ML090751084

2009-02-Final Written Exams, Site-Specific SRO Written Examination

ML090751084
Person / Time
Site:	Diablo Canyon
Issue date:	02/06/2009
From:	NRC Region 4
To:
References
Download: ML090751084 (204)

Text

U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name: Answer Key Date: 02/06/2009 Facility/Unit: Diablo Canyon Power Plant Region: I II III IV 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. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination.

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

Applicants Signature Results RO/SRO-Only/Total Examination Values 75 / 25 / 100 Points Applicants Scores / / Points Applicants Grade / / Percent

Diablo Canyon Power Plant ANSWER KEY Page 1e Multiple Choice (Circle or X your choice) NAME: ANSWER KEY If you change your answer, write your selection in the blank and initial.

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6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 2e Multiple Choice (Circle or X your choice) NAME: ANSWER KEY If you change your answer, write your selection in the blank and initial.

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6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 1 Question 01 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 003 K1.08 Importance Rating 2.7 K/A: Knowledge of the physical connections and/or cause-effect relationships between the RCPS and the following systems: Containment isolation Proposed Question:

Unit 1 is at full power.

A spurious Phase A isolation signal is received.

Which of the following describes the effect, if any, on RCP seal leakoff operation?

A. No effect on seal flow leakoffs will be noted.

B. No. 1 Seal leakoff flow increases.

C. No. 2 Seal leakoff flow increases.

D. No. 3 Seal leakoff flow increases.

Proposed Answer: C. No. 2 Seal leakoff flow increases.

Explanation:

Answer A incorrect - The seal return line isolates, number 1 seal leakoff is directed to the PRT through relief valve RV-8121.

Answer B incorrect - Seal leakoff will decrease.

Answer C correct - the closing of the number 1 seal leakoff causes pressure in the line to increase to the relief valve setpoint (150 psig), and the increased pressure will cause number 2 seal leakoff to increase.

Answer D incorrect - Number 3 leakoff is unaffected.

Technical Reference(s): STG A6, system interfaces 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 2 Proposed references to be provided to applicants during examination: NONE Learning Objective: 35744 - Discuss abnormal conditions associated with the RCP Question Source: Bank # P-0026 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 3 Question 02 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 004 K5.11 Importance Rating 3.6 K/A: 004 K5.11 - Knowledge of the operational implications of the following concepts as they apply to the CVCS: Thermal stress, brittle fracture, pressurized thermal shock Proposed Question:

Which of the following is the concern if Auxiliary Spray is initiated when there is a temperature difference of greater than 320°F across the Auxiliary Spray line?

A. thermal shock of the Pressurizer spray nozzle.

B. thermal stratification of the surge line.

C. thermal shock of the Regenerative Heat Exchanger.

D. thermal stratification of the spray line.

Proposed Answer: A. thermal shock of the Pressurizer spray nozzle.

Explanation:

Answer A correct -

E-3 caution states: Thermal Shock of the PZR will result if the Delta T between PZR Aux Spray and the PZR Steam Space exceeds 320°F. 300°F is the limit for thermal stratification of the pressurizer surge line (associated with pressurizer spray valves).

Answer B incorrect - OP L-1, 6.1.2.i states: At all times during the heatup, maintain the differential temperature between the Pressurizer liquid space and the Loop 2 Hot Leg to less than or equal to 300°F (REFER TO PPC address U0621). This will limit the magnitude of stresses which could be introduced by thermal stratification in the Pressurizer surge line.

Answer C correct - the regen heat exchanger is not the concern, (RCS and charging).

Answer D incorrect - stratification is associated with the surge line, not the spray line.

Technical Reference(s): E-3, caution prior to step 23 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 4 Proposed references to be provided to applicants during examination: NONE Learning Objective: 5093 - Discuss significant precautions and limitations associated with the CVCS Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 10 CFR Part 55 Content: Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 5 Question 03 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 004 A3.03 Importance Rating 2.9 K/A: Ability to monitor automatic operation of the CVCS, including: Ion exchange bypass.

Proposed Question:

Unit 1 is at full power. A letdown mixed bed is in service.

Letdown temperature at the outlet of the Letdown Heat Exchanger rises to 140°F.

What is the status of letdown at this time?

NOTE:

TCV-149 - Letdown Temperature Diversion valve LCV-112A - VCT Level Control Valve A. TCV-149 has repositioned to bypass the demins and direct flow to the VCT.

B. TCV-149 and LCV-112A have repositioned to bypass the demins and divert letdown to the LHUT.

C. LCV-112A has repositioned to divert letdown to the LHUT.

D. No letdown realignment has occurred; the setpoint for auto action has not been reached.

Proposed Answer: A. TCV-149 diverts letdown flow to the VCT.

Explanation:

Answer A correct - divert occurs at 136°F to protect the demins from damage if temperature were to reach 145°F.

Answer B incorrect - Only TCV-149 repositions and diverts to the VCT.

Answer C incorrect - LCV-112A is not affected.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 6 Answer D incorrect - 145°F is the maximum temperature, however, diverts occurs at 136°F Technical Reference(s): STG B1A Proposed references to be provided to applicants during examination: NONE Learning Objective: 40449 - Discuss abnormal conditions associated with the CVCS Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 7 Question 04 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 005 K5.02 Importance Rating 3.4 K/A: 005 K5.02 - Knowledge of the operational implications of the following concepts as they apply the RHRS: Need for adequate subcooling.

Proposed Question:

GIVEN:

• The crew is performing the actions of E-1.2, Post-LOCA Cooldown and Depressurization

• One RCP is running

• Both ECCS CCPs are running

• Both SI pumps are running

• Both RHR pumps are stopped

• RCS temperature is 300°F When evaluating conditions to stop the first SI pump, there is less subcooling than required. In accordance with the RNO step in E-1.2, the operator starts an RHR pump and then stops the SI pump.

What is accomplished by starting an RHR pump prior to stopping the SI pump?

A. Prevents void formation in the reactor vessel head.

B. Ensures the RCS will remain subcooled.

C. Prevents a challenge to the Core Cooling critical safety function.

D. Ensures conditions are maintained for continued RCP operation.

Proposed Answer: B. Ensures the RCS will remain subcooled.

Explanation:

Answer A incorrect - with an RCP running, voids should not form in the reactor vessel head.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 8 Answer B correct - the background document states: If the RCS subcooling criterion is not satisfied, but the RCS hot leg temperatures are less than the saturation temperature corresponding to the low-head (RHR) SI pump head at minimum pump recirculation flow, the charging/SI pump can be stopped if a low-head SI pump is running or can be started. Starting a low-head SI pump for this case ensures that RCS subcooling will be maintained after the charging/SI pump is stopped.

Answer C incorrect - there is adequate ECCS flow with both CCPs and the remaining SI pump to prevent any challenge to core cooling.

Answer D incorrect - RCP operation is not required for current conditions.

Technical Reference(s): E-1.2 background step 13 (DCPP step 15)

Proposed references to be provided to applicants during examination: NONE Learning Objective: 7920 - Explain basis of emergency procedure steps Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 10 CFR Part 55 Content: Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 9 Question 05 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 006 K6.13 Importance Rating 2.8 K/A: 006 K6.13 - Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: Pumps Proposed Question:

A large break LOCA has occurred.

Which of the following would present the greatest challenge to long term core cooling?

A. Loss of ECCS CCPs.

B. Loss of SI pumps.

C. Loss of RHR pumps.

D. Only 2 Accumulators inject.

Proposed Answer: C. Loss of RHR pumps.

Explanation:

E-1 background, After successful initial operation of the ECCS, the reactor core is once again covered with borated water. This water has enough boron concentration to maintain the core in a shutdown condition. Decay heat is removed by a continuous supply of water from the ECCS. This supply initially comes from the refueling water storage tank (RWST). When the RWST level reaches the switchover setpoint the ECCS pumps are transferred into the recirculation mode (using ES 1.3, TRANSFER TO COLD LEG RECIRCULATION) wherein water is drawn from the containment sump and is cooled in the residual heat removal heat exchangers. Thus, long term cooling of the core is maintained by the ECCS in sump recirculation mode. The core is maintained in a shutdown state by borated water.

Also background for step 11: An evaluation of plant equipment available following a LOCA is necessary in determining long-term recovery actions. Hence, this evaluation is initiated at this time and any additional equipment that would assist in the plant recovery is started.

Answer A incorrect - without RHR pumps, CCPs do not have a suction source.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 10 Answer B incorrect - without RHR pumps, SI pumps do not have a suction source.

Answer C correct - RHR pumps provide a source to the ECCS pumps and cooling for water coming out of the containment sump.

Answer D incorrect - Since the accumulators discharge during the blowdown and reflood phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46, though their water volume is credited as part of the long term cooling inventory.

Technical Reference(s): E-1 background, STG B3 Proposed references to be provided to applicants during examination: NONE Learning Objective: 35324 - Explain significant ECCS design features and the importance to nuclear safety Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 10 CFR Part 55 Content: Components, capacity, and functions of emergency systems.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 11 Question 06 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 006 G2.1.30 Importance Rating 4.4 K/A: 006 G2.1.30 - ECCS: Ability to locate and operate components, including local controls.

Proposed Question:

GIVEN:

• A LOCA occurs

• Offsite power is available

• The plant is aligned for Cold Leg Recirculation

• SI has been reset A loss of Start-Up power occurs. The Diesel Emergency Generators start and load their respective vital bus.

Which of the following actions will be taken by the operator?

A. Trip the ECCS CCPs if an RHR pump does not automatically start.

B. Manually start the RHR pumps, then manually start the ECCS CCPs.

C. Hold the switches for the ECCS CCPs in STOP/RESET until RHR is in service.

D. Manually restart the ECCS CCPs after RHR pumps are automatically started.

Proposed Answer: C. Hold the switches for the ECCS CCPs in STOP/RESET until RHR is in service.

Explanation:

Answer A incorrect - the RHR pumps must be started by the operator.

Answer B incorrect - the CCPs will automatically start.

Answer C incorrect - the switches are held in STOP/RESET to prevent the CCPs from running until an RHR pump can be started. Otherwise the pumps would be running without a suction source.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 12 Answer D incorrect - CCPs automatically start. Because SI is reset, the RHR do not.

Technical Reference(s): E-1, Foldout page Proposed references to be provided to applicants during examination: NONE Learning Objective: 42458 Explain basis of emergency steps of E-1.3 Question Source: Bank # X P-50918 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 13 Question 07 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 007 K4.01 Importance Rating 2.6 K/A: 007 K4.01 - Knowledge of PRTS design feature(s) and/or interlock(s) which provide for the following: Quench tank cooling.

Proposed Question:

Unit 1 is at full power in a normal lineup.

Pressurizer Relief Tank (PRT) temperature is increasing.

Which of the following describes how cooling is aligned to the PRT?

A. Outside containment Primary Water valve, 8029, automatically opens if temperature in the PRT reaches 130°F.

B. Inside containment Primary Water valve, 8030, automatically opens if temperature in the PRT reaches 130°F.

C. The operator must open inside containment Primary Water valve, 8030, (outside containment Primary Water valve, 8029, is normally open).

D. The operator must open both the outside containment Primary Water valve, 8029 and inside containment Primary Water valve, 8030.

Proposed Answer: C. The operator must open inside containment Primary Water valve, 8030, (outside containment Primary Water valve, 8029, is normally open).

Explanation:

Answer A incorrect - no auto actions, this is the high temperature alarm setpoint.

Answer B incorrect - no auto actions, this is the high temperature alarm setpoint.

Answer C correct - no auto action and 8029 is normally open.

Answer D incorrect - 8029 is normally open.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 14 Technical Reference(s): STG A4B, section 3.0, section 2.1 and AR PK05-25 Proposed references to be provided to applicants during examination: NONE Learning Objective: 4950 - Explain the operation of PRT system Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

No comments 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 15 Question 08 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 008 A2.02 Importance Rating 3.2 K/A: 008 A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High/low surge tank level Proposed Question:

Unit 1 is at full power.

The following events occur:

• PK01-07, CCW Sys Surge Tk Lvl/Mk-Up alarms

• CCW surge tank level:

o LI 139 - 4% and decreasing o LI 140 - 3% and decreasing

• Makeup is not available

• Containment temperature is 85°F and steady.

• The crew is performing the actions of OP AP-11, Malfunction of Component Cooling Water System, section C, CCW System Outleakage In accordance with the procedure, which of the following will be required if surge tank level continues to decrease?

A. Stopping of all running CFCUs to prevent the units from tripping on overload.

B. Stopping of all running CFCUs to prevent steam binding in the units due to CCW flashing.

C. A reactor trip because the CCW pumps may cavitate resulting in a loss of cooling to the Centrifugal Charging Pumps.

D. A reactor trip and tripping of all RCPs because the CCW pumps may cavitate resulting in a loss of cooling to the RCP motors.

Proposed Answer: D. A reactor trip and tripping of all RCPs because the CCW pumps may cavitate resulting in a loss of cooling to the RCP motors.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 16 Explanation:

Answer A incorrect - CFCUs trip on overload and a loss of cooling could cause temperature to increase. This is not the action to be taken, and the procedure does not address the CFCU temperatures.

Answer B incorrect - CFCUs are cooled by CCW. A loss of CCW should not cause temperature to rise high enough for flashing. In ECA-0.0, CCW is isolated to prevent flashing in the RCP thermal barrier return line.

Answer C incorrect - CCW pumps could cavitate if surge tank level is offscale low, however, a loss of cooling to the CCPs does not require a reactor trip. The step in section C of AP-11 for low surge tank level (step 2) requires a trip and stopping of RCPs and aligning of alternate cooling to the CCPs. However, the trip is required due to the loss of the RCPs not the loss of cooling to the CCPs.

Answer D correct CCW pumps could cavitate if surge tank level is offscale low and they would have to be tripped. This would result in a loss of cooling to the RCP motors.

Therefore, a trip and stopping RCPs is required. Loss of cooling to the motors requires the RCPs tripped in less than 5 minutes (OP AP-28, section E).

Technical Reference(s): AR PK01-07 step 2.2.3, OP AP-11, section C, STG H-2, page 2-6 Proposed references to be provided to applicants during examination: NONE Learning Objective: 35490 - Discuss abnormal conditions associated with the CCW system Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 10 CFR Part 55 Content: Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments: 23 January, 2009 - changed answers to include balanced actions and reason.

Expanded justification, added that containment temperature is stable to remove CFCUs as a problem.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 17 Question 09 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 010 K6.03 Importance Rating 3.2 K/A: 010 K6.03 - Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: PZR sprays and heaters.

Proposed Question:

GIVEN:

• Unit 1 is at full power

• RCS pressure is 2235 psig

• Backup heaters are off

• Both Pressurizer spray valves are closed A failure causes all the Pressurizer heaters to energize.

Which of the following occurs? Assume proper operation of controller HC-455K.

A. Output will increase until the spray valves open, RCS pressure will be returned to approximately 2235 psig.

B. Output will increase until the spray valves open, RCS pressure will stabilize at approximately 2310 psig.

C. Setpoint will increase until the spray valves open, RCS pressure will be returned to approximately 2235 psig.

D. Setpoint will increase until the spray valves open, RCS pressure will stabilize at approximately 2310 psig.

Proposed Answer: A. Output will increase until the spray valves open, RCS pressure will be returned to 2235 psig.

Explanation:

Answer A correct - the output will increase, the spray valves will open enough to balance to input from the heaters and reduce pressure back to setpoint. At setpoint, heaters and spray will be on.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 18 Answer B incorrect - output will increase and open the sprays, RCS pressure will be returned to setpoint by the spray valves.

Answer C incorrect - the setpoint will not change.

Answer D incorrect - the setpoint will not change.

Technical Reference(s): STG A4A, section 2.2 Proposed references to be provided to applicants during examination: NONE Learning Objective: 4560 - Describe the operation of the Pzr, Pzr Pressure and Level Control System Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments: not revised. Removed the word fully from problem statement (heaters to fully energize.)

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 19 Question 10 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 010 A4.01 Importance Rating 3.7 K/A: PZR PCS, Ability to manually operate and/or monitor in the control room: PZR spray valve.

Proposed Question:

GIVEN:

• The plant is in MODE 3

• All RCPs are running

• RCS pressure is 2280 psig and slowly decreasing

• Pressurizer spray valves are open in AUTO A bus fault causes 12 kV bus D to de-energize causing RCPs 12 and 14 to trip. No operator action has been taken.

Which of the following will occur?

A. Both Pressurizer spray valves remain open. Only PCV-455A could lower RCS pressure.

B. Both Pressurizer spray valves remain open. Only PCV-455B could lower RCS pressure.

C. PCV-455A remains open to lower RCS pressure. PCV-455B closes.

D. PCV-455B remains open to lower RCS pressure. PCV-455A closes.

Proposed Answer: A. Both Pressurizer spray valves remain open. Only PCV-455A could lower RCS pressure.

Explanation:

Answer A correct - RCPs 2 and 4 are powered from Bus D. As a result of the lost bus, spray valve PCV-455B would not be effective in reducing pressure. Both would still be open in response to the high pressure, but only PCV-455A would have any effect.

Answer B incorrect - PCV-455B effectiveness is lost when RCP 2 loses power.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 20 Answer C incorrect - Although driving force for spray is lost, controllers for both valves would still open the valves.

Answer D incorrect - Although driving force for spray is lost, controllers for both valves would still open the valves.

Technical Reference(s): OIM A-4-1 & J-1 Proposed references to be provided to applicants during examination: NONE Learning Objective: 36927 - Describe system interrelationships between the Pzr, Pzr Pressure and Level Control System and other plant systems Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments: not revised. Expanded justification for A.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 21 Question 11 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 012 A2.03 Importance Rating 3.4 K/A: RPS, Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Incorrect channel bypassing Proposed Question:

A plant startup is being performed.

At 10% power, the operator places both Intermediate Range Block switches on CC1 in BLOCK.

The operator reports PK08-19, Intermediate Range Trip Blocked light did not light.

Which of the following describes the status of the Intermediate Range High Flux trip?

A. PK08-19 should be lit; the startup should be stopped to prevent a plant trip.

B. The startup may continue; this is the expected condition for PK08-19 above 10%

power.

C. PK08-19 should be lit, however, the startup to full power may continue as long as both Intermediate Range level trip switches are placed in BYPASS.

D. The startup may continue; PK08-19 will not be lit until power is approximately 25%.

Proposed Answer: A. PK08-19 should be lit, the startup should be stopped to prevent a plant trip.

Explanation:

Answer A correct - if both trains are not blocked, the PK will not be lit (both blocked causes the PK to be lit). if both are not blocked, at 20% rod motion would be stopped, at 25%, a reactor trip would occur.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 22 Answer B incorrect - both channels blocked, causes light to be lit. The PK should be lit above 10% power.

Answer C incorrect - the trip could be blocked, (both IR can be inoperable above 10%),

however, the rod stop at 20% would prevent reaching full power.

Answer D incorrect - the PK is lit once both channels are blocked.

Technical Reference(s): B4, figure NIS-14, 15, AR PK08-19 Proposed references to be provided to applicants during examination: NONE Learning Objective: 36971 - Describe controls, indications, and alarms associated with the NIS Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2 55.43 10 CFR Part 55 Content: General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 23 Question 12 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 013 G2.4.31 Importance Rating 4.2 K/A: 012 G2.4.31 - ESFAS - Knowledge of annunciator alarms, indications, or response procedures.

Proposed Question:

Which of the following conditions could cause the PK08-22-"AUTO SI BLOCKED" and PK02-02-"SAFETY INJECTION INITIATE" lights to be on at the same time?

A. Only one train of SI resets.

B. Only one train of SI actuated.

C. The P-4 signal actuates after SI is reset.

D. The reactor trip breakers are closed after SI is reset.

Proposed Answer: A. Only one train of SI resets.

Explanation:

Answer A correct - PK0202 is lit if either train has an SI signal present. PK0822 is lit if either train is reset. If only one train resets, then both PKs would be lit.

Answer B incorrect - if only 1 train actuated, then when reset, PK0202 would be not lit, or if that train didnt reset then PK0822 would be out.

Answer C incorrect - lack of P-4 would prevent PK0822 from actuating and P-4 does not automatically reset SI, the operator must reset SI with P-4 present.

Answer D incorrect - removing P-4 would clear PK0822.

Technical Reference(s): OIM B-6-5 Proposed references to be provided to applicants during examination: NONE Learning Objective: 37049 - Describe controls, indications, and alarms associated with the RPS 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 24 Question Source: Bank # X XA-0555 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Not revised, changed LOD based on NRC review.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 25 Question 13 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 013 K1.10 Importance Rating 2.8 K/A: Knowledge of the physical connections and/or cause effect relationships between the ESFAS and the following systems: CPS (containment purge system)

Proposed Question:

A containment purge is in progress.

A large break LOCA occurs.

Which of the following stops the containment purge supply and exhaust fans?

A. Low flow when the isolation valves close.

B. Containment Isolation Phase B signal.

C. Safety Injection signal.

D. Manually stopping the fans using fan control switches.

Proposed Answer: D. Manually stopping the fans using fan control switches Explanation:

A incorrect. There are fans that stop on low flow, such as Aux Building fans, however, the purge fans do not.

B incorrect. Phase B causes CVI, but does not stop the fans.

C incorrect. SI actuates CVI, however, CVI does not stop the fans.

D correct - STG H-4, Page 3 The supply and exhaust fans are not automatically started or stopped during emergency operations. This must be performed manually if necessary (fans trip on OC or thermal overload) 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 26 CVI actuation causes the following valves to close:

• Close RCV-11 and RCV-12

• Close FCV-679 and 681

• Close containment purge supply valves FCV-660 and 661

• Close containment pressure/vacuum relief FCV-662, 663 and 664.

Technical Reference(s): STG H-4, OIM B-6-9a Proposed references to be provided to applicants during examination: NONE Learning Objective: 5141 - Describe the operation of the Containment Purge System Question Source: Bank # X P-1219 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 27 Question 14 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 022 A4.01 Importance Rating 3.6 K/A: 022 A4.01 - Ability to manually operate and/or monitor in the control room: CCS fans Proposed Question:

GIVEN:

• CFCUs 12, 14, & 15 running in high speed

• CFCUs 11 & 13 off

• CFCU drain valves open What action must be done in order to monitor for RCS leakage using Containment Fan Cooling Unit 12?

A. Close the CFCU 12 drain valve.

B. Shift CFCU 12 to low speed and close the CFCU 12 drain valve.

C. Shift all running CFCUs to low speed and close the CFCU 12 drain valve.

D. Shift all running CFCUs to low speed and close the CFCU 14 and 15 drain valves.

Proposed Answer: B. Shift CFCU 12 to low speed and close the CFCU 12 drain valve.

Explanation:

Answer A incorrect - the fan must be running in low speed.

Answer B correct - to place the drain system in service, the CFCU is placed in low speed and the drain valve closed.

Answer C incorrect - only the one used for drain collection must be in slow speed.

Answer D incorrect - the drain system is placed in service by closing the valve for the CFCU to be used for monitoring.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 28 Technical Reference(s): OP H-2:I section 6.5 Proposed references to be provided to applicants during examination: NONE Learning Objective: 6176 Describe the operation of the CFCUs Question Source: Bank #

Modified Bank # A-0501 (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 10 CFR Part 55 Content: Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments: difficulty based on bank question difficulty rating. Not revised, all caps for GIVEN 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 29 Question 15 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 022 A2.01 Importance Rating 2.5 K/A: 022 A2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Fan motor over-current Proposed Question:

GIVEN:

• A large break LOCA occurs

• The operator is performing Appendix E of E-0, Reactor Trip or Safety Injection

• The operator notes the following indication for the CFCUs:

o CFCUs 14 and 15 are running in HIGH o CFCUs 11, 12 and 13 are running in LOW Which of the following actions will be taken?

A. Shift CFCUs 11, 12 and 13 to HIGH speed to prevent exceeding Containment design pressure.

B. Shift CFCUs 11, 12 and 13 to HIGH speed to prevent the fans tripping on overcurrent.

C. Shift CFCUs 14 and 15 to LOW speed to prevent the fans tripping on overcurrent.

D. Shift CFCUs 14 and 15 to LOW speed to prevent CCW temperature exceeding 140°F.

Proposed Answer: C. Shift CFCUs 14 and 15 to LOW speed to prevent the fans tripping on overcurrent.

Explanation:

Answer A incorrect - appendix E verifies CFCU lights OFF. If lit, the action is to make the light go out by taking the CFCU to low speed. CFCUs run during a LOCA to maintain temperature less than 120°F which keeps containment from overpressurizing, but in low to keep the fans from tripping on overcurrent.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 30 Answer B incorrect - Fans will trip on OC if running in HIGH, not LOW.

Answer C correct - CFCUs are run in slow speed to prevent the fans tripping on overcurrent due to the adverse containment temperature.

Answer D incorrect - CCW is used to cool the CFCUs and design temperature is 140°F.

a design feature of CCW is to provide cooling to ESF equipment, including all CFCUs during a LOCA.

Technical Reference(s): STG F-2, design basis-page 1-5, STG H-2, design basis-page 1-6, 3-6. E-0 appendix E, step 4 Proposed references to be provided to applicants during examination: NONE Learning Objective: 48812 - Discuss abnormal conditions associated with the CFCUs Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 10 CFR Part 55 Content: Components, capacity, and functions of emergency systems.

Comments: not revised, expanded justification for A and changed LOD based on NRC review.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 31 Question 16 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 026 K2.01 Importance Rating 3.4 K/A: 026 K2.01 - Knowledge of bus power supplies to the following: Containment spray pumps.

Proposed Question:

Which of the following lists the power supplies for Containment Spray pumps 11 and 12?

A. 11 - Bus G 12 - Bus H B. 11 - Bus F 12 - Bus G C. 11 - Bus F 12 - Bus H D. 11 - Bus H 12 - Bus G Proposed Answer: A. 11 - Bus G Bus H Explanation:

Answer A correct - power supply to the 11 and 12 pumps are G and H respectively.

Answer B incorrect - G is the power supply to the 11 pump, and no CS are on bus F.

Answer C incorrect - H is the power supply to the 12 pump, no CS pumps on bus F.

Answer D incorrect - power supplies are reversed.

Technical Reference(s): OIM J-1 Proposed references to be provided to applicants during examination: NONE 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 32 Learning Objective: 6022 - State the power supplies to CSS components Question Source: Bank # X S-46995 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 10 CFR Part 55 Content: Components, capacity, and functions of emergency systems.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 33 Question 17 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 039 A3.02 Importance Rating 3.1 K/A: 039 A3.02 - Ability to monitor automatic operation of the MRSS, including:

Isolation of the MRSS Proposed Question:

GIVEN:

• A steam break occurs inside containment on the 11 Steam Generator

• Safety Injection on high containment pressure

• 11 Steam Generator is completely depressurized

• Containment pressure peaks at 19 psig The operator reports all MSIVs are closed.

What was the condition to cause the MSIVs to close?

A. Safety Injection B. Phase A C. Low steam line pressure D. Phase B Proposed Answer: C. Low steam line pressure Explanation:

Answer A incorrect - SI does not close MSIVs (note: low steam pressure causes SI and closes the MSIVs)

Answer B incorrect - Phase A, generated by SI isolates many containment penetrations, but not the main steam lines.

Answer C correct - the MSIVs will close when steam pressure falls to 600 psig (lead-lag compensated) 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 34 Answer D incorrect - phase B will close the MSIVs but the setpoint was not reached.

Technical Reference(s): OIM B-6-10 Proposed references to be provided to applicants during examination: NONE Learning Objective: 37048 - Analyze automatic features and interlocks associated with the RPS Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments: changed conditions from numbered to bullets 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 35 Question 18 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 059 A4.03 Importance Rating 2.9 K/A: 059 A4.03 - Ability to manually operate and monitor in the control room:

Feedwater control during power increase and decrease Proposed Question:

A plant downpower from 100% to 50% is being performed in accordance with OP L-4, Normal Operation at Power.

Approximately how much will the programmed auto delta-P setpoint for the Digital Feedwater Control Feed Pump Speed Control change as power is reduced from 100 to 50% power?

A. 50 psid.

B. 60 psid C. 74 psid.

D. 96 psid.

Proposed Answer: B. 60 psid Explanation: DP changes from 74 to 170 psid (20 to 100% power or 80% range) for a total change of 96 psid.

Answer A incorrect - this approximately half the range (96/2).

Answer B correct - (100-50)/80

• 96 = 60 psid.

Answer C incorrect - this is the bottom of the range.

Answer D incorrect - this is the total change from 20 to 100%.

Technical Reference(s): STG figure DFWCS-16 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 36 Proposed references to be provided to applicants during examination: NONE Learning Objective: 4962 Describe DFWCS components Question Source: Bank # A-0041 Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 10 CFR Part 55 Content: Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

23 January, 2009 - changed conditions to have the candidate determine what the DP will be as power is reduced to align the question with the KA.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 37 Question 19 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 059 K3.04 Importance Rating 3.6 K/A: 059 K3.04 - Knowledge of the effect that a loss or malfunction of the MFW will have on the following: RCS Proposed Question:

Unit 1 is operating at 100% power.

A large Feedwater line break occurs between the outlet of the #1 Main Feedwater Heater and FCV441, Loop 14 Main Feedwater Isolation valve.

As a result, RCS temperature will A. decrease prior to the reactor trip. Post trip, the RCS temperature decrease and Steam Generator 14 level decrease is stabilized by the FWI.

B. decrease prior to the reactor trip. Post trip, temperature continues to decrease until Steam Generator 14 blows down and drys out.

C. increase prior to the reactor trip. Post trip, Steam Generator 14 level will recover due to AFW flow.

D. increase prior to the reactor trip. Post trip, temperature continues to increase due to the unavailability of AFW flow.

Proposed Answer: C. increase prior to the reactor trip. Post trip, Steam Generator 14 level will recover due to AFW flow.

Explanation:

Answer A incorrect - Initially the loss of heat removal will cause the RCS to heatup.

Answer B incorrect - Initially the loss of heat removal will cause the RCS to heatup.

Answer C correct - decreased heat transfer causes RCS temperature to increase. Because the leak is upstream of the isolation valve, AFW will restore level. A break in this location is essentially a loss of feed to the affected steam generator. As such analysis for loss of Feedwater is applicable.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 38 Answer D incorrect - Because the break is upstream of the AFW connection and will be isolated when the Feedwater isolation valve closes, AFW will be available for the steam generator.

Technical Reference(s): FSAR section 15.4.2A.2 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7380 - Explain plant response to loss of feedwater Question Source: Bank # INPO 28092 Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam Robinson 2004 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 10 CFR Part 55 Content: Secondary coolant and auxiliary systems that affect the facility.

Comments:

23 January, 2009 - Changed justification reference to better support question answer.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 39 Question 20 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 061 K1.11 Importance Rating 2.7 K/A: 061 K1.11 - Knowledge of the physical connections and/or cause/effect relationships between the AFW and the following systems: AFW turbine exhaust drains Proposed Question:

A steam generator tube rupture has occurred on Steam Generator 12. The steam generator has not yet been isolated. All AFW pumps are running.

Which of the following releases is occurring from the TDAFW pump exhaust?

A. An unmonitored release to atmosphere.

B. A monitored release to atmosphere.

C. An unmonitored release to the main condenser.

D. An monitored release to the main condenser.

Proposed Answer: A. An unmonitored release to atmosphere.

Explanation:

Answer A correct - the exhaust is to atmosphere and unmonitored.

Answer B incorrect - no rad monitor on the exhaust lines.

Answer C incorrect - Steam drains can be directed to the condenser, but not the exhaust.

Answer D incorrect - Steam drains can be directed to the condenser, but not the exhaust.

Technical Reference(s): STG D-1 page 2-19 Proposed references to be provided to applicants during examination: NONE 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 40 Learning Objective: 8402 - Describe the operation of the AFW system Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 10 CFR Part 55 Content: Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Comments:

Not revised, added information that all AFW pumps are running.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 41 Question 21 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 062 K3.03 Importance Rating 3.7 K/A: 062 K3.03 - Knowledge of the effect that a loss or malfunction of the AC distribution system will have on the following: DC system Proposed Question:

Unit 1 is at full power, in a normal electrical lineup.

Power is lost to 480V Bus H.

What should currently be supplying power to Vital DC distribution panel 13?

A. Battery 11 B. Battery 13 C. Battery Charger 131 D. Battery Charger 132 Proposed Answer: B. Battery 13 Explanation:

Answer A incorrect - Battery 11 supplies distribution panel 11.

Answer B correct - loss of Bus H causes a loss of power to BC 131 and the DC bus will be on battery power from battery 13.

Answer C incorrect - BC 131 can be powered from bus F and supply the bus, but is not normally aligned to supply DC bus 13, operator action is required.

Answer D incorrect - BC 132 is the normal supply and powered from bus H. A loss of 1H results in a loss of power to the normal battery charger and the bus will be supplied by battery 13.

Technical Reference(s): OIM J-1-2, OP J-9:II, page 2 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 42 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7115 - Describe the operation of the DC Power System Question Source: Bank # X A-0051 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 10 CFR Part 55 Content: Components, capacity, and functions of emergency systems.

Comments: difficulty taken from bank question.

23 January, 2009 - Changed to match bank question. Enhanced justification, added OP J-9 as a reference to show the normal and backup supplies.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 43 Question 22 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 063 A1.01 Importance Rating 2.5 K/A: 063 A1.01 - Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including: Battery capacity as it is affected by discharge rate Proposed Question:

During a loss of all AC, what is the design lifetime of a vital 125 VDC battery during a design basis accident coincident with a loss of battery charger?

A. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> D. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Proposed Answer: A. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Explanation:

Answer A correct - design is to provide necessary DC loads from the vital batteries for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during a design basis accident coincident with loss of battery charger.

Answer B incorrect - time is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Answer C incorrect - time is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Answer D incorrect - time is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Technical Reference(s): STG J-9, page 1-8 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7115 Describe the operation of the DC Power System 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 44 Question Source: Bank # X R-1060 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam DCPP 12/2007 (#49)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 10 CFR Part 55 Content: Components, capacity, and functions of emergency systems.

Comments:

PRA Insight, -

CDF Initiator - loss of offsite power Short term RRW - Short term Vital DC power 23 January, 2009 - changed answers to remove 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> as an option. It could be argued, that 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is also a correct answer.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 45 Question 23 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 064 K3.01 Importance Rating 3.8 K/A: 064 K3.01 - Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following: Systems controlled by automatic loader Proposed Question:

GIVEN:

• The plant trips automatically due to SI actuation

• Offsite power is available

• One Diesel Generator fails to start The load sequencer that is associated with the failed diesel generator A. immediately starts all Transfer to SU (SI) loads.

B. sequentially starts all Transfer to SU (SI) loads.

C. immediately starts all Transfer to SU (no SI) loads.

D. sequentially starts all Transfer to SU (no SI) loads.

Proposed Answer: B. sequentially starts all Transfer to SU (SI) loads.

Explanation:

Answer A incorrect - a transfer to startup sequence is initiated when SI actuates. There are transfers that do not result in sequentially loading the bus (12 kv fast transfer),

however, a transfer to startup sequentially loads the bus.

Answer B correct - the transfer to startup and SI results in sequential loading time. An SI signal causes the transfer to Startup to load ECCS equipment. It is not affected diesel start or non-start.

Answer C incorrect - a diesel start signal is required for a transfer to diesel but not for a transfer to startup. The diesel start is a normal, expected response to an SI and it could be 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 46 assumed to be a required action that must occur prior to starting ECCS loads.

Additionally, the loading is sequential, not immediate.

Answer D incorrect - a diesel start signal is required for a transfer to diesel but not for a transfer to startup. The diesel start is a normal, expected response to an SI and it could be assumed to be a required action that must occur prior to starting ECCS loads.

Technical Reference(s): OIM J-5-1c & d and J-6-1 Proposed references to be provided to applicants during examination: NONE Learning Objective: 37857 Describe the operation of the Electric Power Transfer System Question Source: Bank # X P-0553 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Rev 1 - replaced question that did not match KA.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 47 Question 24 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 073 K5.01 Importance Rating 2.5 K/A: 073 K5.01 - Knowledge of the operational implications as they apply to concepts as they apply to the Process Radiation Monitoring (PRM) system: Radiation theory, including sources, types, units, and effects Proposed Question:

Which of the following explains why the steam line radiation monitors, RM-71, 72, 73 and 74 may increase during a large break LOCA?

A. The monitors are responding to containment shine.

B. Radiation from containment cause the radiation monitors to fail high.

C. Particulate from the RCS enters the secondary through the U-tubes.

D. Elevated temperatures in the area cause the radiation monitors to fail high.

Proposed Answer: A. The monitors are responding to containment shine.

Explanation:

Answer A correct - High radiation levels in Containment due to clad or fuel failure can increase process monitor readings outside Containment due to increased background levels. Normally the Containment wall will provide 105 reduction in radiation levels, however, RM-71 to 74 can be exposed to Containment radiation by streaming through the main steam lines causing an increase in ALL steam line radiation monitors. This should not be interpreted as a S/G tube rupture if there are other indications of LOCA or fuel damage that can explain the high level reading in all steam lines.

Answer B incorrect - the monitors have not failed, they are responding to containment radiation.

Answer C incorrect - a tube rupture has not occurred.

Answer D incorrect - the monitors have not failed.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 48 Technical Reference(s): G4A - Radiation Monitoring page 3-15 Proposed references to be provided to applicants during examination: NONE Learning Objective: 8485 Explain the conditions that effect Radiation Monitoring system radiation monitor indications Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 9 55.43 10 CFR Part 55 Content: Shielding, isolation, and containment design features, including access limitations.

Comments: not a rev, added high to distractors B and D to better match the setup of the question.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 49 Question 25 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 076 G2.4.4 Importance Rating 4.5 K/A: 076 G2.4.4 - Service Water: 2.4.4 - Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Proposed Question:

Which of the following plant situations are addressed in OP AP-10, Loss of Auxiliary Saltwater?

a. Failure of both ASW pumps on same unit

b. Loss of suction to ASW pump in service

c. ASW room high temperature

d. Fouling of a CCW heat exchanger

e. Operation of CCW heat exchangers in parallel A. a, b, e B. a, b, d C. b, c, d D. c, d, e Proposed Answer: B. a, b, d Explanation:

Procedure scope:

• Failure of both ASW pumps on the same unit. (a)

• Loss of suction to the pump in service. (c)

• A rupture in the system piping or fouling of a CCW heat exchanger. (d)

• Extensive equipment damage at the intake due to tsunami or other causes.

Answer A incorrect - e (heat exchangers in parallel) covered by CCW procedures Answer B correct - all three are covered by AP-11 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 50 Answer C incorrect - c (ASW high temperature) covered by AR PK01-03 Answer D incorrect - c (ASW high temperature) and e (heat exchangers in parallel) covered by other procedures or AR PKs Technical Reference(s): OP AP-10 scope Proposed references to be provided to applicants during examination: NONE Learning Objective: 3476 - Explain the general purpose/function of abnormal operating procedures Question Source: Bank #

Modified Bank # X P-45940 New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 10 CFR Part 55 Content: Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments: At DCPP Auxiliary Salt Water is the Service Water equivalent.

23 January, 2009 - Changed D to remove b from all answers 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 51 Question 26 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 008 K2.02 Importance Rating 3.0 K/A: 008 K2.02 - Knowledge of bus power supplies to the following: Power supply CCW pump including emergency backup Proposed Question:

GIVEN:

• A loss of offsite power has occurred

• Unit 2 Diesel Emergency Generators are supplying their respective vital AC buses Which of the following Diesel Emergency Generators are supplying power to CCW pumps 22 and 23?

A. CCW pump 22 - DEG 21 CCW pump 23 - DEG 22 B. CCW pump 22 - DEG 22 CCW pump 23 - DEG 23 C. CCW pump 22 - DEG 22 CCW pump 23 - DEG 21 D. CCW pump 22 - DEG 23 CCW pump 23 - DEG 22 Proposed Answer: A. CCW pump 22 - DEG 21 CCW pump 23 - DEG 22 Explanation:

Answer A correct - DEG 21 supplies bus G the power supply for pump 22. DEG 22 supplies bus H, the power supply for 23 Answer B incorrect - DEG 22 supplies pump 22.

Answer C incorrect - power supplies are reversed. (this would be correct for U1) 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 52 Answer D incorrect - DEG 23 supplies bus F which would supply CCW pump 21.

Technical Reference(s): OIM J-1-1 Proposed references to be provided to applicants during examination: NONE Learning Objective: 8129 State the power supplies to CCW system components Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 10 CFR Part 55 Content: Components, capacity, and functions of emergency systems.

Comments: Unit difference 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 53 Question 27 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 078 K4.03 Importance Rating 3.1 K/A: 078 K4.03 - Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following: Securing of S(I)AS upon loss of cooling water Proposed Question:

Instrument Air Compressor 0-5 is running.

The compressor trips on high oil temperature.

Which of the following could have caused the oil temperature to increase to the trip setpoint?

A. Loss of SCW to the compressor.

B. Failure of the unloader to stop the compressor when it was unloaded.

C. A failure of 0-6 to start on decreasing air pressure.

D. SCW temperature control valve, TCV-2 fails closed.

Proposed Answer: A. Loss of SCW to the compressor.

Explanation:

Answer A correct - unlike compressors 01 through 04, 05 and 06 are not air cooled but cooled by SCW. A loss of SCW will cause loss of cooling to the oil cooler, which will cause oil temperature to increase.

Answer B incorrect - the compressors are designed to run for an extended period of time, assuming cooling is available.

Answer C incorrect - either air compressor can handle load.

Answer D incorrect - SCW TCV-2 closing would cause more flow through the heat exchanger, and SCW temperature would decrease.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 54 Technical Reference(s): STG K1, pages 2.1-18 (figure CAS-44C) and 2.1-23, figure CAS-14 Proposed references to be provided to applicants during examination: NONE Learning Objective: 37565 - Analyze automatic features and interlocks associated with the Compressed Air System Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments: The system is Instrument Air (IAS) design and interlocks, while the stated KA is SAS, this appears to be a typo. There is no connection between the two parts. The statement probably should be Securing of IAS upon loss of cooling water. This is the rationale for the question.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 55 Question 28 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 103.K4.01 Importance Rating 3.0 K/A: 103 K4.01 - Knowledge of containment system design feature(s) and/or interlock(s) which provide for the following: Vacuum breaker protection Proposed Question:

Which of the following describes the operation of Containment Vacuum Relief Isolation valve, FCV-664?

A. Manually opened from the control room; automatically closes on an automatic Phase A.

B. Manually opened from the control room; automatically closes on a Containment Ventilation Isolation.

C. Automatically opens on LOW Containment pressure; automatically closes on an automatic Phase A.

D. Automatically opens on LOW Containment pressure; automatically closes on a Containment Ventilation Isolation.

Proposed Answer: B. Manually opened from the control room; automatically closes on a Containment Ventilation Isolation.

Explanation:

Answer A incorrect - Manual Phase A causes a CVI and closes the valve.

Answer B correct - valve manually operated to open and automatically closes on CVI.

Answer C incorrect - valve does not open automatically.

Answer D incorrect - valve does not open automatically.

Technical Reference(s): STG H-4, pages 2-19, 2-20 OIM B-6-9a 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 56 Proposed references to be provided to applicants during examination: NONE Learning Objective: 5119 - Analyze automatic features and interlocks associated with the Containment Purge System Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 57 Question 29 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 001 K2.01 Importance Rating 3.5 K/A: 001 K2.01 - CRDS: Knowledge of bus power supplies to the following: One-line diagram of power supply to M/G sets.

Proposed Question:

Which buses supply power to the Rod Drive MGs?

A. Buses 1F and 1H B. Buses 1F and 1G C. Buses 11D and 11E D. Buses 13D and 13E Proposed Answer: D. Buses 13D and 13E Explanation:

Answer A incorrect - 1F and 1H are vital 480 VAC. Control Rod Drive MG sets have non-vital power, 13D and E Answer B incorrect - 1F and 1G are vital 480 VAC. Control Rod Drive MG sets have non-vital power, 13D and E Answer C incorrect - power supply is 480 VAC non-vital 13D/E not 480 VAC 11D/E Answer D correct - power supplies are 480 VAC non-vital 13D and 13E Technical Reference(s): OIM A-3-1 Proposed references to be provided to applicants during examination: NONE Learning Objective: 40753 - State the power supplies to Rod Control System components 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 58 Question Source: Bank # X S-5638 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 10 CFR Part 55 Content: Design, components, and function of reactivity control mechanisms and instrumentation.

Comments: added explanation that 1F, 1G and 1H are 480 VAC panels.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 59 Question 30 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 011 A4.02 Importance Rating 3.4 K/A: 011 A4.02 - PZR LCS: Ability to manually operate and/or monitor in the control room: Movement of the pressure control valve, using manual controller Proposed Question:

The plant is at full power.

Due to a malfunction, the operator has placed Pressurizer Level Hand Controller HC-459D in Manual.

The operator depresses the increase pushbutton and notes that the output meter increased.

The increase in the output meter indicates that ...

A. FCV-128 has opened further.

B. the magnitude of the control signal being applied to FCV-128 has increased.

C. HCV-142 has opened further.

D. the magnitude of the control signal being applied to HCV-142 has increased.

Proposed Answer: B. the magnitude of the control signal being applied to FCV-128 has increased.

Explanation:

Answer A incorrect - the output meter (0 to 100) does not indicate valve position but demand increase.

Answer B correct - Indicates the magnitude of the control signal that is actually being applied to the level control components.

Answer C incorrect - HCV-142 is not controlled by HC-459D, only by the operator.

Answer D incorrect - HCV-142 is not controlled by HC-459D, only by the operator.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 60 Technical Reference(s): STG A4A, pages 2.3-13 and 2.3-14 Proposed references to be provided to applicants during examination: NONE Learning Objective: 4560 - Describe the operation of the Pzr, Pzr Pressure and Level Control System Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 61 Question 31 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 014 G2.1.32 Importance Rating 3.8 K/A: 014 G2.1.32 - Rod Position Indication: Ability to explain and apply system limits and precautions.

Proposed Question:

A DRPI Urgent failure occurs for a Control Bank D rod. The crew is performing the actions of AR PK03-21, DRPI Failure/Rod Bottom.

The operator places the S106, Accuracy Mode Switch (VB2, back of DRPI Panel), in the A only position and the Urgent Failure clears.

Which of the following is a consequence of operating with only one DRPI channel operable?

A. Increased monitoring of DRPI is required.

B. A flux map is required within one hour.

C. DRPI and demand position indication will always differ by greater than 12 steps.

D. Demand position indication will show greater rod movement before DRPI indication changes.

Proposed Answer: D. Demand position indication will show greater rod movement before DRPI indication changes.

Explanation:

Answer A incorrect - no increased surveillance is required.

Answer B incorrect - rods are returned to AUTO.

Answer C incorrect - while in half accuracy, DRPI and demand will differ but will still agree within required accuracy.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 62 Answer D correct - due to be in half accuracy, one coil will not respond to rod motion while demand will respond immediately, DRPI will only change when the one remaining channel senses rod motion.

Technical Reference(s): AR PK03-21, caution prior to step 2.1.4 Proposed references to be provided to applicants during examination: NONE Learning Objective: 40702 - Discuss abnormal conditions associated with the DRPI system Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 10 CFR Part 55 Content: Design, components, and function of reactivity control mechanisms and instrumentation.

Comments: added indication after demand position. Note did not add to DRPI as that would be digital rod position indication indication (redundant). Changed difficulty based on NRC review.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 63 Question 32 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 028 K5.03 Importance Rating 2.9 K/A: 028 K5.03 - Knowledge of the operational implications of the following concepts as they apply to the HRPS: Sources of hydrogen within containment Proposed Question:

Which of the following could cause hydrogen to pose a challenge to Containment Integrity during a design basis LOCA?

A. One Hydrogen Recombiner is out of service.

B. Zirc-water reaction.

C. Only one train of Containment Spray actuates.

D. Radiolysis of the fuel cladding.

Proposed Answer: B. Zirc-water reaction.

Explanation:

Answer A incorrect - Each Recombiner has 100% capacity and can maintain hydrogen less than 4%.

Answer B correct - sources of hydrogen during a LOCA:

• Hydrogen inventory in the RCS, including the PZR steam space

• Zirc-water reaction

• Metal-water reaction

• Radiolysis of RCS water

• Radiolysis of Containment sump water Answer C incorrect - Containment spray is credited to reduce pressure but not hydrogen, although it will.

Answer D incorrect - radiolysis of water causes hydrogen to be a concern.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 64 Technical Reference(s): LMCDFRZ page 16 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7094 - Explain hydrogen formation, limits, and concerns in Containment.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 2 55.43 10 CFR Part 55 Content: General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 65 Question 33 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 033 K3.03 Importance Rating 2.8 K/A: 033 K3.02 - Knowledge of the effect that a loss or malfunction of the Spent Fuel Pool Cooling System will have on the following: Spent Fuel temperature Proposed Question:

GIVEN:

• A full core offload has just been completed

• Spent Fuel Pool pump 11 is maintaining temperature at 100°F

• Two CCW pumps are in operation A loss of startup power occurs. The diesel generators start and energize the Vital AC buses.

Which of the following will occur?

A. Spent Fuel Pool pump 11 will be automatically restarted and maintain pool temperature.

B. Spent Fuel Pool pump 12 will be automatically started and maintain pool temperature.

C. No Spent Fuel Pool pump will be automatically started and pool temperature will begin to increase.

D. No Spent Fuel Pool pump will be automatically started and there will be a loss of CCW cooling to the Spent Fuel Pool heat exchanger causing pool temperature to begin to increase.

Proposed Answer: C. No Spent Fuel Pool pump will be started and pool temperature will begin to increase.

Explanation: Normal power supply for the SFP pumps is vital power, bus G and H. a pump can have its power supply aligned to bus F. there are no pump auto starts.

Answer A incorrect - the SFP pumps, unlike the CCW pumps, must be manually restarted on a loss of power.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 66 Answer B incorrect - the SFP cooling system does not have preferred pumps, neither pump restarts when the diesels re-energize the vital buses.

Answer C correct - with the core offloaded and no pumps automatically restarting, pool temperature will increase.

Answer D incorrect - CCW is isolated to the heat exchanger by phase B (closing FCV-355), but not for a loss of offsite power.

Technical Reference(s): STG B7, Spent Fuel Pool System pages 2-8, 2-10. F2, CCW, page 2-39 Proposed references to be provided to applicants during examination: NONE Learning Objective: 4818 - Explain the operation of the Spent Fuel Pool Cooling System during abnormal operations Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

rev 1 - resampled KA. Original KA (K3.02) not a testable KA. Not a tie between area and ventilation radiation monitors and a loss of SFP cooling at DCPP.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 67 Question 34 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 034 K6.02 Importance Rating 2.6 K/A: 034 K6.02 - Knowledge of the effect of a loss or malfunction of the following will have on the Fuel Handling System: Radiation monitoring systems Proposed Question:

Spent fuel assemblies are being moved in preparation for an upcoming refueling. Exhaust fan E-4 is running.

Fuel Handling Building (FHB) radiation monitor, RM-58, loses power.

Which of the following automatic actions, if any, occurs?

A. No automatic actions occur. The system is already in Iodine Removal mode.

B. FHB exhaust fan E-4 remains running and BOTH E-5 and E-6 start.

C. FHB exhaust fan E-4 stops and EITHER E-5 or E-6 starts.

D. FHB exhaust fan E-4 stops and BOTH E-5 and E-6 start.

Proposed Answer: C. FHB exhaust fan E-4 stops and EITHER E-5 or E-6 starts.

Explanation:

Answer A incorrect - a loss of power does cause the auto actions (FHB evacuation alarm and FHB building vent shifts to iodine removal). Fan E-4 does not go through the charcoal filter and is therefore shutdown.

Answer B incorrect - only one exhaust (E-5 or E-6) fan starts, E-4 shuts down.

Answer C correct - FHB building vent starts one exhaust (E-5 or E-6) fan starts, E-4 shuts down.

Answer D incorrect - only one of the exhaust fans run.

Technical Reference(s): STG H7, 2 2-13 Proposed references to be provided to applicants during examination: NONE 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 68 Learning Objective: 8469 - Analyze automatic features and interlocks associated with the RMS Question Source: Bank # S-5201 Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 10 CFR Part 55 Content: Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Comments: editorial change to B to match the wording of C and D.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 69 Question 35 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 035 A3.01 Importance Rating 4.0 K/A: 035 A3.01 - Ability to monitor automatic operation of the S/G including: S/G water level control Proposed Question:

GIVEN:

• A Unit 1 startup is in progress

• Turbine power has been increased to 12%

Approximately, what will the program steam generator water level setpoint be at this point?

A. 33%

B. 40%

C. 44%

D. 65%

Proposed Answer: B. 40%

Explanation:

Answer A incorrect - this is no load level for unit 1 Answer B correct - true for unit 1 OSGs. 12% turbine load is 60% of the ramp of 0 to 20%. Steam generator level ramps 11% from 0 to 20% turbine load. 60% of 11 is 6.6 or a level of 39.6 (40%)

Answer C incorrect - this is the unit 1 program value above 20%.

Answer D incorrect - Unit 2 RSGs maintain a constant 65%, regardless of power level.

Technical Reference(s): OIM A-5-1 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 70 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7159 - Describe the operation of the steam generators Question Source: Bank # X P-0189 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 10 CFR Part 55 Content: Secondary coolant and auxiliary systems that affect the facility.

Comments: Unit difference Editorial change, C justification (in)correct.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 71 Question 36 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 015 A1.01 Importance Rating 3.5 K/A: 015 A1.01 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with the NIS controls including: NIS calibration by heat balance Proposed Question:

GIVEN:

• Unit 1 is at full power

• STP R-2B1, PPC Operator Heat Balance indicates calculated calorimetric power is higher than indicated power on the Power Range channels

• The operator is to perform NIS Power Range Channel Gain Adjustment What should be done to prevent tripping a rate trip bistable during the gain adjustment?

A. Slowly adjust the gain pot while decreasing the gain.

B. Hold the Rate Mode Switch in RESET while decreasing the gain.

C. Slowly adjust the gain pot while increasing the gain.

D. Hold the Rate Mode Switch in RESET while increasing the gain.

Proposed Answer: C. Slowly adjust the gain pot while increasing the gain.

Explanation:

Answer A incorrect - while the operator is instructed to turn the pot slowly, if calculated is higher than indicated, the correct action is to raise the gain, not lower it.

Answer B incorrect - The Rate Mode switch is used to reset a bistable that trips during gain adjustment (prior to moving on to the next channel) and gain must be increased.

Answer C correct - the gain pot is turned slowly to avoid tripping the rate trip bistable and the correct action is to raise gain.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 72 Answer D incorrect - gain is increased, but the Rate Mode switch is used to reset a bistable that trips during gain adjustment (prior to moving on to the next channel) .

Technical Reference(s): STP R-2B1, attachment 9.1 Proposed references to be provided to applicants during examination: NONE Learning Objective:

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 10 CFR Part 55 Content: Observe and safely control the operating behavior characteristics of the facility.

Comments:

Rev 1 - Randomly, using the KA program, resampled for a new KA. Original did not apply to DCPP. Using the program to randomly sample assured the balance of the sample plan would be maintained.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 73 Question 37 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 075 K4.01 Importance Rating 2.5 K/A: 075 K4.01 - Knowledge of circulating water system design feature(s) and interlock(s) which provide for the following: Heat Sink Proposed Question:

Unit 1 is in a normal full power lineup.

An automatic fast transfer of 12 kV busses to the start-up power source occurs.

How do the Circulating Water Pumps respond?

A. Both pumps remain running.

B. One pump trips; the pump selected for auto-reclose remains running.

C. Both pumps trip; the pump selected for auto-reclose will restart after a time delay.

D. One pump remains running; the pump selected for auto-reclose will restart after a time delay.

Proposed Answer: B. One pump trips; the pump selected for auto-reclose remains running.

Explanation:

Answer A incorrect - one pump trips due to electrical limitations.

Answer B correct - for fast transfer one pump trips, the other (selected) remains running.

Answer C incorrect - this is the response for slow transfer.

Answer D incorrect - this is a cross between fast and slow transfer.

Technical Reference(s): STG E-4, pages 2-7 & 8 Proposed references to be provided to applicants during examination: NONE 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 74 Learning Objective: 40862 - Analyze automatic features and interlocks associated with the CWS Question Source: Bank # X A-0547 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 75 Question 38 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 079 K1.01 Importance Rating 3.0 K/A: 079 K1.01 - Knowledge of the physical connections and/or cause effect relationships between the SAS and the following systems: IAS Proposed Question:

When the Service Air Compressors are required to supply Instrument Air (IA), the preferred method is to use the:

A. Manual valve which connects to IA upstream of the IA afterfilters.

B. Manual valve which connects to IA downstream of the IA afterfilters.

C. Automatic valve which connects to IA upstream of the IA afterfilters.

D. Automatic valve which connects to IA downstream of the IA afterfilters.

Proposed Answer: A. Manual valve which connects to IA upstream of the IA afterfilters.

Explanation:

OP K-1:I states:

2.5 A normally isolated cross tie line between the Service Air and the Instrument Air Systems ("4568" Tie) can be used to supply Instrument Air in the event of a failure of the instrument air compressors. To maintain instrument air purity, the backfeed air enters the Instrument Air System upstream of the instrument air afterfilters. On failure of the Service Air compressors, the Instrument Air System, by means of a separate normally isolated cross tie line ("PCV 114" Tie), can supply the Service Air header to maintain continuity of control air at the Intake Structure, motive air for resin transfers in the condensate polishers, CVCS and SGBD demins, and FHB movable wall inflatable seals.

Answer A correct - manual valve upstream of the filters.

Answer B incorrect - connection is upstream of the filters.

Answer C incorrect - not an automatic cross tie (automatic, service air refusal valve used to go from IA to SA) 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 76 Answer D incorrect - not an automatic cross tie, connection is upstream (automatic, service air refusal valve used to go from IA to SA).

Technical Reference(s): OP K1:I, step 2.5 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7227 - Describe the basic flow path of the Compressed Air System Question Source: Bank # X S-1234 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments: afterfilters one word.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 77 Question 39 Examination Outline Cross-

Reference:

Level RO SRO Tier # 11 Group # 1 K/A # APE 008 AK1.02 Importance Rating 3.1 K/A: APE 008 AK1.02 - Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: Change in leak rate with change in pressure Proposed Question:

GIVEN:

• A Pressurizer PORV fails open and cannot be isolated

• The plant trips and SI actuates

• All ECCS equipment operates as designed

• 30 minutes after the reactor trip the crew enters E-1.2, Post LOCA Cooldown and Depressurization Which of the following describes the expected plant conditions as the crew enters E-1.2?

A. Break flow is unchanged from its original value; Pressurizer level is off-scale high.

B. Break flow has decreased from its original value; Pressurizer level on scale and decreasing.

C. Break flow is unchanged from its original value; Pressurizer level on scale and decreasing.

D. Break flow has decreased from its original value; Pressurizer level is off-scale high.

Proposed Answer: D. Break flow has decreased from its original value; Pressurizer level is off-scale high.

Explanation:

Answer A incorrect - RCS pressure will be less than NOP as a result break flow will be reduced.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 78 Answer B incorrect - Pressurizer level will be off scale high (even if still on scale, it would be increasing due to increased SI flow and the open PORV).

Answer C incorrect - Break flow will be reduced.

Answer D correct - lower RCS pressure will result in lower break flow and due to vapor space break, level will be off scale high.

Technical Reference(s): LMCD-FRC page 17 and page 37 Proposed references to be provided to applicants during examination: NONE Learning Objective: 41697 - Describe the plant response to a loss of reactor coolant including:

• Vapor Space LOCAs Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 10 CFR Part 55 Content: Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 79 Question 40 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # EPE 009 EA2.12 Importance Rating 2.8 K/A: EPE 009 EA2.12 - Ability to determine or interpret the following as they apply to a small break LOCA: Charging pump ammeter Proposed Question:

GIVEN:

• A small LOCA has occurred thru Pressurizer PORV PCV-474

• RCS pressure is 1200 psig and slowly decreasing

• RVLIS level is 85% and slowly decreasing

• Radiation levels in the Auxiliary building are normal Charging Injection Flow meter, FI-917 has failed low.

Which of the following could be used by the operator to positively confirm the ECCS CCPs are injecting into the RCS through ECCS charging injection lines?

A. ECCS CCPs red lights lit.

B. ECCS CCPs ammeters indicating high in the normal operating band.

C. Increasing pressurizer level.

D. Charging flow indicated on charging header flow indicator, FI-128.

Proposed Answer: B. ECCS CCPs ammeters indicating high in the normal operating band.

Explanation:

Answer A incorrect light indication only verifies the breaker is closed.

Answer B correct - when injecting, amps will increase to approximately 90 amps (max amps). Normal range is approximately 50 to 90 amps, at this pressure, the pump is close to runout, amps will be high.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 80 Answer C incorrect - at this pressure, SI Pumps will be injecting as well. Can not say with certainty that the level increase is due to charging injection (or could be void growth/formation).

Answer D incorrect - FI-128 does not see flow when injecting (does indicate seal injection).

Technical Reference(s): E-0, appendix E step 8.

Proposed references to be provided to applicants during examination: NONE Learning Objective: 40448 - Describe the controls, indications, and alarms associated with the CVCS Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 10 CFR Part 55 Content: Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 81 Question 41 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # EPE 011 EA1.03 Importance Rating 4.0 K/A: EPE 011 EA1.03 - Ability to operate and monitor the following as they apply to a Large Break LOCA: Securing of RCPs Proposed Question:

GIVEN:

• A LOCA has occurred

• Containment pressure peaks at 20 psig

• RCS pressure is 900 psig

• No ECCS CCPs are running

• All other ECCS equipment is operating as designed

• All Diesel Emergency Generators are running unloaded

• The crew has just completed the immediate actions of E-0, Reactor Trip or Safety Injection What is action, if any, should be taken for the Reactor Coolant Pumps (RCPs)?

A. No action, the RCPs should remain running to maintain core cooling.

B. The RCPs should be tripped to prevent overheating due a loss of seal cooling.

C. The RCPs should be tripped to prevent possible severe core uncovery if they trip later in the event.

D. No action, the RCPs have tripped.

Proposed Answer: C. The RCPs should be tripped to prevent possible severe core uncovery if they trip later in the event.

Explanation:

Answer A incorrect - Foldout page requires tripping RCPs if less than 1300 psig and a high head pump running, an SI pump satisfies this requirement.

Answer B incorrect - Phase B isolates cooling, setpoint is 22 psig.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 82 Answer C correct - for a LOCA, because RCP operation cannot be assured for the duration, they are tripped to prevent possible severe core uncovery if they trip later.

Answer D incorrect - offsite power is available (diesels are running unloaded).

Technical Reference(s): Westinghouse Generic Issues - RCP Trip/Restart page 10 Proposed references to be provided to applicants during examination: NONE Learning Objective: 5442 - Explain the conditions affecting RCP trip criteria Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 10 CFR Part 55 Content: Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 83 Question 42 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # EPE055 G2.1.31 Importance Rating 3.6 K/A: APE015/017 G2.2.38 - Station Blackout: Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

Proposed Question:

GIVEN:

• A loss of all vital AC power has occurred on Unit 1

• The crew is performing the actions of ECA-0.0, Loss of All Vital AC Power

• Safeguards loads have been isolated from the deenergized Vital Buses What should be the control switch indications on Vertical Boards 1, 2 and 3 for the 4 kV safeguards loads?

A. No lights for any of the safeguards loads because the breakers are open.

B. No lights for any of the safeguards loads because the DC Control Power switches are open.

C. White lights only for all of the safeguards loads because the breakers are open.

D. White and green lights for all of the safeguards loads because the breakers are open.

Proposed Answer: B. No lights for any of the safeguards loads because the DC Control Power switches are open.

Explanation:

Answer A incorrect - the breakers are open, but there is a loss of indication due to removing control power to allow for controlled loading if a diesel is restored. For 480 VAC loads, such as the CFCUs, the lights are out when the breaker is opened.

Answer B correct - the breakers are opened and control power removed. This will result in not indication on the vertical boards for the loads.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 84 Answer C incorrect - white lights indicate power is available, as would be case if the breaker was open (but the green light would be on also).

Answer D incorrect - the control power is removed for controlled loading, there will be no indication. Both lights would be on if only the breaker is open.

Technical Reference(s): ECA-0.0 step 8 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7920 Explain basis of emergency procedure steps Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments: replaced original KA, APE015/017 G2.2.38. there are no applicable Tech Spec/License requirements (<1 hour) dealing with loss of RCPs for ROs.

Rev 1 - replaced question to better match the KA.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 85 Question 43 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 022 AK1.03 Importance Rating 3.0 K/A: APE 022 AK1.03 - Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: Relationship between charging flow and PZR level Proposed Question:

The plant is at full power, in a normal steady state lineup.

Which of the following failures, without operator action, would cause pressurizer level to decrease to zero?

A. VCT level channel LT-114 fails high.

B. VCT level channel LT-114 fails low.

C. VCT level channel LT-112 fails high.

D. VCT level channel LT-112 fails low.

Proposed Answer: C. VCT level channel LT-112 fails high.

Explanation:

Answer A incorrect - full open signal to LCV-112A. VCT level decreases to 14%, auto makeup from LT-112, makeup is greater than letdown.

Answer B incorrect - no effect, if level on LT-112 decreases to 5%, auto swapover to RWST.

Answer C correct - auto makeup is lost. VCT level will decrease. No auto swapover, at 5% will occur (requires both channels). VCT empties, charging flow decreases.

Pressurizer level decreases and letdown isolates. However, pressurizer level continues to decrease due to seal leakoff. The pressurizer will empty.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 86 Answer D incorrect - VCT makeup initiates, and the VCT fills and diverts. Charging and letdown unaffected.

Technical Reference(s): OIM B-1-4. OP AP-19 appendix A Proposed references to be provided to applicants during examination: NONE Learning Objective: 40581 - Discuss abnormal conditions associated with the reactor makeup control system Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 10 CFR Part 55 Content: Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 87 Question 44 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 025 AA2.07 Importance Rating 3.4 K/A: APE 025 AA2.07 - Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: Pump cavitation Proposed Question:

The plant is in MODE 5, mid-loop.

Which of the following describes the expected indications/plant conditions that would indicate that either vortexing or cavitation is occurring?

Vortexing Cavitation A. Larger amp swings, (8 to 10 amps), Smaller amp swings, (2 to 3 amps),

no discernible increase in noise level. increase in pump noise level.

B. Larger amp swings, (8 to 10 amps), Smaller amp swings, (2 to 3 amps),

increase in pump noise level. no discernible increase in noise level.

C. Smaller amp swings, (2 to 3 amps), Larger amp swings, (8 to 10 amps),

increase in pump noise level. no discernible increase in noise level.

D. Smaller amp swings, (2 to 3 amps), Larger amp swings, (8 to 10 amps),

no discernible increase in noise level. increase in pump noise level.

Proposed Answer: D. Smaller amp swings, (2 to 3 amps), no discernible increase in noise level. Larger amp swings, (8 to 10 amps), increase in pump noise level.

Explanation:

Per OP A-2:III

• RHR pump cavitation can be inferred from either rapid swings of 8 to 10 amps (peak to peak), and/or increased noise level at the pump.

• Cavitation can be differentiated from vortexing in that vortexing will have smaller oscillations of 2 to 3 amps (peak to peak) and no discernible increase in noise level at the pump.

Answer A incorrect - cavitation has larger amp swings.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 88 Answer B incorrect - indications reversed.

Answer C incorrect - noise level increased for cavitation.

Answer D correct - increased noise/large amp swings - cavitation. Smaller amp swings/no discernible noise level increase - vortexing.

Technical Reference(s): OP A-2:III page 10 step 5.12 Proposed references to be provided to applicants during examination: NONE Learning Objective: 5728 - Explain the conditions that affect RHR pump cavitation or vortexing Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 2 55.43 10 CFR Part 55 Content: General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 89 Question 45 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 027 AK2.03 Importance Rating 2.6 K/A: APE 027 AK2.03 - Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners Proposed Question:

The plant is at full power.

The controlling Pressurizer pressure channel begins to fail low.

Which of the following describes the initial expected response of the Master Pressure Controller, HC-455K?

A. Setpoint decreases, output decreases.

B. Setpoint decreases, output increases.

C. Setpoint unchanged, output decreases.

D. Setpoint unchanged, output increases.

Proposed Answer: C. Setpoint unchanged, output decreases.

Explanation:

Answer A incorrect - setpoint is set by the potentiometer, initially not affected by channel failure.

Answer B incorrect - setpoint is set by the potentiometer, initially not affected by channel failure.

Answer C correct - setpoint is initially unaffected. Decreasing channel input causes Pact to be less than Pref and the controller reacts to lowering pressure by lowering output to energize heaters.

Answer D incorrect - output decreases, the controller is not reverse acting.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 90 Technical Reference(s): TG A4A, page 2.2-7 Proposed references to be provided to applicants during examination: NONE Learning Objective: 36926 - Discuss abnormal conditions associated with the Pzr, Pzr Pressure and Level Control System Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 91 Question 46 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # EPE 029 EK2.06 Importance Rating 2.9 K/A: EPE 029 EK2.06 - Knowledge of the interrelations between the following and an ATWS: Breakers, relays, and disconnects Proposed Question:

The plant fails to automatically trip when required.

Which of the following failures would prevent a successful reactor trip and require the crew to enter FR-S.1, Response to Nuclear Power Generation/ATWS from E-0, Reactor Trip or Safety Injection?

A. One reactor trip breaker remains closed and one Rod Drive MG set remains energized.

B. Both reactor trip breakers remain closed and one Rod Drive MG set remains energized.

C. One reactor trip breaker remains closed and both Rod Drive MG sets remain energized.

D. Both reactor trip breakers remain closed and both Rod Drive MG sets are de-energized.

Proposed Answer: B. Both reactor trip breakers remain closed and one Rod Drive MG set remains energized.

Explanation:

Answer A incorrect - reactor trip breakers are in series, either will cause a trip.

Answer B correct - both trip breakers must remain closed, however, only one MG set is required to keep power to the rods.

Answer C incorrect - either breaker open will cause a trip.

Answer D incorrect - Both Rod Drive MG sets would be de-energized per E-0, step 1 RNO, causing all rods to insert.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 92 Technical Reference(s): OIM A-3-1 Proposed references to be provided to applicants during examination: NONE Learning Objective: 40752 - Describe the basic flow path of the Rod Control System Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 10 CFR Part 55 Content: Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 93 Question 47 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # EPE 038 EK1.01 Importance Rating 3.1 K/A: EPE 038 EK1.01 - Knowledge of the operational implications of the following concepts as they apply to the SGTR: Use of steam tables.

Proposed Question:

GIVEN:

• A steam generator tube rupture has occurred

• The crew has just completed an RCS cooldown to the target temperature of 513°F

• RCS pressure is 1300 psig

• Ruptured steam generator pressure is 1040 psig Approximately how much RCS subcooling currently exists?

A. 34°F B. 38°F C. 62°F D. 66°F Proposed Answer: D. 66°F Explanation:

Answer A incorrect - this is the amount of subcooling if the ruptured generator is used and 15 psi is subtracted to calculate the saturation pressure.

Answer B incorrect - this is the amount of subcooling if the ruptured generator is used and 15 psi is added to calculate the saturation pressure.

Answer C incorrect - this is the amount of subcooling if RCS pressure is used and 15 psi is subtracted to calculate the saturation pressure.

Answer D correct - Tsat for 1300 psig + 15 = 1315 psia = 579°F or 66°F of subcooling 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 94 Technical Reference(s): Steam tables Proposed references to be provided to applicants during examination: Steam tables Learning Objective: 65686 - Explain how steam tables are used in the Control Room.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 10 CFR Part 55 Content: Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 95 Question 48 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 040 AK2.02 Importance Rating 2.6 K/A: APE 040 AK2.02 - Knowledge of the interrelations between the Steam Line Rupture and the following: Sensors and detectors Proposed Question:

The plant is at normal operating temperature and pressure.

Which of the following would satisify the MINIMUM coincidence necessary to cause a low steam line pressure Safety Injection actuation? (assume no rate compensation)

A. 2/3 pressure indicators on steam line 11 sense steam pressure of 630 psig.

B. 2/3 pressure indicators on steam lines 11 and 12 sense steam pressure of 630 psig.

C. 2/3 pressure indicators on steam line 11 sense steam pressure of 590 psig.

D. 2/3 pressure indicators on steam lines 11 and 12 sense steam pressure of 590 psig.

Proposed Answer: C. 2/3 pressure indicators on steam line 11 sense steam pressure of 590 psig.

Explanation: The purpose of the Low Steam line Pressure Safety Injection (SI) Actuation signal is to initiate the automatic starting of boron injection and decay heat removal systems and to provide protection against steam line break accidents.

Answer A incorrect - 630 psig is an alarm setpoint of impending SI if RCS pressure just below or above P-11 Answer B incorrect - 630 psig is an alarm setpoint of impending SI if RCS pressure just below or above P-11 Answer C correct - setpoint is 600 psig, lead-lag compensated on 1/4 steam generators.

Answer D incorrect - setpoint is 600 psig on 1 of 4 steam generators. Only need one SG low pressure necessary to initiate SI.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 96 Technical Reference(s): STG B6A, section 2.2, Low Steamline Pressure SI Actuation Proposed references to be provided to applicants during examination: NONE Learning Objective: 37048 - Analyze automatic features and interlocks associated with the RPS Question Source: Bank # S-2762 Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Minimum to all caps 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 97 Question 49 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 054 AA1.01 Importance Rating 4.5 K/A: APE 054 AA1.01 - Ability to operate and/or monitor the following as they apply to the Loss of Main Feedwater (MFW): AFW controls, including the use of alternate AFW sources Proposed Question:

The crew is performing the actions of FR-H.1, Response to Loss of Secondary Heat Sink.

Which of the following describes the setpoint for switching AFW suction to an alternate source and the preferred alternate source?

A. 10% CST level; FWST.

B. 33% CST level; FWST.

C. 10% CST level; Raw Water Reservoir.

D. 33% CST level; Raw Water Reservoir.

Proposed Answer: A. 10% CST level; FWST.

Explanation:

Answer A correct - setpoint for switchover is 10%, preferred source IAW OP D-1:V is the FWST.

Answer B incorrect - setpoint is 10%. 33% is the (RWST) cold leg recirc switchover setpoint.

Answer C incorrect - RWR is the second preferred source.

Answer D incorrect - setpoint is 10%, RWR is the second preferred source.

Technical Reference(s): FR-H.1 FOP and OP D-1:V 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 98 Proposed references to be provided to applicants during examination: NONE Learning Objective: 3850 - Describe system interrelationships between the AFW system and other plant systems Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 10 CFR Part 55 Content: Secondary coolant and auxiliary systems that affect the facility.

Comments: shortened answers 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 99 Question 50 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 056 AA2.19 Importance Rating 4.0 K/A: APE 056 AA2.19 - Ability to determine and interpret the following as they apply to the Loss of Offsite Power: T-cold and T-hot indicators (wide range)

Proposed Question:

GIVEN:

• A reactor trip has occurred due to a loss of offsite power

• Diesel Emergency Generators are supplying the Vital AC buses The crew is checking RCS temperature stable or trending to 547°F at step 1 of E-0.1, Reactor Trip Response.

What indication will be used to perform this check?

A. Average of narrow range Thot and Tcold.

B. Average of wide range Thot and Tcold.

C. Core exit thermocouples.

D. Wide range Tcold.

Proposed Answer: D. Wide range Tcold.

Explanation:

Answer A incorrect - this is used if an RCP is running.

Answer B incorrect - WR Tave is not used.

Answer C incorrect - CETCs are used to monitor for core cooling (or lack of core cooling).

Answer D correct - with no RCPs running (loss of offsite power), wide range Tcold used.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 100 Technical Reference(s): E-0.1 step 1 Proposed references to be provided to applicants during examination: NONE Learning Objective: 40457 - Discuss abnormal conditions associated with the RCS Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2 55.43 10 CFR Part 55 Content: General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 101 Question 51 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 057 G2.4.20 Importance Rating 3.8 K/A: APE 057 G2.4.20 - Loss of vital AC electrical instrument bus: Knowledge of operational implications of EOP warnings, cautions, and notes.

Proposed Question:

The plant is at full power, in a normal full power lineup.

GIVEN:

• PY-12 loses power

• As a result, the MANUAL power supply to the controller for the 12 Motor Driven AFW (MDAFW) pump LCVs, (LCV110 and LCV111) is lost

• 5 minutes later, PY-12 is re-energized Without operator action, what is the status of the controller and AFW pump runout protection once power is restored?

A. The controller remains in AUTO and runout protection was not affected.

B. The controller shifts to AUTO and runout protection is restored.

C. The controller shifts to MANUAL and runout protection is restored.

D. The controller shifts to MANUAL and runout protection for the AFW pump is lost.

Proposed Answer: D. The controller shifts to MANUAL and runout protection for the AFW pump is lost.

Explanation: step 2 of OP AP-4 instructs the operator to refer to OP O-2 for Hagan controller response. The precautions and limitations for O-2, specifically 5.4 addresses the operation of the AFW controllers and its impact on AFW pump runout protection.

Answer A incorrect - loss of manual causes the controller to fail to auto hold.

Answer B incorrect - following power restoration, the controller goes to MANUAL.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 102 Answer C incorrect - the controller is in MANUAL and as a result, run out protection is not available.

Answer D correct - the controller is in MANUAL and as a result, run out protection is not available.

Technical Reference(s): OP AP-4, Loss of Vital or Nonvital Instrument AC, OP O-2, precautions and limitations and attachment 1 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7970 - Identify failure modes associated with Hagan controllers Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

23 January, 2009 - Enhanced the explanation to show the connection between the question and the KA.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 103 Question 52 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 058 AA1.02 Importance Rating 3.1 K/A: APE 058 AA1.02 - AA1.02 - Ability to operate and/or monitor the following as they apply to the Loss of DC Power: Static inverter dc input breaker, frequency meter, ac output breaker, and ground fault detector Proposed Question:

The plant is at full power, in a normal full power lineup.

Which of the following occurs if the vital UPS DC input breaker opens?

A. VITAL UPS TROUBLE alarm with no change in Inverter input voltage.

B. VITAL UPS TROUBLE alarm and a decrease in Inverter input voltage.

C. VITAL UPS FAILURE alarm with no change in Inverter input voltage.

D. VITAL UPS FAILURE alarm and a decrease in Inverter input voltage.

Proposed Answer: C. VITAL UPS FAILURE alarm with no change in Inverter input voltage.

Explanation:

Answer A incorrect - loss of DC causes the failure alarm.

Answer B incorrect - loss of DC causes the failure alarm, additionally, there is no change in dc amps or inverter voltage the AC is the normal source.

Answer C correct - AC is the normal source, the DC at a slightly lower voltage is available if the AC is lost. Loss of DC causes UPS failure alarm.

Answer D incorrect - no change in voltage or amps.

Technical Reference(s): OIM J-1-3 Proposed references to be provided to applicants during examination: NONE 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 104 Learning Objective: 3332 - Discuss abnormal conditions associated with the Instrument AC System Question Source: Bank # X S-2874 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

23 January, 2009 - changed question from modified to bank. No significant difference from the bank question.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 105 Question 53 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 062 AK3.02 Importance Rating 3.6 K/A: APE 062 AK3.02 - Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: The automatic actions (alignments) within the nuclear service water resulting from the actuation of the ESFAS Proposed Question:

The plant is in a normal full power lineup.

Safety Injection actuates.

Which of the following describes the expected response of the ASW system?

A. ASW pumps in AUTO will start immediately to establish an ultimate heat sink for ECCS equipment. An ASW pump in MANUAL will have to be started by the operator.

B. ASW pumps in AUTO or MANUAL will start immediately to establish an ultimate heat sink for ECCS equipment.

C. ASW pumps in AUTO will start after a time delay to prevent excessive starting currents overloading the supply to the bus. An ASW pump in MANUAL will have to be started by the operator.

D. ASW pumps in AUTO or MANUAL will start after a time delay to prevent excessive starting currents overloading the supply to the bus.

Proposed Answer: D. ASW pumps in AUTO or MANUAL will start after a time delay to prevent excessive starting currents overloading the supply to the bus.

Explanation:

Answer A incorrect - the purpose of ASW is to provide an ultimate heat sink, however, the start of the pumps is delayed to prevent overloading the supply (DG or SU) to the bus.

The pumps start in auto or manual.

Answer B incorrect - the purpose of ASW is to provide an ultimate heat sink, however, the start of the pumps is delayed to prevent overloading the supply (DG or SU) to the bus.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 106 Answer C incorrect - The pumps start in auto or manual.

Answer D correct - the pumps start independent of switch position, auto or manual.

Additionally, auto load safeguards equipment on the 4 kV vital buses are started in a sequenced order to prevent excessive starting currents from overloading the supply to the bus (S/U or D/G)

Technical Reference(s): OIM J-6-1, STG J-15, 2.2-35 Proposed references to be provided to applicants during examination: NONE Learning Objective: 5365 - Analyze automatic features and interlocks associated with the ASW system Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments: At DCPP the equivalent to Nuclear Service Water is Auxiliary Salt Water (ASW) 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 107 Question 54 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 065 G2.4.47 Importance Rating 4.2 K/A: APE 065 G2.4.47 - Loss of Instrument Air - Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Proposed Question:

The plant is at full power. Instrument air pressure is 105 psig.

Instrument air pressure indication on PI-380 on VB4, begins to decrease.

Currently instrument air pressure is 78 psig and decreasing at a rate of approximately 3 psig/minute.

Which of the following should occur within the next 1 to 2 minutes?

A. Increasing charging flow.

B. Standby air compressors start.

C. Instrument air to containment isolates.

D. Main Feedwater Reg valves closing.

Proposed Answer: D. Main Feedwater Reg valves closing.

Explanation: based on the observed rate of Instrument Air pressure decrease on PI-380, the pressure should indicate approximately 72 to 75 psig.

Note before step 1 of AP-9 states FCV 584 may begin to close at 85 psig, and the Main Feed Reg Valves may begin to close at 75 psig.

Answer A incorrect - HCV-142 fails closed, charging will decrease, not increase.

Answer B incorrect - the compressors start at higher pressure. The compressors load at approximately 104 psig and all should be running and loaded by 93 psig.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 108 Answer C incorrect -FCV-584 (containment isolation valve) closes at approximately 85 psig.

Answer D correct - Main Feedwater Reg valves begin to close at approximately 75 psig.

Technical Reference(s): AP-9, note - page 2, OIM page K-1-2 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7209 - Discuss abnormal conditions associated with the Compressed Air System Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 77 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

23 January, 2009 - enhanced justification, added OIM page for air compressor start sequence.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 109 Question 55 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # E04 EK3.4 Importance Rating 3.6 K/A: E04 EK3.4 - Knowledge of the reasons for the following responses as they apply to the (LOCA Outside Containment) RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated Proposed Question:

A reactor trip and SI occurred from full power. The crew transitioned from E-0, Reactor Trip or Safety Injection to ECA-1.2, LOCA Outside Containment.

The crew is performing step1 of the procedure to VERIFY proper valve alignment prior to performing RHR isolation in subsequent steps (closing 8809A and B, RHR to Cold Legs).

When directed by the Shift Foreman to verify 8802A, SI to hot legs 1 and 2 closed, the operator observes the valve to be open.

What action should be taken by the operator?

A. Close the valve because it may be the source of the leak. When the valve is closed notify the Shift Foreman that the valve was open and is now closed.

B. Report the position to the Shift Foreman and take action to close the valve because it may be the source of the leak. Notify the Shift Foreman when the valve is closed.

C. Close the valve because 8809A cannot be closed with 8802A open. When the valve is closed notify the Shift Foreman that the valve was open and is now closed.

D. Report the position to the Shift Foreman and take action to close the valve because 8809A cannot be closed with 8802A open. Notify the Shift Foreman when the valve is closed.

Proposed Answer: B. Report the position to the Shift Foreman and take action to close the valve because it may be the source of the leak. Notify the Shift Foreman when the valve is closed.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 110 Explanation:

per OP1.DC10, verify means: Check an indication or status and if it does not match the desired condition, then take steps to make it match the desired condition (i.e., if it's not so, make it so.). According to OP1.DC10, the SFM is notified the valve is open (to allow for correction) then close the valve, once closed, the operator agains reports the valve is closed.

Answer B correct - per OP1.DC10, verify means: Check an indication or status and if it does not match the desired condition, then take steps to make it match the desired condition (i.e., if it's not so, make it so.)

the purpose of the step is to ensure that normally closed valves are closed This step instructs the operator to verify that all normally closed valves in low pressure lines and other plant specific lines that penetrate containment are closed. The valving connecting the RHR System to the RCS is of particular interest in this step since the RHR System is a low pressure system (600 psig) connected to the high pressure reactor coolant system (2500 psig). Therefore, a rupture or break outside containment is most probable to occur in the low pressure RHR System piping.

Answer C incorrect - According to OP1.DC10, the SFM is notified the valve is open (to allow for correction) then close the valve, once closed, the operator agains reports the valve is closed.

Additionally, 8809A/B, unlike other ECCS valves, do not have interlocks preventing operation.

Answer D incorrect - 8809A/B do not have interlocks preventing operation.

Technical Reference(s): ECA-1.2, step 1 background, OP1.DC10, 5.1.2.f.3.

STG B3, pages Proposed references to be provided to applicants during examination: NONE Learning Objective: 42461 Explain basis of emergency steps of ECA-1.2 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 111 Question 56 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # E05 EK3.2 Importance Rating 3.7 K/A: Knowledge of the reasons for the following responses as they apply to the (Loss of Secondary Heat Sink): Normal, abnormal and emergency operating procedures associated with Loss of Secondary Heat Sink Proposed Question:

Why are RCPs tripped during EOP FR-H.1, "Response to Loss of Secondary Heat Sink"?

A. To remove the RCPs as a heat input to the RCS.

B. To delay the pressure rise to the PORV setpoint by lowering core delta P.

C. To allow natural circulation to be established prior to losing the steam generators as a heat sink.

D. To prevent excessive inventory loss through open PORVs if bleed and feed is necessary.

Proposed Answer: A. To remove them as a heat input to the RCS.

Explanation:

Answer A correct - from the WOG background for FR-H.1, RCP operation results in heat addition to the RCS water. By tripping the RCPs, the effectiveness of the remaining water inventory in the SGs is extended, which extends the time at which the operator action to initiate bleed and feed must occur. This extension of time is additional time for the operator to restore feedwater flow to the SGs.

Answer B incorrect - RCP operation may hinder depressurization of the secondary but the reason is not to delay pressure rise to the PORV setpoint.

Answer C incorrect - establishing natural circulation will delay steam generator dryout.

Answer D incorrect - RCPs do cause increased inventory loss for small breaks and is the reason for tripping RCPs at less than 1300 psig, but not for a loss of secondary heat sink.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 112 Technical Reference(s): WOG FR-H.1 background, step 4 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7920 - Explain basis of emergency procedure steps Question Source: Bank # X P-1362 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 10 CFR Part 55 Content: Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

23 January, 2009 - modified distractor B.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 113 Question 57 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # E14 EK2.2 Importance Rating 3.4 K/A: E14 EK2.2 - Knowledge of the operational implications of the following concepts as they apply to the (High Containment Pressure): Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Proposed Question:

Step 6 in FR-Z.1, Response to High Containment Pressure, directs the operators to check if feedwater should be isolated to any steam generator.

Why is this step necessary?

A. To prevent the RCS from challenging the RCS Integrity Critical Safety Function.

B. To prevent a faulted steam generator from further pressurizing containment.

C. To limit RCS cooldown once there is indication of adequate heat sink.

D. To maintain RCP support parameters.

Proposed Answer: B. To prevent a faulted steam generator from further pressurizing containment.

Explanation:

Answer A incorrect - a cooldown will occur, but an RCS integrity challenge is not the concern of Z.1, overpressurizing containment is the concern.

Answer B correct - This step instructs the operator to check for a faulted SG. The operator should then isolate main feedwater and AFW to any faulted SG. Since the source of high containment pressure could result from a steamline break inside containment, this step may eliminate mass and energy releases to the containment.

Answer C incorrect - many procedure steps have the operator reduce AFW flow once there is sufficient steam generator level. In Z.1 the check is for faulted steam generators which could have been the cause for entry.

Answer D incorrect - reducing RCS pressure could cause low DP however, the RCPs are stopped in Z.1 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 114 Technical Reference(s): WOG FR-Z.1, bases step 6 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7920 - Explain basis of emergency procedure steps Question Source: Bank # P-1626 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 10 CFR Part 55 Content: Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments: replaced original KA, APE 003 AK1.17 - theory not site specific.

23 January, 2009 - changed to bank question, does not meet threshold for modified.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 115 Question 58 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # APE 032 AK3.01 Importance Rating 3.2 K/A: APE 032 AK3.01 - Knowledge of the reasons for the following responses as they apply to the Loss of Source Range Nuclear Instrumentation: Startup termination on source-range loss Proposed Question:

GIVEN:

• The plant is in MODE 3

• A reactor startup is in progress

• Shutdown Rods are withdrawn Which of the following would cause termination of the reactor startup?

A. Loss of Instrument Power to a Source Range channel because the P-6 permissive will not function.

B. Loss of Instrument Power to a Source Range channel because the reactor trip breakers will open.

C. Initial Source Range counts less than 50 cps because the reactor will go critical below the RIL.

D. Initial Source Range counts less than 50 cps because the reactor will not be critical with the Control Rods fully withdrawn.

Proposed Answer: B. Loss of Instrument Power to a Source Range channel because the reactor trip breakers will open.

Explanation:

Answer A incorrect - P-6 is operable, both IR are functioning.

Answer B correct - loss of control power will cause the high flux trip bistable to trip, causing a reactor trip (1/2 coincidence).

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 116 Answer C incorrect - while minimum counts should be greater than 1 cps, the amount of reactivity to be added is the same, regardless of where the startup begins, (still 5 to 7 doublings).

Answer D incorrect - while minimum counts should be greater than 1 cps, the amount of reactivity to be added is the same, regardless of where the startup begins, (still 5 to 7 doublings).

Technical Reference(s): OIM B-4-2 Proposed references to be provided to applicants during examination: NONE Learning Objective: 5992 - Discuss abnormal conditions associated with the NIS Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2 55.43 10 CFR Part 55 Content: General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 117 Question 59 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # APE 033 AA2.01 Importance Rating 3.0 K/A: APE 033 AA2.01 - Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Equivalency between source-range, intermediate-range, and power-range channel readings Proposed Question:

A plant startup is in progress.

Intermediate Range N35 is undercompensated.

How will the channels undercompensation affect its operation?

A. The channel will begin to come on scale lower in the Source Range and indicate higher at 100% power.

B. The channel will begin to come on scale lower in the Source Range but indicate correctly at 100% power.

C. The channel will begin to come on scale higher in the Source Range and indicate lower at 100% power.

D. The channel will begin to come on scale higher in the Source Range but indicate correctly at 100% power.

Proposed Answer: B. The channel will begin to come on scale lower in the Source Range but indicate correctly at 100% power.

Explanation:

Answer A incorrect - at power gamma has very little effect on IR indication.

Answer B correct - undercompensation will cause the IR to see gammas and indication to come on scale lower in the source range. At power, gamma will not affect indication.

Answer C incorrect - IR responds earlier.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 118 Answer D incorrect - IR responds earlier.

Technical Reference(s): OIM B-4-3 Proposed references to be provided to applicants during examination: NONE Learning Objective: 5992 - Discuss abnormal conditions associated with the NIS Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2 55.43 10 CFR Part 55 Content: General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 119 Question 60 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # APE 036 AK1.02 Importance Rating 3.4 K/A: APE 036 AK1.02 - Knowledge of the operational implications of the following concepts as they apply to Fuel Handling Incidents: SDM Proposed Question:

While taking logs during core loading the operator notices the following trends.

RCS Temp Boron N31 N32 Initial 62°F 2570 PPM 13 CPS 15 CPS Latest 76°F 2510 PPM 32 CPS 28 CPS The recommendation to the Shift Foreman is that core loading A. should be suspended because of the change in Source Range counts.

B. should be suspended because of the change in RCS temperature.

C. should be suspended because of the change in boron concentration.

D. may continue.

Proposed Answer: C. should be suspended because of the change in boron concentration.

Explanation: step 5.2.1 - The loading procedure will be suspended, pending evaluation by the Refueling SRO and reactor engineer under the following circumstances:

a. If there occurs on any one responding nuclear channel an unexpected increase in count rate by a factor of three (3).

b. An unexpected increase in count rate by a factor of two on all responding channels.

c. An unexpected change in Reactor Coolant System temperature of greater than 20°F.

d. If the measured boron concentration indicates a change of greater than +/- 50 ppm from the nominal value at the start of core loading.

Answer A correct - this would indicate reduced SDM but must be either a factor of 3 on one channel or 2x on both.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 120 Answer B incorrect - this would affect SDM but the temperature change is 20F.

Answer C correct - boron changed by greater than 50 ppm, SDM has been reduced.

Answer D incorrect - must be stopped due to boron change Technical Reference(s): OP B-8DS2, Core Loading Proposed references to be provided to applicants during examination: NONE Learning Objective: 36967 - Discuss abnormal conditions associated with the Fuel Handling system Question Source: Bank # X B-0297 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 10 CFR Part 55 Content: Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

23 January, 2009 - modified answers to remove specific numbers.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 121 Question 61 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # APE 059 G2.2.22 Importance Rating 4.0 K/A: APE 059 G2.2.22 - Accidental Liquid Radwaste Release: Knowledge of limiting conditions for operations and safety limits.

Proposed Question:

GIVEN:

• Reactor power is 45%

• Circulating Water pump 1-1 is running. 1-2 is shutdown

• A liquid radwaste discharge permit has been authorized

• RE-18, Liquid Radwaste Monitor, is determined to be inoperable when adjusting the setpoint Which of the following actions must be taken prior to initiating the liquid radwaste discharge?

A. Start Circulating Water pump 1-2.

B. Install a temporary radiation monitor.

C. Station an operator at RCV-18 to close the valve if necessary.

D. Have a second independent sample analyzed and release rate calculations verified.

Proposed Answer: D. Have a second independent sample analyzed and release rate calculations verified.

Explanation:

A incorrect. Flow from one CWP is adequate and the discharge rate is adjusted accordingly.

B incorrect. Temporary monitor is not required.

C incorrect. No additional operators are required. The sample will be reanalayzed to assure its discharge is allowed.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 122 D correct. Per OP G-1:II, precaution and limitation states:

a. Ensure that the release rate calculations are verified by at least two qualified staff members.

Technical Reference(s): OP G-1:II, precautions and limitations Proposed references to be provided to applicants during examination: NONE Learning Objective: 8443 - State the administrative requirements of Liquid Rad Waste system Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam DCPP 2007 (L051)

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 10 CFR Part 55 Content: Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

23 January, 2009. Action known at the RO level to be taken for a discharge in the OP, which satisfies the LCO for ECG 39.3. Attached the OP as a reference.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 123 Question 62 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # APE 060 G2.4.50 Importance Rating 4.2 K/A: APE 060 G2.4.50 - Accidental Gaseous Radwaste Release: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

Proposed Question:

Several Auxiliary Building radiation alarms are received. It is confirmed that a Waste Gas Decay Tank has ruptured, and is depressurizing into the Auxiliary Building.

What action must be taken to minimize the offsite release of radioactive particulate and iodine?

A. Push "Status Reset" at POV1 and POV2, and reset the "S" signal.

B. Locally close dampers that isolate the Waste Gas Decay Tank rooms.

C. Stop all Aux Bldg supply and exhaust fans, and energize charcoal heaters.

D. Select "S" signal test, secure one Aux Bldg Ventilation train, and energize charcoal heaters.

Proposed Answer: D. Select "S" signal test, secure one Aux Bldg Ventilation train, and energize charcoal heaters.

Explanation:

Answer A incorrect - this is done to recover from the event.

Answer B incorrect - the area is evacuated.

Answer C incorrect - one train left in operation to move air through the charcoal filter.

Answer D correct - action is taken to filter the air and minimize the offsite release.

Technical Reference(s): AP-14 step 2 Proposed references to be provided to applicants during examination: NONE 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 124 Learning Objective: 4767 Discuss abnormal conditions associated with the ABVS Question Source: Bank # X B-0367 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 13 55.43 10 CFR Part 55 Content: Procedures and equipment available for handling and disposal of radioactive materials and effluents.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 125 Question 63 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # EPE 074 EK3.04 Importance Rating 3.9 K/A: EPE 074 EK3.04 - Knowledge of the reasons for the following responses as the apply to the Inadequate Core Cooling: Tripping RCPs Proposed Question:

GIVEN:

• The crew is performing the actions of EOP FR-C.1, "Response to Inadequate Core Cooling"

• The steam generators have been depressurized to inject the accumulators

• The crew is preparing to continue depressurization of the steam generators to atmospheric pressure Prior to the depressurization, the procedure directs the operator to secure all RCPs.

What is the basis for securing the RCPs?

A. To minimize the inventory loss.

B. To facilitate depressurization and establish RHR injection earlier.

C. Continued operation may result in damage to the RCPs due to loss of Number 1 seal requirements.

D. To prevent sweeping accumulator nitrogen into steam generator U-tubes where it could interrupt core cooling.

Proposed Answer: C. Continued operation may result in damage to the RCPs due to loss of Number 1 seal requirements.

Explanation:

Answer A incorrect - while a small break is most likely occurring, inventory loss is a concern for tripping RCPs for a small break LOCA. In FR-C.1, RCPs are left running until depressurizing the SGs to atmosphere.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 126 Answer B incorrect - Core cooling will actually degrade until RHR flow is established, continued operation of the RCPs would result in quicker depressurization.

Answer C correct - Per the background for C.1, the RCPs are stopped to prevent damage due to loss of number 1 seal dp.

Answer D incorrect - accumulator nitrogen is the basis for initially stopping the depressurization.

Technical Reference(s): FR-C.1 step 14, WOG background, C.1 step 13 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7920 Explain basis of emergency procedure steps Question Source: Bank # P-38529 Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 10 CFR Part 55 Content: Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

23 January, 2009 - moved problem statement from setup. Minor editorial changes 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 127 Question 64 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # E08 EK2.2 Importance Rating 3.6 K/A: E08 EK2.2 - Knowledge of the interrelations between the (Pressurized Thermal Shock) and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility Proposed Question:

Why is it desirable to terminate SI in EOP FR-P.1, "Response to Imminent Pressurized Thermal Shock Condition," if the criteria are met?

A. To conserve the water in the RWST.

B. SI flow may have contributed to the RCS cooldown.

C. RCS heat removal is via the steam generators and SI flow is NOT required.

D. The other SI termination criteria will have already been met when FR-P.1 is entered.

Proposed Answer: B. SI flow may have contributed to the RCS cooldown.

Explanation:

Answer A incorrect - this is the basis for other EOPs, such as loss of recirc.

Answer B correct - from the background document: Safety injection flow, if present, is a significant contributor to any cold leg temperature decrease. It can also be a significant contributor to an overpressure condition if the RCS is intact. A check for SI termination is performed early in this guideline based on less restrictive criteria than in other SI termination steps in the ORGs to try to remove its unfavorable PTS effects.

Answer C incorrect - this is not a large break scenario. The procedure is not used for large break.

Answer D incorrect - RVLIS level this low would not satisify other SI termination criteria (doubtful there would be 20 F subcooling).

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 128 Technical Reference(s): WOG FR-P.1 background section 3 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7918 State the bases for relaxing SI termination criteria in response to imminent pressurized thermal shock condition Question Source: Bank # X B-0022 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 10 CFR Part 55 Content: Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

23 January, 2009, minor editorial change. Changed difficulty.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 129 Question 65 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # E13 EA1.1 Importance Rating 3.1 K/A: E13 EA1.1 - Ability to operate and/or monitor the following as they apply to the (Steam Generator Overpressure): Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Proposed Question:

A reactor trip from full power has occurred due to a loss of offsite power.

The crew has entered FR-H.2, Response To Steam Generator Overpressure based upon a YELLOW condition on the Heat Sink CSF Status Tree.

The following conditions exist:

• Steam Generator 11 pressure indicates 1115 psig

• Steam Generators 12, 13, & 14 pressures indicate 1010 psig

• Steam Generator 11 narrow range level is 55%

• PCV-19 indicates CLOSED with full open demand Per FR-H.2, which of the following actions is available to mitigate the steam generator overpressure condition?

A. Initiate steam generator blowdown flow.

B. Open the condenser steam dumps.

C. Cut-in the back-up air bottle for PCV-19 and open PCV-19.

D. Reduce pressure by reducing SG level.

Proposed Answer: C. Cut-in the back-up air bottle for PCV-19 and open PCV-19.

Explanation:

Answer A incorrect - blowdown isolated by AFW start and initiating flow would be an extremely slow evolution.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 130 Answer B incorrect - the condenser dumps are not available (loss of offsite power).

Answer C correct - placing the air in service, allows the operator to open the 10% steam dump valve Answer D incorrect - appropriate if steam generator is nearly full (92%) in FR-H.3 Technical Reference(s): OIM C-2-5 Proposed references to be provided to applicants during examination: NONE Learning Objective: 37810 - Describe controls, indications, and alarms associated with the Steam Dump System Question Source: Bank # X NRC-45739 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 10 CFR Part 55 Content: Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 131 Question 66 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.26 Importance Rating 3.4 K/A: G2.1.26 - Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

Proposed Question:

An operator calls and reports the spill of hydrazine in the vicinity of the Auxiliary Feedwater pump chemical addition skid.

Per CP M-9A, Hazardous Materials Incident Initial Emergency Response/Mitigation Procedure, assistance should be immediately requested from:

A. DCPP Safety.

B. Access Control.

C. DCPP Fire Department.

D. Chemistry & Environmental Operations.

Proposed Answer: C. DCPP Fire Department.

Explanation:

Answer A incorrect - DCPP Hazardous Materials Emergency Response Team (DCPP Fire Department), is dispatched.

Answer B incorrect - DCPP Hazardous Materials Emergency Response Team (DCPP Fire Department), is dispatched.

Answer C correct - DCPP Hazardous Materials Emergency Response Team (DCPP Fire Department), is dispatched.

Answer D incorrect - DCPP Hazardous Materials Emergency Response Team (DCPP Fire Department), is dispatched.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 132 Technical Reference(s): CP M-9A, step 5.2.1 Proposed references to be provided to applicants during examination: NONE Learning Objective: 39579 Discuss the appropriate response to various chemical spills Question Source: Bank # X R-62044 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 10 CFR Part 55 Content: Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 133 Question 67 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.39 Importance Rating 3.6 K/A: G2.1.39 - Knowledge of conservative decision making practices.

Proposed Question:

Operator actions taken in the event of an emergency NOT covered by an approved procedure shall be taken so as to minimize damage to the facility, protect the health and safety of the public, and _____________________.

A. document lessons learned B. minimize personnel injury C. minimize security challenges of the plant Protected Area D. allow continued access by outside agencies (e.g., CDF, Highway Patrol, NRC)

Proposed Answer: B. minimize personnel injury Explanation:

Only B correct:

Per OP1.DC10, step 5.2.3.f.2

f. AOPs and EOPs

2. In the event of an emergency not covered by an approved procedure Operations personnel shall take action so as to:

1) Minimize personnel injury

2) Protect the health and safety of the public

3) Minimize damage to the facility Technical Reference(s): OP1.DC10, step 5.2.3.f.2 Proposed references to be provided to applicants during examination: NONE Learning Objective: 40117 - Discuss DCPP procedures and how they re-enforce the principles for a strong nuclear safety culture.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 134 Question Source: Bank # X R-62671 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 10 CFR Part 55 Content: Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

23 January, 2009 - correction to justification.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 135 Question 68 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # G2.2.2 Importance Rating 4.6 K/A: G2.2.2 - Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

Proposed Question:

GIVEN:

• A reactor startup in accordance with OP L-2, Hot Standby to Startup Mode, is in progress

• The operator has just declared the reactor critical

• Startup rate is +0.25 DPM

• The operator has been instructed to raise power to 10-8 amps How will power be raised to 10-8 amps?

A. Allow power to increase at the current startup rate; withdrawing rods only as necessary to maintain the current startup rate.

B. Continuously withdraw rods until 0.75 DPM startup rate is established.

C. Continuously withdraw rods until a startup rate slightly greater than 0.75 DPM startup rate is established to allow for a slight decrease when the operator stops moving control rods.

D. Withdraw rods in no more than 3 step increments to establish a 0.75 DPM startup rate.

Proposed Answer: D. Withdraw rods in no more than 3 step increments to establish a 0.75 DPM startup rate.

Explanation:

Answer A incorrect - Power is raised by establishing a 0.75 SUR.

Answer B incorrect - Once critical, rod pulls are in 3 step increments.

Answer C incorrect - Once critical, rod pulls are in 3 step increments.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 136 Answer D correct - When manually withdrawing control rods in a critical reactor, the operator at the controls shall stop rod withdrawal at least every three steps and check for expected response on nuclear instruments, rod position and reactor coolant temperature indication.

Technical Reference(s): OP L-2, step 6.1.20 Proposed references to be provided to applicants during examination: NONE Learning Objective:

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 66 55.43 10 CFR Part 55 Content: Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 137 Question 69 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # G2.2.6 Importance Rating 3.0 K/A: G2.2.6 - Knowledge of the process for making changes to procedures.

Proposed Question:

Over the course of an operators shift, the operator finds minor errors (misspellings) in each of the following:

• A normal operating procedure

• A Technical Specification Bases

• A Technical Specification In accordance AD1.ID2, Procedure Process Control, an Editorial Correction should be made to which of the following?

A. The normal operating procedure only.

B. The normal operating procedure and the Technical Specification Bases only.

C. The normal operating procedure and the Technical Specification only.

D. The normal operating procedure, the Technical Specification and the Technical Specification bases.

Proposed Answer: A. The normal operating procedure only.

Explanation:

Answer A correct - Per AD1.ID2, the procedure does not apply to the following nuclear generation documents:

Operating License documents, including Technical Specifications. Refer to XI3.ID1, "Facility License Change Process."

• Technical Specifications Bases. Refer to XI3.ID6, "Technical Specification Bases Control Program."

• Curves and Miscellaneous Data. Refer to CF4.DC3, " Control of Volume 9B - Curves and Miscellaneous Data.".

Answer B incorrect - procedure does not apply to TS Bases.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 138 Answer C incorrect - procedure does not apply to TS.

Answer D incorrect - Procedure does not apply to TS or TS bases.

Technical Reference(s): AD1.ID2 page 2 Proposed references to be provided to applicants during examination: NONE Learning Objective: 3456 - Explain the approval process for plant procedures and policies Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 10 CFR Part 55 Content: Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 139 Question 70 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # G2.2.13 Importance Rating 4.1 K/A: G2.2.13 - Knowledge of tagging and clearance procedures.

Proposed Question:

According to OP2.ID2, Tagging Requirements, what type of tags are used by operations to place equipment in a safe condition for maintenance, repair, or testing?

A. Red tags only.

B. Danger tags only.

C. Either Red or Danger tags.

D. Out of service tags.

Proposed Answer: B. Danger tags only.

Explanation:

Answer A incorrect - Red tags are hung anytime a craft performs work associated within a clearance containing danger tags.

Answer B correct - Danger tags are used to protect personnel by tagging devices used to isolate sources of liquids, steam, gases, or electrical power or place equipment in a safe condition for maintenance, repair, or testing.

Answer C incorrect - a Red tag is used for craft work, after the component has been danger tagged.

Answer D incorrect - Out of service tags are used to identify instruments, controls, or equipment that are affected by maintenance, testing, or a clearance.

Technical Reference(s): OP2.ID2 step 4.1, 4.2 and 4.6 Proposed references to be provided to applicants during examination: NONE 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 140 Learning Objective: 39833 - Describe the Operator's responsibilities when removing equipment from service Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 10 CFR Part 55 Content: Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

23 January, 2009 - editorial enhancement to question.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 141 Question 71 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # G2.3.4 Importance Rating 3.2 K/A: G2.3.4 - Knowledge of radiation exposure limits under normal or emergency conditions.

Proposed Question:

Which of the following describes the 10CFR20 Limit and the Diablo Canyon Administrative Limit for radiation exposure for a calendar year?

10CFR20 Limit DCPP Admin Limit A. 4500 mREM 2000 mREM B. 4500 mREM 4000 mREM C. 5000 mREM 2000 mREM D. 5000 mREM 4500 mREM Proposed Answer: D. 5000 mRem 4500 mRem Explanation:

Answer A is incorrect. 10CFR20 limit is 5000 mREM and not 4500 mREM, 4500 mREM is the Admin limit, DCPP Admin Guideline is 2000 mREM.

Answer B is incorrect. 10CFR20 limit is 5000 MREM and not 4500 mREM , 4500 mREM is the Admin limit, 4000 mREM is 90% of 4500 mREM.

Answer C is incorrect. 10CFR20 limit is 5000 mREM and the DCPP Admin Guideline is 2000 mREM.

Answer D is correct. 10CFR20 limit is 5000 mREM and the DCPP Admin limit is 4500 mREM.

Technical Reference(s): RP1.ID6, Personnel Dose Limits and Monitoring OIM Administrative Radiation Exposure Limits, Page S-1-1, 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 142 Rev. 20.

Proposed references to be provided to applicants during examination: NONE Learning Objective: STATE the DCPP administrative exposure guidelines.

Question Source: Bank # X DCPP RO # 71 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam DCPP 2008 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 10 CFR Part 55 Content: Radiological safety principles and procedures.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 143 Question 72 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # G2.3.13 Importance Rating 3.4 K/A: G2.3.13 - Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters.

Proposed Question:

An accessible area of the Auxiliary Building has a general Area Radiation level of 500 mrem/hr.

Which of the following area postings should be displayed at the entrance to this area?

A. Radiation Area only.

B. High Radiation Area.

C. Very High Radiation Area.

D. Locked High Radiation Area.

Proposed Answer: B. High Radiation Area.

Explanation:

Answer A is incorrect. A Radiation area is an area accessible to personnel with radiation levels greater than 5 mrem/hr.

Answer B is incorrect. A High Radiation area is an area accessible to personnel with radiation levels greater than100 mrem/hr.

Answer C is incorrect. A very high Radiation Area is area accessible to personnel with radiation levels greater than 500 rads/hr measured at one meter from the radiation source or from any surface that the radiation penetrates.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 144 Answer D is incorrect. Locked Hi Radiation Area is an area accessible to personnel with radiation levels greater than 1000 mrem/hr at 30 cm from the radiation source or from any surface that the radiation penetrates.

Technical Reference(s): RCP D-240, Radiological Postings, Pages 1-4, Rev. 18.

Proposed references to be provided to applicants during examination: NONE Learning Objective: DEFINE the following: Comment [TRP1]: Changed High -

High Radiation Area to Locked - High

1. Radiological Controls Area Radiation Area, and Hot Particle Zone as

2. Radiation Area it is now a Red Zone on the CAR rating for Surface Contamination Areas. All this

3. High Radiation Area per Bop Clark - Radiation Protection.

4. Locked-High Radiation Area

5. Very High Radiation Area

6. Surface Contamination Area

7. Airborne Radioactivity Area Question Source: Bank #

Modified Bank # X 2008 DCPP RO #72 New Question History: Last NRC Exam 2008 DCPP Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 10 CFR Part 55 Content: Radiological safety principles and procedures.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 145 Question 73 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # G2.3.14 Importance Rating 3.4 K/A: G2.3.14 - Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Proposed Question:

GIVEN:

• A plant shutdown is in progress due to RCS activity levels greater than allowed by Technical Specifications when a small break LOCA occurs

• The crew initiates Safety Injection and trips the reactor

• The crew has transitioned from E-1, Loss of Reactor or Secondary Coolant to E-1.2, Post LOCA Cooldown and Depressurization

• The crew is now preparing to establish RCP seal return flow in accordance with step 28

• CCW to RCP valves have remained open for the event Prior to the opening of 8100 and 8112, RCP Seal Water Return Stop Valves, an evaluation should be performed to evaluate the consequences of which of the following occurring?

A. Inter-system LOCA.

B. Re-initiation of break flow.

C. Thermal shock to the RCP seals.

D. Increased radiation levels in the Auxiliary Building.

Proposed Answer: D. Increased radiation levels in the Auxiliary Building.

Explanation:

Caution at step 28 states: If excess activity levels in the RCS are suspected, then an evaluation of the consequences of re-establishing seal return flow should be made prior to placing RCP seal return flow in service.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 146 Answer A incorrect - an evaluation is performed due to the activity levels.

Answer B incorrect - an evaluation is performed due to the activity levels.

Answer C incorrect - an evaluation is performed due to the activity levels.

Answer D correct - an evaluation is performed due to the activity levels.

Technical Reference(s): E-1.2 caution at step 28 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7920 - Explain basis of emergency procedure steps Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 10 CFR Part 55 Content: Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 147 Question 74 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # G2.4.26 Importance Rating 3.1 K/A: G2.4.26 - Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment usage.

Proposed Question:

If a Nuclear Operator (NO) is to be used as the Ops Responder, which of the following is a requirement for the NO assigned?

A. The NO must be an NO8.

B. The NO must be the Unit 2 Polisher watch.

C. The NO must be an extra watchstander and not be assigned any watches.

D. The NO must be briefed that post trip duties will take precedence over fire response for fires outside the Protected Area.

Proposed Answer: A. The NO must be an NO8.

Explanation:

Answer A correct - per OP1.DC37, if an NO is used, the NO must be an NO8.

Answer B incorrect - while the Polisher 1 is preferred over Unit 2, it is not mandatory that it be the Polisher, it could be the Aux building or TB watch.

Answer C incorrect - the NO may be a watchstander (but not part of the minimum crew)

Answer D incorrect - another watchstander will perform post trip duties. This is briefed with appropriate operators at the shift brief.

Technical Reference(s): OP1.DC37 step 5.5 Proposed references to be provided to applicants during examination: NONE 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 148 Learning Objective: 5472 - State the administrative requirements for fire brigade Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 10 CFR Part 55 Content: Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 149 Question 75 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # G2.4.32 Importance Rating 3.6 K/A: G2.4.32 - Knowledge of operator response to loss of all annunciators.

Proposed Question:

Both Unit 1 and Unit 2 are at full power, steady state.

The following events occur:

• PK15-22, Main Annunciator System Trouble alarms on Unit 1

• Unit 2 operator reports:

o There are no lit annunciator windows for Unit 2 o The alarm screens (X6000) for Unit 2 are blank Which of the following actions should be taken?

A. Establish continuous surveillance of all parameters in the Control Room for Unit 1.

B. Establish continuous surveillance of all parameters in the Control Room for Unit 2.

C. Trip Unit 1.

D. Trip Unit 2.

Proposed Answer: B. Establish continuous surveillance of all parameters in the Control for Unit 2.

Explanation:

Answer A incorrect - loss of annunciators has occurred for Unit 2.

Answer B correct - per AR PK15-22, for a complete loss of the main annunciator system, the actions are to avoid any load changes and establish continuous surveillance of all parameters in the Control Room Answer C incorrect - no trip required.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 150 Answer D incorrect - no trip required.

Technical Reference(s): AR PK15-22 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7122 - Explain the operation of Alarm Computer system Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 10 CFR Part 55 Content: Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 151 Question 76 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # EPE 007 EA2.06 Importance Rating 4.5 K/A: EPE 007 EA2.06 - Ability to determine or interpret the following as they apply to a reactor trip: Occurrence of a reactor trip Proposed Question:

GIVEN:

• Unit 2 is performing a reactor startup in accordance with OP L-1, Plant Heatup From Hot Shutdown to Hot Standby

• Reactor power is 10-08 amps

• Critical data is being taken A control power fuse blows on Intermediate Range N-35.

What action will be taken?

A. Enter E-0, "Reactor Trip or Safety Injection," at Step 1.

B. Enter AP-5, Malfunction of Eagle 21 Protection or Control Channel, and remove the failed channel from service.

C. Direct the Control Operator to fully insert the Control Rods in accordance with OP L-1.

D. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> either reduce THERMAL POWER to less than P-6 or increase THERMAL POWER to greater than P-10.

Proposed Answer: A. Enter E-0, "Reactor Trip or Safety Injection," at Step 1.

Explanation:

Answer A correct - the loss of control power causes the bistables, specifically IR High Flux, (coincidence 1/2), to trip. Below P-10, this will result in a reactor trip. The appropriate action is to enter E-0.

Answer B incorrect - this is the normal procedure for instrument failures.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 152 Answer C incorrect - in OP L-1, there are criteria during the startup that require the rods to be inserted, (critical below RIL, not critical with rods all out, etc), this failure is not a criteria for rod insertion (other than the rods falling due to the trip).

Answer D incorrect - This is the correct action for one inoperable Intermediate Range channel (Condition F)

Technical Reference(s): OIM B-4-2, TS 3.3.1 Proposed references to be provided to applicants during examination: TS 3.3.1 Learning Objective: 5992 - Discuss abnormal conditions associated with the NIS Question Source: Bank #

Modified Bank # P-1134 (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments: modified distractors of original (bank) question to focus on Tech Specs. If the student fails to realize a trip occurs, the reference provided will appear to apply.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 153 Question 77 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 026G2.2.36 Importance Rating 4.2 K/A: APE 026 G2.2.36 - Loss of Component Cooling Water: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Proposed Question:

GIVEN:

• Unit 1 is at full power

• The 1-1 CCW pump is out of service for maintenance

• Technical Specification 3.7.7 ACTION A has been entered 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later, Emergency Diesel 1-1 is declared inoperable due to the failure of its field to flash, during the performance of STP M-9A, U1 & 2 Diesel Engine Generator Routine Surveillance Test.

For the current plant conditions, the Unit must be in MODE 3 in A. 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

B. 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />.

C. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

D. 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br />.

Proposed Answer: B. 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />.

Explanation:

Answer A incorrect - this is an immediate application of TS 3.0.3.

Answer B correct - This is the correct time for TS 3.0.3. Both trains are inoperable when the second train is inoperable in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, plus the 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> of 3.0.3 for a total of 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 154 Answer C incorrect - Misapplication of 3.7.7. ACTION A has a completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, therefore, there is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> remaining.

Answer D incorrect - misapplication of 3.7.7, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> remaining plus the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from 3.8.1 Technical Reference(s): Technical Specification 3.0.3, 3.7.7, 3.8.1. OIM J-1-1.

OP1.DC38.

Proposed references to be provided to applicants during examination: TS 3.7.7 &

3.8.1 Learning Objective: 9697H - Identify 3.8 Technical Specification LCOs Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 10 CFR Part 55 Content: Facility operating limitations in the technical specifications and their bases.

Comments:

23 January, 2009 - removed Tech Spec 3.0.3 as included reference. Reworded to remove redundant wording in answers.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 155 Question 78 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # EPE 055 G2.4.4 Importance Rating 4.7 K/A: EPE 055 G2.4.4 - Station Blackout - 2.4 - Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Proposed Question:

GIVEN:

• Unit 2 has completed a refueling outage

• The plant is in MODE 3

• L-1, "Plant Heatup from Hot Shutdown to Hot Standby" is in progress A loss of offsite power occurs. None of the Emergency Diesels start.

Which of the following will be performed by the Shift Foreman?

A. Refer to OP AP SD-1, Loss of AC Power.

B. Go to OP AP SD-1, Loss of AC Power.

C. Refer to EOP ECA-0.0, Loss of All AC Power.

D. Go to EOP ECA-0.0, Loss of All AC Power.

Proposed Answer: D. Go to EOP ECA-0.0, Loss Of All AC Power Explanation: With the plant in MODE 3, ECA-0.0 applies.

NOTE: per OP1. DC10, attachment 7.1 Refer to - The designated procedure may be used for guidance at the Operator's discretion.

Go to - Immediately leave the procedure in effect and go to the designated step.

Answer A incorrect - Applies in MODEs 5 and 6.

Answer B incorrect - Applies in MODEs 5 and 6.

Answer C incorrect - The plant is in MODE 3, ECA-0.0 is the applicable procedure but the procedure is entered, not referred to.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 156 Answer D correct - The plant is in MODE 3, ECA-0.0 is the applicable procedure.

Technical Reference(s): WOG EOP users guide (Modes of applicability)

OP1.DC10, attachment 7.1, Emergency Operating Procedures, Synopsis of Rules of Usage OP SD-1, Scope step 1.1 Proposed references to be provided to applicants during examination: NONE Learning Objective: 3552 - Given initial conditions, assumptions, and symptoms, determine the correct Emergency Operating Procedure to be used to mitigate an operational event Question Source: Bank #

Modified Bank # P-45897 (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments: modified conditions to address KA, station blackout.

23 January, 2009 - added reference to define Refer to and Go to, included reference for applicability or OP AP SD-1.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 157 Question 79 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 077 AA2.10 Importance Rating 3.8 K/A: APE 077AA2.10 - Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances: Generator overheating and the required actions Proposed Question:

GIVEN:

• A plant startup is in progress in accordance with OP L-4, Normal Operation at Power

• The plant is at 30% power for a chemistry hold The crew enters AP-30, Main Generator Malfunction based on the following alarms received concurrently in the Control Room:

• PK12-20, Generator Condition Monitor

• PK14-16, RF Monitor What action will be performed by the Shift Foreman?

A. Direct a reactor trip and enter E-0, Reactor Trip or Safety Injection.

B. Direct a unit trip and enter AP-29, Main Turbine Malfunction for a unit trip without a reactor trip.

C. Commence a normal plant shutdown per OP L-4, Normal Operation at Power.

D. Enter AP-25, Rapid Load Reduction or Shutdown and direct a unit trip when power is approximately 15%.

Proposed Answer: B. Direct a unit trip and enter AP-29, Main Turbine Malfunction.

Explanation:

Answer A incorrect - if power is above P-9, a unit trip is warranted (no direction for reactor trip) and E-0 entered. The unit trip actuates a turbine trip. Because power is less than 50%, no reactor trip signal is generated.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 158 Answer B correct - Power is less than P-9 and conditions of generator overheating exists with both the condition monitor and RF monitor in alarm. Action must be taken to immediately remove the unit from service.

Answer C incorrect - the unit must be removed from service for the current plant conditions. The normal shutdown is only performed if the condition monitor is determined to be working properly and plant management concurs a shutdown is appropriate.

Answer D incorrect - a rapid load reduction is not appropriate; the unit must be removed from service. 15% is where the Unit is removed from service in AP-25 Technical Reference(s): AP-30, Main Generator Malfunction, foldout page. OP AP-25, step 22 Proposed references to be provided to applicants during examination: NONE Learning Objective: 7928 - Given initial conditions and assumptions, determine if a unit trip is required Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

23 January , 2009 - modified distractor D.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 159 Question 80 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # W/E11 EA2.2 Importance Rating 4.2 K/A: W/E11 EA2.2 - Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation): Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Proposed Question:

EOP ECA-1.1, Loss of Emergency Coolant Recirculation, is in effect.

Containment pressure rises and the Containment Critical Safety Function turns MAGENTA.

The Shift Foreman will:

A. Remain in ECA-1.1, and direct the operator to start both Containment Spray Pumps to clear the Containment MAGENTA Critical Safety Function.

B. Remain in ECA-1.1, and direct the operator to verify all available CFCUs are running.

C. Transition to FR-Z.1, Response to High Containment Pressure, and direct the operators to operate Containment Spray Pumps as described by FR-Z.1.

D. Transition to FR-Z.1, Response to High Containment Pressure, but direct the operators to operate Containment Spray Pumps as described by ECA-1.1.

Proposed Answer: D. Transition to FR-Z.1, Response to High Containment Pressure, but direct the operators to operate Containment Spray Pumps as described by ECA-1.1.

Explanation:

Answer A incorrect - while in ECA-1.1, the CS pumps are operated based on the number of CFCUs and containment pressure. A transition to address the MAGENTA path is appropriate however, Z.1 will inform the operator to operate the pumps in accordance with ECA-1.1.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 160 Answer B incorrect - CFCUs will cool containment, spray pumps lower pressure. The MAGENTA path must be addressed. Spray pumps are operated to lower pressure but are operated to conserve RWST inventory.

Answer C incorrect - a transition is appropriate, however, the overriding concern to prolong RWST inventory overrides running spray pumps as directed by Z.1.

Answer D correct - a transition is appropriate. Z.1 step 3 b. Operate Containment Spray as directed by EOP ECA 1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, Step 5.

Technical Reference(s): FR-Z.1 Proposed references to be provided to applicants during examination: NONE Learning Objective: 42460 - Explain basis of emergency steps of ECA-1.1 Question Source: Bank # M-0055 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments: no changes following NRC review 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 161 Question 81 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # APE 056 G2.4.6 Importance Rating 4.7 K/A: APE 056 G2.4.6 - Loss of Offsite Power: EOP mitigation strategies.

Proposed Question:

The plant trips from full power due to a loss of all offsite power. All equipment operates as designed.

Which of the following actions will be taken by the Shift Foreman to address the loss of offsite power?

A. Implement ECA-0.3, Restore 4 kV Buses at step 3 of E-0, Reactor Trip or Safety Injection.

B. Go to ECA-0.0, Loss of All AC Power at step 3 of E-0, Reactor Trip or Safety Injection.

C. Implement OP AP-26, Loss of Offsite Power when directed by E-0.1, Reactor Trip Response.

D. Implement OP SD-1, Loss of AC Power when directed by E-0.1, Reactor Trip Response.

Proposed Answer: C. Implement OP AP-26, Loss of Offsite Power when directed by E-0.1, Reactor Trip Response.

Explanation:

Answer A incorrect - vital buses are energized. This is the action in E-0, step 3 if 2 vital buses are de-energized.

Answer B incorrect - this is the immediate action, step 3 if all AC is lost. A total loss of AC has not occurred.

Answer C correct - procedure is implemented to restore power to non-vital buses but only after transitioning to E-0.1, it is not part of the E-0 immediate actions.

Answer D incorrect - procedure is used in MODES 5 and 6.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 162 Technical Reference(s): E-0 immediate actions and E-0.1, step 9 Proposed references to be provided to applicants during examination: NONE Learning Objective: 3552 - Given initial conditions, assumptions, and symptoms, determine the correct Emergency Operating Procedure to be used to mitigate an operational event Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

No changes after NRC review 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 163 Question 82 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # APE 076 G2.2.38 Importance Rating 4.5 K/A: APE 076 G2.2.38 - High Reactor Coolant Activity: Knowledge of conditions and limitations in the facility license.

Proposed Question:

GIVEN:

• The plant is at 70% power when a Main Feedwater Pump trips

• The crew is now stabilizing the plant at approximately 50% power in accordance with AP-15, Loss of Feedwater Flow Why must Chemistry be notified?

A. To determine current boron concentration.

B. To determine if RCS pH adjustment is required.

C. A sample for DOSE EQUIVALENT I-131 specific activity is required by Technical Specifications.

D. A sample for DOSE EQUIVALENT XE-133 specific activity is required by Technical Specifications.

Proposed Answer: C. A sample for DOSE EQUIVALENT I-131 is required by Technical Specifications.

Explanation:

Answer A incorrect - the step specifically instructs the crew to notify chemistry if power reduction exceeded 15% in one hour. This is a Technical Specification requirement. A sample for boron is not required by the procedure and there is no requirement to sample based on the amount of power change.

Answer B correct - no requirement to adjust pH at this time.

Answer C correct - turbine will ramp to 650 MWe at 250 MWe/minute. The power change will take just a few minutes. Step 19 of the AP requires notification if power change is greater than 15% within one hour. This is to satisfy TS 3.4.16 SR 3.4.16.2.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 164 Answer D incorrect - this is a 7 day SR of 3.4.16.2. The surveillance frequency is not affected by power change.

Technical Reference(s): AP-15 step 19 and TS 3.4.16 Proposed references to be provided to applicants during examination: NONE Learning Objective: 3477 - Describe the major actions of abnormal operating procedures Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam DCPP L051 #79 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 10 CFR Part 55 Content: Facility operating limitations in the technical specifications and their bases.

Comments: replacement KA.

23 January, 2009 - distractor A explanation enhancement.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 165 Question 83 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # APE 024 G2.4.11 Importance Rating 4.2 K/A: APE 024 G2.4.11 - Emergency Boration: Knowledge of abnormal condition procedures.

Proposed Question:

GIVEN:

• The crew has entered E-0.1, Reactor Trip Response from E-0, Reactor Trip or Safety Injection

• A 10% steam dump valve is leaking by

• RCS temperature is 540°F and decreasing at a rate of approximately 2 to 3°F/minute

• RCS pressure is stable at approximately 2225 psig

• Pressurizer level is 25% and stable

• AFW flow is throttled to approximately 440 gpm

• MSIVs are closed Which of the following actions will be taken by the Shift Foreman?

A. Direct the operator to initiate Safety Injection and return to step 1 of E-0.

B. Direct the operator to initiate Safety Injection and go to step 4 of E-0.

C. Go to AP-6, Emergency Boration.

D. Implement AP-6, Emergency Boration.

Proposed Answer: D. Implement AP-6, Emergency Boration.

Explanation:

Answer A incorrect - pressurizer level is greater than 6% and subcooling is greater than 20°F, FOP criteria for SI actuation not satisfied.

Answer B incorrect - SI not necessary and the action would be to go to step 1 of E-0.

Answer C incorrect - do not leave an EOP to perform the emergency boration.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 166 Answer D correct - emergency Boration required, the procedure is implemented.

Technical Reference(s): E-0.1 step 1.

Proposed references to be provided to applicants during examination: NONE Learning Objective: 3478 - Given initial conditions, assumptions, and symptoms, determine the correct abnormal operating procedure to be used to mitigate an operational event Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments: Changed from Modified to New question.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 167 Question 84 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # W/E01 EA2.2 Importance Rating 3.9 K/A: W/E01 EA2.2 - Ability to determine and interpret the following as they apply to the (Reactor Trip or Safety Injection Rediagnosis): Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Proposed Question:

The plant trips and Safety Injection actuates automatically.

After exiting E-0 and entering E-1, Loss of Reactor or Secondary Coolant, the Shift Foreman, after review of the entry conditions believes an incorrect transition was made and cannot confirm a proper transition was made.

What action is taken by the Shift Foreman to rediagnose the plant conditions?

A. Perform an update with the crew and return to the diagnosis steps of E-0 to determine the correct procedure.

B. Inform the crew of the suspected incorrect transition, continue in E-1, while monitoring plant conditions via the foldout page.

C. Obtain the concurrence of the Shift Manager to continue in E-1, while referring to the diagnosis steps of E-0.

D. Obtain the concurrence of the Shift Manager and return to the diagnosis steps of E-0 to determine the correct procedure.

Proposed Answer: D. Obtain the concurrence of the Shift Manager and return to the diagnosis steps of E-0 to determine the correct procedure.

Explanation:

Answer A incorrect - Cannot make the unilateral decision to rediagnosis, need SM concurrence.

Answer B incorrect - the appropriate action is to return to the diagnosis steps. Informing the crew during the brief performed upon procedure entry is not appropriate.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 168 Answer C incorrect - SM concurrence must be obtained however it is not appropriate to go to E-1. Per OP1.DC10, if an inappropriate transition is made, the SFM returns to the diagnosis steps of E-0 and would not continue to perform the incorrect procedure.

Answer D correct - per OP1.DC10:

12. If during the implementation of an EOP or AOP, the SFM suspects that the procedure is inappropriate for the existing plant condition, the following steps should be implemented:

a) The Shift Manager (preferred) or another SRO shall concur with the suspected problem.

b) If the required entry conditions are not met, the previous procedure in use shall be reviewed to determine if a correct procedure transition was made.

c) The entry conditions for the procedure in use shall be reviewed.

d) The Shift Manager may authorize return to the Diagnosis Section of EOP E 0 or AP SD 0 for re diagnosis instructions if it still cannot be confirmed that the proper procedure is in use.

Technical Reference(s): OP1.DC10, step 5.2.3.f.12 Proposed references to be provided to applicants during examination: NONE Learning Objective: 41678 - Describe the expectations and standards for abnormal procedure use and adherence, including:

• Actions to be taken in the event that the SFM suspects that and EOP or AOP is inappropriate for existing plant conditions Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

January, 2009 - modified explanation for C 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 169 Question 85 Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # W/E07 EA2.2 Importance Rating 3.9 K/A: W/E07 EA2.2 - Ability to determine and interpret the following as they apply to the (Saturated Core Cooling): Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Proposed Question:

The crew has depressurized the RCS to minimize subcooling in accordance with EOP ECA-3.2, "SGTR With Loss Of Reactor Coolant-Saturated Recovery Desired."

Which of the following CSF YELLOW path procedures, if entered by the Shift Foreman, would direct a return back to ECA-3.2 because actions of the procedure are inconsistent with the actions specified in ECA-3.2?

A. EOP FR-C.3, "Response to Saturated Core Cooling."

B. EOP FR-H.5, "Response to Steam Generator Low Level."

C. EOP FR-P.2, "Response to Anticipated Thermal Shock Condition."

D. EOP FR-S.2, "Response to Loss of Core Shutdown."

Proposed Answer: A. EOP FR-C.3, "Response to Saturated Core Cooling."

Explanation:

Answer A correct - step 1 of FR-C.3 checks if ECA-3.2 is in effect and terminates the performance of FR-C.3 if it is. The goal of ECA-3.2 is to minimize subcooling and reduce leakage, its in conflict with the core cooling yellow path and therefore, the yellow path should not be implemented.

Answer B incorrect - may be performed if desired.

Answer C incorrect - may be performed if desired.

Answer D incorrect - may be performed if desired.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 170 Technical Reference(s): FR-C.3, step 1, FR-C.3 background, step 1 NOTE Proposed references to be provided to applicants during examination: NONE Learning Objective: 3552 - Given initial conditions, assumptions, and symptoms, determine the correct Emergency Operating Procedure to be used to mitigate an operational event Question Source: Bank # NRC-45663 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Wolf Creek 2006 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

No changes 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 171 Question 86 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 003 A2.01 Importance Rating 3.9 K/A: 003 A2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Problems with RCP seals, especially rates of seal leak-off.

Proposed Question:

The plant is at full power. All control systems are in Auto.

The following events occur:

• PK 05-01 alarms, input 1259, RCP 1-1 No. 2 Seal Leakoff Flow High

• RCP 1-1 Number 1 seal leakoff is approximately 1.0 gpm

• RCP 1-2, 1-3, 1-4 Number 1 seal leakoff is approximately 3.0 gpm per RCP

• Charging flow is approximately 87gpm What action will be taken by the Shift Foreman?

A. Per AR PK05-01, direct the operator to increase seal injection.

B. Enter OP AP-28, RCP Malfunction, section B, RCP Number 1 Seal Failure.

C. Enter OP AP-28, RCP Malfunction, section C, RCP Number 2 or 3 Seal Failure.

D. Per AR PK05-01, direct the operator to trip the reactor, stop RCP 1-1 and enter E-0, Reactor Trip or Safety Injection.

Proposed Answer: C. Enter OP AP-28, RCP Malfunction, section C, RCP Number 2 or 3 Seal Failure.

Explanation:

Answer A incorrect - leakoff flow is low, increasing injection would not address the current situation.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 172 Answer B incorrect - the symptoms for number 1 seal failure would be high or low seal leakoff.

Answer C correct - number 2 seal leakoff flow high, coupled with a corresponding decrease in number 1 seal leakoff is an indication of a number 2 seal failure.

Answer D incorrect - a trip at this time is not warranted for a number 2 seal failure.

Technical Reference(s): AR PK05-01, step 2.12, AP-28, section B and C.

Proposed references to be provided to applicants during examination: NONE Learning Objective: 3478 - Given initial conditions, assumptions, and symptoms, determine the correct abnormal operating procedure to be used to mitigate an operational event Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

23 January, 2009 - enhanced distractors A and D. changed reference step to 2.12 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 173 Question 87 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 007 A2.02 Importance Rating 3.2 K/A: 007 A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the PRTS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Abnormal pressure in the PRT.

Proposed Question:

The plant is at full power.

The following events occur:

• PK05-25, PRT Press/Lvl/Temp alarms, input 545 - Pzr Relief Tk Press Hi and Vent Hdr Isol. alarms

• PRT temperature is 110°F and rising at approximately 1°F every 5 minutes

• PRT pressure is 10 psig and rising at approximately 1 psig every 5 minutes

• AR PK05-25 has been entered by the Shift Foreman What action will be taken by the Shift Foreman?

A. Implement AP-1, Excessive RCS Leakage to aid in identifying the source and magnitude of the leakage.

B. Direct the operator to trip the reactor and enter E-0, Reactor Trip or Safety Injection due to an unidentified loss of reactor coolant.

C. Per AR PK05-25, contact maintenance to troubleshoot a possible failure of the PRT N2 Supply regulator, RCS-1-PCV-3035.

D. Per AR PK05-25, vent the PRT to the Waste Gas Header to less than 3 psig using PCV-472, PRT Vent to Vent Hdr, and verify RCS-1-8045, PRT N2 Supply Isolation valve is closed.

Proposed Answer: A. Implement AP-1, Excessive RCS Leakage to aid in identifying the source and magnitude of the leakage.

Explanation:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 174 Answer A correct - increase in pressure with a corresponding increase in level/temperature is an indication of leakage into the PRT. AP-1 is implemented to aid in finding the leakage.

Answer B incorrect - there is no indication that a reactor trip is required at this time.

Answer C incorrect - a failure of the regulator is suspected if only pressure is increasing (no corresponding increase in level or temperature)

Answer D incorrect - PCV-472 cannot be used when PRT pressure is 10 psig or higher.

Technical Reference(s): AR PK05-25 Proposed references to be provided to applicants during examination: NONE Learning Objective: 3552 - Given initial conditions, assumptions, and symptoms, determine the correct abnormal operating procedure to be used to mitigate an operational event Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

23 January, 2009 - enhanced answers to better address both parts of the KA. And modified answer D.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 175 Question 88 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 012 G2.2.38 Importance Rating 4.5 K/A: 012 G2.2.38 - RPS: Knowledge of conditions and limitations in the facility license.

Proposed Question:

A Unit 1 startup is in progress.

While in MODE 3, SG Level Channel LT-527 is determined to be INOPERABLE.

What action(s), if any, must be taken to allow entering MODE 2?

A. No action is required. The startup may continue.

B. The channel must be restored to OPERABLE status prior to entering MODE 2.

C. No action is required until after MODE 2 is entered, then the channel must be tripped within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

D. The channel must be tripped within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and then MODE 2 may be entered.

Proposed Answer: D. The channel must be tripped within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and then MODE 2 may be entered.

Explanation:

Answer A incorrect - Entry in TS 3.3.2 ACTION D is required. Additionally, entry into MODE 2 cannot be made unless the REQUIRED ACTION of TS 3.3.1 and 3.3.2 are satisfied.

Answer B incorrect - MODE change is possible as long as the LCOs are met (channel tripped), therefore, the channel may be inoperable when the mode change is made.

Answer C incorrect - the completion time is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> but the channel must be tripped prior to entry into MODE 2.

Answer D correct - TS 3.3.1 and 3.3.2 require channel tripped in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, then indefinite operation is allowed. AFW actuation, function 5 of 3.3.2 applies in MODES 1, 2 and 3.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 176 Technical Reference(s): AP-5, attachment 4.1, TS 3.3.1 and 3.3.2 Proposed references to be provided to applicants during AP-5, attachment 4.1, TS examination: 3.3.1 and 3.3.2 Learning Objective: 9697C - Identify 3.3 Technical Specification LCOs.

Question Source: Bank # B-0395 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1 10 CFR Part 55 Content: Conditions and limitations in the facility license.

Comments:

23 January, 2009 - editorial changes.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 177 Question 89 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 059 G2.2.38 Importance Rating 4.5 K/A: 059 G2.2.38 - MFW: Knowledge of conditions and limitations in the facility license.

Proposed Question:

The plant is at 100% power.

Which of the following conditions would have the shortest allowable Technical Specification Completion Time?

A. Main Feedwater Reg Valves for all four loops, FCV-510, 520, 530, 540, are declared inoperable.

B. Loop 1 Main Feedwater Reg Valve, FCV-510, and Main Feedwater Reg Bypass valve, FCV-1510, are declared inoperable.

C. Main Feedwater Isolation Valves for all four loops, FCV-438, 439, 440, 441, are declared inoperable.

D. Loop 1 Main Feedwater Reg Valve, FCV-510, and Feedwater Isolation valve, FCV-438, are declared inoperable.

Proposed Answer: D. Loop 1 Main Feedwater Reg Valve, FCV-510, and Feedwater Isolation valve, FCV-438, are declared inoperable.

Explanation:

Answer A incorrect - note in TS 3.7.3 states that separate entry is allowed for each valve, therefore, Condition B may be entered for each valve and the completion time is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Answer B incorrect - condition TS 3.7.3 Condition E with a completion time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> states: Two valves in the same flow path inoperable, resulting in a loss of feedwater isolation capability for the flow path - this could be interpreted as a feed reg/bypass valve combination. However, the bases states the combinations that could result in a loss of capability are:

• a MFRV or MFRV bypass valve and MFIV inoperable, 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 178

• or

• MFWP turbine stop valve (resulting in a loss of MFWP trip function) and MFRV or MFRV bypass valve inoperable A Main Feed and bypass valve are not one of the combinations and therefore, does not apply. The correct application would be condition B (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) for the MFRV and C for the bypass valve (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />).

Answer C incorrect - AD13.DC1 attachment 7.7 states TS 3.6.3 condition C applies.

There is no condition for multiple valves inoperable, however, note 2 states that separate entry is allowed for each valve, therefore, Condition C for each valve allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Also misapplication of 3.6.3 could result in picking condition A, 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> completion time (correct application is condition C, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />).

Answer D correct - Two valves in the same flow path inoperable, resulting in a loss of feedwater isolation capability for the flow path - the bases states: With either a MFRV or MFRV bypass valve and MFIV inoperable, or MFWP turbine stop valve (resulting in a loss of MFWP trip function) and MFRV or MFRV bypass valve inoperable, there may be no redundant system to operate automatically and perform the required safety function.

Condition E completion time is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Technical Reference(s): Tech Spec 3.7.3, and bases, AD13.DC1 attachment 7.7, OVID 106703 sheets 2 and 3.

Proposed references to be Tech Spec 3.7.3 and 3.6.3 AD13.DC1 attachment 7.7 provided to applicants (pages 1 - 4) during examination:

Learning Objective: 9694G - Discuss 3.7 Technical Specification bases Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1 10 CFR Part 55 Content: Conditions and limitations in the facility license.

Comments: 23 January, 2009 - expanded justifications 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 179 Question 90 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 061 A2.03 Importance Rating 3.4 K/A: 061 A2.03 - Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of DC power Proposed Question:

GIVEN:

• MDAFW pump 1-2 is out of service

• The plant tripped in accordance with OP AP-23, Loss of Vital DC Bus due to a loss of DC bus 12

• All steam generator narrow range levels are less than 6%

What action will be taken by the Shift Foreman?

NOTE:

• E Reactor Trip or Safety Injection

• E-0.1 - Reactor Trip Response

• FR-H.1 - Response to Loss of Secondary Heat Sink A. Perform OP AP-23 to stabilize the plant while implementing E-0. Transition to E-0.1 to perform post-reactor trip actions and verify AFW flow of greater than 435 gpm.

B. Perform OP AP-23 to stabilize the plant while implementing E-0. Transition to FR-H.1 will be due to a loss of all AFW.

C. Perform E-0 and transition to E-0.1 to perform post-reactor trip actions.

Implement OP AP-23 to restore AFW flow by manually starting the TDAFW pump and locally open the TDAFW MOVs.

D. Perform E-0 while implementing OP AP-23. Direct the board operator to manually start the TDAFW pump and dispatch an operator to locally control AFW flow until a heat sink is restored.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 180 Proposed Answer: A. Perform OP AP-23 to stabilize the plant while implementing E-0. Transition to E-0.1 to perform post-reactor trip actions and verify AFW flow of greater than 435 gpm.

Explanation:

DC 11 - affects AFW pump 1-3 (bus F)

DC 12 - affects AFW pump 1-1 (TDAFW) (bus G)

DC 13 - affects AFW pump 1-2 (bus H)

Answer A correct - DC bus 12 will prevent a start of the TDAFW pump, however, MDAFW pump 1-3 will start and supply AFW to 2 steam generators. The Heat Sink CSF will not be challenged.

Answer B incorrect - There will be greater than 435 gpm AFW flow available from pump 1-2. Bus 12 affects Bus G loads and the TDAFW. One could assume MDAFW pump 1-3 is powered from bus G and since the TDAFW pump is also affected, would result in a loss of all AFW.

Answer C incorrect - E-0 is implemented and OP AP-23 is performed.

Answer D incorrect - E-0 is implemented and OP AP-23 is performed.

Technical Reference(s): OIM J-1, E-0, AP-23 step 1 and appendix C.

Proposed references to be provided to applicants during examination: none Learning Objective: 7116 - Explain the consequences of loss of DC vital bus 41678 - Describe the expectations and standards for abnormal procedure use and adherence, including:

• Use of concurrent abnormal procedures Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question Cognitive Level: Comprehension or Analysis X 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Rev 1 - 23 January, 2009 - changed DC bus lost in setup and answers to better address KA.

Expanded/corrected justification.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 181 Question 91 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 034 A1.01 Importance Rating 3.2 K/A: 032 A1.01 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fuel Handling System controls including: Load Limits Proposed Question:

While a bowed fuel assembly is being loaded into the core, the Refueling SRO observes the SLACK CABLE light energize.

Which of the following would be an indication that the fuel assembly was properly loaded onto the core plate and not hung up on an adjacent assembly?

A. Minimal load indicated on the fuel cell.

B. Verification that the assembly has been lowered on index.

C. Underload light is also lit.

D. Z-Z tape indicates the full down.

Proposed Answer: D. Z-Z tape indicates the full down.

Explanation:

Answer A incorrect - this indication could indicate either condition.

Answer B incorrect - if the assembly is bowed, it may have to be loaded off-index (as stated in precaution 5.3.5.a:If a fuel assembly being lowered is bowed or out of plumb such that the bottom nozzle is off the core location when the crane is indexed, or if the top of an adjacent assembly is violating the space for an assembly being lowered, it may be necessary to move the crane off index to permit entry into the core.) Additionally, just a verification that the assembly was on index does not mean it is on the core plate.

Answer C incorrect - underload would be encountered for both situations.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 182 Answer D correct - per OP B-8DS2, attachment 9.9 (core loading) step 13, indications are:

• Minimal load is indicated on the load cell

• SLACK CABLE light is ON

• TUBE DOWN light is ON

• Z Z tape indicates full down

• Expected Gemco position for down on core plate Technical Reference(s): OP B-8SD1 attachment 9.7 Proposed references to be provided to applicants during examination: NONE Learning Objective: 36964 - Describe controls, indications, and alarms associated with the Fuel Handling system Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 7 10 CFR Part 55 Content: Fuel handling facilities and procedures.

Comments:

23 January, 2009 - modified question to better align with the KA.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 183 Question 92 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 055 G2.4.11 Importance Rating 4.2 K/A: 055 G2.4.11 - Condenser Air Removal: Knowledge of abnormal condition procedures Proposed Question:

GIVEN:

• Reactor power - 45%

• Gernerator output - decreased approximately 20 MWe to 450 MWe

• PK10-11, Condenser Press/Level, input 620 - Condenser Pressure, in alarm

• The operator reports Condenser pressure is 4.8" Hg Abs and increasing slowly

• The operator reports condenser delta P is 6.5 psid and stable

• CWP 11 and 12 amps are stable Which procedure will be entered by the Shift Foreman to address the current plant conditions?

A. OP AP-7, Degraded Condenser, Section A - Loss of Condenser Vacuum to address the decreasing vacuum.

B. OP AP-7, Degraded Condenser, Section B - Condenser Fouling for guidance on backwashing the condenser halves.

C. OP AP-29, Main Turbine Malfunction for the automatic turbine trip below P-9.

D. OP AP-7, Degraded Condenser, Section C - Traveling Screen Problem to address the fouling of the condenser.

Proposed Answer: A. OP AP-7, Degraded Condenser, Section A - Loss of Condenser Vacuum to address the decreasing vacuum.

Explanation:

Answer A correct - decreasing output and increasing vacuum are entry conditions for section A of OP AP-7.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 184 Answer B incorrect - Increasing DP would indicate a potential entry into Section B, the threshold for entry is all DPs greater than 10 psid or one greater than 10 and increasing rapidly. Because delta P are stable and less than 10 psid, entry is not warranted at this time.

Answer C incorrect - no turbine trip thresholds have been reached at this time (vacuum trip setpoint is 7 inches).

Answer D incorrect - no indication of abnormal screen dp (no bar or rack alarms)

Technical Reference(s): AR PK10-11, AP-7 section A, B, C and attachment 6.2 Proposed references to be provided to applicants during examination: NONE Learning Objective: 3477 - Given an abnormal condition, summarize the major actions of the abnormal operating procedure to mitigate an event in progress.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Rev 1 - 23, January, 2009 - rewrote question to better match KA.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 185 Question 93 Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 041 A2.02 Importance Rating 3.9 K/A: 041 A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the SDS system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Steam valve stuck open Proposed Question:

GIVEN:

• E-3, Steam Generator Tube Rupture is being performed

• The ruptured steam generator has been isolated

• RCS pressure is 1300 psig

• Pressurizer level is 35%

• The crew has just completed an RCS cooldown A 10% steam dump valve on an intact steam generator fails open causing pressure in that steam generator to decrease uncontrollably. RCS pressure and pressurizer level are decreasing slowly.

What action will be taken by the Shift Foreman?

A. Go to ECA-3.1, SGTR With Loss Of Reactor Coolant, Subcooled Recovery Desired.

B. Continue to perform the actions of E-3 unless pressurizer level or subcooling cannot be maintained.

C. Direct the operator to reinitiate SI and return to E-0, Reactor Trip or Safety Injection.

D. Go to E-2, Faulted Steam Generator Isolation.

Proposed Answer: D. Go to E-2, Faulted Steam Generator Isolation.

Explanation:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 186 Answer A incorrect - SI has not at this point be terminated, and ECA-3.1 is for faulted and ruptured steam generators, not separate faulted and ruptured steam generators.

Answer B incorrect - SI reinitiation can only occur once it is terminated. When SI has been terminated, then pressurizer level and subcooling could cause reinitiation.

Answer C incorrect - SI has not been terminated and therefore, cannot be reinitiated.

Additionally, the proper procedure is E-2, not E-0.

Answer D correct - transition to E-2 would be made per the foldout page criteria:

Any S/G Pressure is decreasing in an uncontrolled manner or has completely depressurized, AND has NOT been isolated, unless it is needed for cooldown Technical Reference(s): E-3, foldout page Proposed references to be provided to applicants during examination: NONE Learning Objective: 7336 - State contents of foldout page Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 187 Question 94 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.4 Importance Rating 3.2 K/A: G2.1.4 - Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 10CFR55, etc.

Proposed Question:

Both units are at full power.

How many Nuclear Operators are required to be on watch for both units to comply with Technical Specification 5.2.2, Unit Staff?

A. 1 Nuclear Operator per unit.

B. 3 Nuclear Operators per unit.

C. A total of 3 Nuclear Operators between both units.

D. A total of 4 Nuclear Operators between both units.

Proposed Answer: C. A total of 3 Nuclear Operators between both units.

Explanation: OP1.DC37 states: 5.5.1 The Unit 1 Work Control Lead (WCL) should be responsible for completing the watch list and for notifying the shift manager when the minimum requirements for shift crew composition cannot be met. This responsibility is normally transferred to the non outage unit WCL during outages.

Therefore, knowing the minimum crew composition is an SRO responsibility.

Answer A incorrect - per Tech Spec 5.2.2, a total of 3 NOs between both units is required.

Answer B incorrect - the requirement is 3 total, not 3 per unit.

Answer C correct - per TS 5.2.2 Unit Staff -

The unit staff organization shall include the following:

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 188

a. A non-licensed operator shall be assigned to each reactor containing fuel with a total of three non-licensed operators required for both units.

Answer D incorrect - the requirement is 3 total, not 4.

Technical Reference(s): Tech Spec 5.2.2, OP1.DC37 Proposed references to be provided to applicants during examination: none Learning Objective: 9634 - State the contents of design and admin sections of Technical Specifications Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 1 10 CFR Part 55 Content: Conditions and limitations in the facility license Comments:

Rev 1 - original KA only required basic RO knowledge of radiation levels. Resampled using the KA program to maintain sample plan balance.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 189 Question 95 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G2.1.14 Importance Rating 3.1 K/A: G2.1.14 - Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes, etc.

Proposed Question:

GIVEN:

• PK11-25, Plant Vent Radiation, is in alarm

• PK11-21, High Radiation, is in alarm

• The Nuclear Operator reports LHUT 11 has ruptured Which of the following will be performed by the Shift Foreman?

A. Announce the tank rupture site-wide on the Plant Public Address System.

B. Activate the Site Emergency alarm and follow up with a site wide announcement of the Plant Public Address System.

C. Announce the tank rupture to only the Auxiliary Building on the Plant Public Address System.

D. Activate the Fire alarm and follow up with a site-wide announcement of the Plant Public Address System.

Proposed Answer: A. Announce the tank rupture site-wide on the Plant Public Address System.

Explanation:

Answer A correct - Conditions warrant entry into AP-14. Step 1 is to announce location and source of the problem on the PPAS (and then evacuate personnel from the area).

Answer B incorrect - Site Emergency alarm used if the PPAS is inoperable.

Answer C incorrect - a public announcement to the entire site is warranted to prevent anyone else entering the area and to alert personnel of the possibility of a release.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 190 Answer D incorrect - the fire alarm is not used for this case.

Technical Reference(s): OP AP-14 step 1 Proposed references to be provided to applicants during examination: NONE Learning Objective: 3477 - Given an abnormal condition, summarize the major actions of the abnormal operating procedure to mitigate an event in progress.

Question Source: Bank # P-44269 Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Rev 1 - 23 January, 2009 - modified question to remove possible overlap with sim jpm.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 191 Question 96 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # G2.2.15 Importance Rating 4.3 K/A: G2.2.15 - Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tagouts, etc.

Proposed Question:

The Shift Foreman is reviewing a temporary modification (TMOD) for the Main Feedwater system prior to installation.

How will the Shift Foreman be able to identify the changes to the Main Feedwater system OVID as a result of the TMOD?

A. A NOTE at the bottom of the OVID will describe the TMOD.

B. The changes to the OVID will be identified on the Design Drawing List (DDL).

C. The affected area should be enveloped by a boundary and hatched within the boundary and signed and dated by the system engineer.

D. The changes to the OVID will be clearly indicated as a TMOD.

Proposed Answer: D. The changes to the OVID will be clearly indicated as a TMOD.

Explanation:

Answer A incorrect - A note is used when drawing changes implement corrective action.

It will reference the Quality problem identification number and a requirement that the change shall not be altered or deleted without an approved NCR corrective action change.

(5.1.3.d)

Answer B incorrect - DDL is not applicable to OVIDs (step 5.1.3.g)

Answer C incorrect - this is true for abandoned in place equipment, but no requirement for the system engineer to sign and date. (5.1.3.n)

Answer D correct - the changes should be clearly identified as a TMOD. (5.1.3.m) 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 192 Technical Reference(s): CF3.ID6, step 5.1.3m Proposed references to be provided to applicants during examination: NONE Learning Objective: 9620 - Explain administrative requirements for a temporary modification Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 3 10 CFR Part 55 Content: Facility licensee procedures required to obtain authority for design and operating changes in the facility.

Comments:

23 January, 2009 - replaced distractor A and explanation.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 193 Question 97 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.11 Importance Rating 4.3 K/A: G2.3.11 - Ability to control radiation releases.

Proposed Question:

Given the choice, why is E-3.1, Post-SGTR Cooldown Using Backfill, the preferred procedure for the Shift Foreman to enter from E-3, Steam Generator Tube Rupture?

A. To minimize radiological release and facilitates processing of contaminated primary coolant.

B. To minimize radiological release and it is the fastest of the recovery methods.

C. It is the fastest of the recovery methods and facilitates processing of contaminated primary coolant.

D. To prevent boron dilution of the RCS and facilitates processing of contaminated primary coolant.

Proposed Answer: A. To minimize radiological release and facilitate processing of contaminated primary coolant.

Explanation:

Answer A correct - In general, post-SGTR cooldown using backfill is the preferred method since it minimizes radiological releases and facilitates processing of contaminated primary coolant.

Answer B incorrect - it will minimize radiological release, however, this process will be slow, particularly if no RCP is running. The fastest method is to dump steam (E-3.3)

Answer C incorrect - The fastest method is to dump steam, backfill will be slow, particularly if no RCP is running. It does facilitate processing of the contaminated primary coolant.

Answer D incorrect - Boron dilution is a possibility to be considered when using this method since secondary water will be entering the RCS. It does facilitate processing.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 194 Technical Reference(s): E-3 background Proposed references to be provided to applicants during examination: NONE Learning Objective: 3552 - Given initial conditions, assumptions, and symptoms, determine the correct Emergency Operating Procedure to be used to mitigate an operational event Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 4 10 CFR Part 55 Content: Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Comments:

23 January, 2009 - expanded justification. Modified answers to match background wording.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 195 Question 98 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.15 Importance Rating 3.1 K/A: G2.3.15 - Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Proposed Question:

GIVEN:

• A design basis LOCA has occurred

• RCS pressure is 18 psig

• Steam Generator pressures are 200 psig

• The RHR system has been aligned to cold leg recirculation per E-1.3, Transfer to Cold Leg Recirculation As the crew exits E-1.3, RM-19 (Blowdown radiation monitor, 100 pen area) goes into alarm.

What action, if any, will be taken by the Shift Foreman?

A. None, continue back to E-1, Loss of Reactor or Secondary Coolant.

B. Go to E-3, Steam Generator Tube Rupture.

C. Go to ECA-1.1, Loss of Emergency Coolant Recirculation.

D. Return to E-1.3 to check for proper valve alignment.

Proposed Answer: A. None, continue back to E-1, Loss of Reactor or Secondary Coolant.

Explanation:

Answer A correct - Caution 3 in E-1.3 states: Switchover to cold leg recirculation may cause Very High Radiation Areas in the auxiliary building. RM-19 is in the vicinity of RHR and could be affected by shine from RHR piping.

Answer B incorrect - RCS pressure is less than steam generator pressure.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 196 Answer C incorrect - elevated radiation in the aux building should be expected when on cold leg recirc.

Answer D incorrect - no instruction to return to E-1.3.

Technical Reference(s): E-1.3, caution prior to step 1.

Proposed references to be provided to applicants during examination: NONE Learning Objective: 42458 - Explain basis of emergency steps of E-1.3 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 4 10 CFR Part 55 Content: Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Comments:

No comments 6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 197 Question 99 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.21 Importance Rating 4.6 K/A: G2.4.21 - Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Proposed Question:

GIVEN:

• 20 minutes ago Safety Injection actuated due to a LOCA

• Steam Generator Narrow Range levels are offscale low

• The TDAFW pump has tripped

• All loop Tcolds indicate to the left of Limit A (200°F)

• Four core exit thermocouples indicate 1200°F The crew is preparing to exit E-0, Reactor Trip or Safety Injection.

Which procedure will the Shift Foreman transition to from E-0?

A. E-1, Loss of Reactor or Secondary Coolant.

B. FR-C.1, Response to Inadequate Core Cooling.

C. FR-H.1, Response to Loss of Secondary Heat Sink.

D. FR-P.1, Response to Imminent Pressurized Thermal Shock Condition.

Proposed Answer: D. FR-P.1, Response to Imminent Pressurized Thermal Shock Condition.

Explanation:

Answer A incorrect - the red path for RCS Integrity must be addressed first.

Answer B incorrect - entry into FR-C.1 requires 5 CETCs for entry into FR-C.1, therefore, the entry conditions are not met.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 198 Answer C incorrect - two AFW pumps remain running the entry conditions for FR-H.1 are not met.

Answer D incorrect - entry conditions for RED CSF for RCS Integrity are met. P.1 must be entered before any EOP.

Technical Reference(s): EOP F-0 Proposed references to be provided to applicants during examination: NONE Learning Objective: 9704 - Identify entry conditions for the FRPs Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 10 CFR Part 55 Content: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Rev 1 - 23 January, 2009 - changed question to remove C.1 as the correct answer (now only 4 CETCs) and make entry to P.1 the correct answer.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 199 Question 100 Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.38 Importance Rating 4.4 K/A: G2.4.47 - Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

Proposed Question:

Which of the following describes the responsibility of the Interim Site Emergency Coordinator (ISEC) for upgrading or downgrading an emergency plan classification prior to be being relieved by either the Site Emergency Coordinator or the Recovery Manager?

The ISEC may upgrade NOTE: ECL - Emergency Classification Level A. an ECL and may downgrade an NUE to no ECL.

B. only an NUE to an ALERT and may downgrade an NUE to no ECL.

C. an ECL but does not have the authority to downgrade any ECL.

D. only an NUE to an ALERT but does not have the authority to downgrade any ECL.

Proposed Answer: A. an ECL and may downgrade an NUE to no ECL.

Explanation:

Answer A correct - per EP G-1, the ISEC may upgrade an ECL and may downgrade only an NUE to no ECL.

Answer B incorrect - The ISEC may upgrade any ECL.

Answer C incorrect - The ISEC may upgrade an ECL and has the authority to downgrade an NUE to no ECL.

6 February, 2009

Diablo Canyon Power Plant ANSWER KEY Page 200 Answer D incorrect - The ISEC can upgrade any ECL and has the authority to downgrade NUE to no ECL.

Technical Reference(s): EP G-1, section 6.2 Proposed references to be provided to applicants during examination: none Learning Objective: 5270 - As described in EP G-1, state the responsibilities of Shift Manager as the ISEC Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 3 10 CFR Part 55 Content: Facility license procedures required to obtain authority for design and operating changes in the facility.

Comments:

23 January, 2009 - resampled KA. Original KA not at SRO level.

6 February, 2009

1. AD13.DC1, Attachment 7.7, Containment Isolation Valves (SRO #14)

2. OP AP-5, Attachment 4.1 (SRO #13)

3. Technical Specification 3.3.1, Reactor Trip Instrumentation (SRO #1, SRO#13)

4. Technical Specification 3.3.2, ESFAS Instrumentation (SRO #13)

5. Technical Specification 3.6.3, Containment Isolation Valves (SRO #14)

6. Technical Specification 3.7.3, Main Feedwater (SRO #14)

7. Technical Specification 3.7.7, CCW (SRO #2)

8. Technical Specification 3.8.1, AC Sources - Operating (SRO #2)

9. Steam Tables (not included in package)