closure and the resulting power excursion.

REFERENCE:

EOP Bases ATWS - RPV Control Rev 7 pages 6, 53, and 60.

Distracter 1: A group 1 will occur at 0912. However, Defeat 15 will not prevent the MSIV closure on the High Steam Tunnel Temperature. This is a common error made with Defeat 15. It only prevents closure of MSIV on water level. The other isolation signals still occur.

Distracter 2: Power will increase. However, the MSIVs would have closed at 0912 when Steam Tunnel Temperature reached 200 0 F. This corresponds to EOP 3 300OF max safe Steam tunnel temperature which is sometimes confused with the 200OF trip setpoint.

Distracter 3: The Main Turbine would probably tripped by this point and 4% power is well within the Bypass Valve capacity. However, The MSIVs will be closed and the Bypass valves would have no effect. The candidate may assume that defeat 15 has kept the MSIVs open.

K/A System: 295007 High Reactor Pressure K/A Number: AA2.02 Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Reactor power.

K/A Value: 4.1/4.1 DAEC Objective 95.56.08.04 Number:

95.56.01.02 48.01.01.01 Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -7 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

DAEC Objective Explain the effects, on plant systems or components, of the inability to control Statement: reactor pressure with the bypass valves during an ATWS condition Differentiate between the pressure control strategies of EOP 1 and ATWS Describe how the Main Steam System responds to a Group 1 isolation signal Cognitive Level: 3PEO - The Candidate has to predict a pressure increase and the effect on power.

Source: New Operationally Effects of pressure increase on Reactor power during ATWS conditions.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A F-1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -8 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

5. A SCRAM occurs from full power.

The following actions occur:

"* All RHR (LPCI) and Core Spray Pumps START

"* Primary Containment Group 1 and Group 7 valves CLOSE

"* The "A" and "B" SBDGs START

"* The ADS Timers START Which one of the following would account for this set of indications?

a. Loss of 1A3 and 1A4 essential busses.

b. Drywell Pressure has reached its Hi pressure trip setpoint 2 psig.

c. RPV water level has reached its LO-LO level trip setpoint 119.5 inches.

d. RPV water level has reached its LO-LO-LO level trip setpoint 64 inches.

ANSWER: d Answer: All these action occur at 64 inches RPV level IAW ARP 1C05 A-1. The candidate must interpret from the conditions that level has dropped below 64 inches in the RPV.

REFERENCE:

ARP 1C05 A-1 Rev 5 Distracter 1: This is a plausible choice because the SBDGs both start if a Group 1 occurs, also Loss of RPS will give a false LO-LO-LO RPV level annunciator and drywell 2 psig annunciator. However, ADS times will not initiate. RHR and Core Spray pumps load shed.

Distracter 2: This is a plausible choice because several systems that start at 64 inches RPV level also start on 2 psig Drywell pressure. However, ADS does not start and Group 1 and 7 do not occur.

Distracter 3: This is a plausible choice because lower RPV level auto starts several systems.

LPCI LOOP selects at this level.

K/A System: 295009 K/A Number: AA2.01Ability to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL: Reactor water level K/A Value: 4.2/4.2 DAEC Objective 42.08.01.07 Number:

DAEC Objective List the signals, which cause Primary Containment and Containment Atmosphere Statement: Monitoring and Control System isolations. Describe their purpose, setpoints, and logic. Describe how they are bypassed and how they are reset Cognitive Level: 2DR - The candidate has to recognize the relationship between RPV LO-LO -LO level (64 inches) and various system responses based on that level dropping below the setpoint.

Source: New Operationally EOP break points and ARP actions. EOP actions required by operators Validity:

OE: There have been occurrences of systems not responding on isolation signals and thus requiring operator action to compensate for the failure.

Estimated Completion Time: EB#

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -9 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Time Validation: N/A F1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc 10 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

6. Review the indication given in the Support Material Booklet The plant was at power.

10 minutes ago annunciator 1C04C D-3 "Drywell Equip Drain Sump Hi Temp" alarmed followed shortly by 1C04C C-3 "Drywell Equip Drain Sump Hi Leakage.

The ARP automatic actions occurred properly with 1P37B "B DW Equipment Drain Pump" RUNNING recirculation through the 1E-34 "DW Equipment Drain Sump HX".

The STA reports Drywell pressure rising.

Drywell pressure has reached 3.5 psig and is slowly rising.

Annunciator 1C04C B-1 "Drywell Equipment Drain Sump HI-HI Level is in alarm.

You observe the Drywell Drain system indications shown on the next page.

Which action is correct for the current plant conditions?

(Assume indicated annunciators are still in alarm)

a. Take HS for CV-3728 and CV-3729 "Drywell Equipment Drain Sump Isolation Valves" to CLOSED and check that both 1P37A and 1P37B AUTOMATICALLY STOP.

b. Take HS for 1P37A and 1P37B to STOP and check CV-3728 and CV-3729 "Drywell Equipment Drain Sump Isolation Valves" AUTOMATICALLY CLOSE.

c. Take HS for CV-3728 and CV-3729 "Drywell Equipment Drain Sump Isolation Valves" to CLOSED and check that 1P37A and 1P37B remain RUNNING.

d. Take HS for CV-3728 and CV-3729 "Drywell Equipment Drain Sump Isolation Valves" to CLOSED and check that 1P37A or 1P37B remains RUNNING.

ANSWER: a Answer: At 2 psig in the drywell a group 2 isolation will close CV-3728 and CV-3729 Drywell Equipment Drain Discharge valves. This will trip 1P37A and 1P37B if running. In the given plant conditions the Group 2 has failed to occur and with the Hi-Hi sump level both pumps will be running. The correct operator action is to complete the isolation that failed.

REFERENCE:

SD 920.1 Rev 3 pages 11-14, ARP 1C04C B-3/C-3/D-3 Revs.4/4/5. 1C05B B-1 Rev 6.

Distracter 1: There is an interlock between the isolation valves and the sump pumps. However, the interlock is that if the valves close the pumps trip not the other way round.

Distracter 2: With a Hi-Hi sump level still in alarm there will be a signal for both sump pumps to be running. However, because there is a group 2 signal the isolation CVs should have closed and the interlock between the isolation valves and the sump pumps will override the Hi-Hi level alarm signal.

Distracter 3: With a Hi sump temperature still in alarm there will be a signal for at least one sump pump to be running. However, because there is a group 2 signal the isolation CVs should have closed and the interlock between the isolation valves and the sump pumps will override the Hi temp alarm signal.

K/A System: 295010 K/A Number: AA1.02 Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE: Drywell Floor and Equipment Drain Sumps K/A Value: 3.6/3.6 Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc 11 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

DAEC Objective 89.03.01.06 Number:

z DAEC Objective Describe the Drywell Sump system interlocks, including purpose, setpoints, pump Statement: and valve logics Cognitive Level: 2RI - The operator must determine the correct expected system response for the Drywell sump system when a Group 2 PCIS isolation occurs at 2 psig.

Source: New Operationally Required action to verify automatic action and isolations complete Validity:

OE: During a recent SCRAM the sump system had a system malfunction which was caught by the operating crew and was incorporated into lessons learned during LOR Requal training.

Estimated Completion Time: EB#

Time Validation: N/A m* (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 12 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

7. The plant is at full power.

You are the IC05 operator and notice the following indications:

"* Reactor power on the APRMs is noted to be slowly rising.

"* Main Generator load is rising.

"* Final Feedwater temperature is lowering.

"* Condenser backpressure has risen.

Which of the following would account for this set of indications?

a. CV-1139 "Feedwater Heater lE-5A Dump to Condenser" is OPEN.

b. CV-1158A"HP Turbine Extraction Steam Outlet to 1E-6A" is OPEN.

c. MO-1546 "LP Feedwater Heater Bypass Isolation Valve" is CLOSED.

d. CV-1237 "High Pressure Turbine Extraction Steam Bypass Valve " is CLOSED.

ANSWER: a Answer: The indications given are for a positive reactivity addition event.CV-1139 being open will cause a loss of feed water heating which will increase inlet subcooling and add positive reactivity to the core.

REFERENCE:

SD 646 Rev 5 pages 37 and 47. SD 644 Rev 5 page 42. 01 646 Rev 33 page 4. AOP 255.2 Rev. 22 page 4.

Distracter 1: CV 1158 could cause the indications given. However, the valve would have to go closed. Open is the normal position of this valve.

Distracter 2: MO-1546 could cause the indications given. However, the valve would have to be open. Closed is the normal position of this valve.

Distracter 3: CV-1237 could cause the indications given. However, the valve would have to be open. Closed is the normal position of this valve.

K/A System: 295014 Inadvertent Reactivity Additions K/A Number: AA1.07 Ability to operate and/or monitor the following as they apply to INADVERTENT REACTIVITY ADDITIONS: Cold water injection K/A Value: 4.0/4.1 DAEC Objective 46.00.00.07 Number:

DAEC Objective Evaluate plant conditions and control room indications to determine if the Statement: Extraction Steam and Feedwater Heating system is operating as expected, and identify any actions that may be necessary to place the Extraction Steam and Feedwater Heating system or the plant in the correct condition Cognitive Level: 2RI Source: Bank - Slightly modified.

Operationally AOP actions to identify cause of loss of Feedwater heating..

Validity:

OE: DAEC and other plants have experienced a loss of Feedwater Heating.

Estimated Completion Time: EB#

STime Validation: N/A Vt (time) Incorrect Ratio Data: (ratio)  %

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 13 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Question Developed By: Peer Checked By:

Operator Validated By:

SApproved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 14 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

8. An electrical ATWS has occurred with reactor power remaining at 7%.

The 1C05 operator is directed to manually drive control rods.

1) Should the Rod Worth Minimizer be PLACED IN BYPASS or should it REMAIN IN OPERATE?

AND

2) What is the CORRECT reason for the answer to part 1 of this question?

a. 1) PLACED IN BYPASS

2) To disable possible Select Blocks.

b. 1) PLACED IN BYPASS

2) To disable possible Insert Blocks.

c. 1) REMAIN IN OPERATE

2) To enforce insertion of the highest worth rods FIRST.

d. 1) REMAIN IN OPERATE

2) To transmit rod positions to the Plant Process Computer.

ANSWER: b Answer: If bypass is not selected the rods will not fully insert.

REFERENCE:

EOP Bases, ATWS Rev. 7, page 87; SD 878.8 Rev 5 page 33; RIP 103.3 Rev 3 page 1

~ Distracter 1: Select blocks come from RSCS.

Distracter 2: Placed in Bypass per RIP 103.3. RWM enforcement is not desirable.

Distracter 3: Placed in Bypass per RIP 103.3. RWM does not transmit to PPC.

K/A System: 295015 Incomplete SCRAM K/A Number: AK3.01 Knowledge of the reasons for the following responses as they apply to INCOMPLETE SCRAM: Bypassing rod insertion blocks.

K/A Value: 3.4/3.7 DAEC Objective 95.00.00.20 Number:

DAEC Objective For any step from EOP Support Procedures:

Statement: 1. Explain the basis for the statement

2. Explain how it accomplishes the goals of EOP Support Procedure Cognitive Level: 1-B - The purpose of the Bypass feature is to allow rod motion outside the programmed sequence which is necessary under these conditions.

Source: Bank - 1999 RO Exam, 2001 RO Audit Exam Operationally Operator action required to shutdown the reactor as directed in the EOPs.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A m] (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

• , Operator Validated By:

Approved By: Date: Date:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 15 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 16 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

9. As primary containment pressure exceeds 55 psig, which of the below correctly describes the operational implications of this pressure?

a. The containment vent valves will not open due to high delta pressure.

b. The HPCI malfunctions due to high backpressure.

c. The SRVs will not remain open.

d. Containment failure.

ANSWER: d Answer: For DAEC the PCPL has been derived from several curves and determined to be 53 psig for pressure and 95 ft water level in containment.

REFERENCE:

EOP Bases Document Curves and Limits B.5 Rev 5 page 35-40 Distracter 1: The vent valves are a concern at 56 psig. However, the pressure in containment at this pressure has exceeded the PCPL of 53 psig.

Distracter 2: RCIC will trip at 50 psig containment backpressure. However, HPCI trips at 150 psig. RCIC and HPCI set points are often confused because several set points are the same.

Distracter 3: SRVs take about 50 psig between RPV pressure and containment pressure to operate. They will open at a higher RPV pressure than normal, in this case about 105 psig.

K/A System: 295024 High Drywell Pressure K/A Number: EK1.01 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE: Drywell integrity K/A Value: 4.1/4.2 DAEC Objective 95.00.00.18 Number:

DAEC Objective Evaluate the possible consequences of exceeding any EOP Curve or Limit on the Statement: mitigation of an event Cognitive Level: 1B - The containment design limit is the bases for the 53 psig setpoint.

Source: Bank 1998 NRC Exam Operationally Primary containment failure criteria and operators knowledge of EOP Validity: breakpoints.

OE:

Estimated Completion Time: EB#

Time Validation: N/A F-1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By: --

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 17 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

10. The plant is at full power when annunciator 1C07A B-4 "EHC Fluid reservoir 1T-33 LO Level" is received.

The in-plant operator reports an unisolable EHC leak.

The OSS directs Reactor SCRAM and IPOI 5 actions to be carried out.

EOP 1 is entered on LO RPV water level after the SCRAM.

The OSS then directs securing both EHC pumps.

Which of the following is correct concerning Bypass valve operation for this event and subsequent Reactor pressure control?

a. Bypass Valve operation will NOT be available for long term RPV pressure control under these conditions. ADS/SRVs, HPCI, or RCIC can be used for RPV pressure control.

b. The Bypass Valves will NOT control with Pressure Set. However, the Bypass Valve Opening Jack will still function.

c. The installed Bypass Valve accumulators provide 30 minutes of Bypass Valve operation. At this point decay heat will be within the capacity of the MSL Drain Valves.

d. The Bypass Valves will NOT be available for RPV pressure control. Use Chest Warming to control RPV pressure until decay heat is within the capacity of the MSL Drains.

ANSWER: a Answer: With loss of EHC header pressure the Bypass Valves have about 40 seconds accumulator reserve before the valves go closed. The Bypass Valves will be unavailable for RPV pressure control for EOP 1. Reactor pressure will increase to the high pressure setpoint and LLS will be the pressure control system unless operator action is taken to control pressure.

REFERENCE:

SD 693.2 Rev3 page 43, EOP 1 Distracter 1: This is a plausible choice because the valves will not respond to pressure set and we do use The Bypass Valve Opening Jack in the EOPs. However, EHC pressure is required.

Distracter 2: This is a plausible choice because there are accumulators installed in the plant for 30-min. operations. However, this is not the case with loss of EHC header pressure the Bypass valves. They have about 40 seconds accumulator reserve before the valves go closed.

Distracter 3: Chest warming would reduce pressure. However, with the loss of EHC pressure the valves used for chest warming will not open.

K/A System: 295025 High Reactor Pressure (EOP)

K/A Number: EK2.08 Knowledge of the interrelationships between HIGH REACTOR PRESSURE and the following: Reactor/turbine pressure regulating system.

K/A Value: 3.7/3.7 DAEC Objective 95.46.09.08 Number:

DAEC Objective Explain the effects, on plant systems or components, of the inability to stabilize Statement: reactor pressure with the bypass valves.

Cognitive Level: 11 - The candidate has to determine the system response to a loss of EHC pressure.

S>Source: New Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W-xm.doc - 18 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operationally EHC system response and system knowledge.

Validity:

'i OE: Yes - DAEC has experienced this event.

Estimated Completion Time: EB#

Time Validation: N/A I-] (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 19 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

11. The Plant is at 90% power.

ATWS conditions have occurred.

Which of the following is correct with regard to automatic initiation of "ATWS ARI/RPT" and what does this accomplish?

Actions are automatically initiated ...

a. at 119.5 inches RPV Water Level which rapidly reduces power due to a rapid increase in voids.

b. at 1055 psig RPV Pressure which OPENS (vents to atmosphere) the ARI solenoid valves, depressurizing the Scram Air Header.

c. at <90% OPEN on the Turbine Stop Valves which rapidly reduces power due to a rapid increase in voids.

d. on a "Fast Closure Signal" to the Turbine Control Valves which OPENS (vents to atmosphere) the ARI solenoid valves, depressurizing the Scram Air Header.

ANSWER: a Answer: At 119.5 inches the ATWS ARI/RPT will trip the Recirc pump RTP breakers open.

This will cause the core to void adding negative reactivity to the core and causing a large power reduction.

REFERENCE:

SD 264 Rev 6 pages 43-47.

\- Distracter 1: The ARI solenoid valves open on an ATWS ARI/RPT trip. However, the ATWS ARI/RPT RPV pressure setpoint is 1140 psig. 1055 is the SCRAM setpoint for RPS.

Distracter 2: TSV closure to <90% will cause an RPT and power reduction due to-voids.

However, this is an EOC RPT for thermal limit protection and not an ATWS trip.

Distracter 3: On a fast closure of the TCVs an RPT trip will occur and the ARI solenoid valves do open. However, this is an EOC RPT for thermal limit protection and not an ATWS trip.

K/A System: 295031 Reactor Low Water Level K/A Number: 2.1.28 Knowledge of the purpose and function of major system components and controls K/A Value: 3.2/3.3 DAEC Objective 12.00.00.03c Number:

12.01.01.12 DAEC Objective Describe the operation of the following principle Recirc system components:

Statement: c. RPT breakers Describe the Recirc system interlocks, including purpose, setpoints, logic, and when/how they are bypassed Cognitive Level: 1I Source: New Operationally System knowledge S'Validity:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 20 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

OE:

Estimated Completion Time: EB#

Time Validation: N/A mý (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -21 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

12. A Group 1 isolation and ATWS have occurred from full power.

0 Reactor power was 15% after the Recirc pumps were tripped.

0 LLS is controlling RPV pressure.

0 RPV injection was terminated and prevented for Level/Power Control.

As the 1C05 operator, you CLOSED the Feed Reg Valves and are monitoring critical parameters.

You report the following parameters to the OSS:

"* RPV level is at 150".

"* Reactor power is at 2%.

At this point the OSS directs you to reestablish injection with feedwater.

Is this direction correct? (YES or NO)

If YES, identify the reason it is correct.

If NO, identify the additional considerations necessary to reestablish injection.

a. YES Injection may be reestablished when power lowers to <5%. There is no restriction on RPV level.

b. YES Injection may be reestablished when power lowers to <5% and RPV level is <158".

c. NO Injection must remain terminated until RPV level lowers to less than +119.5".

d. NO Injection must remain terminated until RPV level lowers to less than +87".

ANSWER: d Answer: This is a recent change to the Level/Power control strategy. A 1C05 operator should know that level must be lowered to < +87" in this situation.

REFERENCE:

ATWS /L-1 & 2 Rev 10. DAEC EOP Bases Document Rev 7 ATWS - RPV Control page 25 - 34.

Distracter 1: Reactor level must be < +87" and power <5% to re-establish injection.

Distracter 2: Reactor level must be < +87" and power <5% to re-establish injection. When in an ATWS but not L/P control, operators often use the Lo-Lo level as a limit.

Distracter 3: Reactor level must be < +87" and power <5% to re-establish injection. Operators are often directed to maintain level at 158" during an ATWS scenario.

K/A System: 295037 SCRAM Condition Present And Power Above APRM Downscale Or Unknown (EOP).

K/A Number: EK1.02 Knowledge of the operational implications of the following as they apply to SCRAM CONDITION PRESENT AND POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Reactor water level effects on reactor power.

K/A Value: 4.1/4.3 DAEC Objective 95.51.03.05 Number:

DAEC Objective Differentiate between the entry conditions and strategies employed in power/level Statement: control to protect against thermal-hydraulic instability and to protect primary containment Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W-xm. doc - 22 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Cognitive Level: 3SPK - The plant Level/Power control. The candidate must determine that the OSS has given an incorrect order for the plant conditions. They must also determine this is incorrect based on RPV level above 87 inches.

Source: Bank 1999 Retake exam, 2001 Practice Audit exam Operationally EOP actions and Point of Emphasis in EOP Bases Validity:

OE:

Estimated Complet tion Time: EB#

Time Validation: N/A F (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 23 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

13. The plant has suffered a major event and SAG entry was required.

Hydrogen concentrations in the Drywell have reached 5.1%.

You are directed to perform SEP 303.3 "CAD Purge for H2 Control In SAGs" The following plant conditions currently exist:

"* "B" RHR Loop is being used for Drywell and Torus Sprays.

"* MO-2010 "RHR Loops A/B Cross-Tie Header Isolation Valve" is CLOSED.

"* "A" RHR Loop is in Torus Cooling.

"* Drywell Pressure is 25 psig and steady.

"* MO-2000 "RHR Loop "A" Inboard Drywell Spray Valve" is bound CLOSED.

"* Torus Level is 11.2 ft and stable.

Can N2 from the CAD system be purged into the containment with the current plant condition?

If YES, where can it be injected?

If NO, why can it not be injected?

a. YES. It can be injected into the "A" RHR Torus Spray Header.

b. YES. It can be injected into the "A" RHR Drywell Spray Header.

c. NO. MO-2000 must be OPENED.

d. NO. Drywell pressure above the High Drywell pressure interlock.

ANSWER: b Answer: With "B" RHR in Drywell and Torus Spray mode the "B" RHR Loop can not be used for CAD injection. The "A" RHR Loop is available for Drywell injection only.

The Torus spray header is a single header and common to both "A" and "B" RHR Loops unlike the Drywell Spray headers which are separate (Common misconception). Drywell pressure is below 30 psig, which is the pressure at which CAD will isolate to the Drywell. MO-2000 is upstream of the CAD injection valves (Common misconception).

REFERENCE:

SD 573 Rev 4 pages 42 and 43. SEP 303.3 Rev 4 pages 1-4. P&IDs Bech M119 and 120.

Distracter 1: CAD can be injected through the Torus Spray Header. However, it can not be injected while the header is being used for spring the Torus. The "B" RHR Loop is using the common header for sprays so this makes this injection path unavailable.

Distracter 2: CAD does inject at MO-2000. However, it injects down stream and being closed would not effect the injection of CAD.

Distracter 3: There is a Drywell pressure interlock at 30 psig. However, pressure is below the setpoint and will allow CAD injection.

K/A System: 50000 High Containment Hydrogen Concentration K/A Number: 2.1.23 Ability to perform specific system and integrated plant procedures during different modes of plant operation.

.* K/A Value: 3.9/4.0 Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 24 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

DAEC Objective 42.01.01.06 Number:

,* DAEC Objective Describe the Primary Containment and Containment Atmosphere Monitoring and Statement: Control System interlocks, including purpose, setpoints, logic, and which valves are affected.

Cognitive Level: 2RI - The candidate must understand the interactions between Drywell pressure, RHR, and CAD based on given plant conditions.

Source: New Operationally EOP Defeat action and system knowledge requirements for EOP action.

Validity:

OE: Design concerns during Noble Metal addition and engineering determination which followed to address the concerns.

Estimated Completion Time: EB#

Time Validation: N/A F1 (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 25 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

14. Which of the following is designed to initiate AUTOMATIC actions to protect the Main Turbine from over heating of the last stage buckets?

(Assume NO operator actions)

a. High Condenser Backpressure, 7.5" Hg absolute.

b. MSR high level, 3 inches below the MSR Shell.

c. HI-HI exhaust hood temperature, 225 0 F

d. Main Turbine HI-HI Vibration, 10 mils.

ANSWER: a Answer: This is the main reason and the setpoint given in the SD for the 7.5" Hg Abs. HI Condenser backpressure trip.

REFERENCE:

SD 693.1 Rev 4 page 40. ARP IC07A A-2 Rev 6 page 3 Distracter 1: This is a turbine trip signal. However, this is not the purpose of this trip. This will prevent damage to the turbine from water entering the turbine.

Distracter 2: This is an indication the last stage buckets are over heating and calls for operator actions to trip the turbine manually. However, there are no automatic actions other than an annunciator and the ARP for this alarm require manually tripping the turbine on a confirmed high temp. This is also referenced in the Turbine Trip ARP.

Distracter 3: Turbine high vibration could occur as a result of last stage over heating and the alarm comes in at 10 mils. However, there is no automatic action that occurs. This was a turbine trip at one time but was removed and the turbine HI-HI vibration ARP requires a manual turbine trip for this alarm. This is also referenced in the Turbine Trip ARP.

K/A System: 295002 Loss of Main Condenser Vacuum K/A Number: 2.1.27 Knowledge of system purpose and/or function K/A Value: 2.8/2.9 DAEC Objective 49.01.01.09b Number:

DAEC Objective List the setpoints associated with Main Condenser Backpressure and describe the Statement: relationship between those setpoints and each of the following components or events:

b. Main Turbine Trip Cognitive Level: 1B - This is the bases behind the Turbine Trip from Main Condenser Backpressure at 7.5" Hg Absolute.

Source: New Operationally ARP setpoint and system knowledge.

Validity:

OE: Turbine trips on loss of vacuum and high vibration has occurred at DAEC.

Estimated Completion Time: EB#

Time Validation: N/A LI (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

SApproved By: Date: Date:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 26 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 27 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

15. The plant was at full power when a "Station Blackout" occurred.

Assume the OSS directed you to perform each of the following EOP actions.

Which one of the following actions can be accomplished AND will perform the system function directed by the OSS?

(Assume manual operation of MOTOR OPERATED valves is performed when required)

a. Install Defeat 16 "Containment Atmosphere Monitoring Sample Line Isolation Defeat" to re establish H2 sampling of the Drywell.

b. Lower Torus water level by draining the Torus to Radwaste per 01 149 "RHR System" section 12.0 "Draining to Radwaste".

c. Install Defeat 4 "Drywell Cooler Isolation Defeat" to reduce Drywell air temperature.

d. Install AIP 404 "Injection with Fire Water" then initiate Torus and Drywell sprays to reduce Torus and Drywell pressure.

ANSWER: d Answer: Although the RHR pumps and MOVs are AC powered and would be deenergized, Spray the Drywell and Torus can be accomplished by lining up the Diesel Fire pump to the RHR header and manual operation of the MOVs. This would require considerable time but it is achievable.

REFERENCE:

EOP 2 Rev 9. EOP 2 Bases Document Rev 8 pages 44 and 46. AIP 404 Rev 5 pages 1-5. Defeat 16 Rev 2pages 2 and 3. Defeat 4 Rev 6 page 5. 01 149 Rev 78 pages 72 and 73.

1: Defeat 16 is directed in the PC/H leg of EOP 2. However, with the loss of AC SDistracter power the SVs and sample pumps would not function.

Distracter 2: The valves used to accomplish this are MOVs. However, draining to Radwaste required RHR in service and with a loss of AC power this is not possible.

Distracter 3: Defeat 4 is directed in the EOPs for elevated Drywell temperatures that are expected during SBO. However, with loss of AC fans and dampers would not function.

K/A System: 295003 Partial or complete loss of AC Power K/A Number: 2.4.6 Knowledge of symptom based EOP mitigation strategies K/A Value: 3.1/4.0 DAEC Objective 2.01.01.06a Number:

34.05.01.02 DAEC Objective Given an RHR system operating mode and various plant conditions, predict how Statement: the RHR system will be impacted by operation, or failure of the following support system(s):

a. Essential 4160/480 VAC electrical power supplies Identify the appropriate procedures that govern the fire protection system operation, include operator responsibilities during all modes of operation, and any actions required by personnel outside of the control room Cognitive Level: 2R1 - The candidate must understand the relationship between the Fire suppression system and RHR to come to this conclusion.

~'* Source: New Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO -Wxm.doc - 28 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operationally Validity:

, OE:

Estimated Completion Time: EB#

Time Validation: N/A J] (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 29 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

16. A plant transient occurred from full power.

A SCRAM was successfully inserted.

RPV injection is secured with the exception of CRD.

The MSIVs are open.

Reactor Water level is 280 inches on the Floodup Level indicator and slowly rising.

Reactor pressure is 145 psig and slowly raising.

Which of the following is required for the CURRENT plant conditions?

a. Initiate Shutdown Cooling and drain to Radwaste.

b. Manually open SRVs as needed to provide a drain path.

c. Reduce Load Set until the Main Steam Bypass Valves open.

d. Close the MSIVs until RPV level is restored and MSLs drained.

ANSWER: d Answer: ARP 1C05A D-1 requires closure of MSIVs prior to flooding the MSLs. This is at about 258 inches as indicated on the floodup instrumentation.

REFERENCE:

ARP 1C05A D-1 Rev 4 Distracter 1: This is a plausible choice because this is one method to lower RPV level. However, The SDC interlock will not allow this until RPV pressure is less than 135 psig.

Also this would not drain the MSL.

Distracter 2: This could be possible because the SRVs would drain water to the Torus at this pressure. The SRVs are used for water flow during some EOP conditions.

However, this is not a condition that would warrant water flow through the SRVs and the 01 P&Ls specifically say not to manually open the SRVs if the potential for two-phase flow exist unless directed by the EOPs.

Distracter 3: With the MSIVs open the Bypass valves could possibly drain to the condenser.

However, using Load Set would not work because the lowest pressure the Bypass Valves will control on Load Set is 150 psig.

K/A System: 295008 High Reactor Water Level:

K/A Number: AA1.03 Ability to operate and/or monitor the following as they apply to HIGH REACTOR WATER LEVEL: Main steam system K/A Value: 3.1/3.1 DAEC Objective 48.04.01.01 Number:

DAEC Objective Evaluate the precautions and limitations, operating cautions, or procedural notes Statement: of 01 683 to any component or Main Steam System operating status Cognitive Level: 1P - The candidate has to recognize the MSLs are flooded and ARP actions are required for high RPV level.

Source: New Operationally ARP action required by ARP and an industry event at another NMC plant.

Validity:

OE: Yes - Monticello Estimated Completion Time: EB#

Time Validation: N/A F1 (time) Incorrect Ratio Data: (ratio) 0%

Question Developed By: Peer Checked By:

Validated By: S'Operator Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 30 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.cdoc -31 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

17. The plant is at full power.

Drywell temperature is 130OF and rising due to a Well Water problem.

The OSS directs you to, "Vent the Drywell IAW 01 573 "Containment Atmosphere Control System" section 6.1 "Normal Containment Venting".

"B" SBGT is started and Containment Rad levels are normal.

You have lined up and are venting the Drywell and maintaining Drywell pressure as directed.

RIM-7606A "A RB Vent Shaft Rad Monitor" fails downscale.

Which one of the following is correct concerning the venting of the Drywell?

Drywell venting ...

a. is NOT effected because RIM-7606A is not associated with "B" SBGT.

b. can NOT be re-established with either a Group 3A or B Isolation signal present.

c. can be re-established once RIM-7606A input to Group 3A logic is bypassed and the Containment Vent Path is selected to Drywell on 1C05.

d. can continue but only (the small valve) CV-4310 "Inboard DW Vent CV-4302 Bypass Valve" is available with the Group 3A isolation signal present.

ANSWER: c Answer: The rise in Drywell air temperature will cause Drywell pressure to rise. The ARP for Hi pressure directs investigation as to why pressure is increasing. It also directs venting the Drywell to maintain <1.5 psig. After vent starts the Group 3A isolation occurs. The venting can be re-established as long as no 2 psig Drywell pressure is received.

REFERENCE:

01 573 Rev 61 pages 17, 18, and 20. SD 573 Rev 4 page 11-13, 60, and 61.ARP 1C05B C-8 Rev 16 pages 1, 5, and 8.

Distracter 1: It is correct that RIM 7606A effects the "A" SBGT train. However, It will cause Group 3A isolation and even though the "B" SBGT was running the vent path is secured on either an "A" or "B" isolation signal.

Distracter 2: The venting will stop upon receipt of the Group 3A isolation. However, there are provisions to override most of these signal and in this case it is possible to recover the venting.

Distracter 3: CV 4302 (Large 18 inch valve)is isolated on a Group 3A and CV-4310 (Small 2 inch valve) also isolates on a Group 3A signal. CV-4303 is down steam of both of these valves and it is isolated on a Group 3B logic. The arrangement of A and B logic is often confused. In any case if either an "A" or "B" isolation occurs venting is stopped. In this case venting though CV-4310 can not continue unless overrides are inserted.

K/A System: 295012 High Drywell Temperature K/A Number: 2.4.31 Knowledge of annunciators, alarms, and indications, and use of the response instructions.

K/A Value: 3.3/3.4 DAEC Objective 68.01.01.08b Number:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 32 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

DAEC Objective Given a Primary Containment Ventilation System operating mode and various Statement: plant conditions, predict how the Primary Containment Ventilation System will be impacted by failures in the following support systems: b. Primary Containment Isolation System Cognitive Level: 2RI - The candidate will have to understand the interrelationships between PCIS, SBGT, and Containment venting.

Source: New Operationally Expected operator knowledge of elevated Drywell temperature and its effect on Validity: the plant along with operator actions.

OE:

Estimated Completion Time: EB#

Time Validation: N/A 111 (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -33 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

18. The plant is at full power.

One loop of RHR is in the Torus Cooling mode with full RHR and RHRSW flow.

At this point a Safety Relief Valve (SRV) fails FULL OPEN.

The other loop of RHR is quickly placed in Torus Cooling and flows maximized.

Which of the following CORRECTLY describes the expected iesponse of Torus water temperature if the SRV CAN NOT be closed?

a. Torus water temperature will still be LOWERING with only one loop of Torus Cooling on and LOWER EVEN FASTER when the second loop of Torus Cooling is placed in service.

b. Torus water temperature will STABILIZE with only one loop of Torus Cooling on and BEGIN TO LOWER when the second loop of Torus Cooling is placed in service.

c. Torus water temperature will RISE with only one loop of Torus Cooling on and BEGIN TO LOWER when the second loop of Torus Cooling is placed in service.

d. Torus water temperature will RISE with only one loop of Torus Cooling on and CONTINUE TO RISE when the second loop of Torus Cooling is placed in service.

ANSWER: d Answer: An operator should know that the RHR system Torus Cooling mode is not designed to keep up with a stuck open SRV. This is the most limiting Torus cooling event with Max Temperature reaching 194°F.

REFERENCE:

UFSAR 6.2.1.3.3.3 Distracter 1: Temp goes up and continues to go up.

Distracter 2: Temp goes up and continues to go up.

Distracter 3: Temp goes up and continues to go up.

K/A System: 295013 High Suppression Pool Temperature.

K/A Number: AK2.01 Knowledge of the interrelations between HIGH SUPPRESSION POOL TEMPERATURE and the following: Suppression Pool Cooling.

K/A Value: 3.6/3.7 DAEC Objective 2.01.01.08 Number:

DAEC Objective State the purpose of the RHR system Statement:

Cognitive Level: 1-B Source: Bank 2001 NRC Exam Operationally dominate operator action on PRA Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A [] (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

\_* Operator Validated By:

Approved By: Date: _ Date:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 34 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 35 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

19. AOP 915 "Shutdown Outside Control Room" has been entered and the transfer has been completed.

Which of the following is correct?

a. Div I RHR and Core Spray logics are disabled. LLS and ADS are disabled.

b. Div I RHR and Core Spray logics are disabled. ADS and Group 1 are disabled.

c. Div II RHR and Core Spray logics are disabled. LLS and ADS are disabled.

d. Div II RHR and Core Spray logics are disabled. LLS and Group 1 are disabled.

ANSWER: c Answer: Note in AOP 915 explains the functions which are disabled

REFERENCE:

AOP 915 TAB lRev 22 Distracter 1: LLS and ADS Are disabled. However Div II RHR and CS are disabled not Div I Distracter 2: ADS is disabled. However, we expect a Group 1 and purposely leave the Mode Switch in Run to ensure it occurs. Div II RHR and CS are disabled not Div I Distracter 3: LLS and Div II RHR and CS are disabled. However we expect a Group I and purposely leave the Mode Switch in Run to ensure it occurs.

K/A System: 295016 Control Room Abandonment K/A Number: AK2.02 Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following: Local Control Stations K/A Value: 4.0/4.1 DAEC Objective 94.28.06.02 Number:

~ DAEC Objective State the effect of manipulating each of the emergency transfer switches, for the Statement: Remote Shutdown Panel system, including the bases for operating the YELLOW transfer switches last Cognitive Level: 2RI Source: New Operationally AOP action which removes control from the control room Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A [] (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 36 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

20. Review the indication given in the Support Material Booklet The plant was at power.

A transient has occurred requiring entry into EOP 4.

An Off-site release is in progress.

Review the attached panel indications.

Which one of the following systems has failed to isolate and what is the source of the steam leak?

(Assume in each case the indicated system is the only failure and all other systems have responded correctly)

a. HPCI steam leak in Radwaste.

b. RWCU steam leak in Steam Tunnel.

c. MSL drain steam leak in Turbine Building.

d. RCIC steam leak at Barometric Condenser.

ANSWER: c Answer: MSL drains have failed to isolate and are leaking in the Turbine Building. A Group 1 has isolated the MSL.

REFERENCE:

SD 683 Rev 3 page 63. BECH-M114 Rev 64.

Distracter 1: HPCI is in a normal Standby line up. CV2234 to RW is normally open. This is actually from the HPCI Turbine condensate pump and would be low energy water a break down stream would still be inside the Reactor Building.

Distracter 2: RWCU has isolated. However, RWCU is indicating an outboard isolation. The leak in this case would be in the Drywell or RB 2nd floor not the Steam Tunnel. Also a Steam tunnel leak would have isolated HPCI, RCIC and MSL drains.

Distracter 3: RCIC is showing normal running indications. A leak at the Barometric Condenser is possible with these indications be it would not lead to a leak outside secondary containment unless there is a failure in Secondary Containment. All others systems were assumed to have functioned so a leak would not be possible.

K/A System: 295017 High Off-Site Release Rate K/A Number: AA2.04 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: Source of off-site release.

K/A Value: 3.6/4.3 DAEC Objective 95.71.06.06 Number:

DAEC Objective Use control room indications and plant personnel to determine the source and Statement: location of a primary system discharge outside of containment Cognitive Level: 2RI - The Candidate must recognize the relationship between PCIS and various other Systems and what the expected response would be to a leak.

Source: New Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A Fm (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 37 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm. doc - 38 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

21. The plant is at Full power.

A partial loss of GSW occurs.

Which of the following components has exceeded an AUTOMATIC trip point and is required to be tripped manually?

a. "A" RFP with oil temperature at 2450 F.

b. The Main Generator with Isolation Bus Temperature at 185 0 F.

c. The "A" Recirc MG Set with oil temperature at 220 0 F.

d. Both "A" and "B" EHC pumps with EHC Tank temperature at 140 0 F.

ANSWER: C Answer: The Recirc MG set should have received a Trip signal A 210OF and operator action is require if a trip did not occur but should have to prevent equipment damage.

The reason the operator has to secure the "A" MG is because the oil temperature has exceeded the trip set point and operation standards require the operator to

REFERENCE:

take the action if an automatic action has failed.

AOP 411 Rev 14, ARP 1C04A A-4 Rev 14 Distracter 1: The LO temp will rise on loss of GSW. However operator actions are required because there is no Automatic trip associated with RFP lube oil temp high.

Distracter 2: Loss of GSW will cause ISO Bus duct temperatures to rise. However there is no Automatic feature to trip the Main Generator. Operator action is required.

Distracter 3: Loss of GSW will cause EHC tank temperatures to rise. However there is no pump trip associated with high oil temp. Operator action is required.

S'K/A System: 295018 Partial or Complete Loss of Component Cooling Water K/A Number: AK3.03 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

Securing individual components (prevent equipment damage)

K/A Value: 3.1/3.3 DAEC Objective 28.01.01.06c/d/e/h Number:

94.16.01.04/4 DAEC Objective Given a General Service Water System operating mode and various plant Statement: conditions, predict how each supported system will be impacted by failures in the General Service Water System:

c. Recirc MG System

d. Condensate and Feedwater System

e. Main Generator System

h. EHC System State when AOP-411, GSW Abnormal, directs the following:

4. One or Both Recirc MG Sets Tripped Cognitive Level: 2RI Source: New Operationally AOP and ARP action Validity:

OE:

- Estimated Completion Time: EB#

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 39 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Time Validation: N/A V- (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 40 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

22. A complete loss of the service and instrument air systems occurs while at full power.

Operators are trying to restart a compressor as air header pressure lowers, but they may be required to SCRAM the reactor.

Which one of the following is a potential effect on the Control Rod Drive (CRD) Hydraulic System?

a. The scram discharge volume may fail to isolate when the scram occurs.

b. Cooling flow could be lost to the control rod drive mechanisms before the scram occurs.

c. The running CRD pump could trip due to operating at no flow conditions.

d. The Backup scram valves could fail to vent the Scram Air Header when the scram occurs.

ANSWER: b Answer: The Flow control valves fail closed on a loss of air. This will cause a loss of cooling flow and drive water flow. This ensures all CRD flow will be directed to the HCUs through the charging water header for insert flow through the SCRAM Valves.

REFERENCE:

AOP 518, revision 19 Distracter 1: SDV vents and drains fail CLOSED, not OPEN.

Distracter 2: Flow control valve fail closed, but water is still supplied to the charging water header CVs and the minimum flow line.

Distracter 3: Backup scram valves are solenoid valves.

K/A System: 295019 Partial or Complete Loss of Instrument Air K/A Number: AA2.02 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Status of safety-related instrument air system loads K/A Value: 3.6/3.7 DAEC Objective 10.01.01.05.f Number:

DAEC Objective Given a Control Rod Drive Mechanisms and Hydraulic System operating mode Statement: and various plant conditions, predict how the Control Rod Drive Mechanisms and Hydraulic System will be impacted by the following support system failures:

f. Instrument and Service Air Cognitive Level: 2RI Source: Bank 1999 NRC Exam, 2001 ILC Audit Exam Operationally AOP. ARP, and system knowledge.

Validity:

OE: DAEC has experienced several Instrument Air System failures in the past.

Estimated Completion Time: EB#.

Time Validation: N/A -] (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -41 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

23. Review the indication given in the Support Material Booklet The plant is currently at 90% power.

I&C Technicians are performing a TIF on a problem in 1C43 "Division 1 Core Spray Vertical Board".

A CIMS alarm is received.

Which of the following is correct concerning the indications provided?

a. The "B" Drywell Well Water isolation valves have isolated. The "A" Drywell Well Water isolation valves have not isolated. If no operator actions are taken the Reactor will scram.

b. Both "A" and "B" Drywell Well Water isolation valves have isolated. If no operator actions are taken the Reactor will scram.

c. A signal is present for the Well Water Isolation but not all valves have isolated. If no operator actions are taken the reactor may scram depending on the current valve positions.

d. A signal is present for the Well Water Isolation but the signal has an override in effect. A Reactor scram will not occur.

ANSWER: b (NOTE: Attach a picture of the PCIS Status showing a Group 7 isolation with no override in effect and no other isolations)

Answer: The indications are for an inadvertent Group 7 isolation. All valves are indicating closed and an isolation signal is present. The candidate must understand a reactor SCRAM will occur under these conditions on Drywell pressure of 2 psig unless operator action is taken.

REFERENCE:

BECH Ell sheet 16 Rev9, BECH E113 sheet 94 Rev 16, SD 959.1 Rev 3 Distracter 1: Group 7 has only a "B" side logic. The Rx will scram on 2 psig Drywell if no actions are taken. However, all valves "A" and "B" are closed on the isolation signal.

Distracter 2: The green light signifies all valves have positioned properly Distracter 3: The amber light indicates an isolation signal is present there is another amber light to the left which indicates an override is in effect.

K/A System: 295020 Inadvertent Containment Isolation K/A Number: AA1.02 Ability to operate and/or monitor the following as they apply to INADVERTENT CONTAINMENT ISOLATION: Drywell ventilation/cooling system K/A Value: 3.2/3.2 DAEC Objective 68.01.01.08a/b Number:

DAEC Objective Given a Primary Containment Ventilation System operating mode and various Statement: plant conditions, predict how the Primary Containment Ventilation System will be impacted by failures in the following support systems:

a. Well Water System

b. Primary Containment Isolation System Cognitive Level: 3SPR Source: New Operationally Control Board indications, system response and plant effects Validity:

OE: Plant transients have occurred during I&C troubleshooting at DAEC and other stations.

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 42 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Estimated Completion Time: EB#

Time Validation: N/A - (time) Incorrect Ratio Data: (ratio) 0/

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 43 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

24. Given the following Conditions:

The plant is operating at full power.

The "A" CRD pump Tripped.

The "B" CRD pump has NOT been started.

Which one of the following statements describes the effect the loss of CRD pumps has on the plant?

a. The Control Rods will NOT fully SCRAM with RPV pressure alone due to pressure equalization across the drive piston.

b. The Control Rods can still be SCRAMMED, but the insertion time will be significantly longer.

c. Multiple Control Rods will begin to drift if a pump is NOT returned to service within 15 minutes.

d. Control Rod Drive mechanism temperatures will begin to rise.

ANSWER: d Answer: If the CRD pump remains off an extended time the CRD high temperature annunciator will eventually come in as CRD temperatures rise. With a loss of cooling water flow at power CRD high temperature alarms are expected

REFERENCE:

ARP 1C05A E-6 Rev 3. SD 255 Rev 7 Pages 44-48.

Distracter 1: There would actually be a large DP across the drive piston that would drive the rod in at normal SCRAM speed. The Rods would fully insert.

Distracter 2: At power rod insertion time would not be longer due to high RPV pressure. At low RPV pressure this could be a concern if accumulator pressure was allowed to lower.

Distracter 3: A CRD Accumulator pressure will start to lower. However There is no reason for a Rod drifts to occur.

K/A System: 295022 Loss of CRD Pumps K/A Number: AA2.03 Ability to determine and/or interpret the following as they apply to LOSS OF CRD PUMPS: CRD Mechanism Temperatures K/A Value: 3.1/3.2 DAEC Objective 10.01.01.04 Number:

DAEC Objective Evaluate plant conditions and control room indications to determine if the Control Statement: Rod Drive Mechanisms and Hydraulic System is operating as expected, and identify any actions that may be necessary to place the Control Rod Drive Mechanisms and Hydraulic System in the correct lineup.

Cognitive Level: 3PEO Source: Bank - 2000 Clinton unit 1 NRC Exam slightly modified for DAEC.

Operationally Validity:

OE:

Completion Time: SEstimated EB#

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 44 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Time Validation: N/A L-- (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

S'Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 45 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

25. Review the indication given in the Support Material Booklet The plant is currently at 45% power.

An SRV is leaking by to the Torus.

Which of the following is correct concerning the Torus Temperature indicator on 1C03?

(Assume the ONLY parameter changing is Torus Temperature)

a. The indicator is receiving a SINGLE input signal. Before the RED band is reached an EOP 1 entry is required.

b. The indicator is receiving a SINGLE input signal. Before the YELLOW band is reached an EOP 1 entry is required.

c. The indicator is receiving MULTIPLE input signals. Before the RED band is reached an EOP 1 entry is required.

d. The indicator is receiving MULTIPLE input signals. Before the YELLOW band is reached an EOP 1 entry is required.

ANSWER: c (Attach a picture of 1C03 of the Torus water temp indicator.)

Answer: This indicator has multiple inputs and if inputs are lost the amber light next to the indicator would be on indicating the inputs are not averaged. The red band starts at 110OF and EOP 2 requires entry into EOP 1 prior to 110OF

REFERENCE:

EOP 2 Rev 9 step T/T-4 Distracter 1: The red band does require EOP 1 entry but the indicator is receiving multiple input based on the amber light being out.

Distracter 2: The yellow band requires EOP 2 entry on Torus Temperature of 95 0 F and the indicator is receiving multiple inputs based on the amber light being out.

Distracter 3: The indicator is receiving multiple inputs but the yellow band requires EOP 2 entry on Torus Temperature of 95 0 F.

K/A System: 295026 High Suppression Pool Water Temperature K/A Number: EA1.03 Ability to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL WATER TEMPERATURE: Temperature monitoring.

K/A Value: 3.9/3.9 DAEC Objective 95.00.00.17 Number:

DAEC Objective Evaluate plant status and control room indications to determine the applicability Statement: and affect of any EOP Curve or Limit Cognitive Level: 2RI Sourcar------- New Operationally -c÷EOP break point for SCRAM required Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A r-1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

. Operator Validated By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 46 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-W-xm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 47 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

26. A loss of Drywell cooling occurred, causing the crew to SCRAM (successfully) from Full power.

When Average Drywell Air temperature could NOT be restored and maintained <280'F the crew Emergency Depressurized the RPV.

The OSS has performed a crew brief to inform the crew that the "Sat Curve" has been entered.

(EOP-2, Graph 1)

The following applicable parameters are at these stable values:

Both Recirc pumps are tripped.

S Indicated NR GEMAC RPV level 190" S Indicated WR Yarway RPV level 200" S Indicated WR GEMAC (Floodup) RPV level 200" 0 Indicated Fuel Zone RPV level 190" Which value for RPV water level, if any, should be used for EOP decision-making purposes?

a. 167"

b. 190"

c. 200"

d. None, all RPV level indicators should be considered unreliable.

ANSWER: a Answer: This strategy has changed drastically since over the past several years. See training material for full explanation.

REFERENCE:

EOP-2; EOP-2 Bases Rev 8 page 38.

Distracter 1: Indicators can be used but only with a -23" penalty.

Distracter 2: WR Yarway cannot be used after ED; WR GEMAC cannot be used until TSC evaluates.

Distracter 3: Possible misconception; this was the correct answer for years until EOP-2 was revised and system modifications were performed. Level indication reliability should be questioned, but indications are stable, thus OK.

K/A System: 295028 High Drywell Temperature K/A Number: EK1.01 Knowledge of the operational implications of the following as they apply to HIGH DRYWELL TEMPERATURE: Reactor water level measurement K/A Value: 3.5/3.7 DAEC Objective 95.59.03.01 Number:

DAEC Objective For any given entry condition, step, Caution, or Continuous Recheck Statement in Statement: EOP 2, explain the bases for the statement Cognitive Level: 2RI Source: BANK 1999 Retake Exam, 2001 Practice Audit Exam Operationally EOP caution and adequate core cooling concern if the wrong level instruments are Validity: used for level control.

OE:

Estimated Completion Time: EB#

• Time Validation: N/A F11 (time) Incorrect Ratio Data: (ratio)  %

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 48 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Question Developed By: Peer Checked By:

Operator Validated By:

SApproved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 49 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

27. Review the indication given in the Support Material Booklet EOP 2 directs entry into EOP 1 if Torus Water Level CANNOT be maintained BELOW 13.5 ft.

Review the attached picture.

At what point does Torus water level reach the level that requires EOP 1 entry from EOP 2?

a. A

b. B

c. C

d. D ANSWER: b (Attach a picture of the Torus to DW vacuum breakers with point A-D to choose from)

Answer: EOP 2 directs EOP 1 entry when level cannot be maintained below 13.5 ft which corresponds to the Bottom of the Torus-DW Vacuum Breakers. The candidate will have to understand the reason a SCRAM is required is that at this water level containment integrity could be challenged and that while there is no SCRAM signal on Torus level entering EOP 1 requires the Reactor to be shutdown.

REFERENCE:

EOP 2 Rev 9, EOP Bases Document EOP Breakpoints Rev 4 page 8 Distracter 1: Bottom of ring header is another setpoint for ED. However, the point shown is not the bottom elevation of the ring header. 13.8 ft is the level of the ring header the point shown is a transition piece from the DW downcomer to the ring header.

There is no EOP level associated with this point Distracter 2: This is when the Vacuum breakers would be covered with water and Containment integrity could be challenged. There is no EOP level associated with this point Distracter 3: This is when the Vacuum breakers would be covered with water and Containment integrity could be challenged. There is no EOP level associated with this point K/A System: 295029 High Suppression Pool Water Level K/A Number: EK3.03 Knowledge of the reasons for the following responses as they apply to HIGH SUPPRESSION POOL WATER LEVEL: Reactor SCRAM K/A Value: 3.4/3.5 DAEC Objective 93.00.00.16 Number:

DAEC Objective Evaluate plant status and control room indications and determine when a manual Statement: scram shall be initiated.

Cognitive Level: 2DR Source: New Operationally EOP Break point containment failure will Torus level above this point is possible Validity: under some accident conditions.

OE:

Estimated Completion Time: EB#

Time Validation: N/A F-1 (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

By: SApproved Date: Date:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 50 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -51 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

28. Which of the following curves, if followed, will prevent system damage due to air entrainment?

a. ECCS Vortex Limit

b. RHR NPSH Limit

c. Heat Capacity Limit

d. Core Spray NPSH Limit ANSWER: a Answer: Only the Vortex limit is concerned with air entrainment.

REFERENCE:

EOP Bases EOP curves and limits Rev 5 page 67 Distracter 1: Viable suppression pool limit. However does not protect against air entrainment.

Distracter 2: Viable suppression pool limit. However does not protect against air entrainment.

Distracter 3: Viable suppression pool limit. However does not protect against air entrainment.

K/A System: 295030 Low Suppression Pool Water Level K/A Number: EK1.02 Knowledge of the operational implications of the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: Pump NPSH K/A Value: 3.5/3.8 DAEC Objective 95.00.00.18 Number:

DAEC Objective Evaluate the possible consequences of exceeding any EOP Curve or Limit on the Statement: mitigation of an event Cognitive Level: 1B Source: Bank: Quad Cities 2000 NRC exam (restructured)

• ,> Operationally EOP curve and equipment failure prevention.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A L-- (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 52 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

29. The plant is at full power.

During transfer of RWCU resin, a pipe break released resin to the RB 2nd Floor North area outside the RWCU Phase Separator Room.

EOP 3 was entered on Radiation Levels in ONE area above MAX SAFE.

No other EOP 3 entry conditions exist.

The resin has been contained in this area.

The following High Radiation Areas have been posted due to this event:

• RB 1st Floor Northwest 0 RB 2nd Floor North and SBGT Room RB 3rd Floor Northwest

"* RB North Stairwell below RB 4th Floor

"* ALL other area radiation levels are normal Which alarm is consistent with this event?

a. 1C03A D-9 "RBCCW RM-4820 HI RAD"

b. 1C04C A-1 "RADWASTE EFFLUENT RIS-3972 HI RAD"

c. 1C04B B-6 "NEW FUEL STORAGE AREA ARM HI RAD"

d. 1C04B D-7 "RHRSW AND ESW EFFLUENT RM-1997 OR RM-4268 HI RAD" ANSWER: a Answer: The RBCCW process rad monitor is located in the overhead area where this resin spill would be. The alarm would be expected with elevated rad levels.

REFERENCE:

SD 414 Rev 4 page 10, ARP 1C03A D-9 Rev 6 Distracter 1: The Radwaste building is located next to the RB West wall, which is near this area. However the Radwaste Effluent Process Rad Monitor is not near this area.

Distracter 2: This ARM is located in the RB. However, the detector is located on the refuel floor and would not be alarming under the given conditions.

Distracter 3: ESW piping is located near this area. However the process rad monitor is located outside the TB and would not be effected by this event.

K/A System: 295033 High Secondary Containment Radiation Levels K/A Number: EK2.02 Knowledge of the interrelations between HIGH SECONDARY CONTAINMENT RADIATION LEVELS and the following: Process radiation monitoring system.

K/A Value: 3.8/4.1 DAEC Objective 85.00.00.05 Number:

DAEC Objective Evaluate plant conditions and control room indications to determine if the PRM Statement: system is operating as expected and identify any actions that may be necessary to place the PRM system or the plant in the correct condition Cognitive Level:- -S Source: New Operationally EOP indications. The candidate should be able to distinguish normal alarms based Validity: on rad monitoring indications and given rad conditions in the plant.

OE:

Estimated Complet-ion Time: EB#

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 53 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Time Validation: N/A F[1 (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 54 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

30. Which of the following requires operator verification of Secondary Containment Isolation and entry into EOP 3?

a. RB Vent Shaft Rad Monitor Rad levels reading 10 mR/hr.

b. Offgas Vent Pipe above the HI-HI Rad Trip setpoint.

c. Fuel Pool Exhaust Rad levels reading 10 mR/hr.

d. MSL HI-HI Rad at 500 mR/hr.

ANSWER: c Answer: EOP 3 entry is required if the Fuel Pool Exhaust Rad levels reach 9 mR/hr and this also corresponds to a Group 3 isolation and SBGT start.

REFERENCE:

EOP 3 Rev 15, ARP 1C05B C-8 Rev 16 Distracter 1: RB Vent Shaft Rad Monitor Rad levels at 11 mR/hr will cause the conditions stated. However the reading given is below that setpoint. Neither action would occur.

Distracter 2: Offgas Vent Pipe above the HI-HI Rad Trip setpoint causes a Group 3 isolation and SBGT start. However this is not an EOP 3 entry. Note this was an EOP 3 entry at one time.

Distracter 3: MSL HI-HI Rad will cause a Group 1 isolation and may lead to an EOP 3 entry.

However this will not generate a Group 3 and is not an EOP 3 entry condition.

K/A System: 295034 Secondary Containment Ventilation High Radiation.

K/A Number: EA1.03 Ability to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Secondary Containment Ventilation K/A Value: 4.0/3.9 DAEC Objective 95.00.00.22 Number:

7.02.01.02 DAEC Objective For any given EOP, state the entry condition for the EOP Statement:

List the signals which cause a SBGT System auto initiation including setpoints and logic. Describe how they are bypassed and how they are reset.

Cdghitive Level: 11 Source: New Operationally EOP entry and knowledge of proper system response during a plant transient.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A FI (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 55 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

31. EOP 4" Radioactive Release Control" requires Emergency Depressurization before certain conditions are reached.

What purpose does Emergency Depressurization achieve AS IT RELATES TO EOP 4?

a. This establishes or maintains adequate core cooling.

b. This places the primary system in a low energy condition and reduces the driving head to the leak.

c. Reduces the energy in the RPV before reaching conditions where the primary containment will not accommodate an SRV opening.

d. This places the RPV in a low energy condition before reaching conditions where a loss of coolant accident could not be adequately contained in the primary containment.

ANSWER: b Answer: The purpose of ED during EOP 4 is to remove the driving head to the leak outside secondary containment and place the RPV in a low energy condition.

REFERENCE:

EOP Bases EOP 4 Rev 4 page 9, EOP Bases ED Rev 6 page 1 Distracter 1: This is a purpose for ED. However, this is not the purpose associated with EOP 4 Distracter 2: This is a purpose for ED. However, this is not the purpose associated with EOP 4 Distracter 3: This is a purpose for ED. However, this is not the purpose associated with EOP 4 K/A System: 295038 High Off-Site Release Rate K/A Number: EK3.04 Knowledge of the reasons for the following responses as they apply to HIGH OFF-SITE RELEASE RATE: Emergency Depressurization.

K/A DAECValue:

Objective 3.6/3.9 95.71.01.04 Number:

DAEC Objective Explain how the mitigation strategies used in EOP 4 accomplish the purpose of Statement: EOP 4 Cognitive Level: 1B Source: New Operationally EOP bases Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A F-l (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 56 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

32. A fire occurs in the Cable Spreading Room.

(Assume the Fire Protection System is lined up for AUTOMATIC ACTUATION and Cable Spreading Room access is restricted)

The installed fire protection system automatically actuates.

The room must be entered to determine if the fire has been extinguished.

(1) What is the classification of the fire that is expected in this area?

AND (2) What safety hazard, from the automatic system actuation, should be considered prior to operators entering the Cable Spreading Room?

(1) (2)

a. Class B Electrical Shock from water spray

b. Class C Electrical Shock from water spray

c. Class B Suffocation from oxygen depletion due to the discharge of C02 in the area

d. Class C Suffocation from oxygen depletion due to the discharge of C02 in the area ANSWER: d

'* Answer: This is a Class C electrical fire and the installed fire suppression system is CARDOX.

REFERENCE:

01 513 Rev 59 pages 4 and 6. Safety Rule book.

Distracter 1: Class B is an Oil type fire and not an expected type fire in the cable spreading room and there is no installed water systems in the Cable Spreading Room Distracter 2: There is no installed water systems in the Cable Spreading Room Distracter 3: Class B is an Oil type fire and not an expected type fire in the cable spreading room.

K/A System: 600000 Plant Fire On Site K/A Number: AK1.01Knowledge of the operation applications of the following concepts as they apply to PLANT FIRE ON SITE: Fire Classifications by type.

K/A Value: 2.5/2.8 DAEC Objective 34.05.01.01 Number:

Fire Brigade Training DAEC Objective Relate the precautions and limitations, operating cautions, or procedural notes of Statement: 01-513 to any component or fire protection system operating status Fire Classification Cognitive Level: 11 Source: Bank from Dresden 2000 NRC Exam (modified)

Operationally Personnel safety and plant fire system knowledge. Also all operators are trained Validity: as part of the fire brigade.

OE: The CARDOX system has inadvertently actuated at DAEC and there are industry events for C02 hazards.

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W-xm.doc -57 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Estimated Completion Time: EB#

Time Validation: N/A I-I (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 58 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

33. SRO ONLY The plant is shutdown.

The "B" side of Shutdown Cooling (SDC) is in service with "B" RHR and "B" RHRSW pumps running.

MO1940 "RHR HX 1E-201B Bypass" and MO1939 "RHR HX 1E-201B Inlet Throttle" valves are THROTTLED in mid position.

MO 1904 and MO 1905 "RHR Loop "B" Inject Isolation Valves" are OPEN.

MO 1908 and MO 1909 "RHR Shutdown Cooling Suction Isolation Valves" are OPEN.

Annunciator 1C03B B-4 "RHR SHUTDOWN COOLING SUCTION HEADER HI PRESSURE" alarms and SDC Header pressure is reported to be 105 psig and rising at 2 psig/min.

You direct the 1C03 operator to raise the cooldown rate.

Several minutes later the 1C03 operator reports RHR suction header pressure is 145 psig and MO1940 is not responding.

(No other plant conditions have changed and no alarms have occurred)

Which of the following is the correct action?

a. Direct the 1C03 operator to secure SDC per 01 149" Residual Heat Removal System". Enter AOP 302.2 "Loss of Alarm Power".

b. Direct the 1C03 operator to throttle OPEN more on MO1939 and start the "D" RHR pump if necessary. Enter T.S. for LPCI.

c. Direct the 1C03 operator to CLOSE M01905, Trip the "B" RHR pump and CLOSE MO1908 and MO 1909. Enter AOP 149 "Loss of Decay Heat Removal".

d. Direct the 1C03 operator to CLOSE M01939, Start the "D" RHR pump and then re-establish SDC flow. Enter AOP 149 "Loss of Decay Heat Removal" until SDC is re-established.

ANSWER: c Answer: The initial alarm indicates an increase in RPV temperature and pressure. The ARP directs increasing the cooldown rate to lower pressure, which was directed.

At 135 psig a PCIS group 4 should have occurred but did not. ARP 1C05B D-8 "PCIS Group 4 Isolation" should be in alarm and SDC secured. The OSS should direct the actions from the ARP that did not occur. In this case securing SDC is appropriate. Also entry into AOP 149 is directed.

REFERENCE:

1C05B D-8 Rev 9 page 1-3. 1C03B B-4 Rev 5 page 1.

Distracter 1: ARP 1C03B B-4 directs securing SDC per 01 149 if an isolation is imminent. An alarm has not been received for a Group 4 isolation. This should have occurred at 135 psig. The ARP for a Group 4 should be carried out and there is no reason to enter AOP 302.2. The loss of alarm power would not prevent the isolation and plant condition would have changed.

Distracter 2: ARP 1C03B B-4 directs increasing cooldown with MO 1939 and another pump would help flow. T.S. should be entered on failure of M01940. However, the plant is above the PCIS Group 4 pressure and SDC should be promptly removed and isolated.

Distracter 3: Starting a second RHR pump would increase flow. AOP 149 entry is correct when SDC is lost and recovery of SDC will be the goal. However, the plant is above the PCIS Group 4 pressure and SDC should be promptly removed and isolated as directed in ARP 1C05B D-8.

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 59 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

K/A System: 295021 Loss of Shutdown Cooling K/A Number: 2.4.50 Ability to verify system alarm setpoints and operate controls identified in

/ the alarm response manual.

K/A Value: 3.3/3.3 10 CFR 55.43(b)(5)

DAEC Objective SRO 5.01.08.01 Number:

DAEC Objective Evaluate plant conditions and respond to problems relating to the decay heat Statement: removal method being used.

Cognitive Level: 3SPK Source: New Operationally Recognition of loss of SDC and failure of PCIS.

Validity:

OE: DAEC has experienced failure of PCIS that required operator actions. We have also lost SDC.

Estimated Completion Time: EB#

Time Validation: N/A I-I (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 60 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

34. Refueling operations are in progress with the Mode Switch in "REFUEL".

Which of the following allows the withdrawal of a control rod AND ensures the reactor will remain shutdown?

a. The One-Rod permissive interlock and adequate shutdown margin designed into the core.

b. The RSCS Group Selector switch selected to the correct sequence and adequate shutdown margin designed into the core.

c. Refueling Rod Block Interlocks and the proper rod withdrawal sequence loaded into the RWM.

d. The RSCS Mode Selector Switch placed in "WITHDRAW" and the proper rod withdrawal sequence loaded into the RWM.

ANSWER: a Answer: The One-Rod Permissive along with adequate SDM designed into the core ensures the reactor will remain shut down with any rod withdrawn from the core during refueling operations and refueling accident conditions. Operationally the One-Rod Permissive is designed to prevent inadvertent criticality during refueling operations. However, this feature depends on the appropriate SDM designed into

REFERENCE:

the core.

UFSAR/DAEC-1 Rev 14 page 15.4-3. SD 856.1 Rev 4 pages 28and 29.

Distracter 1: Adequate shutdown margin is correct and the RSCS Group select switch is used during rod withdrawals. However, RSCS provides back lighting only.

Distracter 2: Refuel Floor interlocks are part of the One-Rod permissive logic. However, the RWM proper rod sequence is used during power operations.

Distracter 3: This sounds appropriate but the RSCS switch does not provide for this function.

K/A System: 295023 Refuel Accidents K/A Number: AK1.02 Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS: Shutdown margin.

K/A Value: 3.2/3.6 DAEC Objective 72.02.01.05 Number:

72.03.01.08 DAEC Objective Describe the Reactor Manual Control System interlocks, including purpose and Statement: setpoints Evaluate reactor status and control room indications to determine which of the different methods of control rod movement is allowed, including an explanation of the criteria used in the evaluation Cognitive Level: 2RI Source: New Operationally The one-rod permissive is an important feature to prevent inadvertent criticality Validity: during shut down conditions and is based on having adequate SDM designed into the core.

OE: Error in SDM have occurred in the industry.

Estimated Completion Time: EB#

Time Validation: N/A F] (time) Incorrect Ratio Data: (ratio) 0/

Question Developed By: Peer Checked By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -61 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 62 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

35. Review the indication given in the Support Material Booklet The plant is at full power.

Annunciator 1C04B B-4 "Steam Leak Detection Ambient HI Temp" was received.

RCIC has failed to isolate and temperatures were noted to be rising slowly.

The operator manipulates the Steam Leak Detection panel as Shown.

Which of the following is correct concerning the RCIC Room?

a. Room Differential Temperature is 50 0 F.

b. Room Differential Temperature is >50 0 F.

c. Room Ambient Temperature is 155 0 F.

d. Room Ambient Temperature is >175 0 F.

ANSWER: d Answer: RCIC room ambient is in alarm, which occurs at 175°F as indicated by TS-2450B light. With RCIC failure to isolate and temperatures increasing the room temperature must be somewhere above 175 0 F.

REFERENCE:

SD 858 Rev 3 pages 14 and 15.

Distracter 1: The operator has taken the switch to set, which displays the setpoint. If the student mistakes the reading for 50OF (upper scale) then they would incorrectly read this as 50 0 F. This is a possible error that has been made with this type indicator and can lead to misinformation being communicated to the OSS.

Distracter 2: The operator has taken the switch to set, which displays the setpoint. If the student mistakes the reading for 155 0 F (lower scale) then they would incorrectly read this as 155 0 F. This is a possible error that has been made with this type indicator and can lead to misinformation being communicated to the OSS.

Distracter 3: The indication on TDS-2445B is for its "setpoint" of 50°F Delta T on the upper scale. The operator has taken the switch to set, which displays the alarm setpoint.

Someone not familiar with this type instrument may assume the set position would set the meter for the current room temperature reading, in this case 155 0 F.

However, the room temperature is > 175 0 F.

K/A System: 295032 High Secondary Containment Area Temperatures K/A Number: EA2.01 Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURES: Area Temperature K/A Value: 3.8/3.8 DAEC Objective 75.00.00.02 Number:

75.01.01.04 DAEC Objective Evaluate plant conditions and control room indications to determine if the Steam Statement: Leak Detection System is operating as expected, and identify any actions that may be necessary to place the Steam Leak Detection System in the correct lineup Describe the operation of the following Steam Leak Detection System components:

a. Temperature Selector Switches (TSS)

b. Differential Temperature Selector Switches (TDSS)

Cognitive Level: 2DR Source: New Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 63 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operationally Validity:

K OE: Error in reading Steam Leak Detection has occurred in the simulator and caused crews to take incorrect actions based on the indication communicated by the operator at the Steam Leak Detection Panel. In this case if the operator had read the RCIC room temperature the same way he would have reported RCIC room temperature at 175 0 F and steady. In reality room temperature could be in excess of 300OF MAX Safe and would require EOP 1 and SCRAM.

Estimated Completion Time: EB#

Time Validation: N/A F-1 (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 64 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

36. Main Plant Exhaust Plenum pressure is rising.

Which of the following will AUTOMATICALLY initiate to provide the indicated protective function?

a. The Main Plant Exhaust Fans sequentially SHIFT to High Speed to prevent collapsing the Main Plant Exhaust Plenum.

b. The Reactor Building Exhaust Fans START to prevent Refuel Floor blow out panels from lifting.

c. The Reactor Building Supply fans TRIP to prevent over pressurizing Secondary Containment.

d. SBGT STARTS to draw at least a -.25 inch WG in the Reactor building.

ANSWER: c Answer: As Secondary Containment pressure increases the Main plant supply fans will start to shutdown at their start permissive setpoint. At -0.20 Inches WG all the RB supply fans will be off. This prevents overpressurization of Secondary Containment.

REFERENCE:

1C23C A-6 Rev 6 pages 1 and 2 Distracter 1: .--The Main plant exhaust fan starting is directed if needed. However, this is a manual operation and they do not have a fast speed position. There are fans in plant (Turbine Building Exhaust) that do have similar functions). DAEC has had damage to the Main Plant Exhaust Plenum in the past.

Distracter 2: Starting the Reactor Building exhaust fans would lower Secondary Containment pressure. However, these fans are manually started and their inlet dampers modulate to control the pressure in the Exhaust Plenum. The blow out panel will lift if RB pressure gets to high.

Distracter 3: SBGT is designed to maintain the RB negative at -.25 WG with an isolation.

However, there is no automatic start of SBGT on RB pressure although the OSS could direct this action if warranted but it would have to be done manually.

K/A System: 295035 Secondary Containment High Differential Pressure K/A Number: EK3.02 Knowledge of the reasons for the following responses as they apply to the SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: Secondary containment ventilation response.

K/A Value: 3.3/3.5 DAEC Objective 67.01.01.09 Number:

DAEC Objective Describe the Reactor Building fan control system including interlocks, starting Statement: permissive, auto operation, and trips on RB Supply and Exhaust fans, and Refuel Floor fan Cognitive Level: 2RI Source: New Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A F-' (time) Incorrect Ratio Data: (ratio) 9/

Question Developed By: Peer Checked By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 65 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 66 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

37. The plant has SCRAMMED from full power. While stabilizing the plant, an operator observes the following indications:

"* The running CRD pump current is above normal running amps.

"* CRD Cooling Water flow is DOWNSCALE.

Select the answer that CORRECTLY explains these indications and identifies the appropriate operator response.

a. The flow control station has failed; action must be taken as soon as possible to protect the CRDMs.

b. The in service CRD suction filter is clogged; the standby filter must be placed in service as soon as possible.

c. CRD pump flow is being diverted through the Scram Inlet valves to the CRDMs and the reactor; no further actions are necessary.

d. The running CRD pump is air bound and must be tripped as soon as possible.

ANSWER: c Answer: Under these condition water is being diverted to the accumulators send a high flow signal to the flow control valves causing them to close. This is expected under these conditions and no operator action is required.

REFERENCE:

SD 255 Rev 7 page 47 Distracter 1: Failure of the flow control station will also result in LOW current.

Distracter 2: A clogged drive filter will result in LOW amps on the motor.

Distracter 3: An air bound pump will have LOW current.

K/A System: 201001 CRD Hydraulic K/A Number: K5.02 Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD DRIVE HYDRAULIC SYSTEM: Flow indication.

K/A Value: 2.6/2.6 DAEC Objective #: 10.01.01.04 DAEC Objective Evaluate plant conditions and control room indications to determine if the Statement: Control Rod Drive Mechanisms and Hydraulic System is operating as expected, and identify any actions that may be necessary to place the Control Rod Drive Mechanisms and Hydraulic System in the correct lineup.

Cognitive Level: 2RI Source: Bank- 1999 NRC exam, 2001 RO Audit exam.

Estimated Completion Time: EB#

Time Validation: N/A n'I (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 67 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

38. SRO ONLY K>j An unisolable leak on the RWCU system has caused radiation levels to increase in the Reactor Building.

The BOP operator reports the "RWCU PUMP AREA" is above Maximum Safe for BOTH Temperature and Radiation levels.

Multiple areas indicate radiation readings above Maximum Normal.

Which of the below correctly list the required actions and the basis for those actions?

a. Commence a normal reactor shutdown, due to multiple areas above the Maximum Normal Radiation and one area being greater than Maximum Safe Temperature.

b. Insert a manual SCRAM, due to one area being greater than Maximum Safe and a primary system discharging into the secondary containment.

c. Perform an Emergency Depressurization, based on the potential of exceeding Maximum Safe Radiation in two areas.

d. Perform an Emergency Depressurization, based on exceeding two Maximum Safe conditions.

ANSWER: b Answer: Radiation levels are at the point the EOP directs isolating the leak from The RPV.

However, they have not gotten to the point where a SCRAM is required although it is an option the OSS could exercise.

THE CANDIDATE IS PROVIDED A COPY OF EOP 3

REFERENCE:

EOP 3 Rev 15

~ Distracter 1: A normal shutdown may be commenced. However, there is no EOP 3 requirement at this point to shutdown.

Distracter 2: If the same parameter reaches max safe in more than one area an ED would be required. However, this is a wait until statement not a before statement. The OSS is to wait until the parameter is exceeded.

Distracter 3: ED is required if max safe is exceeded in two or more areas. However, this is if it is the same parameter. In this case the max safe rads and temperature are in the same area and ED is not warranted.

K/A System: Generic K/A Number: 2.4.1 Knowledge of EOP entry conditions and immediate action steps.

K/A Value: 4.3/4.6 10 CFR 55.43(b)(5)

DAEC Objective SRO 6.67.01.06 Number:

DAEC Objective Evaluate plant conditions and control room indications and determine actions Statement: directed by EOP 3 Cognitive Level: 3SPR Source: Bank - 1998 NRC Exam slightly modified.

Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A I-] (time) Incorrect Ratio Data: (ratio)

• ., Question Developed By: Peer Checked By:

Operator Validated By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 68 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 69 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

39. SRO ONLY The plant is at power.

The "Rod Select Power Switch" 51 is to be immediately replaced in response to a GE SIL for problems with CR2940 type switches.

It is determined that power to the switch will have to be tagged out and de-energized for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

You are the OSS and you are reviewing the work prior to release.

1) Which person(s) is/are the minimum level of approval required to approve this work prior to the tagout being hung AND

2) what procedure directs this approval?

a. 1) Plant Manager

2) 01 856.1 "Reactor Manual Control System"

b. 1) Plant Manager

2) Technical Specifications Bases

c. 1) Operations Manager and Reactor Engineer

2) 01 856.1 "Reactor Manual Control System"

d. 1) Operations Manager and Reactor Engineer

2) Technical Specifications Bases ANSWER: a S'Answer: The tagout will de-energize RMCS by removing power from the ROD SELECT MATRIX and providing a Select Block. This will prevent any rod motion except by SCRAM. The Plant Manager has to give prior approval to perform work which remove the RMCS from service. The only place that gives this direction is in 01

REFERENCE:

856.1.

01 856.1 Rev 21 pages 4 and 16. SD 856.1 Rev 4 pages 10 and 16.

Distracter 1: The Plant Manager must approve this work. However, T.S. Bases does not cover RMCS.

Distracter 2: 01 856.1 is the correct location for this information. However, the Plant Manager must give the approval to de-energize RMCS.

Distracter 3: T.S. Bases does not cover RMCS and the Plant Manager must give the approval to de-energize RMCS.

K/A System: 201002 Reactor Manual Control System K/A Number: A2.03 Ability to (a) predict the impacts of the following on the REACTOR MANUAL CONTROL SYSTEM: and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Select block.

K/A Value: 2.9/2.8 10 CFR 55.43(b)(5)

DAEC Objective SRO 1.12.02.01 Number:

DAEC Objective Evaluate the request for its potential to adversely affect plant operation.

Statement:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 70 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Cognitive Level: 3PEO - The candidate must recognize the effect de-energizing RMCS power will have on RMCS and then determine there is a P&L to have the Plant Manager pre approve this work.

Source: New Operationally Validity:

OE: The plant removed the rod select matrix at one time during power operations. A commitment was made to only remove RMCS from service if needed for RMCS failures that potentially restricts/impact plant operations and then only with Plant Managers approval.

Estimated Completion Time: EB#

Time Validation: N/A F (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc 71 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

40. Both Reactor Recirculation pumps were running at 70% speed when an internal component failure in the "B" MG SET SPEED CONTROL caused the controller speed demand output signal to instantaneously fail to the MAXIMUM value.

Which of the following CORRECTLY describes the expected affect of this failure on core flow?

Core flow will rise until...

a. the "B" Recirc Scoop Tube Positioner reaches its ELECTRICAL STOP.

b. the "B" Recirc Scoop Tube Positioner reaches its MECHANICAL STOP.

c. a "B" Recirc Scoop Tube Positioner LOCK-UP occurs due to high Milliamp output signal from the Controller.

d. a "B" Recirc Scoop Tube Positioner LOCK-UP occurs due to high deviation between the Controller speed demand and the Positioner position.

ANSWER: d Answer: The scoop tube will automatically lockup if High Speed Demand vs. Positioner Feedback reaches .05 vdc. This is the condition indicated.

REFERENCE:

SD 264 Rev 5, ARP 1C04B C-2 Rev 6 Distracter 1: Positioner has an electrical stop, but the deviation lockup should occur much sooner.

Distracter 2: Positioner has an mechanical stop, but the deviation lockup should occur much sooner.

Distracter 3: Milliamp output is something that is checked after a lockup but it does not cause the lockup.

K/A System: 202002 Recirculation Flow Control System K/A Number: A3.03 Ability to monitor automatic operations of the RECIRCULATION FLOW CONTROL SYSTEM including: Scoop tube operation.

K/A Value: 3.1/3.0 DAEC Objective 12.00.00.02c Number:

DAEC Objective Identify the conditions that allow or cause the following events to occur Statement: c. Scoop tube lockup Cognitive Level: 1I Source: New Operationally Understanding of reactivity control system reason during an inadvertent Validity: reactivity addition event.

OE:

Estimated Completion Time: EB#

Time Validation: N/A F1 (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

• / Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 72 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

41. The plant is being shut down.

All Control Rods have been inserted.

RPV pressure is 220 psig.

RHR is in Standby/Readiness Condition per 01 149 "RHR System" with NO pumps running.

The "B" Recirculation pump is RUNNING.

The "A" Recirculation pump has been SECURED.

The "A" Recirculation pump Suction, Discharge, and Discharge Bypass Valves are OPEN.

Which of the following will occur if RPV water level were to drop to 119.5"?

(The DEFAULT loop is selected for injection and Drywell pressure is NOT rising)

a. The "A" Recirc Loop Discharge and Discharge Bypass valves will receive a CLOSE signal.

b. The "A" RHR Loop LPCI Inboard and Outboard Injection valves will receive a CLOSE signal that is sealed in for 10 minutes from the time that a loop is selected.

c. The "B" Recirc pump will trip after the Discharge and Discharge Bypass valves CLOSE.

d. The "B" RHR Loop Inboard LPCI Inject valve will OPEN and RHR will immediately inject into the RPV.

ANSWER: b Answer: This is normal system operation for a break detected in the "A" Loop or no breaks detected.

REFERENCE:

SD-149 Rev 8 pages 20 and 26 Distracter 1: --This would be true if the "A" loop where the default loop or a break was detected in the "B" loop. The Loop Select logic place the Recirc system in a position to detect which Loop has a break or if there is no break the "B" Loop will select by default.

Distracter 2: The "B" Recirc pump will trip and during most cases this is the trip function that occurs. However, in single loop operation LPCI Loop Select can not determine the Loop with a break if unless both Recirc pumps are secured. The logic will trip the running pump if it doesn't sense both pumps running.

Distracter 3: Th "B" LPCI inject valve will open and be ready to inject and RPV pressure is below the shutoff head of the RHR pumps. However, the RHR pumps do not start on a LPCI Loop Select signal. There has not been an initiation signal to start the pumps. There is no 2 psig drywell pressure and RPV level is above 64 inches.

K/A System: 203000 RHR/LPCI: Injection Mode K/A Number: A4. 11 Ability to manually operate and/or monitor the control room: Indicating lights and alarms.

K/A Value: 3.7/3.5 DAEC Objective RO 2.03.01.22 Number:

DAEC Objective Describe how the RHR system responds to a LPCI Loop Select/LPCI auto Statement: initiation signal during various modes of operation and any actions that are required to verify proper response.

Cognitive Level: 3PEO - The Candidate will have to predict the system response of both Recirc and RHR based on system response to RPV water level reaching 119.5 inches.

Source: ILC Exam Bank - slightly modified Operationally ECCS system response to a ECCS signal.

Validity:

S OE: Systems have failed to respond to automatic signals at DAEC and in the industry.

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W-xm. doc - 73 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Estimated Completion Time: EB#

Time Validation: N/A 1] (time) Incorrect Ratio Data: (ratio) 0/

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-W xm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 74 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

42. SRO ONLY The plant is at full power.

Review the following time line.

2005: The RO reports Drywell Identified leakage is 27.3 gpm and Unidentified leakage is 0.5 gpm.

2009: 1C05B B-2 "Primary Containment HI/LO Pressure" alarms on high pressure.

2014: The OSS directs a SCRAM that is successful.

2015: 1C05B A-1 "Primary Containment Hi Pressure Trip" due to a steam leak in the Drywell.

2020: Torus and Drywell Sprays are direct, but are unsuccessful. Drywell pressure is rising.

2045: RPV level is at +10 inches and lowering at 1 inch every 5 minutes with all available injection sources injecting.

2046: Emergency Depressurization is directed. When the first ADS/SRV is opened Torus pressure increases rapidly to 58 psig.

2050: Containment venting per SEP 301.1 "Torus Vent Via SBGT" is directed and successful.

Excluding EC/OSM Judgement, at what time does plant conditions FIRST meet the requirement(s) to sound the plant Evacuation Alarm?

a. 2005

b. 2015

c. 2045

d. 2050 ANSWER: b Answer: EPIP 1.3 requires plant evacuation at an Alert or greater. OSM Judgement is allowed for earlier sounding but this was not allowed in this case. The in-plant personnel will be notified when Drywell pressure reaches 2 psig due FA1 Alert conditions being satisfied and the required plant assembly.

THE CANDIDATE IS PROVIDED A COPY OF EAL TABLES

REFERENCE:

EPIP 1.3 Rev 9 page 3. 1C05B A-1 Rev 3 page 1. EAL 2 Rev 0 Fission Barrier Table.

Distracter 1: The plant is at an Unusual Event with Identified leakage at >25 gpm. The OSM could sound the evacuation alarm due to OSM Judgement. However, this was not allowed in the question and plant evacuation is not required at this time.

Distracter 2: The plant has reached conditions of a Site Area Emergency with the 2 psig Drywell condition and RPV level <+15 inches. If a Site Area Emergency were entered immediately then this would be the point the evacuation alarm would be sounded first. However, the conditions have progressed slowly enough that the evacuation would have already occurred.

Distracter 3: The plant is at a General Emergency due to the venting of the containment per the EOPs and the other conditions that exist. If the event were extreme the plant could reach this point before the evacuation alarm is sounded and it would be required at this point. However, the conditions have progressed slowly enough that the evacuation would have already occurred.

K/A System: 295010 High Drywell Pressure K/A Number: 2.1.14 Knowledge of system status criteria which require the notification of plant personnel.

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 75 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

K/A Value: 2.5/3.3 10 CFR 55.43(b)(4/5)

DAEC Objective SRO 3.01.03.01 SRO 3.01.03.02 DAEC Objective Explain the Responsibilities and Instructions contained in EPIP 1.3 Statement:

Demonstrate the ability to use EPIP 1.3, Note-04/06, to direct a Plant Assembly/Evacuation.

Cognitive Level: 3SPR - The candidate must determine when the plant has meet the conditions for an ALERT based on Drywell pressure and understand that a plant evacuation is required.

S ource: NEW Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A R] (time) Incorrect Ratio Data: (ratio) 0%

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 76 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

43. Following a full Group I Isolation at full power, HPCI received an auto initiation signal on RPV low level.

During the Group I Isolation, the HPCI pump flow signal was lost to the flow controller, sensing a constant ZERO gpm flow signal.

Which of the following describes the HPCI system response if NO operator action is taken?

The HPCI turbine will...

a. remain at minimum speed.

b. trip on high RPV water level.

c. trip as soon as the signal is lost.

d. trip on overspeed and remain shutdown.

ANSWER: b Answer: With no flow signal to the flow controller the controller will send a signal to the HPCI Turbine for max flow. This will try to run HPCI at max speed to increase the flow. RPV level will go to 211" at which time the HPCI turbine will trip on high level.

REFERENCE:

SD 154 Rev 4 Distracter 1: With no sensed flow, flow demand will cause HPCI to run at max speed to increase flow. This is the indication of a lost inverter.

Distracter 2: The flow signal is not required to initiate the HPCI start up.

Distracter 3: If the HPCI reaches the overspeed limit it will not stay shutdown. Speed will reduce to a point < the overspeed setpoint and the turbine will restart.

K/A System: 206000 High Pressure Coolant Injection K/A Number: K3.01 Knowledge of the effect that a loss of the HIGH PRESSURE COOLANT INJECTION SYSTEM will have on the following: Reactor water level control K/A Value: 4.0/4.0 DAEC Objective 5.00.00.02a Number:

5.02.01.03 DAEC Objective Describe the operation of the following principle HPCI System components: a.

Statement: HPCI Turbine Evaluate plant conditions and control room indications to determine if the HPCI System is operating as expected, and identify any actions that may be necessary to place the HPCI System in the correct lineup.

Cognitive Level: 3PEO / 2RI Source: Bank 2000 Dresden NRC exam, 1998 Cooper NRC exam Operationally Understanding of ECCS system design and operations.

Validity:

OE: HPCI failures challenge operators during scenarios.

Estimated Completion Time: EB#

Time Validation: N/A FI (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

K* Operator Validated By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc 77 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Approved By: _ Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 78 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

44. The plant is at full power.

Which of the following Core Spray pressure annunciators/indications describe the operator's indication that a leak has occurred in the Core Spray piping BETWEEN the Reactor Vessel and Core Shroud?

a. Core Spray Sparger HI A P (3.6 psig increasing)

b. Core Spray Sparger LO A P (2.46 psid decreasing)

c. Core Spray Discharge line low pressure (47.5 psig decreasing)

d. Core Spray Discharge line high pressure (100 psig increasing)

ANSWER: b Answer: This indicator is in place for this purpose and would be the only way to detect this condition prior to system initiation.

REFERENCE:

ARP 1C03A C-8 Rev 7; SD-151 Distracter 1: Common misconception that A P goes up. Break detection alarms at 2.46 psid decreasing Distracter 2: Indicates a leak but not in the described location.

Distracter 3: Possible misconception that the described leak pressurized CS discharge header.

K/A System: 209001 Low Pressure Core Spray K/A Number: A1.02 Ability to predict and/or monitor changes in parameters associated with operating the LOW PRESSURE CORE SPRAY SYSTEM controls including: Core Spray Pressure K/A Value: 3.2/3.4 DAEC Objective 4.01.01.02 Number:

DAEC Objective Evaluate plant conditions and control room indications to determine if the Core Statement: Spray System is operating as expected, and identify any actions that may be necessary to place the Core Spray System in the correct lineup.

Cognitive Level: 11 Source: Bank - 2001 NRC RO Exam Operationally Knowledge of ECCS designs which provide operators indications of system failure.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A F-* (time) Incorrect Ratio Data: (ratio) 0%

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 79 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

45. A transient has occurred while at power, which has resulted in a Bus 1A3 lockout followed by a hydraulic ATWS.

The 1C05 operator initiates Standby Liquid Control (SBLC) as directed.

Which of the following CORRECTLY describes the expected condition of the SBLC Squib valves based on their respective power supplies?

a. Both will have actuated; they are powered from respective divisions of 125 VDC.

b. Both will have actuated; they are powered from Uninterruptible AC.

c. Only one will have actuated; it is powered from B RPS.

d. Only one will have actuated; it is powered from its associated pump breaker on 1B44.

ANSWER: d Answer: Power for the SQUIB valves comes from the pump power supply line. In this case 1B44.

REFERENCE:

SD 153 Rev 4 page 29, ARP 1C05A F-3 Rev 3 Distracter 1: No power to A SBLC pump/Squib valve due to 1A3 lockout; power supplies are 1B34 &1B44.

Distracter 2: No power to A SBLC pump/Squib valve due to 1A3 lockout; power supplies are 1B34 &1B44.

Distracter 3: Power supply is 1B44 K/A System: 211000 Standby Liquid Control K/A Number: K2.02 Knowledge of electrical power supplies to the following: Explosive valves K/A Value: 3.1/3.2 DAEC Objective 6.00.00.03 Number:

DAEC Objective Statement:

Cognitive Level: IF Source: Bank - 1999 NRC Retest exam series A Operationally Knowledge of power supplies to safety equipment required for Reactor Shutdown Validity: in the event of an ATWS.

OE:

Estimated Completion Time: EB#

Time Validation: N/A Ft (time) Incorrect Ratio Data: (ratio) 0/

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 80 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

46. A turbine trip from full power has caused a reactor scram.

RPV level lowered to 110" during the initial transient but has been restored to the normal operating band for two minutes.

The scram has NOT been reset?

Select the answer that correctly describes the status of the RPS Backup Scram valves under these plant conditions.

Both Backup Scram valves should be...

a. energized and OPEN (venting)

.ý-b. energized and CLOSED (not venting)

c. deenergized and OPEN (venting)

d. deenergized and CLOSED (not venting)

ANSWER: a Answer: Note: ARI solenoids would have also have energized to vent, but would reset after 45 seconds. However, the backup scram valves do not reset until the scram is reset.

REFERENCE:

SD 358 Rev 5 pages 12 and 33 Distracter 1: Backup scram valves would be energized and open.

Distracter 2: Backup scram valves would be energized and open Distracter 3: Backup scram valves would be energized and open K/A System: 212000 Reactor Protection System K/A Number: K1.08 Knowledge of the physical connections and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the following: Control rod and drive mechanism.

K/A Value: 3.7/3.9 DAEC Objective 22.00.00.07 Number:

DAEC Objective Describe the operation of the following principle Reactor Protection System Statement: components:

f. Backup Scram Valves Cognitive Level: 1I Source: Bank - 1998 ILC Retest (Cover page indicates 1998 footer indicates 1999)

Operationally Power supplies to safety equipment necessary to shutdown the plant.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A Fm (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -81 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 82 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

47. You are directed by 01 878.2 "Intermediate Range Neutron Monitoring System" to closely monitor IRMs when switching between Range 6 and Range 7.

What is the reason for this direction?

a. The affects of Power Range Gamma's are starting to overcome the affects of Decay Gamma's and will cause power to show a faster rate of rise.

b. The SRM Rod Blocks are bypassed at this point and IRM spiking has occurred due to noise when the SRM circuitry bypasses the SRMs.

c. The Mean Square Analog Unit is switched into the circuit at this point and starts applying the Campbelling calculations to the IRM output.

d. The IRMs shift between Low and High Frequency Amplifiers at this point and could affect the IRM output signal.

ANSWER: d Answer: -This is described in the System Description for IRMs and is noted in the 01 because the shift from Low to High Frequency Amps could show an unexpected IRM reading.

REFERENCE:

01 878.2 Rev 19 page 6. SD 878.2 Rev 5 pages 13, 14, 16, 17, 19, and 27 Distracter 1: Power Gamma's do start to overcome the Decay and Background Gamma's at higher power. However, the IRMs are gamma compensated through all ranges and there is no significant effect at this point.

Distracter 2: .. The SRMs are automatically bypassed by IRM range switch position. However, this occurs at Range 8 and there is no noise from SRM circuitry from this switching. There is a caution about driving multiple SRMs and noise induced SCRAMs.

Distracter 3: The Mean Square Analog Unit does perform the Campbelling function for the IRMs, which is for gamma compensation. However, this is always functioning and there is no switching that occurs.

K/A System: 215003 Intermediate Range Monitor IKA Number: A4.07 Ability to manually operate and/or monitor in the control room: Verification of proper functioning/ operability K/A Value: 3.6/3.6 DAEC Objective 79.00.00.02 Number:

79.01.01.01 DAEC Objective Evaluate plant conditions and control room indications to determine if the IRM Statement: System is operating as expected, and identify any actions that may be necessary to place the IRM System in the correct lineup relate the precautions and limitations, operating cautions, or procedural notes of 01-878.2 to any component or IRM System operating status Cognitive Level: 1P Source: New Operationally 01 note and system knowledge.

Validity:

OE:

K... Estimated Completion Time: EB#

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 83 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Time Validation: N/A J-] (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 84 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

48. Review the indication given in the Support Material Booklet SRO ONLY The plant was at full power when a transient occurred that required insertion of a SCRAM.

The plant experienced an electrical ATWS.

Control Rods were successfully inserted with the individual test switches per the RIPs.

Plant conditions were stabilized.

Alarm 1C05A D-5 "SRM Downscale" is received.

The SCRAM has not been reset.

As OSS what direction should you give based on the indications given?

a. Direct the STA to write an AR for the "C" SRM downscale alarm.

b. Direct the BOP operator to perform AOP 375 "Loss of +/-24 VDC Power"

c. Direct the 1C05 operator to complete the IPOI 5 "SCRAM" Immediate Actions.

d. Direct the RO to have the Second Assistant check the status of the SRM Drive Motor relays.

ANSWER: c Answer: During an ATWS the 1C05 operator normally will drive the IRMs in but leave the SRMs out of the core until power is lowered. This often is missed following an ATWS. IPOI 5 directs this as an Immediate operator action.

REFERENCE:

IPOI 5 Rev 33 step 3.2 (7) page 5.

Distracter 1: SRM down scale comes in at 3 cps which is where the indication is currently.

There is sometimes confusion on downscale alarms several come in at 5% of scale or 5/120 scale. The indications are concurrent with the SRMs not inserted into the core and the "C" SRM reaching 3 cps.

Distracter 2: 24 VDC does supply the SRMs. However, the indications do not support loss of

+/-24VDC.

Distracter 3: There have been problems with drive relays burning up in the past. However, it probably would only effect one relay at a time and the SRMs show no sign of being selected to be driven into the core.

K/A System: 215004 Source Range Monitor System K/A Number: 2.4.49 Ability to perform without references to procedure those actions that require immediate operation of system components and controls.

K/A Value: 4.0/4.0 10 CFR 55.43(b)(5)

DAEC Objective SRO 4.21.03.01 Number:

DAEC Objective Verify the performance of IPOI 5 steps/sub-steps and evaluate plant conditions Statement: that will impact their performance. ( i.e.: Group 1 affects on monitoring condenser vacuum )

Cognitive Level: 1P - The Candidate has to determine that IPOI 5 Immediate action required SRMs to be inserted and following an ATWS this may need to be directed if the 1C05 operator does not perform the actions.

Source: New Operationally Validity:

OE:

Estimated Completion Time: EB#

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 85 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Time Validation: N/A I-] (time) Incorrect Ratio Data: (ratio) V%

Question Operator Developed By: Peer Checked By:

Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 86 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

49. The plant is at 85% power.

1C05A D-2 "APRM DOWNSCALE" annunciator alarms.

The "A" APRM indicates a downscale condition.

When does the APRM alarm occur AND with NO other plant changes what other annunciator will alarm?

a. Alarm at 12% and 1C05B D-6 "RWM ROD BLOCK"

b. Alarm at 5% and 1C05B D-6 "RWM ROD BLOCK'

c. Alarm at 12% and 1C05B A-6 "ROD OUT BLOCK'

d. Alarm at 5% and 1C05B A-6 "ROD OUT BLOCK" ANSWER: d Answer: At 5% APRM scale the Downscale alarm occurs. 6 sec later the Rod Block annunciator occurs. The Block is enforced immediately.

REFERENCE:

1C05B A-6 Rev 9, 1C05A D-2 Rev 3 Distracter 1: 12% power is the upscale Rod Block with the Mode Switch not in Run and the RWM Rod Block does not interface with the APRM at this point.

Distracter 2: 5% power is correct. However, the RWM Rod Block does not interface with the APRM at this point.

Distracter 3: A Rod Block will occur. However, 12% power is the upscale Rod Block with the Mode Switch not in Run K/A System: 215005 Average Power Range Monitor/Local Power Range Monitor K/A Number: K4.01 Knowledge of AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR design feature(s) and/or interlocks which provide for the following: Rod withdrawal blocks K/A Value: 3.7/3.7 DAEC Objective 81.01.01.03 Number:

81.01.01.07 82.00.00.04 82.01.01.02b Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 87 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

DAEC Objective Evaluate plant conditions and control room indications to determine if the APRM Statement: system is operating as expected, and identify any actions that may be necessary to place the APRM system or the reactor plant in the correct condition List the signals or conditions which cause an APRM system Downscale Alarm, Upscale Alarm, Inop or Upscale Trip, including setpoints and logic. Describe how they are bypassed List the signals which cause a RBM Rod Block including purpose, setpoint, and logic. Describe how they are bypassed and how they are reset Given a Rod Block Monitor System operating mode and various plant conditions, predict how the Rod Block Monitor System will be impacted by failures in the following support systems:

b. APRM System Cognitive Level: 1I Source: New Operationally ARP setpoint and reactivity monitoring system operation.

Validity:

OE:

Estimated Complettion Time: EB#

Time Validation: N/A FI (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 88 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

50. SRO ONLY The plant was at full power.

A SCRAM was manually inserted due to the loss of 120VAC Uninterruptible power.

RPV Level is being controlled at 160 inches.

SBLC was NOT initiated.

The RPS Trip Test Switches were placed in the TRIP position per RIP 101.1 "RPS Trip Test Switches".

The 1C05 operator reports IRMs are inserted and are reading 35 on range 6 and lowering.

Which of the following is correct for the given plant conditions?

a. Exit ATWS and enter EOP 1.

b. Remove RIP 101.1 then exit ATWS and enter EOP 1.

c. Exit the "Q" flowpath of ATWS and enter IPOI 5 "SCRAM".

d. Remain in all legs of ATWS and continue inserting the RIPs.

ANSWER: c Answer: The loss of UPS prevents us from determining Control Rod position. The EOPs have been written to allow for times when we can not verify all rods are fully inserted. However, we have other indications that the Reactor is Shutdown. The first CRS under the Q leg of ATWS allows exiting that leg and entering IPOI 5 if the Reactor is shutdown and no Boron was injected. IPOI 5 has allowances for use of RIPs that allows leaving RIP 101.1 in place and still exiting the Q leg. The EOP bases defines the "Reactor is Shutdown" as Reactor power < POAH which corresponds to about 20 on range 8. Here we are on range 6 and well below the POAH. The EOP bases explain that we do not want to take drastic actions if there is indication the reactor is shutdown.

THE CANDIDATE IS PROVIDED A COPY OF EOP ATWS and EOP1

REFERENCE:

EOP ATWS Rev 10. EOP ATWS Bases Rev 7 Page 80. EOP Bases Breakpoints Rev 4 page 2. AOP 357 Rev 27 page 3 Distracter 1: The first CRS in ATWS allows exiting ATWS when it has been determined the "Reactor is shutdown under ALL conditions without boron". The Reactor is shutdown at this point. However, the reactor is not shutdown under ALL conditions. Exiting ALL ATWS legs is not authorized because rod position is still not known.

Distracter 2: When exiting ATWS removal of RIPs/SEP/defeats is desirable but not a requirement.

Distracter 3: If the candidate does not understand the term "Reactor is Shutdown" then this would be the action taken. However, the CRS in the Q leg directs exiting that leg under the conditions given.

K/A System: 295015 Incomplete SCRAM K/A Number: AA 2.02 Ability to determine and/or interpret the following as they apply to INCOMPLETE SCRAM: Control rod position K/A Value: 4.1/4.2 SRO 10 CFR 55.43(b)(5)

DAEC Objective 6.50.01.04 Number:

DAEC Objective Evaluate plant conditions and control room indications and determine the actions

"\./ Statement: directed by ATWS Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 89 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Cognitive Level: 3SPR - The candidate must determine the procedure step that directs the action and have understanding of terminology for ATWS. They musts also understand how IPOI 5 actions complement ATWS for these conditions.

Source: New Operationally Validity:

OE: Based on a plant event, a loss of UPS, which resulted in a revision to EOP ATWS to allow return to IPOI 5 if the Reactor is Shutdown.

Estimated Completion Time: EB#

Time Validation: N/A FI (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 90 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

51. SRO ONLY A normal plant Shutdown from full power operations is in progress.

The GEMAC Reference Leg Backfill System has been out of service for 3 weeks.

RPV level is 190 inches and stable.

Which of the following is correct based on the GEMAC Reference Leg Backfill System status?

a. Direct cooldown at less than 40°F per hour to prevent "Notching" IAW IPOI 4 "Shutdown".

b. When RPV pressure reaches 400 psig stop the cooldown and direct the 1C05 panel operator to shift RPV level control to "Single Element" IAW 01 644 "Condensate and Feedwater Systems".

c. When RPV pressure reaches 500 psig direct the operating crew NOT to use the Yarway instruments on 1C05 for level indication IAW Plant Specific Technical Guidance from the BWR Owners Group.

d. Direct enhanced RPV Level monitoring during the Shutdown IAW 01 880 "Non-Nuclear Instrumentation System" Section J "Operation of the GEMAC Reactor Level Instruments Reference Leg Backfill System".

ANSWER: d Answer: Anytime the Reference Leg Backfill system is out of service for 7 days performance of 01-880, J-1 section 6.1 is required. Notching is expected when reducing RPV pressure during a shutdown. The Narrow range GEMAC level instruments are susceptible.01-880 directs these actions for the given plant conditions.

NOTE: This answer is longer than the others. However due to our exam writing guidelines when we give the document number we give the name of the dic-effent and in this case the document name is exceptionally long. If you remove the document information the answer is comparable in size and characteristic as that of the distracters.

REFERENCE:

01-880 Rev 9 pages 4, 15 and 16. IPOI 4 Rev 56 page 6 Distracter 1: Slowing the cooldown rate may reduce the magnitude and/or frequency of the notching. However, there is no direction in IPOI 4 for this cooldown rate. IPOI 4 does direct action be performed IAW 01 880 for these conditions.

Distracter 2: At 400 psig the cooldown is stopped and shifting to single element would effect the RPV level control process. However, this is not directed at this time and would not help with notching. IPOI 4 directs this action at 35% power.

Distracter 3: During a rapid cooldown greater than 100OF per hour we are directed to not use the Yarways on 1C05 when RPV pressure reaches 500 psig and decreasing.

However, during a normal shutdown these instruments would not be effected and could be used. Also notching would not render these instruments inoperable.

K/A System: 216000 Nuclear Boiler Instrumentation K/A Number: A2. 11 Ability to (a) predict the impacts of the following on the NUCLEAR BOILER INSTRUMENTATION: and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Heatup or cooldown of the reactor vessel.

K/A Value: 3.2/3.3 SRO 10 CFR 55.43(b)(5)

DAEC Objective 4.18.03 Number:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc -91 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

DAEC Objective Direct operator actions to control RPV level throughout the cooldown.

Statement:

SCognitive Level: 2RI Source: NEW Operationally Knowledge of conditions, which can affect the reliability of level instrumentation, Validity: is an important operator knowledge concern.

OE: Notching has occurred at DAEC during normal plant cooldowns before the Reference leg back fill system was installed.

Estimated Completion Time: EB#

Time Validation: N/A R] (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-W-xm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm. doc - 92 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

> 52. The plant is at 75% power.

The RCIC turbine has been started CST to CST using the flow-indicating controller.

RCIC is at 3000 RPM and stable.

Other plant conditions are normal with the exception of those systems lined up to support the RCIC testing.

Annunciator 1C04C C-9 " RCIC INVERTER POWER FAILURE" Alarms and RCIC Inverter failure is confirmed.

Which of the following is correct concerning the RCIC turbine?

RCIC will ...

a. remain running at 3000 rpm.

b. tripped on electrical overspeed.

c. tripped on mechanical overspeed.

d. be running at idle speed (approx. 1000 rpm).

ANSWER: d Answer: With an inverter failure the RCIC Turbine will reduce speed to idle about 1000 rpm if running on the flow controller.

REFERENCE:

1C04C C-9 Rev 2, SD 150 Rev 4 pages 36 and 37 Distracter 1: This is indication that would occur if the RCIC turbine were running on the "Test Pot".

Distracter 2: Possible error if candidate assumes a loss of inverter will cause the flow controller to send a zero flow signal and call for max flow. However, speed control is lost on loss of inverter causing the turbine to run at idle.

Distracter 3: Possible error if candidate assumes a loss of inverter will cause the flow controller to send a zero flow signal and call for max flow and the electrical overspeed power is lost. However, speed control is lost on loss of inverter causing the turbine to run at idle and the electrical overspeed is not affected.

K/A System: 217000 Reactor Core Isolation Cooling System K/A Number: K6.01 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR CORE ISOLATION COOLING SYSTEM: Electrical power K/A Value: 3.4/3.5 DAEC Objective 3.01.01.02 Number:

DAEC Objective given a RCIC system operating mode and various plant conditions, predict how Statement: the RCIC System will be impacted by the following support system failures:

o. RCIC Inverter Cognitive Level: 3PEO Source: New Operationally ARP notes and system operational knowledge.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A [m (time) Incorrect Ratio Data: (ratio)  %

- Question Developed By: Peer Checked By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 93 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 94 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

53. Review the indication given in the Support Material Booklet The Plant was at full power.

A loss-of-coolant accident has occurred resulting in a successful reactor scram.

RPV water level lowered to 110 inches before the trend was reversed.

A transient has occurred involving the RCIC system.

The only operator action taken with RCIC was to cycle the handswitch for MO-2405, RCIC Turbine Stop Valve, to the fully CLOSED position and then to hold it in the OPEN position for three seconds.

What is the status of RCIC based on these indications?

a. A RCIC Auto Isolation trip has occurred.

The RCIC Turbine trip is RESET.

b. A RCIC High RPV Level trip has occurred.

The RCIC Turbine trip is RESET.

c. A RCIC Electrical Overspeed trip has occurred.

The RCIC Turbine trip is NOT RESET.

d. A RCIC Mechanical Overspeed trip has occurred.

The RCIC Turbine trip is NOT RESET.

ANSWER: d

• Answer: The indications given are for a RCIC mechanical overspeed condition.

MO-2405 will close on electrical or mechanical overspeed but a limit switch in the open circuit prevents moving the valve in the open direction when the trip is due to a mechanical overspeed condition. This is thr0bnly indication in the control room of the type of trip that occurred.

REFERENCE:

ARPs 1C04C, A-5 and A-6 Distracter 1: Incorrect because MO-2404 OPEN indicates that a 211" trip has NOT occurred and MO-2405 is indicating that a mechanical overspeed trip has occurred, and the valve motor operator indications indicate that the trip cannot be reset from the control room.

Distracter 2: Incorrect because MO-2404 OPEN indicates that a 211" trip has NOT occurred and MO-2405 is indicating that a mechanical overspeed trip has occurred, and the valve motor operator indications indicate that the trip cannot be reset from the control room.

Distracter 3: Incorrect because MO-2404 OPEN indicates that a 211" trip has NOT occurred and MO-2405 is indicating that a mechanical overspeed trip has occurred, and the valve motor operator indications indicate that the trip cannot be reset from the control room..

K/A System: 217000 Reactor Core Isolation Cooling System K/A Number: A3.01 Ability to monitor automatic operations of the REACTOR CORE ISOLATION COOLING SYSTEM including: Valve operation K/A Value: 3.5/3.5 DAEC Objective 3.02.01.05 Number:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 95 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

DAEC Objective Evaluate plant conditions and control room indications to determine if the RCIC Statement: System is operating as expected, and identify any actions necessary to place the RCIC System in the correct lineup.

Cognitive Level: 2DR Source: Bank 2001 NRC RO Exam Operationally ARP actions and knowledge of level control equipment.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A R-1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 96 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

54. The plant was operating at power.

A transient occurred, a SCRAM was successfully inserted, and the MSIVs were closed.

RPV level is 190 inches and being controlled in Auto by the Feedwater Level Control (FWLC)

System.

SRVs are being used to manually control pressure between 800 and 1000 psig.

The 1C03 operator reports he is opening an SRV to reduce RPV pressure.

Which of the following would be the FIRST expected plant response in regards to cycling open the SRV?

a. An RPV High Level alarm due to void increase.

b. An RPV High Level alarm due to FWLC System response to increased steam flow.

c. An RPV Low Level alarm due to void increase.

d. An RPV Low Level alarm due to increased mass flowrate NOT sensed by the FWLC System.

ANSWER: a Answer: Under these conditions when the SRV is opened RPV pressure will drop causing the voids in the core region to expand. This will increase the flow resistance in the core. Level will increase initially then will decrease due to loss of mass and FWLC system response to the indicated level increase.

REFERENCE:

50007 IG 8 Rev 7 pages 30 and 37. RO Task 8.04 requires manual operation of SRVs and cycling of SRVs is performed for ARP 1C03A C-5 for an open SRV.

Distracter 1: An RPV High level will occur. However, the FWLC system will not increase level at a rate that would cause the alarm.

Distracter 2: Voids do increase and the candidate may incorrectly assume this would cause lower density and a corresponding lower indicated level.

Distracter 3: The mass flow rate loss is not sensed by the FWLC system because the SRVs are upstream of the steam flow sensors. However, there will be an initial increase in indicated water level due to the voiding.

K/A System: 218000 Automatic Depressurization System K/A Number: A1.05 Ability to predict and/or monitor changes in parameters associated with operating the AUTOMATIC DEPRESSURIZATION SYSTEM controls including:

Reactor water level.

K/A Value: 4.1/4.1 DAEC Objective 95.80.10.06 Number:

99.28.01.08 DAEC Objective Evaluate the RPV and containment response to opening an SRV, and determine if Statement: containment and the SRV and its support systems are functioning as expected.

Identify the actions to be taken if it is determined that an SRV is indicating open.

Cognitive Level: 3PEO Source: New Operationally Validity:

OE:

Estimated Completion Time: EB#

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 97 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Time Validation: N/A - (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 98 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

55. Review the indication given in the Support Material Booklet The plant is at full power.

A Group 3B isolation has occurred and verified to be complete.

You are ordered to override CV-4371A and the system responds as indicated in the picture on the following page.

What are the long-term implications of this system response?

a. All of the SRVs will eventually lose the ability to lift on their relief setpoint.

b. V2 of the SRVs will eventually lose the ability to lift on their relief setpoint.

The other Y2 are NOT effected.

c. All of the SRVs have lost N2 from outside the Drywell.

d. /2 of the SRVs have lost N2 from outside the Drywell.

The other % are NOT effected.

ANSWER: c Answer: With the indications given all SRVs have-a limited amount of N 2 available based on the capacity of the accumulators. The Group 3 isolated CV-4371A which goes to all the SRVs and MSIVs.

REFERENCE:

SD Distracter 1: The MSIVs would eventually go closed on a loss of N 2 and not be able to be opened and the SRVs will loss the external N2 source. However, this will not effect the relief setpoint of the SRVs, which is a common misconception.

Distracter 2: The MSIVs would eventually go closed on a lose of N2and not be able to be opened and the SRVs will loss the external N2 source. However, this will not effect the relief setpoint of the SRVs, which is a common misconception, and CV-4371A goes to all the SRVs not /2.

Distracter 3: The "B" side of Group 3 closes CV-4371A the "A" side closes CV-4371C that isolates N2 to the DW-Torus vacuum breakers. CV-4371A goes to all SRVs.

K/A System: 223001 Primary Containment System and Auxiliaries K/A Number: K1.08 Knowledge of the physical connections and/or cause-effect relationships between PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES and the following: Relief/safety valves K/A Value: 3.6/3.8 DAEC Objective 95.11.01.03 Number:

DAEC Objective Evaluate the effect of installation or restoration of each of the EOP defeats on Statement: plant systems and equipment Cognitive Level: 3PEO Source: New Operationally System knowledge and EOP action requirements.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A - (time) Incorrect Ratio Data: (ratio)

- Question Developed By: Peer Checked By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 99 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 100 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

56. Review the indication given in the Support Material Booklet A plant transient has occurred from full power.

Defeat 4 was inserted and verified correct.

Other EOP actions have been taken and some time later, the BOP observes the panel indications on the following page.

Which of the following BY ITSELF would account for the indications observed?

a. Drywell pressure is below 2 psig.

b. Drywell sprays have been initiated.

c. Drywell H2 or 02 concentration is >4%.

d. Drywell temperature is below the over temperature alarm setpoint.

ANSWER: b Answer: In this case defeat 4 has been inserted at some point and verified. When Drywell sprays are initiated the Drywell fans will automatically shutdown to prevent damaging the fans.

REFERENCE:

Defeat 4 Rev 5, SD 760 Rev 4 page 6 Distracter 1: Drywell pressure going above 2 psig will shift fans to slow but with the Defeat 4 installed fans would run in fast. Dropping below 2 psig would have no effect on the Defeat 4 the fans would still be in fast.

Distracter 2: Drywell H2 and 02 may be viewed as explosive hazard. However, there are no automatic actions connected with drywell cooling.

Distracter 3: Drywell temperature above the alarm setpoint will shift fans to fast and if temperatures return to normal the fans would go back to their original speed.

However, defeat 4 would over ride this function and the fans would not secure as shown.

K/A System: 223001 Primary Containment System and Auxiliaries K/A Number: A4.12 Ability to manually operate and/or monitor in the control room: Drywell coolers/chillers.

K/A Value: 3.5/3.6 DAEC Objective 68.01.01.06 Number:

DAEC Objective Describe the Primary Containment Ventilation System interlocks, including Statement: purpose, setpoints, logic, and when/how they are bypassed Cognitive Level: 3SPK Source: New Operationally Ability to verify EOP actions and system response.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A [1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 101 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 102 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

57. The plant is at full power and the "A" RPS bus becomes de-energized.

Which of the following correctly describes the plant response to this event?

a. Inboard MSIV control power and position indication will be lost.

b. PCIS Div 1 Groups 1 through 5 (excluding MSIVs) will isolate.

c. A Scoop Tube Lockup will occur on the "A" Recirc MG set.

d. The "A" SBDG 1G31 will automatically start.

ANSWER: b Answer: On loss of "A" RPS the Div 1 PCIS loses power an isolates with the exception of the MSIVs. The DC powered solenoids will hold the MSIVs open.

REFERENCE:

AOP 358 Rev 18 page 3 Distracter 1: This is a possible choice. However, this will occur if 125 vdc is lost.

Distracter 2: This is a possible choice. However, this occurs if 1Y1 1 is lost Distracter 3: This is a possible choice. However, this will occur if 1A3 is lost which can also lead to losing the "A" RPS.

K/A System: 223002 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off K/A Number: K6.08 Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF: Reactor protection system K/A Value: 3.5/3.7 DAEC Objective 22.02.01.07 Number:

94.11.01.04 DAEC Objective Given a Reactor Protection System operating mode and various plant conditions, Statement: predict how each supported system will be impacted by the following Reactor Protection System failures:

b. trip of a Motor-Generator

c. Trip of an EPA Breaker Relate the automatic actions to the immediate and follow-up actions directed by AOP 358 Cognitive Level: 2RI Source: Bank: Fermi 98 NRC exam (modified)

Operationally AOP response and system knowledge.

Validity:

OE: Loss of RPS has occurred at DAEC.

Estimated Completion Time: EB#

Time Validation: N/A [- (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 103 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 104 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

58. Review the indication given in the Support Material Booklet The plant was at full power when a spurious Group 1 inboard isolation occurred.

The following annunciators are currently ALARMING:

"* 1C03A C-5 "SRV/SV Tailpipe HI Pressure or HI Temp"

"* 1C03A D-5 "LLS "A" or "B" Armed"

"* 1C05A C-4 "Reactor Vessel HI Pressure Trip" Which of the following is correct in regards to the panel indications on the following page?

a. PSV 4407 has failed to open.

b. The operator has reset the "B" LLS logic on 1C03.

c. The initial pressure reduction by LLS has just started.

d. The initial LLS pressure reduction is complete and LLS is cycling normally.

ANSWER: a Answer: LLS has armed and is currently RPV pressure is >1055 psig. PSV 4407 should have opened at 1110 psig for the initial LLS Blowdown but this has not occurred.

REFERENCE:

SD 183.1 Rev 4, ARP 1C03A D-5 Rev 5, 1C03A C-5 Rev 11, IC05B C-4 Rev 4 Distracter 1: The "B" LLS logic does control PSV 4407. However, even if the reset push button is held in reset under these conditions PSV 4407 would be open.

'* Distracter 2: A LLS blowdown has commenced. However, PSV 4407 should have opened first to initiate the blowdown. With indication of RPV pressure >1025 psig and 25 psig in any SRV tailpipe, PSV 4407 should be opened.

Distracter 3: PSV 4401 is indicating normally for LLS cycling. However, with the alarms given the RPV pressure is > 1055 psig and PSV 4407 should be open.

K/A System: 239002 Safety Relief Valves K/A Number: A3.08 Ability to monitor the automatic operation of the RELIEF/SAFETY VALVES including: Lights and alarms.

K/A Value: 3.6/3.6 DAEC Objective 8.03.01.03 Number:

DAEC Objective Evaluate plant conditions and control room indications to determine if the ADS Statement: System or the Low Low Set System is operating as expected, and identify any actions that may be necessary to place the ADS/LLS Systems in the correct lineup.

Cognitive Level: 3SPK Source: New Operationally Automatic actions for ECCS system operation.

Validity:

OE: SRVs and-a'rtomatic system actions have failed in the industry and at DAEC.

Estimated Completion Time: EB#

Time Validation: N/A F1 (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

K-/ Operator Validated By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 105 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 106 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

59. The plant is currently at 540 MWe.

LOAD SET was inadvertently placed at 550 MWe.

All other Turbine controls are normal.

As Reactor power is raised which of the following is the correct plant response?

a. The Turbine Supervisory Panel will sense a load imbalance at 575 MWe and a Turbine trip will occur.

b. The 100% Load Limit is in effect at this power level which will allow the Turbine to be loaded to full power.

c. When Turbine load reaches 550 MWE number 1 Bypass Valve will start opening to control turbine inlet pressure.

d. Turbine Control Valves will reach 90% open when Turbine load reaches 561 MWe and a SCRAM will occur.

ANSWER: c Answer: With pressure set at 550 MWe, Reactor pressure may be increased until turbine out put reach the 550MWe limit. At that time the LVG will pass the signal from Load Set, limiting any further increase in turbine load. The #1 Bypass valve will open to control pressure and an alarm will come in when the Bypass valve reaches 20% open.

REFERENCE:

SD 693.2a Rev 3 Distracter 1: There is a load reject input to the EHC logic. However, this is NOT the input signal that provides the trip.

• -' Distracter 2: The 100% load limit is in effect at this power. However, this would NOT allow the turbine to go to full load because the Load set is at a lower value and would be the controlling signal.

Distracter 3: As reactor power and steam throttle pressure increase the control valves will close down. 561Mwe is about 90% between a load set of 550Mwe ahd normal load set of 666Mwe. However, there is no trip function to RPS from these valves. There is a trip function from the Stop valve and if any 3 of them reach 90% open there will be a scram signal generated.

K/A System: 241000 Reactor/Turbine Pressure Regulating System K/A Number: K5.05 Knowledge of the operational Implications of the following as they apply to REACTOR/TURBINE PRESSURE REGULATING SYSTEM: Turbine inlet pressure vs. turbine load.

K/A Value: 2.8/2.9 DAEC Objective 52.00.00.05d Number:

DAEC Objective Describe the operation of the following principle EHC Logic System components:

Statement: d. Load Control Unit Cognitive Level: 3PEO Source: New Operationally System operational knowledge and RPV reactivity control by pressure set.

Validity:

OE:

Estimated Completion Time: EB#

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 107 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Time Validation: N/A r-1 (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

• ' Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-W_xm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 108 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

60. SRO ONLY

'The plant was operating at power.

A plant transient occurred resulting in fuel damage.

Groups 3 "A/B" were verified complete.

1C35A C-3 "Reactor Building (RB) KAMAN 3, 4, 5, 6, 7 & 8 Hi Rad or Monitor Trouble" alarms.

The 1C03 operator reports that the RB KAMANs are above the yellow "hi" line and shortly after reports they are at the red "hi-hi" line and still going up.

How would you direct the following fans to be lined up?

1V-EF- 1/2/3 1V-SF-22A/B/C 1V-EF-21A/B/C Main Plant Exhaust Fans Turbine Building Supply Fans Turbine Building Exhaust Fans

a. All Secured All Secured All Secured

b. All Secured All Secured At Least One Running in Hi Speed

c. At Least One Running All Secured All Secured

d. All Secured At Least One Running At Least One Running in Hi Speed ANSWER: b Answer: The ARP 1C35A C-3 directs this lineup to ensure an elevated release point from the Turbine Building and prevents bypassing SBGT in the RB.

THE CANDIDATE IS NOT PROVIDED A COPY OF EOP 4. This omission makes this question a memory level question based on 1C35 Red KAMAN annunciator actions.

REFERENCE:

1C35A C-3 Rev 10 pages 1-3.

Distracter 1: This is plausible if you do not want to bypass the SBGT system. However, to provide for an elevated release point in the TB and maintain TB habitability the TB exhaust fans are directed to be running.

Distracter 2: This would provide for an elevated release point. However, the SBGT system may be bypassed which would reduce the filtering capability of SBGT.

Distracter 3: This would provide for max flow through SBGT and ventilate the TB. However, to ensure the TB is maintained at a negative pressure to prevent a ground release the TB exhaust fan should be running alone.

K/A System: 295017 High Off-Site Release Rate K/A Number: 2.4.31 Knowledge of annunciators, alarms, and indications / and use of the response instructions.

K/A Value: 2.5/3.7 10 CFR 55.43(b)(4 & 5)

DAEC Objective 6.72.03 Number:

DAEC Objective Direct operator actions to stop Main Plant Exhaust Fans and stop Turbine Statement: Building Supply Fans and verify at least 1 Turbine Building Exhaust Fan operating in Hi speed and verify Group 3 isolation complete.

Cognitive Level: 1P Source: NEW Operationally Operators have not taken these actions during EPIP drills as required and Validity: training and annunciator modifications were made to ensure when these conditions are present the operators are more aware of the required actions.

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 109 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

OE: This was a major concern during a recent EP drill when operators failed to take correct actions. The annunciators for these windows were color coded red to signify

. the importance of this condition.

Estimated Completion Time: EB#

Time Validation: N/A [3 (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-W xm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W-xm.doc 110 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point K.> 61. The Narrow Range GEMAC level transmitters (LT-4559, 4560, and 4561) are used in the Reactor Water Level Control system.

1) Are these transmitters calibrated HOT or COLD?

And

2) What type of compensation, if any, do they use?

a. 1) COLD

2) Electronic pressure compensation

b. 1) COLD

2) None

c. 1) HOT

2) Temperature compensation

d. 1) HOT

2) None ANSWER: d Answer: The Narrow Range GEMACs are calibrated HOT and they have no compensation.

REFERENCE:

SD 880 Rev 8 page 30 Distracter 1: RPV level control GEMACs are not calibrated cold and are not pressure compensated. This describes Fuel Zone indicators.

Distracter 2: RPV level control GEMACs are calibrated cold. This describes the Floodup GEMACs.

3: RPV level control GEMACs are not temperature compensated. This describes Wide Range Yarways.

K/A System: 259002 Reactor Water Level Control System K/A Number: K5.03 Knowledge of the operational implications of the following as they apply to REACTOR WATER LEVEL CONTROL SYSTEM: Water level measurement.

K/A Value: 3.1/3.2 DAEC Objective 88.00.00.02 Number:

DAEC Objective (Describe the operation of the following non-nuclear instrument system Statement: components including range, control room location, calibration condition, any compensation and any instruments that share the same lines: 1 Level)

Cognitive Level: 1F Source: Bank - 2001 NRC exam Operationally RPV level indication knowledge.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A F-1 (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 111 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 112 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

62. SRO ONLY The plant is at power.

"A" SBGT is tagged out for charcoal replacement.

1C24B A-5 ...B" SBGT Control Power Failure" alarms.

The Second Assistant reports 1D21 Ckt 19 ("B" SBGT Control Power) has tripped OPEN and looks damaged.

One attempt is made to close 1D21 Ckt 19 and it will NOT close in.

What is the longest the Mode switch can be left in run if plant conditions do not change?

a. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

b. 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />

c. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

d. 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> ANSWER: b Answer: With the "A" SBGT inop for filter replacement and loss of control power to the "B" SBGT system both trains are inop which requires immediate entry of LCO 3.0.3 which requires the plant to be in mode 2 in 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. The mode switch can not be in run in mode 2.

-THE CANDIDATE IS PROVIDED A COPY OF THE T.S. LCO BOOK

REFERENCE:

T.S. LCO 3.6.4.3 condition D. ARP 1C24B A-5 Rev Distracter 1: LCO 3.0.3 allows an extra hour to get to mode 2 from mode 1. This is to allow an orderly shutdown. However, one hour is not when the mode switch has to be placed out of run.

Distracter 2: 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> would be the time if the candidate assumes condition "A" is not met and the mode switch can not be in run in mode 3. However, even though there is an LCO for condition "A" this is not the most limiting.

Distracter 3: 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> would be the time if the candidate assumes condition "A" is not met, adding an hour for a controlled shutdown as in LCO 3.0.3.and the mode switch can not be in run in mode 3. However, even though there is an LCO for condition "A" and directions are given to immediately enter 3.0.3 this is not the most limiting.

K/A System: 261000 Standby Gas Treatment System K/A Number: 2.2.22 Knowledge of limiting conditions and safety limits.

K/A Value: 3.4/4.1 SRO 10 CFR 55.43(b)(2 &5)

DAEC Objective SRO 1.02.03.01 Number:

DAEC Objective Explain the requirements of Conditions, Required Actions, and Completion Times, Statement: when entering planned and unplanned LCOs Cognitive Level: 2RI - The candidate has to recognize how the loss of control power effects SBGT and how this effects TS with the given conditions.

Source: New Operationally Validity:

OE:

>Estimated Completion" Time: EB#

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W xm.doc - 113 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Time Validation: N/A E] (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 114 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

63. Which one of the following conditions would PREVENT the 1G31, "A Diesel Generator", from being shutdown using the Engine Mode Selector Switch on Panel 1C93 "SBDG 1G31 Control Panel"?

a. The DG automatically started due to bus undervoltage on 1A3.

b. The DG automatically started due to a 2 psig in the Drywell.

c. The DG was started by "Fast Manual Start" from the DG Room.

d. The DG was manually started from the Control Room.

ANSWER: b Answer: A LOOP signal prevents a shutdown from the DG room

REFERENCE:

SD-324 Rev. 4 pages 39-43 Distracter 1: Undervoltage on the bus will start the SBDG. However, low voltage does not energize the ESA and ESB relays and the SBDG can be shutdown from 1C93.

Distracter 2: Starting the SBDG with a "Fast Manual Start" will not override the 1C93 shutdown. The SBDG engine mode selector switch can still energize the stopping relay.

Distracter 3: Starting the SBDG from the control room will not override the 1C93 shutdown.

The SBDG engine mode selector switch can still energize the stopping relay.

K/A System: 264000 Emergency Generators (Diesel)

K/A Number: K4.07Knowledge of EMERGENCY GENERATORS (DIESEL) design feature(s) and/or interlocks which provide for the following: Local operation and control K/A Value: 3.3/3.4 DAEC Objective 19.00.00.02 Number:

DAEC Objective Describe how the SBDG responds to a trip signal.

Statement:

Cognitive Level: 1I Source: Bank - Slightly modified LOR Operationally Operation of emergency equipment and system knowledge.

Validity: 3/14 OE:

Estimated Completion Time: EB#

Time Validation: N/A [-I (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 115 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

64. SRO ONLY Refueling is in progress.

The Mode switch is in Refuel.

The Refueling Bridge is positioned to move the next fuel bundle from the Core to the Spent Fuel Pool.

The Fuel Bundle is identified and grappled.

The Refuel Floor SRO reports the bundle is being lifted out of the core.

About 5 seconds later 1C05B A-6 "Rod Out Block" alarms.

Which of the following is correct for these indications?

a. Cycle Rod select power OFF the ON to clear the Rod Block.

b. Suspend fuel moving immediately due to the failure of a refueling interlock.

c. Indications are normal the time delay is expected due to the Rod Block time delay.

d. Inform the Refuel Floor SRO of the annunciator because this will prevent bridge travel.

ANSWER: b Answer: With the bridge over the core and grapple not full up a Rod Block should have occurred. This has an indication that the Grapple not full up Refueling Equipment Interlock has failed. TS 3.9.1 condition "A" requires immediate suspension of fuel movement in the vessel.

REFERENCE:

THE CANDIDATE IS PROVIDED A COPY OF THE T.S. LCO BOOK T.S. 3.9.1 condition "A". T.S. bases B 3.9.1.

Distracter 1: This is an action statement in the Rod Block ARP and must be done during refueling to clear Rod Blocks. However, This would not clear the Rod Block.

Distracter 2: This is correct the Rod Block does have a time delay. However, a Rod Block should have occurred when the grapple left the full up position.

Distracter 3: There are bridge interlocks that prevent travel under some conditions. These are often confused with the refueling rod block interlock. The indicated conditions would not prevent bridge travel and are expected for fuel movement.

K/A System: 295023 Refueling Accidents K/A Number: 2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

K/A Value: 3.4/4.0 10 CFR 55.43(b)(1,2,5,6, and 7)

DAEC Objective SRO 1.02.03.01 Number:

SRO 1.04.02.19 DAEC Objective Explain the requirements of Conditions, Required Actions, and Completion Times, Statement: when entering planned and unplanned LCOs.

Describe the circumstances that would require core alterations and fuel handling activities to be suspended.

Cognitive Level: 3SPR - The candidate has to recognize a failure with the Refuel Equipment interlocks and then use T.S. to determine the correct course of action.

Source: New Operationally Recognition of improper indications during fuel movement which would not be Validity: available to the Refuel Floor SRO.

S' OE:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 116 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Estimated Completion Time: EB#

Time Validation: N/A I] (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc 117 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

65. The plant is at 40% power.

Control Rod 14-31 is at position 12.

Which of the following components is designed to prevent OUTWARD rod motion if the insert line were to rupture on the CRD Mechanism to Control Rod 14-31?

a. Collet assembly.

b. Coupling spud.

c. Drive piston.

d. Stop piston.

ANSWER: a Answer: If the insert line fails the under piston area would start to vent allowing outward rod motion. The Collet fingers are normally engaged except during rod motion.

They would hold the rod at its current position.

REFERENCE:

SD 255 Rev 7 page 14 Distracter 1: The coupling spud is attached to the control rod and if the rod were at position 48 it would prevent further out motion. This is observed during coupling checks.

However, the rod is at Position 12 and this feature would not prevent rod motion.

Distracter 2: The dive piston with normal system flows provides for a larger surface area on the bottom of the piston than the top and the even with a loss of CRD drive flow with the RPV at pressure the rod will drive in to the core. However, under these conditions the insert line is depressurized and this would tend to cause outward rod motion.

- Distracter 3: *The stop piston will stop the inward rod motion when the rod is inserted. This component would not prevent outward rod motion.

K/A System: 201003 Control Rod and Drive Mechanism K/A Number: K4.07 Knowledge of CONTROL ROD AND DRIVE MECHANISM design feature(s) and or interlocks, which provide for the following: Maintaining the control rod at a given location.

K/A Value: 3.2/3.2 DAEC Objective 10.07.01.05 Number:

DAEC Objective Describe the construction and operation of the CRD Mechanism including the Statement: following components:

b. collect assembly Cognitive Level: 1B Source: New Operationally Knowledge of system design features.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A [-D (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

\.t Approved By: Date: __ Date:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 118 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 119 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

66. A startup is in progress.

Control Rod 18-15 is being notched out.

Position indication for position 18 has been lost.

The RWM-OD (Operator Display) was taken to BYPASS.

Rod 18-15 was driven IN to position 16, IAW the pull sheet, and position 16 was verified operable.

The RWM-OD was taken to OPERATE.

Rod 18-15 is again withdrawn to position 18.

Which of the following is correct AND will allow rod withdrawal of Control Rod 18-15 to position 48 to continue?

a. A RWM WITHDRAW ERROR will occur. After position 18 is SUBSTITUTED on the RW1MI-OD the withdraw error will clear.

b. A RWM WITHDRAW ERROR will occur. After Control Rod 18-15 is BYPASSED on the RWM- CC (Computer Chassis) the withdraw error will clear.

c. Both RWM INSERT and WITHDRAW BLOCKS will occur. After position 18 is SUBSTITUTED on the RWM-OD both rod blocks will clear.

d. Both RWM INSERT and WITHDRAW BLOCKS will occur. After Control Rod 18-15 is BYPASSED on the RWM-CC (Computer Chassis) both blocks will clear.

ANSWER: c Answer: When position indication is lost the RWM will show IB and WB. Substituting the position will clear these blocks and rod withdrawal can continue.

REFERENCE:

SD 878.8 Rev 5 Distracter 1: Substitution of position 18 is a correct action. However, a RWM withdraw error will not occur. When blocks and errors from RWM occur is a common misconception.

Distracter 2: Bypassing the control rod is an option and done at the RWM-CC. However, a RWM withdraw error will not occur. When blocks and errors from RWM occur is a common misconception.

Distracter 3: Both IB and WB will occur. However, bypassing the rod will not remove the withdraw block and continued rod withdrawal will not be allowed.

K/A System: 201006 Rod Worth Minimizer System K/A Number: A1.02 Ability to predict and/or monitor changes in parameters associated with operating the ROD WORTH MINIMIZER SYSTEM controls including: Status of control rod movement blocks.

K/A Value: 3.4/3.5 DAEC Objective 84.00.00.05a Number:

DAEC Objective Given a Rod Worth Minimizer System operating mode and various plant Statement: conditions, predict how the Rod Worth Minimizer System will be impacted by failures in the following support systems:

a. RPIS Cognitive Level: 3PEO Source: New Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W-xm.doc - 120 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operationally Reactivity control and proper system response.

Validity:

OE: Problems with RWM operations have be seen in the simulator with crews.

Positions of rods have been lost due to broken sensors.

Estimated Completion Time: EB#

Time Validation: N/A Fm (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 121 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

67. Review the indication given in the Support Material Booklet The plant was at Full power.

"A" RWCU Pump running with both "A and B" RWCU Beds in Service.

The "A" SBLC pump is tagged out to replace the motor.

A transient occurred resulting in an Electrical ATWS.

You have been directed to inject with SBLC.

You have placed the Handswitch (HS-2613) for SBLC in the position shown in the Support Material Booklet and observe the indications shown.

The following plant conditions currently exist:

LLS is controlling RPV pressure.

0 RPV level control is set at 158 inches in auto and controlling level.

0 SBLC system flow indicator is indicating 0 GPM SBLC pump discharge pressure is indicating 1375 psig SBLC tank level is reading 88% and stable What is the status of the following?

MO-2701 "RWCU Suction Outboard Isolation Valve" MO-2740 "RWCU Return Header Outboard Isolation Valve" "A" RWCU pump

a. MO-2701 - CLOSED MO-2740 - OPEN "A" RWCU pump - OFF

b. MO-2701 - OPEN MO-2740 - CLOSED "A" RWCU pump - ON

c. MO-2701 - OPEN MO-2740- OPEN "A" RWCU pump - ON

d. MO-2701 - CLOSED MO-2740 - CLOSED "A" RWCU pump - OFF ANSWER: d Answer: The indications given are of a "B" SBLC logic failure to fire the Squib valves. The effect on the RWCU system is not effected by this logic failure. The RWCU isolation will occur based on the positioning HS-2613 to the "Pumps A&B Run" position.

REFERENCE:

SD 261 Rev 4 pages 10,17, 55: SD153 Rev 4 pages 45,47: 01 153 QRC Rev 0 page 1 Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 122 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Distracter 1: -MO-2701 will be closed and the pump off. A candidate may assume that only one side of the logic functioned leaving MO-2740 open. However, the logic is initiated by the Handswitch position and closes both valves. The pump trip will occur on MO-2701 leaving full open.

Distracter 2: MO-2740 will go closed. A candidate may assume that only one side of the logic functioned leaving MO-2701 open. However, the logic is initiated by the Handswitch position and closes both valves. The pump trip will occur on MO-2701 leaving full open and in this case with MO-2701 open the pump would not receive a trip signal.

Distracter 3: If the candidate incorrectly assumes the logic that closes the valves comes from the logic that controls the SQUIB valve and that is the part of the logic that failed then this would be the correct indication. However, the logic is initiated by the Handswitch position and closes both valves.

K/A System: 204000 Reactor Water Cleanup System K/A Number: K6.07 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER CLEANUP SYSTEM: SBLC Logic K/A Value: 3.3/3.5 DAEC Objective 11.01.01.05 Number:

DAEC Objective List the signals that cause a RWCU pump or filter/demineralizer auto trip Statement: including setpoints and logic. Describe how they are bypassed and how they are reset Cognitive Level: 3SPK - The candidate is presented with indications that the SBLC system has a logic failure in both SQUIB valve firing circuits and then understand that the logic that trips the RWCU has not been effected.

Source: New Operationally System knowledge and interlocks.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A El (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 123 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point S68. Assuming a normal SDC alignment with the "B" RHR loop in SDC.

The "B" RHR pump is running at 4300 gpm.

The "B" RHRSW pump is running at 2000 gpm.

MO 1909 "RHR Shutdown Cooling Outboard Isolation Valve" indicates OPEN.

MO 1939 "RHR HX 1E-201B Inlet Throttle Valve" indicates DUAL.

MO 1940 "RHR HX 1E-201B Bypass Valve" indicates DUAL.

MO 1947 "RHR HX 1E-201B Service Water Outlet Isolation Valve" indicates DUAL.

The STA reports a cooldown rate of 70°F/hr.

The OSS directs you to achieve a cooldown rate between 40 and 6 0 °F/hr.

Which of the following panel manipulation would be appropriate in achieving this order?

(Assume you place the handswitch for the indicated MO in the given direction for one second and then recalculate the cooldown rate.)

a. CLOSE on MO 1909.

b. OPEN on MO 1939.

c. OPEN on MO 1940.

d. OPEN on MO 1947.

ANSWER: c Answer: The cooldown rate must be slowed to comply with the direction given. From the above there are two ways to achieve this. Throttle the bypass around the HX open MO 1940 to reduce flow through the HX or Throttle closed on the HX inlet MO-1939. MO 1940 is the only option given to choose from.

REFERENCE:

01149 Rev 78 page 40 Distracter 1: Taking the HS for M01909 to the closed position will slow the Cooldown rate because the valve will fully close and trip the Running RHR pump. The candidate must understand this is not a throttling valve. This would cause a Loss of SDC.

Distracter 2: This is a valve that is approved for controlling the cooldown rate and would be the choice if the candidate incorrectly diagnoses which way the cooldown rate needs to be adjusted. However, opening MO 1939 would increase the cooldown rate.

Distracter 3: Throttling on the RHRSW HX outlet valve on the RHRSW side is a technique used for fine temperature rate control. Procedures allow the RHRSW flow rate to be varied. However, opening this valve would cause the cooldown rate to increase, which is not the correct direction.

K/A System: 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)

K/A Number: K1. 14 Knowledge of the physical connections and/or cause-effect relationships BETWEEN SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) and the following: Reactor Temperature (moderator, vessel, flange)

K/A Value: 3.6/3.6 DAEC Objective 2.01.01.02 Number:

2.01.01.02 Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 124 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

DAEC Objective Identify the appropriate procedures that govern the RHR system operation, Statement: include operator responsibilities during all modes of operation, and any actions required by personnel outside of the Control Room.

Describe the major flowpaths and purpose for each of RHR system operation, including:

c. Shutdown Cooling (SDC)

Cognitive Level: 2RI Source: New Operationally SDC control Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A ['] (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 125 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

69. A non-selected control rod at position 36 is uncoupled.

The RO will be selecting and WITHDRAWING this control rod fully.

Which of the following CORRECTLY describes when the uncoupled control rod can be identified using RPIS indications only?

a. As soon as the control rod is selected.

b. When the selected control rod is withdrawn from position 36 to position 38.

c. After the RMCS Timer times out with the selected control rod at position 48.

d. When the selected control rod has withdrawn past position 48 independent of the RMCS Timer.

ANSWER: d Answer: The rod position will still indicate on 1C05 as the CRD mechanism is moved. If the rod were coupled the rod would not withdraw past position 48.

REFERENCE:

SD 856.1 Rev 4 page 31 Distracter 1: A Rod Drift alarm would be received when the RMCS timer timed out and the rod was not on an even numbered position. RMCS timing out is not necessary to get the overtravel alarm.

Distracter 2: No way to detect this failure until rod is at position 48 and a coupling check is performed.

Distracter 3: No way to detect this failure until rod is at position 48 and a coupling check is performed.

K/A System: 214000 Rod Position Information System S- K/A Number: A3.03 Ability to monitor automatic operation of the ROD POSITION INFORMATION SYSTEM including: Verification of proper functioning/operability K/A Value: 3.5/3.7 DAEC Objective 10.01.01.04 Number:

DAEC Objective Evaluate plant conditions and control room indications to determine if the Control Statement: Rod Drive Mechanisms and Hydraulic System is operating as expected, and identify any actions that may be necessary to place the Control Rod Drive Mechanisms and Hydraulic System in the correct lineup.

Cognitive Level: 2RI Source: Bank - 2001 RO audit, 1999 RO NRC (minor word changes)

Operationally Reactivity manipulation and system knowledge.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A FI (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By: -----

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 126 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Question Usage (exams): 57_2002 ILC-RO-W xm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 127 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

70. Which of the following is utilized to automatically select the appropriate RBM upscale trip setpoint when the RBM is required to be in operation?

a. Steam Flow

b. Reference APRM

c. Averaging Circuit

d. Turbine Ist Stage pressure ANSWER: b Answer: The Reference APRM sends a power level to the RBM setpoint circuit which direct it to the RBM upscale trip unit.

REFERENCE:

SD 878.5 Rev 5 Pages 18 and 33.

Distracter 1: Steam flow is used as power level indications in some systems. However, RBM does not use this signal Distracter 2: The averaging circuit looks at the LPRM inputs at the current power level and sends them to be monitored by RBM. However, this signal does not interface with the RBM setpoint circuit to select the upscale trip setpoint.

Distracter 3: Turbine 1st Stage pressure is used in systems to monitor power. However, this is not an input to the RBM system.

K/A System: 215002 Rod Block Monitor System K/A Number: K4.02 Knowledge of ROD BLOCK MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following: Allows stepping up of rod block setpoint K/A Value: 2.9/3.0 DAEC Objective 82.00.00.02h Number:

DAEC Objective Describe the purpose and operation of the following principle Rod Block Monitor Statement: System components:

h. RBM Setpoint Circuits Cognitive Level: 11 Source: New Operationally System knowledge and reactivity monitoring interlocks/features.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A '-I (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 128 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

71. The plant is at full power.

HPCI was running for a surveillance test with RHR, RHRSW, and ESW running as required for support.

Torus temperature reached the EOP 2 entry condition during the surveillance.

The OSS entered EOP 2 on Torus water temperature and Torus Cooling was MAXIMIZED with all RHR and RHRSW pumps running at rated flows.

Currently:

HPCI is secured and Torus Water temperature is 2 0 F above the EOP 2 entry temperature and slowly lowering.

Torus Cooling is still MAXIMIZED.

Bus 1A3 receives a bus Lockout.

Which of the following would be the expected status of Torus Cooling with no operator action?

a. "B" RHR Loop would be operating within limits.

b. The "A" RHR Loop header would be below the allowed pressure.

c. Torus Cooling would no longer be MAXIMIZED as directed by EOP 2.

d. "B" and "D" RHR pumps would be above their maximum allowed flow rates.

ANSWER: d Answer: With Torus Cooling maximized and normal system lineup the RHR loops will be setup with 9600 gpm in both loops. With the loss of 1A3 the "A" loop pumps trip but the Torus Cooling inject valves are still open. The "B" loop pumps will now be in a runout condition with a flow rate of around 12,000 gpm which is well above the required 9600 gpm limit.

REFERENCE:

01 149 Rev 78 pages 5, 28-32. SD 149 Rev 8 pages 35 and 36.

Distracter 1: The loss of 1A3 will generate false 2 psig Drywell and 64 inch RPV water level annunciators. However, this will not cause a LPCI Loop select and LPCI initiation signal that would effect the RHR Loops as it would if it were actual signals. Torus Cooling can be restored to normal for these plant conditions. However, Torus Cooling has not been lost because the "B" side RHR pumps are running at max flow but is not within limits.

Distracter 2: This would be possible if the loops were not cross-tied. Which the Torus inject valves both open and pumps secured the header would depressurize. However, the RHR loops are normally cross-connected and with the "B" loop in service the header will not depressurize.

Distracter 3: Maximized is defined as all available RHR cooling not required for adequate core cooling being supplied to the Torus. Even though half of the RHR system is unavailable the direction to Maximize Torus Cooling can still be accomplished with the "B" RHR Loop.

K/A System: 219000 RHR/LPCI: Torus/Suppression Pool Cooling Mode K/A Number: K1.04 Knowledge of the physical connections and/or cause-effect relationships between RHR/LPCI: TORUS/SUPPRESSION POOL COOLING MODE and the following: LPCI/RHR pumps K/A Value: 3.9/3.9 DAEC Objective 2.01.01.06a s Number:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W-xm. doc - 129 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

DAEC Objective Given an RHR system operating mode and various plant conditions, predict how Statement: the RHR system will be impacted by operation, or failure of the following support system(s):

a. Essential 4160/480 VAC electrical power supplies Cognitive Level: 3PEO- The Candidate must recognize the effects of the loss of the essential bus on Torus cooling and realize that the "B" and "D" pumps are discharging into both loops to the Torus and this will cause the pumps to be in runout conditions.

Source: New Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A F-1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 130 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

72. SRO ONLY The plant was at full power.

A transient occurred and a SCRAM was successfully inserted.

RPV water level is 195 inches and stable.

CV4371A is overridden OPEN.

Containment Sprays have been unavailable.

The 1C03 operator reports Containment Sprays are now available.

The OSS requests Containment parameters.

The 1C03 operator reports the following:

"* Drywell pressure 40 psig rising slowly

"* Drywell temperature upscale (> 350OF)

"* Torus pressure 40 psig rising slowly

"* Torus temperature 97 OF rising slowly

"* Torus level 10.5 feet rising slowly The OSS directs the 1C03 operator to initiate Torus and Drywell Sprays.

Why is the direction for sprays CORRECT/INCORRECT AND what other actions are required?

a. CORRECT due to Torus pressure greater than 11 psig.

Direct the 1C05 operator to Maximize Torus Cooling.

b. CORRECT due to Drywell temperature above 280 0 F.

Enter Emergency Depressurization (ED) and direct the IC05 operator to OPEN 4 ADS/SRVs.

c. INCORRECT due to possible brittle fracture failure of Recirc piping with current containment conditions.

Direct the 1C05 operator to Maximize Torus Cooling.

d. INCORRECT due to possible containment damage with current containment conditions.

Enter Emergency Depressurization (ED) and direct the 1C05 operator to OPEN 4 ADS/SRVs.

ANSWER: d Answer: Under these plant conditions spraying the Drywell could cause the Drywell to Torus dP to increase to >10 psid. While Maximizing Torus cooling is an action which is directed in EOP 2 it is not the highest priority under these plant conditions. ED is the correct action to be taken at this time. It is required in both the DW/T and PC/P legs of EOP 2. The candidate should recognize the implication of Drywell sprays under these conditions.

THE CANDIDATE IS PROVIDED A COPY OF EOP 2 and EOP ED

REFERENCE:

EOP 2 Rev 9, EOP 2 Curves and Limits Rev 5 pages 23-27, EOP 2 Rev 8 pages 44, 45, 48, 63, 64, 69, and 70.

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 131 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Distracter 1: Torus Cooling is allowed and the Torus pressure is above 11 psig, which would require Drywell sprays if within the Spray Initiation Curve. However, initiation of sprays at this time is NOT allowed due to possible containment failure.

Distracter 2: ED is required. Drywell temperature is above 280 0 F, which would require Drywell sprays if within the Spray Initiation Curve. However, initiation of sprays at this time is NOT allowed due to possible containment failure.

Distracter 3: Drywell Sprays are not allowed and Torus cooling is allowed. However, Not spraying the Drywell due to Brittle fracture of the Recirc piping is NOT correct.

The Recirc piping is not a factor in the EOP bases for the Drywell Spray Initiation Curve.

K/A System: 226001 RHR/LPCI: Containment Spray Mode K/A Number: A2.17 Ability to (a) predict the impacts of the following on the RHR/LPCI:

CONTAINMENT SPRAY MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High containment/drywell temperature.

K/A Value: 3.2/3.2 SRO 10 CFR 55.43(b)(5)

DAEC Objective 6.58.01.05 Number:

DAEC Objective Evaluate plant conditions and control room indications and determine the actions Statement: directed by EOP 2.

Cognitive Level: 3SPR - The Candidate must understand the bases for the Drywell Spray Initiation Curve and determine what EOP actions are appropriate.

Source: New Operationally EOP actions and containment integrity.

Validity:

'> OE: Operators have sprayed the containment in violation of the DWSIL in simulator scenarios.

Estimated Completion Time: EB#

Time Validation: N/A [] (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: __ Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 132 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

73. SRO ONLY The plant has entered ATWS from full power.

Plant conditions are as follows:

"* RPV pressure is 900 psig.

"* Torus level is 9.9 ft. and stable.

"* Torus temperature is 145 0 F and rising.

Which one of the following is the LOWEST Torus water temperature at which an Emergency Depressurization must be performed per the EOPs for these conditions?

a. 150OF

b. 160°F

c. 170OF

d. 180OF ANSWER: b Answer: EOP 2 must be entered on Torus low water level and high Torus water temperature. EOP 2 Graph 4 requires ED for these plant conditions at 160OF and even though Torus Water level is low it is still bounded by the curve in Graph 4.

EOP ATWS references this graph and directs pressure reduction to stay below it.

However, the candidate has to connect that EOP 2 is also-a concern- and is the document, which states, ED is required. This then meets the first CRS in ATWS /P leg which says that if ED is required exit the /P leg and enter ED.

THE CANDIDATE IS PROVIDED A COPY OF EOP 2 and EOP ED

REFERENCE:

EOP ATWS Rev 10. EOP Bases ATWS Rev 7 pages55, 61, and 62. EOP 2 Rev 9.

EOP 2 Bases Rev 8 pages 34 and 35.

Distracter 1: Homogeneous but ED is not required at this temperature.

Distracter 2: Homogeneous but ED should have been ordered prior to this point.

Distracter 3: Homogeneous but ED should have been ordered prior to this point.

K/A System: 295030 Low Suppression Pool Water Level K/A Number: EA2.02 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: Suppression pool temperature K/A Value: 3.9/3.9 10 CFR 55.43(b)(5)

DAEC Objective SRO 6.50.01.04 Number:

DAEC Objective Evaluate plant conditions and control room indications and determine the actions Statement: directed by ATWS Cognitive Level: 3SPR Source: Bank Hope Creek Unit 1 12/18/1995 Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A F] (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

• -/ Operator Validated By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 133 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 134 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

74. SRO ONLY The plant was at full power when a transient required EOP 1 and EOP 2 entry.

All RPV Water level instrumentation was subsequently lost.

The following plant conditions exist:

"* All Control Rods are inserted.

"* All SRV handswitches are in open and only 3 SRVs are open.

"* All available injection sources are injecting into the RPV at maximum flow.

"* RPV pressure is 60 psig and stable.

"* Torus pressure is 3 psig and stable.

"* Drywell Temperature is 275 0 F increasing slowly.

With the indicated plant conditions what actions are required?

a. Adequate core cooling is NOT assured. Enter the SAGs.

b. ADS/SRV operation is NOT assured. Initiate Drywell Sprays per EOP 2.

c. Containment integrity is NOT assured. Initiate Torus Sprays per EOP 2.

d. Core submergence is NOT assured. Enter ED and open Main Steam Line Drains.

ANSWER: a Answer: The candidate must recognize Adequate Core Cooling (ACC) assurance is lost at this point with only 3 SRVs open. RPV Flood directs entry into the SAGs. The SAGs may direct flooding the containment by sprays to regain ACC. However, the EOPs do not address this and sprays are not authorized if ACC is not assured.

THE CANDIDATE IS PROVIDED A COPY OF EOP 1, 2, ALC, ED, RPV FLOOD.

REFERENCE:

EOP RPV Flooding Rev 6. EOP RPV Flooding Bases Rev 5 pages 29 - 33.

Distracter 1: ADS/SRV may be challenged at higher Drywell temperatures and spraying the Drywell would reduce the temperature. However, diverting sprays to DW sprays is not allowed unless ACC is assured.

Distracter 2: Containment integrity will be challenged as pressure rises and sprays will reduce the pressure. However, diverting sprays to Torus sprays is not allowed unless ACC is assured.

Distracter 3: Core submergence is not assured at this time. However, when level indication is lost in EOP 1 you have to exit EOP 1 and enter RPV Flooding. There are only two ways to get to ED one is in EOP 1 and the other is in ATWS. Under the above conditions neither EOP1 nor ATWS apply. RPV Flooding essentially performs the ED without the alternate depressurization system available for pressure control.

In fact you are directed to close the MSL Drains in RPV/F.

K/A System: 295031 Reactor Low Water Level K/A Number: EA2.04 Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL: Adequate core cooling K/A Value: 4.6/4.8 SRO 10 CFR 55.43(b)(5)

DAEC Objective 6.85.01.02 Number:

DAEC Objective Evaluate plant conditions and control room indications and determine the actions Statement: directed by RPV/F.

Cognitive Level: 3SPR Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 135 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Source: New Operationally v-*' Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A I-l (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 136 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

75. SRO ONLY The plant is at power.

Annunciator 1C08B B- 11 "250V DC Battery 1D4 Disconnected" alarms.

The Auxiliary Operator reports "1D40 Ckt 01 has tripped OPEN".

AOP 388 "Loss Of 250 VDC Power" is entered by direction of the OSS.

One attempt is made to reclose the breaker 1D40 Ckt 01 but it will NOT CLOSE in.

All 250 VDC buses remain energized from 1D43 "250 VDC Battery Charger".

1D44 "250 VDC Battery Charger" is operable.

Which of the following is correct in regards to the given conditions?

a. With no 250 VDC buses lost AND 1D44 operable no LCO entry is required.

b. Enter LCO 3.0.3 due to the loss of safety function based on the total loss of 250 VDC.

c. Enter LCO 3.8.4 "DC Sources- Operating" Condition "C" due to the loss of 250 VDC being able to supply peak power during all events.

d. Due to the large load imposed by the HPCI system on 250 VDC within I hour deenergize "MCC 1D41" and enter LCO 3.5.1 "ECCS - Operating" Condition "F" for HPCI. All other 250 VDC loads remain operable.

ANSWER: c Answer: Although the 250 VDC buses remain energized they will not be able to handle peak current during DBS conditions with the battery charger alone supplying the buses. The T.S. bases identify this as being inop.

THE CANDIDATE IS PROVIDED A COPY OF THE T.S. LCO BOOK

REFERENCE:

T.S. 3.8.4 condition "C" Amendment 223. T.S. bases B 3.8.4 Amendment 223 and TSCR-020. AOP 388 Rev 15 pages 4 and 6. ARP 1C08B B-11 Rev 9 page 1.

Distracter 1: Bus voltage is available and conditions appear to be normal. However, the candidate must understand T.S. bases to know that the charger alone, during DBA events, can not power the bus. This makes the 250 VDC source inop.

Distracter 2: All 250 VDC is inop under these conditions. However, T.S. allow the total loss of 250 VDC and not consider it a loss of safety function. Where as a total loss of 125 VDC would be a loss of safety function and entry into 3.0.3 would be required.

Distracter 3: De-energizing 1D41 to HPCI would reduce the load on the bus if HPCI were used.

However, HPCI is not the only system that would challenge peak bus loading. All supported systems have to be declared inop even if 1D41 is deenergized.

K/A System: 295004 Partial or Complete Loss of D.C. Power K/A Number: AA2.01 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Cause of partial or complete loss of D.C. power.

K/A Value: 3.2/3.6 SRO 10 CFR 55.43(b)(2)

DAEC Objective 5.13.04.01 Number:

DAEC Objective Evaluate plant conditions and control room indications and determine actions Statement: directed by AOP 388 for restoration of 250vdc power Cognitive Level: 3SPR - The candidate must determine from T.S. that 3.8.4 is the correct LCO and from Bases know that even though the 250 VDC buses still have power the given conditions render 250 VDC sources inop.

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 137 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Source: -t*NEW Operationally v-' Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A E] (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 138 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

76. SRO ONLY The plant is at full power.

CV1062 "MSR 1E-18A 1st Stage Scavenging Steam Isolation Valve" fails CLOSED.

Upon investigation it is determined the valve will NOT OPEN and will have to be repaired.

(Assume the repair will be completed when the parts arrive in 4 days)

Which one of the following procedural directions is intended to prevent the undesirable effects of "CHUGGING" within the tube bundles of the MSR(s) for these plant conditions?

a. AOP 255.2 "Power/Reactivity Abnormal Change" directs Generator load reduction to < 450 MWe.

b. 01 646 "Extraction Steam System" directs the SHUTDOWN of First and Second stage reheat on both MSRs.

c. IPOI 3 "Power Operations (35% - 100% Rated Power)" directs power change be limited to a step change of 5% and overall power changes to <7 MWe/minute.

d. ARP 1C07B B-3 "MSR Ist Stage Drain Tank 1T-91A LO Level" directs taking manual level control as needed to maintain 1T-91A above the LO Level alarm setpoint.

ANSWER: b Answer: 01 646 has a caution to not operate MSR Reheaters without scavenging steam due to the possibility of "Chugging" which can damage the tube bundle. There is also a caution to not run Second Stage reheat without First stage reheat in service this is also due to the loss of scavenging steam to the Second stage Reheaters. This valve closing requires a generator load reduction to <400 MWe and the removal of both 1st and 2nd stage Reheaters on both MSRs.

REFERENCE:

01 646 Rev 33 pages 4-6, 10, 17, and 51.

Distracter 1: There will be a power change due to loss of efficiency and AOP 255.2 may have been entered when power was lost. However, AOP 255.2 does not require the power change indicated to prevent the effects of "chugging" in the MSRs.

Distracter 2: IPOI 3 does require these power change guidelines to be followed. However, these are for the fuel and not to prevent "chugging" in the MSRs.

Distracter 3: ARP 1C07B B-3 does allow manual control to maintain level and the candidate may assume this could be required due to non-condensable gases building up in the 1st stage Drain Tank due to the loss of scavenging steam. However, this will not prevent the "chugging" effect which would occur.

K/A System: 239001 Main and Reheat Steam System K/A Number: 2.1.32 Ability to explain and apply system limits and precautions.

K/A Value: 3.4/3.8 SRO 10 CFR 55.43(b)(5)

DAEC Objective 5.03.04 Number:

DAEC Objective Direct operator actions to control the malfunctioning system or component.

Statement:

Cognitive Level: 2DR - The candidate must understand the relationship between the 1st and 2nd stage reheat and what to do to correct the problem.

Source: NEW Operationally S-Validity:

OE:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 139 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Estimated Completion Time: EB#

Time Validation: N/A F1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 140 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

77. A trainee is synchronizing 1G 31 "A SBDG" to the 1A3 bus.

The following conditions are present during synchronizing the SBDG:

"* The incoming voltage is slightly HIGHER than running voltage.

"* The synchroscope is rotating slowly in the clockwise direction.

The trainee places the "A" SBDG output breaker to the close position when the synchroscope is at the 3 o'clock position.

Which of the following describes the expected breaker response?

The "A" SBDG output breaker will...

a. close and then trip open due to "A" SBDG overspeed trip.

b. close and then trip open due to an instantaneous overcurrent trip.

c. remain open due to the sync-check relay sensing excessive current differential.

d. remain open due to the sync-check relay sensing excessive incoming to running phase angle differential.

ANSWER: d Answer: The sync-check relay prevents closing in the SBDG output breaker if too large a phase difference is sensed. This protects the electrical plant from inadvertent paralleling of power sources that are not synchronized and the resulting damage that could occur.

REFERENCE:

01 324 Rev 55 Pages 17 and 18.

Distracter 1: The SBDG could possibly over speed if the breaker closed in but this is unlikely.

The breaker will not try to close due to the sync-check relay action.

Distracter 2: There could be an instantaneous overcurrent condition due to the large phase difference if the breaker were to close. The breaker will not try to close due to the sync-check relay action.

Distracter 3: The breaker will remain open. However, it is not due to the excessive current differential.

K/A System: 262001 A.C. Electrical Distribution K/A Number: A4.05 Ability to manually operate and/or monitor in the control room: Voltage, current, power, and frequency on A.C. buses.

K/A Value: 3.3/3.3 DAEC Objective 19.00.00.03 Number:

DAEC Objective Evaluate plant conditions and control room indications to determine if the SBDG Statement: is operating as expected, and identify any actions that may be necessary to place the SBDG in the correct lineup Cognitive Level: 3PEO - The candidate will have to predict the expected plant response due to the positioning of the SBDG output breaker at the incorrect time.

Source: Bank - Lasalle 2000 NRC Exam Operationally Validity:

OE:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 141 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Estimated Completion Time: EB#

Time Validation: N/A I-] (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 142 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

78. Of the systems listed below which one meets the requirements needed for the following situation?

The plant is installing a new component.

The component is NOT safety related or essential to plant safety.

However, power interruptions to this component should be avoided.

The plant power supply to the component must have 3 power sources, one normal source and two alternate sources.

Upon loss of power to the normal power supply the system must be able to AUTOMATICALLY align to either of the two alternate sources WITHOUT prolonged loss of power (less than 2 seconds).

a. The 125 VDC Power Supply System

b. The Instrument A.C. Control System.

c. The Uninterruptible A.C. Control Power System.

d. Reactor Protection System Distribution Panel 1Y30 ANSWER: c Answer: This is the purpose of Uninterruptible AC. The candidate must understand the purpose of the system and understand how the purpose is achieved. Many electrical systems at DEAC have multiple power sources. However, the Uninterruptible system is unique in the fact that there are two alternate sources which will quickly supply power to the bus automatically with a minimal delay in restoring bus power.

REFERENCE:

SD 357 Rev 5 page 4 Distracter 1: 125 VDC does not have three power sources and there are none safety-related components on this bus. The essential bus, which supplies 125 VDC, does have 3 sources but if the SBDG transfers to the buss it will take longer than 2 seconds.

Distracter 2: Instrument AC has three power sources and there are none safety-related components on this bus. However, it only automatically switches between 2 Distracter 3: RPS has two sources and no automatic switching. The essential bus, which supplies RPS, does have 3 sources but if the SBDG transfers RPS will trip and have to be manually recovered.

K/A System: 262002 Uninterruptible Power Supply (A.C./D.C)

K/A Number: 2.1.27 Knowledge of system purpose and/or function.

K/A Value: 2.8/2.9 DAEC Objective 21.00.00.01 Number:

DAEC Objective State the purpose of the Uninterruptible AC system Statement:

Cognitive Level: 2RW Source: New Operationally System purpose.

Validity:

OE:

Completion Time: SEstimated EB#

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 143 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Time Validation: N/A E[ (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W-xm.doc - 144 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point 1 79. The plant is at power.

Annunciator 1C35A "Offgas Stack KAMAN 9 & 10 HI RAD or Monitor Trouble" alarms.

Offgas Stack flow has not changed.

The Chemist reports the Normal Range sample pump is not working.

All other Offgas Stack KAMAN equipment is functional.

Which of the following is correct?

a. The KAMAN Normal Range is INOPERABLE. The Accident Range is OPERABLE.

b. The KAMAN Normal Range is OPERABLE. The Accident Range is INOPERABLE.

c. Both KAMAN Normal and Accident Ranges are INOPERABLE.

d. Both KAMAN Normal and Accident Ranges are OPERABLE.

ANSWER: c Answer: .*The KAMAN has to be running to allow the Accident Range Sample pump to function properly. The Normal Range Sample Pump uses an isokinetic probe to sample process flow and the Accident Range Sample pump uses an isokinetic probe to sample the flow going to the Normal Range Monitors. If Normal Sample flow stops the Accident Range Sample will not be an accurate sampling.

REFERENCE:

SD 879.3 Rev 5 pages 12 and 29.

Distracter 1: The candidate may assume that because the Accident Range Sample pump is functional this will allow the Accident Range Monitor to still be operable.

However, the Accident Range Sample pump depends on the Normal Range Sample pump for sample flow.

S Distracter 2: The candidate may assume because Offgas has flow the isokinetic probe will still have flow to the Normal Range monitor but the Accident Range Monitor would be inaccurate.

Distracter 3: The candidate may assume because Offgas has flow the isokinetic probe will still have flow to the Normal Range monitor and the Accident Range Monitor still has a pump available for sampling.

K/A System: 272000 Radiation Monitoring System K/A Number: K3.02 Knowledge of the effect that a loss or malfunction of the RADIATION MONITORING System will have on the following: Station gaseous effluent release monitoring.

K/A Value: 3.2/3.8 DAEC Objective 87.00.00.02a/b Number:

87.00.00.04 DAEC Objective Describe the operation of the following principle KAMAN System components:

Statement: a. Accident Range Monitor

b. Normal Range Monitor State when the KAMAN System is required to be operable by Technical Specifications and describe the bases of the KAMAN System LCOs Cognitive Level: 2DR - The candidate must recognize the relationship and interdependence of the Accident Range Monitor on the Normal Range Monitor and understand how the isokinetic probe functions.

SSource: New Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 145 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operationally System knowledge.

Validity:

, OE:

Estimated Completion Time: EB#

Time Validation: N/A m (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 146 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point K.. 80. During shiftly annunciator checks the annunciators do not alarm on panel 1C40 "Fire Protection".

AOP 302.2 " Loss of Alarm Power" is entered.

There are NO other annunciators in alarm.

Which of the following panels has the breaker that supplies power to 1C40 and would give these indications if the breaker had tripped open?

a. 1D13

b. 1D50

c. 1Yll

d. 1Y23 ANSWER: a Answer: 1D13 ckt 6 125 VDC Div 1 is the power supply to 1C40. This is one of the fire detection systems major annunciator panels. The loss of this panel would require several fire watches covering most of the plant areas.

REFERENCE:

AOP 302.2 Rev 11 page 3 and 38. 01 302 attachment 3 Rev 0 page 7 Distracter 1: -This is Div 1 of 24 VDC. Alarm power comes from 125 VDC.

Distracter 2: 1Yll supplies power to many components in the control room but no alarm power.

Distracter 3: 1Y23 supplies power to many components in the control room but no alarm power.

K/A System: 286000 Fire Protection System K/A Number: K2.03 Knowledge of electrical power supplies to the following: Fire detection system.

K/A Value: 2.5/2.7 DAEC Objective NSPEO 9.00.00.02 Number:

DAEC Objective Identify power supplies for the Fire Protection system components Statement:

Cognitive Level: 1F Source: New Operationally AOP actions and annunciator power supplies.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A I] (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 147 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

81. SRO ONLY The plant is at 85% power.

Preparations are under way for a refueling outage.

Contractors mistakenly lifted the RCIC Room Equipment Hatch covers located in the yard outside the Reactor Building South side.

While attempting to replace the RCIC hatch cover, the rigging broke and the hatch plug fell into the RCIC room and is currently at the bottom of the RCIC Room stairs.

The Second Assistant reports that the RCIC system was NOT damaged.

Which of the following action AND reason is correct?

a. Place the mode switch out of RUN within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> due being outside analyzed design bases.

b. Enter T.S. LCO 3.0.3 immediately due to a loss of safety function for both SBGT systems being inoperable based on the inability to meet SR 3.6.4.1.3 to maintain Ž0.25 inches of vacuum water gauge in Secondary Containment.

c. Declare Secondary Containment Inoperable and restore the RCIC Equipment hatch within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This allows time to fix the RCIC Equipment hatch and takes into consideration that without this hatch installed a DBA LOCA could release radioactivity at ground level.

d. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate administrative controls to maintain RCIC Room HELB doors CLOSED except for essential access to prevent unmonitored contamination release to the environment in the event of a failure of primary containment in the Reactor Building.

ANSWER: c

• Answer: This is a breach of Secondary Containment and requires correction within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The reason for Secondary Containment is to provide a hold up volume for radioactivity and along with SBGT to reduce the radioactivity concentration prior to releasing it at an elevated point. This prevents a high rad level release at ground level.

THE CANDIDATE IS PROVIDED A COPY OF THE T.S. LCO BOOK

REFERENCE:

Tech. Spec. 3.6.1.4 Rev Amendment 237 page 3.6-35 and 36. T.S. Bases B 3.6.4.1 Rev TSCR-037 pages 3.6-78 to 82.

Distracter 1: This is an unusual but plausible situation. This is analyzed in that secondary containment can be inop for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with out having to start a shutdown.

T.S. discusses the reasoning for a 4-hour limit.

Distracter 2: If both SBGT systems are inop this would be correct. However, there is no reason to suspect SBGT to be inop. The STP we run with SBGT to ensure it can pull a 0.25 in VAC. WG is to check Secondary Containment not SBGT.

Distracter 3: There is a P&L for the HELB (High Energy Line Break) doors in RCIC and HPCI.

However, this is for a steam line break in the room to protect EQ equipment in the RB. This is a plausible distracter if the candidate does not understand these doors are not for containment from RB to RCIC but from RCIC to the RB.

K/A System: 290001 Secondary Containment K/A Number: A2.02 Ability to (a) predict the impacts of the following on the SECONDARY CONTAINMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Excessive outleakage.

ki K/A Value: 3.5/3.7 10 CFR 55.43(b)(2)

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W-xm. doe - 148 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

DAEC Objective SRO 1.02.01.01 Number:

K_ DAEC Objective Explain the Technical Specification Condition, Required Action, and Completion Statement: Time for any given instrument, component, structure, or system.

Cognitive Level: 3SPR - The Candidate must use T.S. to determine the time requirement and understand the bases and plant design to answer this question.

Source: NEW Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A r-1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-W xm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 149 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

82. SRO ONLY The Plant was at 60% power.

A lightning strike caused the following loads on 1A1 to trip:

"* "A" Recirc MG Set

"* "A" Circ Water Pump The plant has been stabilized with RPV Water level controlling in the normal level band.

The 1C05 operator reports that we are in the Buffer Region on the "Power to Flow Map".

Which procedure and direction is correct for these conditions?

a. Enter AOP 255.2 "Power/Reactivity Abnormal Change" and direct the 1C05 operator to monitor APRMs for undamped oscillations.

b. Enter IPOI 3 "Power Operations (35% to 100% Power)". IPOI 3 directs inserting a Manual SCRAM when the plant enters the Buffer Region on the "Power to Flow Map".

c. Enter AOP 304.1 "Loss of 4160V Non-Essential Electrical Power" and conduct a Rod Withdrawal briefing in preparation for Withdrawing Control Rods to exit the Buffer Region.

d. Enter IPOI 4 "Shutdown" and direct the 1C05 operator to run the "B" Recirc pump to Minimum. When "B" Recirc is at Minimum perform a Rod movement briefing in preparation for Reactor Shutdown.

ANSWER: a Answer: Although power has stabilized IPOI 3 directs you to AOP 255.2 if you enter the Buffer Region on the Power to Flow Map. AOP 255.2 will direct you to SCRAM if undamped oscillations are observed on any APRM. With the plant at low power and stabilized after this transient an immediate plant shutdown is not required unless the OSS determines it appropriate.

REFERENCE:

IPOI 3 Rev 54 page 3. AOP 255.2 Rev 22 pages 3-5.

Distracter 1: Entry into IPOI 3 is appropriate at this power. However, IPOI 3 does not direct inserting a SCRAM based on entry into the Buffer Region. It does require a SCRAM if the core becomes unstable.

Distracter 2: Entry into AOP 304.1 could be made if the OSS directs it. However, entry is not required because the bus was not lost. Rod withdrawal at this time would cause the plant to g'o deeper into the Buffer Region and possibly into the Exclusion Region. This action would cause the core to be more unstable.

Distracter 3: Entry into IPOI 4 could be directed if the OSS determines that we would have to shutdown to correct the problem. However, running the "B" Recirc to minimum would cause the plant to go deeper into the Buffer Region and possibly into the Exclusion Region. AOP 255.2 directs inserting Control Rods if you are going to exit the Buffer Region for shutdown.

K/A System: Generic K/A Number: 2.1.7 Ability to evaluate plant performance and to make operational judgements based on operating characteristics / reactor behavior / and instrument interpretation.

K/A Value: 3.7/4.4 10 CFR 55.43(b)(5)

DAEC Objective 5.03.06.01 Number:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 150 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

DAEC Objective Explain how component/system malfunction and corrective actions influence Statement: power constraints imposed following a power/reactivity abnormal change.

Cognitive Level: IP Source: NEW Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A I-I (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 151 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point 1 83. SRO ONLY The plant is at power.

A phone call is received in the Control Room with the report of a loud noise in the Condensate Pump area.

The following annunciator are received on 1C07B:

"* B-9 "Air Compressor Facility Trouble"

"* C-9 " Service Air Header LO Pressure"

"* A-10 "Instrument Air Dryers 1T-265A/B Trouble" 1K-90 A, B, and C are running and loaded.

The Auxiliary Operator reports the problem is confined to the in service Air Dryer and he will be shifting to the standby dryer.

Instrument Air Header pressure reached 75 psig and is starting to recover.

Besides AOP 518 "Failure of Instrument and Service Air" what would cause entry into another AOP based on these plant conditions and which AOP would be entered?

a. CV3039 "Instrument Air Header Isolation Valve" isolates at 80 psig causing the CRD Flow Control valves to CLOSE. Enter AOP 255.1 "Control Rod Movement/Indication Abnormal".

b. Cooling Tower Basin level fails high causing the River Water Makeup valves to CLOSE.

Enter AOP 410 "Loss of River Water Supply".

c. Partial loss of Extraction Steam to the Feedwater Heaters occurs. Enter AOP 646 "Loss of Feedwater Heating".

d. CV 4108 "Discharge to Offgas Stack Isolation Valve" CLOSES. Enter AOP 672.1 "Loss of Offgas System".

ANSWER: c Answer: When Air pressure is lowered to <80 psig air is isolated to the feedwater heater dump valves and they will fail open to the condenser. This will cause a loss of Feedwater heating and AOP 646 entry would be appropriate.

REFERENCE:

AOP 518 Rev 22 pages 7 and 8. 1C05A E-3 Rev 2 page 1 and 2.

Distracter 1: CV 3039 has closed due to pressure reaching 80 psig and this would cause the CRD drive water controllers to close. Under normal conditions loss of drive water would prompt enter into AOP 255.1. However, the candidate must realize that although the CRD Flow control valves are listed in AOP 518 as a load on Instrument air they are on an unisolable leg and 75 psig will maintain them in the correct position.

Distracter 2: Cooling water level will actually increase due to loss of air but not because of loss of air to the cooling tower level indicators. The increase is due to RWS Valves failing open on loss of air. The Cooling towers have their own independent air compressors to supply the level indicators and will not fail under these conditions.

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 152 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Distracter 3: The Offgas system is affected by loss of air and needs to be monitored. However, CV4108 does not fail closed on loss of air. This is a change in the way the system worked in the past. CV4108 use to go closed on loss of air or power and would cause a loss of Offgas. SCRAM freq. Reduction removed this feature and CV4108 now fails open on loss of air or power. CV4150 "Main Steam Supply to Offgas does fail closed and so a loss of air will affect condenser backpressure if not monitored and controlled.

K/A System: 300000 Instrument Air System K/A Number: 2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

K/A Value: 4.0/4.3 10 CFR 55.43(b)(5)

DAEC Objective SRO 5.17.03 Number:

DAEC Objective Direct operators to perform follow-up actions of AOP 518 Statement:

Cognitive Level: ip Source: NEW Operationally AOP entry conditions and system response.

Validity:

OE: Loss of Instrument Air is an industry and DAEC event.

Estimated Completion Time: EB#

Time Validation: N/A JI (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

/ Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-W xm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 153 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point k, . 8 4 . The Traversing In-core Probe System was in service when a Group 2 Containment Isolation signal is received.

One probe does NOT automatically retract.

The key lock switch on the TIP Valve Control Monitor has been placed to the FIRE position.

Which of the below correctly describes the control room indication(s) of this condition?

a. The SHEAR VLV MONITOR light is OFF.

b. The GROUP 2 ISOLATION light is OFF.

c. The BALL VALVE CLOSED light is ON.

d. The SQUIB MONITOR light is ON.

ANSWER: d Answer: After firing the squib valves the light on will indicate a successful actuation.

REFERENCE:

SD 878.6 Rev 5 pages 26 and 27, 01 878.6 Rev 23 pages 17 and 18.

Distracter 1: After firing the squib valves the Shear Valve Monitor light is ON, indicating a successful actuation. This is a common misconception opposite of the SBLC system.

Distracter 2: The Group 2 isolation light will be ON. However, the group 2 completed light will be OFF.

Distracter 3: The Ball valve will be open and the open light on. The shear valve is upstream of the valve and the TIP cable will not be able to be withdrawn after it is sheared.

This is a common misconception.

K/A System: 215001 Traversing In-Core Probe K/A Number: A3.03 Ability to monitor automatic operations of the TRAVERSING IN-CORE PROBE including: Valve operation.

K/A Value: 2.5/2.6 DAEC Objective 83.01.01.07 Number:

83.03.01.05 DAEC Objective Evaluate plant conditions and control room indications to determine if the TIP Statement: System is operating as expected, and identify any actions that may be necessary to place the TIP System in the correct lineup.

Describe how the TIP System responds to a group 2 PCIS isolation signal, include how to reset the TIP System Cognitive Level: 1I Source: Bank - 1998 NRC Operationally Automatic actions required for PCIS and the action did not occur. The operator is Validity: responsible to complete the action and verify the correct system response.

OE:

Estimated Completion Time: EB#

Time Validation: N/A Fm (time) Incorrect Ratio Data: (ratio) %_

Question Developed By: Peer Checked By:

Operator Validated By:

• / Approved By: Date: Date:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 154 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 155 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

85. SRO ONLY The plant is at power.

You are the OSS and HPCI is being returned to service following repairs to 1P218 "HPCI Auxiliary Oil Pump".

In what procedure do you find general guidance on checks that should be performed prior to declaring the HPCI system operable?

a. ACP 1403.3 "Modification Acceptance Test Control Program"

b. ACP 1415.3 "Inspection Trending Program"

c. MD-24 "Post Maintenance Testing Program"

d. MD-5 "Conduct of Maintenance" ANSWER: c Answer: MD-24 provides generic guidance on post maintenance testing. The Maintenance Coordinator TOSS is responsible for ensuring adequate PMT is assigned.

REFERENCE:

MD-24 Rev 15 page 2. Attachment 1 of MD-24 Rev 16 page 2.

Distracter 1: This is not considered a modification and this procedure would not apply.

Distracter 2: This is a QA procedure that has been deleted.

Distracter 3: This is a generic procedure that outlines the fundamental maintenance directives.

K/A System: Generic K/A Number: 2.2.21 Knowledge of pre and post maintenance operability requirements.

K/A Value: 2.3/3.5 10 CFR 55.43(b)(5)

DAEC Objective SRO 1.13.04.01 Number:

'*' DAEC Objective Explain how to use MD-24 to determine operability testing requirements.

Statement:

Cognitive Level: 1P Source: NEW Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A l1 (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm. doc - 156 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

86. SRO ONLY Which of the following can be used during venting the containment irrespective of radiation release but does NOT require plant entry?

a. SEP 301.1 "Torus Vent Via SBGT'.

b. SEP 301.3 "Torus Vent Via Hardpipe Vent".

c. SEP 303.1 "Air Purge for H2 Control in SAGs".

d. SEP 303.2 "N 2 Purge for H2 Control in SAGs".

ANSWER: a Answer: SEP 301.1 is performed when venting is required irrespective of rad release. This also can be done from the Control Room with no plant entry.

REFERENCE:

THE CANDIDATE IS PROVIDED A COPY OF EOP 2 EOP 2 Rev 9. EOP 2 bases Rev 8 pages 71 and 72.

Distracter 1: SEP 301.3 is performed when venting is required irrespective of rad release and SEP 301.1 can not be performed. However, this would require plant entry.

Distracter 2: The air purge could be used and venting could be required irrespective of rad release. However, this would require plant entry.

Distracter 3: N2 purge could be used and venting could be required irrespective of rad release.

However, this would require plant entry.

K/A System: Generic K/A Number: 2.3.8 Knowledge of the process for performing a planned gaseous radioactive release.

K/A Value: 2.3/3.2 10 CFR 55.43(b)(5)

DAEC Objective SRO 6.28.01.04 Number:

DAEC Objective Explain the criteria that will determine if the torus or the drywell is to be vented.

Statement:

Cognitive Level: 1P Source: New Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A r_1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 157 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

87. During vessel re-assembly the Head Bolts for the Steam Separator and Shroud Head Assembly were NOT tightened.

Which of the following indications would operators see?

a. The Reactor will go critical several rods early.

b. Recirc Pump flow mismatch will be greater at high power.

c. At higher Recirc flow rates as flow is increased power will NOT increase as expected.

d. The GEMAC Level instruments will start to deviate from each other until they are no longer within tolerance allowed by the TRM.

ANSWER: c Answer: This is an industry event from Dresden 3 1991. The operators observed these indications at 88% power when raising Recirc flow.

REFERENCE:

Event 249-910324-1 02/24/1991 Dresden unit 3.

Distracter 1: Pressure in the core region may be marginally affected. However, this would have little or no effect on when criticality occurs.

Distracter 2: Recirc flow may be effected at higher power but the mismatch under these conditions will be controlled by the operators and not effected by the event.

Distracter 3: The GEMAC level instruments would possibly be affected when the shroud head lifts. However, all the instruments will be affected equally.

K/A System: 290002 Reactor Vessel Internals K/A Number: K3.03 Knowledge of the effect that a loss or malfunction of the REACTOR VESSEL INTERNALS will have on the following: Reactor power.

K/A Value: 3.3/3.4 DAEC Objective 12.00.00.04 Number:

12.00.00.05 12.03.01.10 DAEC Objective Evaluate plant conditions and control room indications to determine if the Recirc Statement: system is operating as expected, and identify any actions that may be necessary to place the Recirc system in the correct lineup Identify the applicability of, and explain the significance of, any given SOER or IndustryEvent to the Recirc system State the parameters that must be monitored during power changes Cognitive Level: 3PEO Source: New Operationally Reactivity monitoring and vessel internal failure indications.

Validity:

OE: Event 249-910324-1 02/24/1991 Dresden unit 3.

Estimated Completion Time: EB#

Time Validation: N/A 11 (time) Incorrect Ratio Data: (ratio)  %

S Question Developed By: Peer Checked By:

Operator Validated By: _

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W-xm.doc - 158 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 159 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

88. Review the indication given in the Support Material Booklet The plant is at power.

The "B" Recirc MG Set Scoop tube is locked.

You are directed to take Local control of the "B" Recirc MG Set Scoop tube.

Review the Picture on the following page.

From the labeled components choose which one performs each of the following functions?

1) Deenergizes the brake.

2) Mechanically releases brake tension to allow manual turning of the hand crank.

3) Proper Hand Crank location when adjusting scoop tube position.

a. 1) B

2) A

3) D

b. 1)B

2) F

3) D

c. 1) C

2) A

3) E

d. 1) C

2) F

3) E ANSWER: a Answer: The locations indicated are the correct location to perform the indicated function.

REFERENCE:

OI264 Rev 70 page 33 and 34.

Distracter 1: B and D are correct. F is not correct per procedure and it will not perform the intended function.

Distracter 2: A is correct. E would function to control the scoop tube but is not the designated place for installation of the hand crank. C is not correct and would not perform the intended function.

Distracter 3: E would function to control the scoop tube but is not the designated place for installation of the hand crank. C and F are not correct and would not perform the intended functions.

K/A System: Generic K/A Number: 2.1.30 Ability to locate and operate components / including local controls.

K/A Value: 3.9/3.4 DAEC Objective 12.02 Number:

12.01.01.01 DAEC Objective Adjust Speed controls from MG Set Room Statement:

Identify the appropriate procedures that govern the Recirc system operation, include operator responsibilities during all modes of operation, and any actions required by personnel outside of the Control Room Cognitive Level: IF Rev. 0 Reactor Operator, 50007 57_2002- ILC- SRO-W-xm. doe - 160 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Source: New Operationally This is a reactivity manipulation performed out side the control room and would s-' Validity: only be performed by a Licensed operator during power operations.

OE:

Estimated Completion Time: EB#

Time Validation: N/A [m (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 161 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

89. SRO ONLY The plant is at power.

A problem with plant valving allows Condensate Storage Tank (CST) level to drop to 7.4 feet in both CSTs.

Level is being restored at 0.1 ft every 15 minutes.

Which of the following is correct concerning the operability of both HPCI and RCIC?

Both HPCI and RCIC are ...

a. inoperable due to less than 75,000 gallons of water available for RPV makeup.

b. inoperable since there is no assurance the discharge piping can be maintained full using the CSTs.

c. operable as long as the CST level is restored to greater than 75,000 gallons of water within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

d. operable as long as their discharge piping is lined up to the CRD system for keep fill within one hour.

ANSWER: b Answer: With CST level <8 ft the potential exist for voids and pockets of air to form in the discharge line and subsequent water hammer on an initiation.

THE CANDIDATE IS PROVIDED A COPY OF THE T.S. LCO BOOK

REFERENCE:

01 152 Rev 51 page 4. 01 150 Rev 42 page 4. T.S. Bases B 3.5.3 Amend. 223 pages B 3.5-27, 28, and 30. T.S. 3.5.1 Amend 223 page 3.5-4 Distracter 1: HPCI and RCIC are inop. However, the 75,000 gal requirement is for CS shutdown.

Distracter 2: The 75,000 gal requirement is for CS shutdown and both HPCI and RCIC are currently inop.

Distracter 3: The discharge piping on HPCI could be lined up to CRD. However, HPCI is still inop until that is accomplished and RCIC has no keep fill from CRD.

K/A System: Generic K/A Number: 2.1.32 Ability to explain and apply system limits and precautions.

K/A Value: 3.4/3.8 10 CFR 55.43(b)(2)

DAEC Objective SRO 5.18.01.01 Number:

DAEC Objective Evaluate plant conditions and determine Technical Specification compliance and Statement: identify any actions required.

Cognitive Level: 3SPR - The candidate must reference T.S. and the SR for the systems and apply the bases to assess the status of the system.

Source: New Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A [1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W-xm.doc - 162 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 163 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

90. SRO ONLY Operators have been responding to problems in the Offgas system. The following conditions have been observed:

• High range Offgas flow peaked at 100 scfm and is now stable at 50 scfm.

• The in-service Recombiner temperatures are at 600'F and have been rising slowly throughout the transient.

• Offgas pressure is at 20 in. Hg Abs and steady.

• Hydrogen concentration as read by both Offgas Hydrogen Analyzers is 4.5% and stable.

What, if any, actions must be taken, in response to the above listed situation?

a. There are no actions to be taken at this time.

b. Increase steam pressure to the operating SJAE to about 300 psig.

c. Place all of the loop seal drain valve handswitches in OPEN within 5 minutes to reduce system pressure to normal.

d. Reduce the Offgas hydrogen concentration to <4% in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

ANSWER: d Answer: ---The TRM requires us to reduce the H2 concentration to <4% in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

.HE CANDIDATE IS PROVIDED A COPY OF THE TRM LCO BOOK

REFERENCE:

ýJRRM 3.10.A Distracter 1: TRM 3.10 applies.

Distracter 2: This action is for a premature recombination event. There is no evidence, other than flow, of this event taking place. Recombiner temperatures contraindicate the presence of a premature recombination event.

Distracter 3: In the case of a loss of Offgas system, HIC4151 is used to lower system pressure, not the loop seal drain valves. Opening the drain valves at this time would cause a release of airborne contamination into the Power Block and Retention Building.

K/A System: Generic K/A Number: 2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

K/A Value: 3.4/4.0 10 CFR 55.43(b)(2 / 5)

DAEC Objective SRO 5.18.01.01 Number:

DAEC Objective Evaluate plant conditions and determine Technical Specification compliance and Statement: identify any actions required.

Cognitive Level: 1F Source: Bank - 1999 NRC Exam Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A I-- (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

,/ Operator Validated By: _

Approved By: Date: Date:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-W xm.doc - 164 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 165 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

91. SRO ONLY Reactor power is 40% when an EHC failure causes the Bypass and Control valves to go full open.

The MSIVs fail to automatically close, when required, but are manually closed.

Prior to MSIV closure, which of the following combinations of reactor power and reactor pressure would indicate that a safety limit violation had occurred?

a. Reactor power is 35% and RPV pressure is 790 psig.

b. Reactor power is 30% and RPV pressure is 775 psig.

c. Reactor power is 20% and RPV pressure is 770 psig.

d. Reactor power is 10% and RPV pressure is 750 psig.

ANSWER: b Answer: Per T.S. SL 2.1.1.1. This SL has been recently changed to *<21.7% RTP with reactor pressure <785 due to power uprate.

THE CANDIDATE IS PROVIDED A COPY OF THE T.S. LCO BOOK

REFERENCE:

-T.S. 2.1.1 Amend. 243 page 2.0-1 Distracter 1: Not a violation per T.S. 2.1.1 pressure >785 Distracter 2: Not a violation per T.S. 2.1.1 power <21.7%

Distracter 3: Not a violation per T.S. 2.1.1 power <21.7%

K/A System: Generic K/A Number: 2.2.22 Knowledge of limiting conditions for operations and safety limits.

K/A Value: 3.4/4.1 10 CFR 55.43(b)(2)

DAEC Objective SRO 5.18.01.01

' Number:

DAEC Objective Evaluate plant conditions and determine Tech Spec compliance and identify any Statement: actions required.

Cognitive Level: 1P Source: Bank - 1998 NRC Operationally T.S. and Safety Limit.

Validity:

OE: Recent change in T.S.

Estimated Completion Time: "EB#

Time Validation: N/A i-1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 166 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

92. A new Al control rod sequence is to be loaded into the RWM.

1) From what RWM panel is the new sequence loaded into RWM memory AND

2) from what panel is the Al sequence selected for RWM to enforce?

a. 1) RWM-OD (1C05 Operator Display)

2) RWM-OD (1C05 Operator Display)

b. 1) RWM-OD (1C05 Operator Display)

2) RWM-CC (1C28 Computer Chassis)

c. 1) RWM-CC (1C28 Computer Chassis)

2) RWM-OD (1C05 Operator Display)

d. 1) RWM-CC (1C28 Computer Chassis)

2) RWM-CC (1C28 Computer Chassis)

ANSWER: d Answer: The 01 directs both of these actions from the RWM-CC in the back panel area. The system description also describes where the two functions occur.

REFERENCE:

01 878.8 Rev 15 page 4-5. SD 878.8 Rev 5 pages 22-27 Distracter 1: These functions are not selectable at the RWM-OD.

Distracter 2: RWC-CC is correct. However, this function is not selectable at the RWM-OD.

"-.' Distracter 3: RWC-CC is correct. However, this function is not selectable at the RWM-OD.

K/A System: Generic K/A Number: 2.2.33 Knowledge of control rod programming.

K/A Value: 2.5/2.9 DAEC Objective 84.00.00.07 Number:

DAEC Objective Identify the appropriate procedures that govern the Rod Worth Minimizer System Statement: operation, include operator responsibilities during all modes of operation, and any action required by personnel outside of the Control Room Cognitive Level: 1P or S Source: New Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A I-1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 167 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.cdoc - 168 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

93. A reactor start up is in progress.

The IRMs are being ranged up.

The SRMs have been withdrawn.

The IRMs show Reactor power has peaked and is starting to LOWER.

What is occurring?

a. A Recirc run back has started.

b. Feed flow is increasing for makeup.

c. The Bypass Valves are starting to close.

d. The Reactor has reached the point of adding heat.

ANSWER: d Answer: Normal conditions for startup when reaching the POAH.

REFERENCE:

IPOI 2 Rev 70 page 13. Generic Fundamentals.

Distracter 1: A Recirc runback will cause power to go down. However, the Recirc pumps are already at minimum speed at this power level.

Distracter 2: Feed flow maybe increasing for makeup. However, an increase would cause power to go up.

Distracter 3: The bypass valves will affect reactivity. However, closing would cause power to go up.

K/A System: Generic K/A Number: 2.2.34 Knowledge of the process for determining the internal and external effects on core reactivity.

K/A Value: 2.8/3.2 DAEC Objective 93.00.00.09 Number:

93.03.01.11 DAEC Objective Explain the techniques used to heatup/pressurize the reactor during startup Statement: including a prediction of the response to a change in any controlling factor Evaluate plant conditions and control room indications to determine if the plant is exhibiting proper response during the startup, and identify any necessary actions if the response is not proper Cognitive Level: 2DR Source: New Operationally Power response during reactor startup to the point of adding heat.

Validity:

OE: Operators have misdiagnosed 1C05 indications during low power operations.

Estimated Completion Time: EB#

Time Validation: N/A F1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops

• -'- TMARs:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 169 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 170 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

, 94. SRO ONLY Who has the responsibility to approve radiological exposures in excess of 10CFR20 limits?

a. OSC Supervisor.

b. Emergency Coordinator.

c. Health Physics Supervisor.

d. Site Radiological Protection Coordinator.

ANSWER: b Answer: The Emergency Coordinator has this responsibility and it can not be delegate.

REFERENCE:

EPIP 4.3 Rev 10 pages 5, 6, 10, and 22.

Distracter 1: High level EPIP position. However, not authorized.

Distracter 2: High level EPIP position. However, not authorized.

Distracter 3: High level EPIP position. However, not authorized.

K/A System: Generic K/A Number: 2.3.4 Knowledge of radiation exposure limits and contamination control /including permissible levels in excess of those authorized.

K/A Value: 2.5/3.1 10 CFR 55.43(b)(?)

DAEC Objective SRO 3.01.04.01 Number:

DAEC Objective Explain the Responsibilities and Instructions of EPIP 4.3.

Statement:

Cognitive Level: 1P Source: Bank - 1999 LaSalle County Station NRC Exam Operationally EPIP Validity:

OE: Excessive exposure.

Estimated Completion Time: EB#

Time Validation: N/A F-1 (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 171 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

95. The 1C03 operator is performing an Air Purge (De-inerting) the Primary Containment.

Which of the following radiation-monitoring systems monitor the atmosphere that is exhausted through the Drywell/Torus vent valves?

a. Reactor Building Vent KAMAN monitors (KAMAN 3 through 8)

b. Offgas Vent Pipe Rad Monitors (RM-4116A & B) only

c. Offgas Stack KAMAN monitors (KAMAN 9 & 10) only

d. Offgas Vent Pipe Rad Monitors (RM-4116A & B)

AND Offgas Stack KAMAN monitors (KAMAN 9 & 10)

ANSWER: d Answer: This question measures knowledge of how many Radiation Monitors are involved during Containment purge, which comes from ventilation and goes past all 4 monitors. The Vent pipe monitors can send a Group 3 isolation signal.

REFERENCE:

P&ID M-14land M-176 Distracter 1: Possible misconception because it is not obvious that SBGT in the reactor Bldg.

exhausts out the Offgas Stack Distracter 2: Possible misconception because it is not obvious that the KAMAN monitors are downstream of the Offgas flow to the offgas stack.

Distracter 3: Common misconception that Vent pipe rad monitors are part of the Offgas system.

K/A System: Generic K/A Number: 2.3.9 Knowledge of the process for performing a containment purge.

Value: 2.5/3.4 SK/A DAEC Objective 85.00.00.03 Number:

87.00.00.05 DAEC Objective Describe in detail the subsystem of the PRM system, including methods Statement:

State the effluent types monitored by the KAMAN system.

Cognitive Level: 1S Source: Bank - 2001 NRC Operationally Rad release to atmosphere.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A E] (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 172 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

96. The plant is being started up following a refueling outage.

Inspectors are in the Drywell performing the 400-psig inspection.

Plant conditions are as follows:

"* RPV pressure is 350 psig

"* Reactor Power is 6%

Which of the following evolutions would NOT be allowed while the inspectors are still in the Drywell?

a. Perform a heavy lift with the Reactor Building Crane.

b. Close the inboard Drywell Personnel Hatch.

c. Movement of fuel within the Fuel Pool.

d. Raise Pressure set to 390 psig.

ANSWER: d Answer: IPOI 7 requires the operator verify RPV pressure to be <400 psig and to allow no significant increases in RPV Pressure.

REFERENCE:

IPOI 7 Rev 67 page 33 Distracter 1: There are no restrictions on the RB crane with personnel in the Drywell.

Distracter 2: The inboard hatch could be left open with an operator watching it. However, there is no requirement for this.

Distracter 3: There are requirements during outages with respect to fuel movement and personnel in the Drywell. However, this would have no effect on the inspectors in the Drywell under these conditions.

K/A System: Generic K/A Number: 2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

K/A Value: 2.9/3.3 DAEC Objective 93.25.01.01 Number:

93.25.01.02 DAEC Objective Explain the bases for each Precaution and Limitation and each Caution in IPOI 7 Statement:

Explain the restrictions on performing a Primary Containment entry Cognitive Level: 1P Source: New Operationally Radiation exposure and personnel safety. Also IPOI 7 caution.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A F-] (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 173 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 174 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

97. Which of the procedures listed below would require a DCF to change if you determine that performing one of the steps would lead to serious personnel injury?

a. ODI 19 "Tagging Practices".

b. AOP 301 "Loss of Essential Electrical Power".

c. EOP 2 "Primary Containment Control" Flow Chart.

d. SAG 1 "Primary Containment Flooding" Flow Chart.

ANSWER: b Answer: AOPs are covered in ACP 106.1 and required a DCF for the indicated problem.

REFERENCE:

ACP 106.1 Rev 22 pages 4, 5, and 10.

Distracter 1: Department instructions do not fall under the DCF requirements and can be changed at a lower level of review. These instructions are not covered by ACP106.1.

Distracter 2: EOP flow charts that require changed have to undergo a different process for changing and a DFC is not authorized for the change. These instructions are not covered by ACP106.1.

Distracter 3: SAG flow charts that require changed have to undergo a different process for changing and a DFC is not authorized for the change. These instructions are not covered by ACP106.1.

K/A System: Generic K/A Number: 2.4.5 Knowledge of the organization of the operating procedures network for normal / abnormal / and emergency evolutions.

K/A Value: 2.9/3.6 DAEC Objective 96.05.01.11 Number:

96.05.01.12 DAEC Objective Differentiate between when a Document Change Form (DCF) would be used and Statement: when a Procedure Work Request (PWR) would be used, Describe the temporary revision process in accordance with ACP 106.1, including actions required if the computer system is unavailable, Cognitive Level: 1P Source: New Operationally Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A I-- (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 175 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm. doc - 176 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

98. SRO ONLY Plant conditions are as follows:

"* SCRAM conditions exist but control rods did not insert.

"* Reactor power is 15%.

"* SBLC is injecting.

"* The MSIVs are CLOSED and LLS is controlling pressure.

"* RPV water level is 125 inches and is being deliberately lowered.

"* Torus temperature is 115 0 F.

Which one of the following is the primary reason for deliberately lowering RPV water level under these conditions?

a. To promote boron mixing.

b. To prevent exceeding a thermal limit.

c. To minimize the challenge to the fuel cladding.

d. To minimize the threat to primary containment integrity.

ANSWER: d Answer: Under these condition primary containment integrity is being challenged and lowering RPV level will reduce power which will then reduce the energy addition

REFERENCE:

to containment.

EOP bases ATWS Rev 7 page 16.

Distracter 1: Boron mixing is a consideration in EOPs. However, this is not the reason to lower RPV level.

Distracter 2: Thermal limits are a concern. However, this is not the reason to lower RPV level and ACC is still assured.

Distracter 3: Fuel Cladding is a concern. However, this is not the reason to lower RPV level and ACC is still assured.

K/A System: Generic K/A Number: 2.4.6 Knowledge symptom based EOP mitigation strategies.

K/A Value: 3.1/4.0 10 CFR 55.43(b)(5)

DAEC Objective SRO 6.51.03 Number:

DAEC Objective Direct operator actions to continue to lower RPV level until reactor power drops Statement: below 5%, or RPV level reaches +15", or all SRVs remain closed and drywell pressure remains below 2.0 psig.

Cognitive Level: 1B Source: Bank - 1998 Hope Creek NRC Audit Exam Operationally EOP bases Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A E] (time) Incorrect Ratio Data: (ratio)  %

Question Developed By: Peer Checked By:

-' ... Operator Validated By:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 177 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-SRO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 178 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point

99. The EOPs have been entered and plant conditions have degraded such that SAG entry is required.

The TSC is NOT ready to assume control.

Which of the following is correct?

The operating crew should ...

a. continue implementing the current EOP actions until the TSC is ready to transition to the SAGs.

b. exit the EOP which directs the entry into the SAGs and continue to implement all other EOPs which are entered.

c. exit the EOP leg that is directing the SAG entry and continue to implement all other EOPs legs in effect.

d. enter the SAG that is directed and when the TSC is ready, turnover all actions which were directed from the SAGs entered.

ANSWER: a Answer: Until the TSC is ready the operating crew is directed to continue to use the EOP strategies to combat the event.

REFERENCE:

EOP Bases Document - EOP Flow chart use and logic Rev 4 page 43 Distracter 1: Exiting the EOPs is correct. However, the TSC must be ready to take control and ALL EOPs are exited at that time.

Distracter 2: Exiting the EOPs is correct. However, the TSC must be ready to take control and ALL EOPs are exited at that time.

Distracter 3: Entering the SAGs would be correct if the TSC is ready. The crews do not enter the SAGs without the TSC being ready.

K/A System: Generic K/A Number: 2.4.14 Knowledge of general guidelines for EOP flowchart use.

K/A Value:... 3.0/3.9 DAEC Objective 95.74.16.01 Number:

95.74.16.02 DAEC Objective Explain the transition process from EOPs to SAGs Statement:

Explain the concept of default actions as it pertains to the actions to take while still in EOPs and waiting to make the transition to SAGs Cognitive Level: 1B Source: New Operationally EOP Bases.

Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A [] (time) Incorrect Ratio Data: (ratio)

Question Developed By: Peer Checked By:

vY Operator Validated By:

Approved By: Date: Date:

Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 179 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30)

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 180 2002 ILC SRO Written Exam

QF-1030-02 Rev. 0 (FP-T-SAT-30) 1 Point 100. To place the Reactor Core in a low energy state, reduce RPV temperature, and maintain RPV water level above the top of active fuel to prevent Fuel Cladding temperature from exceeding 1500 0 F, is the bases for which EOP?

a. EOP 1

b. EOP 2

c. EOP 3

d. EOP 4 ANSWER: a Answer: EOP 1 is utilized to shutdown, cooldown, and maintain ACC.

REFERENCE:

EOP Bases Document - EOP 1 RPV Control Rev 7 Page 1 Distracter 1: EOP 2 does have the reactor shutdown if conditions warrant. However, this is accomplished by entry into EOP 1 from EOP 2 at entry point 1.

Distracter 2: EOP 3 does have the reactor shutdown if conditions warrant. However, this is accomplished by entry into EOP 1 from EOP 2 at entry point 1.

Distracter 3: EOP 4 does have the reactor shutdown if conditions warrant. However, this is accomplished by entry into EOP 1 from EOP 2 at entry point 1.

K/A System: Generic K/A Number: 2.4.18 Knowledge of the specific bases for EOPs.

K/A Value: 2.7/3.6 DAEC Objective 95.00.00.03 Number:

DAEC Objective Explain the overall mitigation strategy of the EOPs

• -' Statement:

Cognitive Level: 1B Source: New Operationally EOP Bases Validity:

OE:

Estimated Completion Time: EB#

Time Validation: N/A M] (time) Incorrect Ratio Data: (ratio) 9/

Question Developed By: Peer Checked By:

Operator Validated By:

Approved By: Date: Date:

Operations Manager Trng Supervisor-Ops TMARs:

Question Usage (exams): 57_2002 ILC-RO-Wxm.doc Rev. 0 Reactor Operator, 50007 57_2002-ILC-SRO-Wxm.doc - 181 2002 ILC SRO Written Exam