ML12325A398

2012-09 Draft Retake Written Exam

ML12325A398
Person / Time
Site:	Fort Calhoun
Issue date:	09/21/2012
From:	Vincent Gaddy Operations Branch IV
To:	Omaha Public Power District
References
50-285/12-09 50-285/OL-12
Download: ML12325A398 (457)

Text

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 001 A reactor trip has occurred from 100% power. While performing the standard post trip actions, the BOPO observes that the main turbine has not tripped and attempted to trip the turbine by pressing both turbine trip pushbuttons.

Which one of the following additional actions is directed by EOP-00, "REACTOR TRIP,"

to trip the main turbine?

A. Place both EHC pump control switches in after-trip.

B. Place both EHC pump control switches in pull-to-lock.

C. Open the generator output breakers.

D. Open the generator field breaker.

Thursday, August 02, 2012 09:53:14 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 1 Rev 0 KA #: 000007 EA2.02 Tier 1 Group 1: Reactor Trip Ability to determine or interpret the following as they apply to a reactor trip: Proper actions to be taken if the automatic safety functions have not taken place Importance 4.3 / 4.6 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective LIST the sixteen (16) immediate actions of EOP-00 in the order they are required to be performed Question Pedigree Bank question used on 2002 NRC Exam, reworded and two choices changed. Not considered to be modified question.

K/A Fit:

Applicant is required to describe actions to be taken if the turbine does not trip as designed following a reactor trip.

Choice A:

Distractor; Plausible because both EHC control switches are repositioned, Incorrect because EOP-00 specifies that both EHC pump control switches be placed in pull-to-lock, not after trip.

Choice B:

Correct answer; EOP-00 specifies that both EHC pump control switches be placed in pull-to-lock if the turbine has not tripped.

Choice C:

Distractor; Plausible because tripping the generator output breakers should result in a turbine trip signal, but incorrect per EOP-00 Choice D:

Distractor; Plausible because loss of excitation should result in a turbine trip signal, but incorrect per EOP-00 KA#: 000007 EA2.02 Bank Ref #: 07-18-10 056 LP# / Objective: 0718-10 01.10 Exam Level: RO-10 Cognitive Level: LOW Source: BANK

Reference:

EOP-00 Handout: NONE Thursday, August 02, 2012 09:53:14 2

EOP-00 Page 8 of 44 INSTRUCTIONS CONTINGENCY ACTIONS

2. Verify the Turbine is tripped as 2.1 Trip the Turbine by performing Step a, indicated by Stop and Intercept b or c:

Valves indicating closed.

a. Trip the Turbine (CB-10,11).

b. Stop the EHC Pumps by placing BOTH of the following control switches in "PULL TO LOCK":

EHC-3A EHC-3B

c. (LOCAL) Trip the Turbine using 1A-3/ST-1 and 1B-3/ST-1, "TURBINE TRIP PUSHBUTTONS" (Turbine Bldg Stairway to Cable Spread Room).

R29

SOURCE QUESTION for Question #1 QUESTIONS REPORT for ILO EXAM BANK 07-18-10 056 A reactor trip has occurred from 100% power. While performing the standard post trip actions, the RO observes that the main turbine has not tripped.

Which one of the following is a procedurally directed action that may be performed to trip the main turbine?

A. Stop the EHC pumps.

B. Deenergize excitation.

C. Reduce load limit pot to zero.

D. Open generator output breakers.

Thursday, August 02, 2012 09:46:56 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 002 The following sequence of events occurred:

A loss of load event resulted in a high pressure reactor trip PORVs and Pressurizer Safety Valves opened One of the Safety Valves failed to close when pressurizer pressure lowered below its setpoint PPLS actuated All RCPs were tripped by the operators per EOP-00, "STANDARD POST TRIP ACTIONS" Current plant conditions are:

Pressurizer pressure is 900 psia and steady Representative CET temperatures are 532°F and steady Reactor Vessel Level is 63%

Pressurizer level is 90% and rising rapidly LPSI pumps SI-1A and SI-1B are running HPSI Pump SI-2A is running All charging pumps are running What action should be taken in response to these plant conditions?

A. Stop all but one charging pump and throttle HPSI loop injection valves.

B. Stop all charging pumps and throttle HPSI loop injection valves.

C. Start an additional HPSI pump and restore letdown.

D. Start an additional HPSI pump and begin a RCS cooldown Thursday, August 02, 2012 09:56:01 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 2 Rev 0 KA #: 000008 AA1.06 Tier 1 Group 1: Pressurizer Vapor Space Accident Ability to operate and / or monitor the following as they apply to the Pressurizer Vapor Space Accident: Control of PZR level Importance 3.6 / 3.6 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

GIVEN a copy of the HPSI Stop and Throttle Criteria floating step, EXPLAIN the four indications used to determine that HPSI Stop and Throttle Criteria are met.

Question Pedigree New Question.

K/A Fit:

Question addresses the use of HPSI stop and throttle to help control pressurizer level with a stuck open safety valve.

Choice A:

Distractor: Plausible because this would be correct if HPSI Stop and throttle criteria were all met.

Choice B:

Distractor: Plausible because this would help control pressurizer level, but HPSI stop and throttle criteria are not met and even if they were, all charging pumps should not be tripped.

Choice C:

Distractor. Plausible because an additional HPSI pump should be started but incorrect because letdown should not be restored.

Choice D:

Correct answer. HPSI stop and throttle criteria are not met due to inadequate subcooling, EOP-03 directs ensuring 2 HPSI pumps are running and performing a plant cooldown.

KA#: 000008 AA1.06 Bank Ref #: N/A LP# / Objective: 0718-13 03.01 Exam Level: RO-10 Cognitive Level: HIGH Source: NEW

Reference:

EOP-03 Handout: NONE Thursday, August 02, 2012 09:56:01 2

EOP/AOP FLOATING STEPS Page 21 of 115 2.0 FLOATING STEPS A. HPSI STOP AND THROTTLE CRITERIA INSTRUCTIONS CONTINGENCY ACTIONS CAUTIONS

1. If emergency boration is required then at least one charging pump must remain running.

2. As natural circulation develops, the expected rise in TH will reduce subcooling. This may jeopardize HPSI Stop and Throttle Criteria.

3. Reducing SI flow should be approached cautiously.

4. The purpose of HPSI stop and throttle is to prevent an over pressurization of the RCS and a solid PZR, however, maintaining RCS inventory is more important than pressure control.

1. Verify ALL of the following stop and throttle criteria are satisfied:

RCS subcooling is greater than or equal to 20F PZR level is greater than or equal to 10% and not lowering At least one S/G is available for RCS heat removal RVLMS indicates level is at or above the top of the Hot Leg (43%, ERF "I" display)

R3

EOP-03 Page 13 of 82 INSTRUCTIONS CONTINGENCY ACTIONS

8. IF SIAS has actuated, 8.1 IF Safety Injection and Charging flow is THEN maximize Safety Injection and NOT maximized, Charging flow to the RCS by operating THEN restore Safety Injection and ALL of the following available pumps: Charging flow by performing the following:

• Either HPSI Pumps, SI-2A/B or SI-2B/C

• LPSI Pumps, SI-1A/B

a. Restore electrical power to valves

• Charging Pumps, CH-1A/B/C and pumps.

b. Ensure ALL of the following SI valves are open to align SI for injection:

• HPSI Loop Injection Valves

• LPSI Loop Injection Valves

• HPSI Discharge Cross-Connect Valves

• HPSI Header Isolation Valves

• SI Pump Suction SIRWT Isolation Valves

c. Start ALL of the following idle pumps:

• HPSI Pumps, SI-2A/B/C

• LPSI Pumps, SI-1A/B

• Charging Pumps, CH-1A/B/C

d. Verify SI flow is acceptable PER Attachment 3, Safety Injection Flow vs. Pressurizer Pressure.

Continuously Applicable or Non-Sequential Step R36

EOP-03 Page 31 of 82 INSTRUCTIONS CONTINGENCY ACTIONS

22. Place ALL Containment Spray Pumps, SI-3A/B/C in PULL TO LOCK.

NOTE Voiding of the RCS is indicated by the inability to depressurize to SDC entry pressure.

Attachment 14, RCS Void Elimination, provides guidance to correct this condition.

CAUTION

1) When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr.

When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

2) No more than three RCPs shall be in operation when RCS temperature is less than 500°F.

23. Commence a Steam Generator 23.1 IF manual operation of Steam Dump cooldown by manually operating the and Bypass Valves is NOT available, Steam Dump and Bypass Valves. THEN commence a Steam Generator cooldown by performing step a or b:

Time:

a. Steam the S/Gs as follows:

1) Ensure MS-164, "MAIN (continue)

STEAM LINE "A" STEAM DUMP TO ATMOSPHERE ISOLATION VALVE", is open (Room 81).

(continue)

Continuously Applicable or Non-Sequential Step R36

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 003 If the primary system temperatures all drop below saturation temperature corresponding to the steam generator pressures during a loss of coolant accident, it indicates that:

A. The ECCS flow is inadequate to remove decay heat.

B. Reflux boiling is taking place in the hot legs.

C. The break flow is adequate to remove decay heat.

D. The safety injection tanks are injecting into the RCS.

Thursday, August 02, 2012 10:03:57 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 3 Rev 0 KA #: 000009 EK2.03 Tier 1 Group 1: Small Break LOCA Knowledge of the interrelations between the small break LOCA and the following: S/Gs Importance 3.0 / 3.3 CFR Number: 55.41(b)(5)

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.

Fort Calhoun Objective:

EXPLAIN how the decay heat removal capacity of the break affects plant response.

Question Pedigree Bank question used on 2002 NRC Exam. Distractor B changed from source question.

Not counted as modified question.

K/A Fit:

Question addresses when S/G's are no longer needed for heat removal in a LOCA.

Choice A:

Distractor: Plausible if Applicant believes primary system saturation pressure is less than S/G pressure due to inadequate ECCS flow. Incorrect because decay heat is being removed by the break.

Choice B:

Distractor: Plausible if Applicant believes reflux boiling improves heat transfer and lowers primary temperatures. Incorrect because primary temperature must be greater than secondary temperature for heat transfer to the steam generators to occur.

Choice C:

Correct answer. If primary temperatures are less than secondary temperatures, there will be no heat transfer to the S/G's. The break is removing all decay heat.

Choice D:

Distractor: Plausible if Applicant believes that SIT injection is causing primary temperature to drop. Incorrect, because primary temperatures fall below secondary temperature at a pressure much higher than 250 psig where the SIT's inject.

KA#: 000009 EK2.03 Bank Ref #: 07-15-23 027 LP# / Objective: 0715-23 01.02 Exam Level: RO-5 Cognitive Level: HIGH Source: BANK

Reference:

SHB 0715-23 Handout: NONE Thursday, August 02, 2012 10:03:57 2

IHB: 07-15-23 Revision: 10 OUTLINE OF INSTRUCTION COMMENTS II. B. 1.

b. In a large break LOCA, pressurizer pressure will decrease very rapidly. It may decrease to match containment pressure in 30 seconds or less.

c. For most breaks, the depressurization rate will slow down considerably once saturated conditions are reached in the RCS.

d. If the break is too small to remove decay heat, the RCS pressure will not fall below the steam generator pressure. This is because the steam generator is needed to remove heat. If the break is large enough to remove decay heat, the primary pressure will drop below the steam generator pressure.

e. For a very small break LOCA, a pressure balance may develop where the break flow is equal to the charging plus HPSI flow.

In this case, some amount of subcooling will remain. If charging or HPSI flow is reduced, the RCS pressure will lower to a new balance point where break flow again equals charging plus HPSI flow. The amount of subcooling will then be reduced.

2. Pressurizer Level

a. Pressurizer level is NOT a good indicator of RCS inventory in a Loss of Coolant Accident.

b. Anytime subcooling has been lost anywhere in the RCS except the pressurizer, steam voids may exist in the RCS. When steam voids are forming or growing, pressurizer level may be increasing while RCS inventory is decreasing.

II. B. 2.

7

Question 3 Source Question QUESTIONS REPORT for ILO EXAM BANK 07-15-23 027 If the primary system temperatures all drop below saturation temperature coresponding to the steam generator pressures during a loss of coolant accident, it indicates that:

A. The ECCS flow is inadequate to remove decay heat.

B. HPSI "Stop and Throttle" should be performed..

C. The break flow is adequate to remove decay heat..

D. The safety injection tanks are injecting into the RCS.

Thursday, August 02, 2012 10:01:46 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 004 5.5 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a large loss of coolant accident which resulted in RAS, the HPSI injection flow is split between the cold legs and the pressurizer auxiliary spray.

Why is this done?

A. To minimize the temperature difference between the pressurizer and the RCS loops.

B. To maintain the RCS pressure low to minimize break flow.

C. To prevent boron precipitation on the fuel assemblies.

D. To equalize boron concentration between the pressurizer and the loops.

Wednesday, August 08, 2012 15:56:29 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 4 Rev 0 KA #: 000011 EK3.12 Tier 1 Group 1:Large Break LOCA Knowledge of the reasons for the following responses as the apply to the Large Break LOCA: Actions contained in EOP for emergency LOCA (large break)

Importance 4.4 / 4.6 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

EXPLAIN how boric acid precipitation in the reactor vessel is prevented following a Loss of Coolant Accident.

Question Pedigree Bank Question used on 2005 NRC exam.

K/A Fit:

A large break LOCA will require simultaneous hot and cold leg injection 5.5 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after LOCA initiation to prevent boron precipitation on the fuel assemblies. Hot leg injection uses pressurizes auxilary spray at Fort Calhoun.

Choice A:

Distractor: Plausible because auxiliary spray will minimize the temperature difference between the pressurizer and the loops. Incorrect because this is not the purpose.

Choice B:

Distractor: Plausible because auxiliary spray can be used to lower pressure. Incorrect because this is not the purpose.

Choice C:

Correct answer: Simultaneous hot and cold leg injection, using auxiliary spray is performed to prevent boron precipitation on the fuel assemblies.

Choice D:

Distractor: Plausible because pressurizer spray is maximized during boration and dilution to equalize boron concentration between the pressurizer and the RCS loops.

Incorrect because this is not the purpose following a LOCA.

KA#: 000011 EK3.12 Bank Ref #: 07-15-23 001 LP# / Objective: 0715-23 02.07 Exam Level: RO-10 Cognitive Level: LOW Source: BANK

Reference:

TBD-EOP-03 Handout: NONE Wednesday, August 08, 2012 15:56:29 2

TBD-EOP-03 Page 73 of 119 INSTRUCTIONS CONTINGENCY ACTIONS CAUTION Simultaneous hot and cold leg injection must be implemented between five and one-half and six hours after a LOCA with RCS pressure less than 140 psia if SDC can not be established.

43. IF BOTH of the following criteria are satisfied:

RCS Pressure is less than 140 psia Five and one-half hours may elapse before SDC can be established THEN IMPLEMENT Attachment 9, Simultaneous Hot and Cold Leg Injection.

Time:

EPG Step: 40 Deviation: 1) The EOP specifies establishment of simultaneous hot and cold leg injection between 5.5 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> into the event if normal shutdown cooling system operation is not available. The EPG uses a band of 2-4 hours and lists conditions instead of stating that if SDC operation can not be established.

2) The simultaneous hot leg injection path that is discussed in Attachment 9 uses the auxiliary Pressurizer spray line and the Pressurizer surge line.

When an unisolated LOCA event exists, simultaneous hot leg and cold leg injection should be implemented if the plant cannot be placed on shutdown cooling within six hours of the LOCA initiation and RCS pressure is less than 120 psig, 140 psia is used as a conservative value to meet this condition (R4). The procedure is implemented at five and one-half hours to provide adequate time to align simultaneous hot/cold leg injection before the six hour time limit.

Injecting to each side of the Reactor Vessel at an injection rate of 147 gpm to the cold legs and 134 gpm to the hot legs, ensures that fluid from the Reactor Vessel (where the boric acid is being concentrated) flows out of the break regardless of the break location and is replenished with a dilute solution of borated water from the other side of the Reactor Vessel (R3).

Continuously Applicable or Non-Sequential Step R36

TBD-EOP-03 Page 74 of 119 INSTRUCTIONS CONTINGENCY ACTIONS The action is taken between 5.5 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the LOCA in order to ensure that the buildup of boric acid is terminated well before the potential for boric acid precipitation occurs. Once the RCS is refilled, the boric acid is dispersed throughout the RCS via natural circulation. If entry into shutdown cooling system operation is anticipated before the 5.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> limit, then the realignment to hot/cold leg injection is unnecessary. This justifies deviation 1.

Hot leg injection path that is used during simultaneous hot and cold leg injection is via containment sump to available HPSI pump configuration to auxiliary Pressurizer spray line to Pressurizer to Pressurizer surge line to RCS hot leg. This justifies deviation 2.

44. WHEN ALL of the following SDC entry conditions are established:

44.1 IF SDC can NOT be established, THEN GO TO Step 45.

PZR level is greater than or equal to 45% and constant or rising RCS subcooling is greater than or equal to 20°F RCS pressure is less than or equal to 300 psia.

RCS TC less than 350°F THEN initiate SDC operation PER ONE of the following Attachments:

Attachment 4, SDC Without RAS Attachment 7, SDC with RAS Attachment 8, Cooled SI Flow with RAS Time:

EPG Step: 45, 46 Deviation: 1) For unisolated LOCA entry into Shutdown Cooling, the EPG criterion on RCS activity level is not included.

2) The EOP specifies a temperature band less than 350°F for RCS TC. The EPG specifies using TH.

3) The EOP is not exited to implement a long term heat removal method that utilizes the Shutdown Cooling Heat Exchangers. Rather several Attachments that look at several different configurations are used. The EPG directs exiting to an operating instruction.

Continuously Applicable or Non-Sequential Step R36

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 005 Annunciator CB-1/2/3, A6, "REACTOR COOLANT PUMP RC-3A VIBRATION HI" came into alarm when Reactor Coolant Pump RC-3A was started.

What action must be taken to reset the high vibration alarm once the vibration is below the alarm setpoint?

A. Reset the alarm on PC-69 located in the plant computer room (Room 73).

B. Reset the alarm on PC-69 located in the upper electrical penetration room (Room 57).

C. Reset the alarm at AI-270 located in the plant computer room (Room 73)

D. Reset the alarm at AI-270 located in the upper electrical penetration room (Room 57)

Wednesday, August 08, 2012 15:57:30 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 5 Rev 0 KA #: 000015 2.1.30 Tier 1 Group 1 Reactor Coolant Pump Malfunctions Ability to locate and operate components, including local controls.

Importance 4.4 / 4.0 CFR Number: 55.41(b)(3)

Mechanical components and design features of the reactor primary system.

Fort Calhoun Objective:

DISCUSS the prerequisites and precautions for operating an RCP.

Question Pedigree Bank question used on the 2001-1 NRC exam. Reworded stem and changed 3 distractors to include location. Not counted as modified question.

K/A Fit:

AI-270 is a local panel where RCP vibration alarms can be reset.

Choice A:

Distractor: Plausible because RCP vibration alarms can be acknowleged on PC-69 and the plant computer room is a plausible location. Incorrect location and alarm can not be reset on PC-66.

Choice B:

Distractor: Plausible because PC-69 is located in room 57 and can be used to acknowledge RCP vibration alarms. Incorrect because the alarms can not be reset on PC-66.

Choice C:

Distractor: Plausible because the alarm can be reset on AI-270. Incorrect because AI-270 is not located in room 73.

Choice D:

Correct Answer: The alarm can be reset on AI-270. AI-270 is located in room 57.

KA#: 000015 2.1.30 Bank Ref #: 07-11-20 031 LP# / Objective: 0711-20 03.02A Exam Level: RO-3 Cognitive Level: LOW Source: BANK

Reference:

ARP CB-1,2,3/A6 Handout: NONE Wednesday, August 08, 2012 15:57:30 2

FORT CALHOUN STATION CONTINUOUS USE ARP-CB-1,2,3/A6 ANNUNCIATOR RESPONSE PROCEDURE PAGE 7 OF 61 Panel: CB-1/2/3 Annunciator: A6 Window: A-4 REACTOR COOLANT PUMP A VIBRATION HIGH Page 1 of 3 REACTOR SAFETY RELATED COOLANT PUMP RC-3A VIBRATION HI Tech Spec

References:

None Initiating Device AI-270A Setpoint REF OI-RC-13 Power AI-42B OPERATOR ACTIONS NOTE: This annunciator may be deactivated if RC-3A vibrations are monitored every three hours.

NOTE: OI-RC-13, Attachment 3, "Alarm Response and Monitoring/Trending of RCP Vibrations",

should be referenced in conjunction with this ARP guidance.

1. Dispatch an operator to AI-270 (Room 57) to report vibration indications and to determine the validity of the vibration alarm.

1.1 If the vibration alarm is due to an electronic problem in the monitoring system, the dual vibration unit that caused the alarm will be in "Bypass" (AI-270A) and/or the "OK" light will be out.

1.2 Vibration indications can be validated by:

1.2.1 Verify the redundant probe for the monitoring point has a similar trend in vibration level.

1.2.2 Corresponding changes in vibration levels on other monitoring points for RC-3A. One point in alarm with no corresponding change in vibration levels of other points indicates a monitoring equipment problem.

(continue)

PROBABLE CAUSES Pump Startup/Shutdown Plant Transient/Trip Loose pump or motor parts Worn or damaged bearings REFERENCES OI-RC-9 OI-RC-13 D-23866-210-111 Sh 1 10473 R46

FORT CALHOUN STATION CONTINUOUS USE ARP-CB-1,2,3/A6 ANNUNCIATOR RESPONSE PROCEDURE PAGE 8 OF 61 Panel: CB-1/2/3 Annunciator: A6 Window: A-4 REACTOR COOLANT PUMP A VIBRATION HIGH Page 2 of 3 SAFETY RELATED OPERATOR ACTIONS (continued) 1.2.3 Points monitored by velocity transducers (seal flange and motor casing) are for diagnostic purposes only. Response to an alarm monitored by velocity transducers should be validated by a review of the other points monitored by proximity probes which indicate actual shaft displacement.

2. IF an alarm is proven to be valid and has reached the alert setpoint during power operation, THEN perform the following:

NOTE: RCP vibration alarm setpoints have been set to monitor RCPs during normal RCS temperature and pressure with 4 RCP's running. Alarms can be expected during RCP starting, stopping, during plant transients (plant trip) or other changes in RCS temperature and pressure.

2.1 Reset the alarm on AI-270A (Room 57).

2.2 If the alarm does not clear, the responsible System Engineer should be contacted to begin diagnostic efforts and trending.

2.3 Reference OI-RC-13 for vibration Baseline, Action Level and Shutdown Limit values.

3. IF an alarm is proven to be valid and has reached the danger setpoint during power operation, THEN perform the following:

3.1 Reset the alarm on AI-270A (Room 57).

(continue)

R46

STM-RCS - Volume 37 Information Use Page 41 of 149 Reactor Coolant System Rev. 44 2.4.2 Design/Specifications A. The following detectors are installed on each reactor coolant pump:

Table 1 - Reactor Coolant Pump Detectors Detector Quantity Seal Pressure 3 Seal Temperature 1 Controlled bleed off flow 1 Controlled bleedoff 1 temperature Motor oil level 2 Motor bearing temperature 4 Motor stator temperature 6 (one in service)

Oil lift pressure 1 Pump vibration (AI-270) 10 Gasket bleedoff pressure 1 Pump rotation 4 Low pressure oil flow 1 (except RC-3B)

Oil pump suction filter 1 pressure Main impeller flow 1 Motor amps 1 B. The reactor coolant pump seal system is monitored by three pressure detectors and two temperature detectors. The three full-pressure seals (lower, middle, and upper) are mounted in series, with the middle and upper seals having a pressure detector that senses the pressure below the seal Additionally, bleed off temperature and pressure detectors are provided.

C. The reactor coolant pump motor is provided with stator winding temperature detectors, bearing temperature detectors, and oil reservoir level detectors which provide alarms to alert the operators to any abnormal conditions.

D. The Reactor Coolant Pump (RCP) Vibration Monitoring System trends, logs, and analyzes vibration on the four reactor coolant pumps (RC-3A/RC-3B/RC-3C/RC-3D) and their associated motors. This system allows for predictive and preventive maintenance to be performed in order to avoid possible equipment damage. The system consists of vibration and speed detection devices, signal conditioners, vibration monitoring instrumentation panels, and a system computer and printer.

STM-RCS - Volume 37 Information Use Page 42 of 149 Reactor Coolant System Rev. 44 2.4.2 (continued)

E. The vibration detection equipment is located on the pumps and pump motors. The detectors consist of proximity and velocity probes. The proximity probes monitor shaft vibration at each bearing location. The velocity probes measure overall pump casing vibration displacement.

The vibration data is monitored on panels AI-270A/AI-270B/AI-270C/

AI-270D located in the upper electrical penetration area, Room 57.

F. Reactor coolant pump speed detection is also performed by a proximity probe which monitors holes drilled in the speed ring on the motor shaft for input to a tachometer.

Figure 2 RCP Speed Sensors G. The system computer monitors the system, records measured values, and manipulates the data to provide trending as well as graphs and plots of data manipulations to allow a visual display of the vibration analysis.

The system monitor and system computer are also located in the upper electrical penetration area, Room 57.

STM-RCS - Volume 37 Information Use Page 50 of 149 Reactor Coolant System Rev. 44 2.4.3M (continued)

12. Each RCP oil reservoir level detector actuates a computer alarm at high and low levels. For RC-3A, the upper oil reservoir is monitored and has a normal range of 75% to 95%, with alarm setpoints that may vary after each calibration. The computer range for this value is 0 to 100%, corresponding to 0 to 15 inches of oil. The lower oil reservoir of RC-3A has a normal range of 80% to 110% and alarm setpoints that vary. The computer range for this value is 0 to 100%,

corresponding to 0 to 5 inches of oil.

13. The RCP Vibration Monitoring System is programmed with setpoints to alarm when vibration limits are exceeded or when set speed limits are met. The alarms are local at panel AI-270. The system also provides input to the REACTOR COOLANT PUMP RC-3A (RC-3B, RC-3C, RC-3D) VIBRATION HI alarm on CB-1,2,3, panel A6. To reset an alarm, the vibration level must be below the setpoints and the alarm must be reset locally at the respective AI-270 panel.

Instructions on how to reset the alarm and other operating information is detailed in procedure OI-RC-13, Operation of the RCP Vibration Monitoring System.

14. The RCP Oil Lift System pressure switch (PCS-3109 for RC-3A) also actuates the REACTOR COOLANT PUMP RC-3A (RC-3B, RC-3C, RC-3D) OIL LIFT PRESS HI alarm at 750 psig on CB-1,2,3, panel A6.

15. When the oil lift pumps suction filter needs cleaning, a differential pressure switch actuates a computer alarm. The filter has a lever on the side that points to a "Needs Cleaning" label. When flow through the cooler is less than 10 gpm, low-pressure switch FAS-3182 actuates the LUBE OIL LOW FLOW alarm on the computer.

16. Each reactor coolant pump has a tachometer located at AI-270, Room 57, which provides direct indication of RCP rotational speed.

The RCP speed tachometer provides contacts which will illuminate the green zero speed light on CB-1,2,3 when the rotation speed approaches zero rpm.

17. Reactor coolant pump RC-3A's reverse rotation switch OCS-3112 is actuated when a change in the direction of the oil being flung off the thrust runner is sensed. A flow of 3.5 gpm causes the switch to actuate the REACTOR COOLANT PUMP RC-3A (RC-3B, RC-3C, RC-3D) REVERSE ROTATION alarm on CB-1,2,3, panel A6 (RC-3B setpoint is greater than 5 rpm in the reverse direction).

Question 5 Source Question QUESTIONS REPORT for ILO EXAM BANK 07-11-20 031 Annunciator CB-1/2/3, A6, "REACTOR COOLANT PUMP RC-3A VIBRATION HI" is in alarm. What action, if any, must be taken to reset the high vibration alarm once the vibration is below the alarm setpoint?

A. The alarm will automatically reset.

B. The alarm can be manually reset at CB-1/2/3.

C. The alarm can be reset at AI-270.

D. The alarm can be reset at an ERFCS terminal.

Thursday, August 02, 2012 10:19:23 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 006 The following Plant Conditions exist:

The plant is operating at full power Letdown control valves are in manual Charging Pump CH-1A is in operation Charging Pumps CH-1B and CH-1C are in the After-Stop position The standby selector switch is in the CH-1B/1C position Pressurizer level control channel LRC-101X is selected as the controlling channel and is in CASCADE Charging flow is 32 gpm and slowly lowering due to a pump malfunction Assuming no operator actions are taken and the pressurizer level continues to lower, when will CH-1B start?

A. When the "charging flow lo" alarm comes in.

B. When the level error is -2.9%

C. When the level error is -3.3%

D. When the level error is -6.0%

Thursday, August 02, 2012 10:24:21 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 6 Rev 0 KA #: 000022 AK1.03 Tier 1 Group 1 Loss of Reactor Coolant Makeup 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 Importance 3.0 / 3.4 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN the interlocks and control functions associated with RCS Instrumentation.

Question Pedigree New question.

K/A Fit:

Question addresses how additional charging pumps increase RCS makeup flow as pressurizer level lowers following a pump malfunction.

Choice A:

Distractor: Plausible if Applicant believes 2nd pump starts at the charging flow low setpoint. Incorrect, because 2nd pump starts due to level deviation.

Choice B:

Correct answer: CH-1B is selected as the first standby charging pump. The first standby pumps starts with a level deviation of -2.9%.

Choice C:

Distractor: Plausible because the standby charging pumps will autostart based on level deviation. Incorrect because this is where the second backup pump, CH-1C, will start.

Choice D:

Distractor: Plausible because the standby charging pumps will autostart based on level deviation. Incorrect because this is where a backup start signal will occur to start all charging pumps.

KA#: 000022 AK1.03 Bank Ref #: N/A LP# / Objective: 0711-20 04.04 Exam Level: RO-7 Cognitive Level: HIGH Source: NEW

Reference:

STM 37 Handout: NONE Thursday, August 02, 2012 10:24:21 2

FORT CALHOUN STATION TDB-III.5 TECHNICAL DATA BOOK PAGE 2 OF 2 R3

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 007 The following plant conditions exist:

The plant operated for 11 months at full power before shutting down to replace a reactor coolant pump seal.

The reactor has been shutdown for 5 days The RCS is currently at Mid-Loop for the RCP seal maintenance The plant is on shutdown cooling with RCS Temperature at 126°F Core Exit Thermocouple and Heated Junction Thermocouple indications are NOT available How long would it take for the RCS to begin to boil if shutdown cooling was lost and not restored?

A. 9 -11 minutes B. 16 - 18 minutes C. 22 - 24 minutes D. 37 - 39 minutes Thursday, August 02, 2012 10:49:12 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 7 Rev 0 KA #: 000025 AK1.01 Tier 1 Group 1: Loss of Residual Heat Removal System Knowledge of the operational implications of the following concepts as they apply to Loss of Residual Heat Removal System: Loss of RHRS during all modes of operation Importance 3.9 / 4.3 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Describe how the plant responds to a Loss of Shutdown Cooling in terms of how specific equipment is affected and how it affects overall plant operation and reliability.

Question Pedigree New question K/A Fit:

Question addresses the consequences of a loss of shutdown cooling while in cold shutdown, mid-loop conditions.

Choice A:

Distractor: This is the correct answer for 1 day after shutdown but incorrect because the stem condition is 5 days after shutdown.

Choice B:

Correct Answer: This is the correct answer for conditions given in the stem per AOP-19 Attachment B.

Choice C:

Distractor: This is the correct answer for 10 days after shutdown but incorrect because the stem condition is 5 days after shutdown.

Choice D:

Distractor: This is the correct answer for 30 days after shutdown but incorrect because the stem condition is 5 days after shutdown.

KA#: 000025 AK1.01 Bank Ref #: N/A LP# / Objective: 0717-19 01.02 Exam Level: RO-10 Cognitive Level: HIGH Source: NEW

Reference:

AOP-19 ATT B Handout: AOP-19 ATT B Thursday, August 02, 2012 10:49:12 2

AOP-19 Page 19 of 110 Attachment B Time to Boil Determination Worksheet

1. Time Shutdown Cooling was lost:

2. Last known RCS/SDCS temperature: °F from instrument number:

3. Record the following information and inform the Shift Manager on 10 minute intervals.

TIME CET/HJTC F HEATUP RATE TIME TO BOIL Alternate Method: Tb = Ta + T0 - Tc Where: Tb is the remaining time to boil Ta is the approximate time to boil from the appropriate curve To is the time SDC was lost Tc is the current time R17

AOP-19 Page 20 of 110 Attachment B Time to Boil Determination Worksheet Time to Boil (RCS at Mid Loop) 70 65 30 DAYS AFTER SHUTDOWN 60 55 50 10 DAYS AFTER SHUTDOWN 45 40 Time 5 DAYS AFTER SHUTDOWN to Boil 35 Min.

30 25 20 15 10 1 DAY AFTER SHUTDOWN 5

0 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 INITIAL TEMPERATURE (Deg F)

R17

AOP-19 Page 21 of 110 Attachment B Time to Boil Determination Worksheet Time to Boil (RCS at Reactor Vessel Flange Steam Generators Unavailable) 120 110 30 DAYS AFTER SHUTDOWN 100 90 80 10 DAYS AFTER SHUTDOWN 70 Time 5 DAYS AFTER SHUTDOWN to Boil 60 Min.

50 40 30 20 10 1 DAY AFTER SHUTDOWN 0

60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 INITIAL TEMPERATURE (Deg F)

R17

AOP-19 Page 22 of 110 Attachment B Time to Boil Determination Worksheet Time to Boil (RCS Open, RV Head on, RCS Level > 1013, Steam Generator U-tubes Filled, One Steam Generator in Wet Layup) 650 600 30 DAYS AFTER SHUTDOWN 550 500 450 10 DAYS AFTER SHUTDOWN 400 Time 350 to Boil 5 DAYS AFTER SHUTDOWN Min.

300 250 200 150 100 50 1 DAY AFTER SHUTDOWN 0

60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 INITIAL TEMPERATURE (Deg F)

R17

AOP-19 Page 23 of 110 Attachment B Time to Boil Determination Worksheet Time to Boil (RCS at Normal Operating Volume Steam Generators Unavailable) 120 110 100 30 DAYS AFTER SHUTDOWN 90 80 10 DAYS AFTER SHUTDOWN 70 Time to Boil 60 Min.

5 DAYS AFTER SHUTDOWN 50 40 30 20 1 DAY AFTER SHUTDOWN 10 0

60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 INITIAL TEMPERATURE (Deg F)

R17

AOP-19 Page 24 of 110 Attachment B Time to Boil Determination Worksheet Time to Boil (Refueling Cavity Flooded) 2000 1500 TIME TO BOIL (MIN) 1000 500 0

60 80 100 120 140 160 180 200 INITIAL TEMPERATURE (DEGREES F) 1 DAY AFTER SHUTDOWN 5 DAYS AFTER SHUTDOWN 10 DAYS AFTER SHUTDOWN 30 DAYS AFTER SHUTDOWN End of Attachment B R17

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 008 CCW flow to Containment Cooler, VA-1A automatically isolated due to low CCW flow following a CIAS actuation. CCW Pump discharge pressure and flow are normal.

What must be done to reestablish flow to the containment cooler?

A. The Control switch for HCV-400A/C must be placed in the "ISOL" position momentarily and then returned to the "NOR" position.

B. The Control switch for HCV-400A/C must be placed in the "CIRC" position momentarily and then returned to the "NOR" position.

C. The Control switch for HCV-400A/C must be held in the "ISOL" position until the low flow signal has reset.

D. The Control switch for HCV-400A/C must be held in the "CIRC" position until the low flow signal has reset.

Thursday, August 02, 2012 14:14:47 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 8 Rev 0 KA #: 000026 AA1.02 Tier 1 Group 1: Loss of Component Cooling Water Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water: Loads on the CCWS in the control room Importance 3.2 / 3.3 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN the operation of controls associated with the CCW System valves operated from the Control Room.

Question Pedigree Bank question used on 2007 NRC exam.

K/A Fit:

Question addresses operation of CCW valves to restore flow to a containment cooler following a low flow isolation.

Choice A:

Distractor: Plausible because this would be a possible logic combination. Incorrect because the control switch must be placed in the CIRC position until the low flow signal resets.

Choice B:

Distractor: Plausible because the control switch must be placed in the CIRC position.

Incorrect because it must be held in that position until the low flow signal resets.

Choice C:

Distractor: Plausible because the control switch must be held until the low flow signal resets. Incorrect because it must be held in the CIRC position.

Choice D:

Correct Answer: The control switch must be held in the CIRC position until the low flow signal resets.

KA#: 000026 AA1.02 Bank Ref #: 07-11-06 063 LP# / Objective: 0711-06 01.02 Exam Level: RO-7 Cognitive Level: LOW Source: BANK

Reference:

STM 08 2.188 Handout: NONE Thursday, August 02, 2012 14:14:47 2

SYSTEM TRAINING MANUAL VOL. 8 COMPONENT COOLING WATER SYSTEM

• • Location 2.184 The containment air cooling and filtering unit CCW isolation valves HCV-400A/B/C/D through HCV-403A/B/C/D are on the 1025 foot elevation of the Auxiliary Building at the east end of Room 69, southwest of the CCW pumps.

• • Power Supplies 2.185 Control power is supplied from opposite 125 VDC buses to ensure isolation of CCW to a component is possible coincident with a loss of one DC bus.

• • Instrumentation and Controls 2.186 The following is a detailed description of the Instrumentation and Controls of the Containment air cooling and filtering unit CCW isolation valves.

• • • Local 2.187 Not applicable

• • • Remote 2.188 Each cooling/filtering unit has two control switches that are on the vertical section of CB-1/2/3. Each control switch operates one pair of CCW isolation valves.

2.189 The control switch positions for the A/C sets of valves are ISOL, NOR, and CIRC. The control switches for the B/D sets of valves also have three positions: CLOSE, NOR, and OPEN. The switches spring return to center (NORMAL). Placing a switch in ISOL will close the A/C set of valves if CIAS is reset. Placing the switch in CIRC opens the inlet valve and enables the controller for modula-tion of the outlet valve. The B/D valves are either opened or closed; the outlet valves are not modulated.

2.190 Holding the control switch in CIRC will override a low-flow sig-nal and allow the operator to reopen the valves. This low-flow sig-nal is only enabled if a CIAS is initiated. The low-flow signal is designed to allow isolation of CCW to a cooling and filtering unit that might develop a leak or rupture in the cooling coils.

• • • Interlocks 2.191 The inlet and outlet valves open on a CIAS. A flow signal closes the valves if a CIAS is present and there is no flow out of the cooler after a variable time delay.

• • • Alarms and Indications 2.192 The CC WATER FROM COIL VA-1A (1B) NO FLOW alarm (two windows) actuates if the flow sensed by FC-416A (FC-417A) for VA-1A (VA-1B) is less than the setpoint of approximately zero Level: Basic > Intermediate > Advanced Revision - 39

• 36

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 009 The reactor tripped 20 minutes ago. The following conditions are observed:

• "PRESSURIZER PRESSURE OFF NORMAL HI-LO" channel X and Y are in alarm.

• "PRESSURIZER LEVEL HI-LO" channel X and Y are in alarm.

• PRC-103X (controlling channel) indicates 2020 psia and stable

• PRC-103Y indicates 2150 psia and rising

• All backup heaters in AUTO and energized

• LRC-101X indicates 40% and rising

• LRC-101Y (controlling channel) indicates 41% and rising

• Letdown flow is 26 gpm

• All charging pump are running

• T-cold indicates 533°F, T-hot indicates 534°F, both are stable Select the probable cause and the action that should be taken to restore RCS pressure:

A. Low level on LRC-101X is maintaining the B/U heaters on, place the pressurizer heater cutout channel select switch in channel Y.

B. The bistable for the B/U heaters needs to be reset, place the control switches for all B/U heaters to reset and back to auto.

C. LRC-101Y has malfunctioned causing the B/U heaters to remain on, place LRC-101X in service.

D. PRC 103X has malfunctioned causing the B/U heaters to remain on, place PRC-103Y in service.

Wednesday, August 08, 2012 15:58:11 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 9 Rev 0 KA #: 000027 AK2.03 Tier 1 Group 1:Pressurizer Pressure Control System Malfunction Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners Importance 2.6 / 2.8 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Given a current copy of ARP, EXPLAIN the alarms associated with the RCS Instrumentation System and the required actions.

Question Pedigree The source question is a bank question that was used on the 2004 NRC exam. The question was modified by changing the conditions in the stem to make another choice correct.

K/A Fit:

Question addresses a pressure control malfunction of a pressurizer pressure controller.

Choice A:

Distractor: Plausible because a low level on either channel will actuate an interlock.

Incorrect because the setpoint is 32% and the interlock deenergizes the heaters.

Choice B:

Distractor: Plausible because there are pressurizer level bistables that can be reset.

Incorrect because the bistables affect the charging pumps not the heaters.

Choice C:

Distractor: Plausible because LRC-101Y is the controlling channel and a level deviation of 5% for program (program is 48%) will cause heaters to turn on. Incorrect because the level deviation must be 5% above programmed level.

Choice D:

Correct Answer: PRC-103X has failed at 2020 psia causing the heaters to remain on.

The correct pressure is indicated by the non-controlling channel, PRC-103Y.

KA#: 000027 AK2.03 Bank Ref #: 07-11-20 008 LP# / Objective: 0711-20 05.04 Exam Level: RO-7 Cognitive Level: HIGH Source: MODIFIED

Reference:

STM 37 Handout: NONE Wednesday, August 08, 2012 15:58:11 2

FORT CALHOUN STATION CONTINUOUS USE ARP-CB-1,2,3/A4 ANNUNCIATOR RESPONSE PROCEDURE PAGE 7 OF 52 Panel: CB-1/2/3 Annunciator: A4 Window: A-4 PRESSURIZER PRESSURE DEVIATION Page 1 of 2 PRESSURIZER SAFETY RELATED PRESSURE OFF NORMAL HI-LO CHANNEL X Tech Spec

References:

2.10.4(5)(a)(ii)

Initiating Device PR-103X Setpoint HI > 2145 psia Setpoint LO < 2080 psia Power PQ-103X/AI-40A OPERATOR ACTIONS

1. Check RCS Pressure using all available indications.

2. IF actual pressure is low, THEN perform the following:

2.1 Apply Technical Specification LCO Action Statement if pressure is less than or equal to 2075 psia.

2.2 Check Pressurizer Heaters are energized.

2.3 Ensure the Pressurizer Spray Valves are closed.

2.4 Ensure the Pressurizer Heater Control Switches are in AUTO.

2.5 Check Pressurizer Level for decreasing trend (LR-101X, LR-101Y, CB-2).

2.6 Check VCT Level for level trend (YR-4100, CB-2).

(continue)

PROBABLE CAUSES Plant Power transient Pressurizer Spray Valve closed Pressurizer Spray Valve open Turbine Load/Reactor Power Mismatch Pressurizer Heaters on REFERENCES 136B3081 Sh 21 06245 161F561 Sh 3 09478 161F561 Sh 19 09503 EM-103 01567 IC-CP-01-0103X E-23866-210-110 10475 R30

FORT CALHOUN STATION CONTINUOUS USE ARP-CB-1,2,3/A4 ANNUNCIATOR RESPONSE PROCEDURE PAGE 8 OF 52 Panel: CB-1/2/3 Annunciator: A4 Window: A-4 PRESSURIZER PRESSURE DEVIATION Page 2 of 2 SAFETY RELATED OPERATOR ACTIONS (continued)

3. IF actual pressure is high, THEN perform the following:

3.1 Ensure the Pressurizer Spray Valves are open and controlling Pressurizer Pressure.

3.2 Ensure the Pressurizer Heaters are not energized.

4. Monitor Pressurizer Pressure and the operation of PC-103X.

4.1 IF PC-103X is not controlling pressure, THEN shift pressure control to PC-103Y (HC-103, CB-3).

4.2 If necessary, take manual control of Pressurizer Pressure.

R30

FORT CALHOUN STATION CONTINUOUS USE ARP-CB-1,2,3/A4 ANNUNCIATOR RESPONSE PROCEDURE PAGE 12 OF 52 Panel: CB-1/2/3 Annunciator: A4 Window: A-8 PRESSURIZER LEVEL DEVIATION Page 1 of 2 FROM PROGRAMMED LEVEL PRESSURIZER SAFETY RELATED LEVEL HI-LO CHANNEL X Tech Spec

References:

2.1.4 Initiating Device LA-101X Setpoint HI 8.0% > program LCA-101X Setpoint LO 6.0% < program Power LQ-101X / AI-40A OPERATOR ACTIONS

1. Check Pressurizer Level on LR-101X and Y (CB-3).

2. IF Pressurizer Level is Low, THEN perform the following:

2.1 Ensure all available Charging Pumps are running and supplying flow to the RCS (FIA-236, CB-2).

2.2 Ensure Letdown Flow is at minimum (26 gpm).

2.3 Check VCT Level for trend (YR-4100, CB-2).

2.4 Check the position of LCV-101-1 and LCV-101-2.

2.4.1 IF LCV-101-1 or -2 are not at minimum position, THEN place HC-101 Pressurizer Level Chan Selector Switch to the Alternate Channel.

(continue)

PROBABLE CAUSES Pressurizer Level Channel failed High or Low Turbine Load/Reactor Power Mismatch Improper functioning of the Letdown Level Control Valves Loss of Makeup capabilities Charging Pump fail to start Leakage in the Reactor Coolant or Chemical and Volume Control System REFERENCES 136B3081 Sh 23 06247 161F561 Sh 3 09478 161F561 Sh 27 09512 EM-101 10227 IC-ST-RC-0004 E-23866-210-110 10475 R30

FORT CALHOUN STATION CONTINUOUS USE ARP-CB-1,2,3/A4 ANNUNCIATOR RESPONSE PROCEDURE PAGE 13 OF 52 Panel: CB-1/2/3 Annunciator: A4 Window: A-8 PRESSURIZED LEVEL DEVIATION Page 2 of 2 FROM PROGRAMMED LEVEL SAFETY RELATED OPERATOR ACTIONS (continued) 2.5 Monitor Tavg versus Pressurizer Level for indication of leakage.

2.5.1 IF low level is due to system leak, THEN GO TO AOP-22.

3. IF Pressurizer Level is High, THEN perform the following:

3.1 Check VCT Level for trend (YR-4100, CB-2).

3.2 Ensure only one Charging Pump is running.

3.3 Ensure Letdown Flow is at maximum.

3.4 Ensure Pressurizer Backup Heaters are Energized.

3.5 Check the position of LCV-101-1 and LCV-101-2.

3.5.1 IF LCV-101-1 or -2 is not fully open, THEN place HC-101 Pressurizer Level Chan Selector Switch to the Alternate Channel.

4. IF Auto PZR Level Control does not control Level, THEN take manual control.

5. IF LI-101X has failed, THEN notify I&C and refer to Technical Specification 2.15.5.

5.1.1 Place HC-101-1, Pzr Heater Cutout Channel Select Switch, to Chan Y.

R30

SYSTEM TRAINING MANUAL VOL. 36 RCS INSTRUMENTATION & REACTOR REGULATING SYSTEMS channel fails, the switch is used to select the controlling (operable)

DETAILED SYSTEM DESCRIPTION channel (X or Y). NOTE: When control via one channel is selected, the opposite channel is defeated.

2.125 The manual/automatic transfer station and bias station, HIC-101, receives the signal from LC-101X/Y through an isolator module (IY-101) with a time delay and a signal converter (IY-101-1) to control the position of the letdown throttle valves. The output of HIC-101 is adjusted with the bias dial so that pressurizer level may be controlled at the program setpoint; whether one, two, or three charging pumps are in operation. A detailed description of HIC-101 is in the STM for CVCS.

2.126 Signal limiter LM-101 limits the letdown flow to between 26 and 116 gpm. Both the manual and auto signals from HIC-101 are lim-ited by LM-101X. The limiter bypass switch allows letdown flow to vary from no flow to 126 gpm.

2.127 Selector switch HC-101-3 selects one or both letdown throttle valves to accept a signal from the level control system. If both valves are in service, a letdown flow rate greater than 126 gpm is possible.

2.128 The PRESSURIZER LEVEL HI-LO CHANNEL X (Y) alarm actuates at +8 percent or -6 percent level deviation. High deviation starts the backup heaters, sends a backup signal to stop the standby charging pumps and increases letdown flow. The low setpoint starts all charging pumps and decreases letdown flow. The PRES-SURIZER LEVEL LO-LO CHANNEL X (Y) alarm actuates at a level of 32% and de-energizes all pressurizer heaters.

2.129 There are four banks of backup heaters, and each bank is com-prised of two or three groups of heaters. Each bank has a control switch on CB-1/2/3 with four positions:

• OFF, de-energizes the heaters.

• RESET, allows heater reset after a trip.

• AUTO, allows heater control by the selected pressure controller and level controller.

• ON, energizes the heaters.

2.130 The backup heater control switches must be placed in RESET momentarily:

• After power to the heaters is lost.

• To energize one group of heaters in each bank during 480 volt load shed, after a time delay, or to operate that group of heaters in each bank after the load shed signal has cleared (if that group was not operated during the load shed condition). The other one or two groups of heaters in each bank are locked out during 480 Level: Basic > Intermediate > Advanced Revision - 23

• 27

SYSTEM TRAINING MANUAL VOL. 36 RCS INSTRUMENTATION & REACTOR REGULATING SYSTEMS volt load shed and cannot be operated until the load shed signal DETAILED SYSTEM DESCRIPTION has cleared, at which time they will return to their pre-load shed condition.

• To operate bank four backup heaters after the transfer switch is returned to REMOTE following LOCAL control at alternate shutdown panel AI-185.

2.131 Each backup heater control switch has four indicating lights. The green light indicates the heaters in that bank are de-energized. The amber light indicates that the control switch for that bank is in AUTO. The red light indicates that all groups of heaters in that bank are energized. The small red light indicates that only the group that is not locked out during 480 volt load shed is energized within that bank.

2.132 The ON/OFF control switches on MCC-4C1 for each group of bank four backup heaters control the heaters if the transfer switch on AI-185 is in LOCAL. This switch bypasses all normal control signals, including the low-low level heater cutout.

2.133 The proportional heater control switch on CB-1/2/3 has two posi-tions: OFF and AUTO. The heaters remain de-energized when selected to OFF. AUTO energizes the heaters through the silicon controlled rectifier (SCR proportional controller).

SCR "ON" SCR "OFF" 37.5KW 4.16 msec SCR "ON" SCR "OFF" 56KW 6.25 msec FIGURE 2-11: PROPORTIONAL HEATER SCR POWER TIMING Level: Basic > Intermediate > Advanced Revision - 23

• 28

SYSTEM TRAINING MANUAL VOL. 36 RCS INSTRUMENTATION & REACTOR REGULATING SYSTEMS Two single-pen pressure recorder controllers, PRC-103X and DETAILED SYSTEM DESCRIPTION 2.134 PRC-103Y, are located on CB-1/2/3. Channel selector switch HC-103 selects which channel controls the heaters and spray valves.

The setpoint, nominally 2100 psia, is adjustable by depressing the respective setpoint pushbutton (up/down) on each pressure con-troller. The selected controller generates an analog signal that modulates the proportional heater power and spray valve opening.

A HI-LO PRESSURE alarm actuates at 2145 psia (HI) or 2080 psia (LO).

2545 2nd CODE SAFETY RE SSU PRE 2500 1st CODE SAFETY 2450 DIVERSE SCRAM TRIP G

SIN REA 2400 DIVERSE SCRAM DEC PRE-TRIP 2350 TRIP & PORV'S OPEN 2300 HI PRESS PRETRIP 2225 FULLY OPEN SPRAY VALVE MODULATION BAND 2175 FULLY CLOSED 2145 HI PRESS ALARM 2115 PROP HTRS OFF 2100 SETPOINT 2080 LO PRESS ALARM 2075 BACKUP 2075 PROP HTRS FULLY ON HEATERS OFF ALL HEATERS 2060 BACKUP HTRS ON PRES FULLY ON SUR E INC REA TM/LP TRIP 1750 SING PPLS BLOCK PERMISSIVE 1700 LO LO ALARM and PPLS 1600 FIGURE 2-12: RCS PRESSURE CONTROL WITH 2100 PSIA SETPOINT 2.135 Low-range pressure channel P-118 provides Control Room indi-cation on CB-1/2/3 from zero to 1600 psia. The low-range pressure channel actuates a close signal to the shutdown cooling suction valves if pressure exceeds 300 psia. P-118 must be energized to allow opening of HCV-347 and HCV-348. A loss of power over-ride switch is provided to allow shutdown cooling operation during a loss of power from AI-40B. The redundant shutdown cooling valve interlock is provided by P-115, which requires power to pro-vide the interlock, so no override switch is needed.

Level: Basic > Intermediate > Advanced Revision - 23

• 29

Question 9 source question QUESTIONS REPORT for ILO EXAM BANK 07-11-20 008 The reactor tripped 20 minutes ago. The following conditions are observed:

• "PRESSURIZER PRESSURE OFF NORMAL HI-LO" channel X and Y are in alarm.

• PRC-103X (controlling channel) indicates 2160 psia and stable

• All backup heaters in AUTO and energized

• LRC-101Y (controlling channel) indicates 60% and stable

• LRC-101X indicates 43% and increasing slowly

• LI-106 indicates 28%

• Letdown flow is 26 gpm

• One charging pump is running

• Tcold indicates 533°F, Thot indicates 534°F, both are stable Select the probable cause and the action that should be taken to restore RCS pressure:

A. Low level on LRC-101X is maintaining the B/U heaters on, place the pressurizer heater cutout channel select switch in channel Y.

B. The bistable for the B/U heaters needs to be reset, place the control switches for all B/U heaters to reset and back to auto.

C. LRC-101Y has malfunctioned causing the B/U heaters to remain on, place LRC-101X in service.

D. PRC 103X has malfunctioned causing the B/U heaters to remain on, place PRC-103Y in service.

Thursday, August 02, 2012 11:14:33 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 010 The plant is operating at full power near the end of an operating cycle when a loss of feedwater without scram (ATWS) event occurs.

Which one of the following reactivity mechanisms causes a POSITIVE reactivity change during this event?

A. Fuel temperature coefficient.

B. Moderator temperature coefficient.

C. Void coefficient.

D. Boron concentration change.

Thursday, August 02, 2012 11:49:13 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 10 Rev 0 KA #: 000029 EK1.01 Tier 1 Group 1: Anticipated Transient Without Scram (ATWS)

Knowledge of the operational implications of the following concepts as they apply to the ATWS: Reactor nucleonics and thermo-hydraulics behavior Importance 2.8 / 3.1 CFR Number: 55.41(b)(1)

Fundamentals of Reactor Theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

Fort Calhoun Objective:

EXPLAIN the primary and the secondary plant response to a loss of feedwater ATWS.

Question Pedigree Bank question used in the 2007 NRC exam.

K/A Fit:

Question addresses reactivity change due to temperature changes during a loss of feedwater ATWS.

Choice A:

Correct answer. In a loss of feedwater ATWS, the moderator will heat up significantly providing negative reactivity feedback and a reduction in reactor power. The fuel temperature will decrease due to the power reduction causing a positive reactivity change.

Choice B:

Plausible because the change in moderator temperature will cause a reactivity change but incorrect because the reactivity change will be negative.

Choice C:

Plausible because a change in moderator voiding will cause a reactivity change but incorrect because the reactivity change will be negative.

Choice D:

Plausible, because an ATWS may result in automatic emergency boration due to safeguards actuation due to high containment pressure. Incorrect because the reactivity change will be negative.

KA#: 000029 EK1.01 Bank Ref #: 07-15-17 002 LP# / Objective: 0715-17 01.06 Exam Level: RO-1 Cognitive Level: HIGH Source: 2007 NRC EXAM

Reference:

LP 0715-17 Handout: NONE Thursday, August 02, 2012 11:49:13 2

IHB: 7-15-17 Revision: 6 OUTLINE OF INSTRUCTION COMMENTS II. A. 5. (continued)

The loss of feedwater without scram is the classic ATWS event since it produces the highest pressure in the RCS. Peak RCS pressure is allowed to exceed 2750 psia for this event and indeed it does!

A loss of feedwater without scram produces a very large mismatch between core heat production and secondary heat removal.

Primary temperatures and pressures increase rapidly. The pressurizer quickly fills with water and the PORVs and safety valves open and remain open for an extended period of time.

As the temperature of the water in the core increases, reactor power will decrease due to the negative MTC. The negative reactivity added by the MTC is partially offset by positive reactivity added by the doppler coefficient as the power decreases.

Eventually the power is reduced to the point that the rate of volumetric expansion of the primary fluid equals the volumetric flow out the PORVs and the pressurizer safety valves.

This is when the system pressure reaches its peak of over 4000 psia. Combustion Engineering has stated that at this pressure, the reactor vessel head bolts will stretch and water will be relieved between the reactor vessel and the head.

Figures 6 - 10 show the primary plant response to Figures 6-10 this event. Note that this is a generic CE analysis and is not specific to Fort Calhoun.

An excessive heat removal event occurring at End of Cycle will result in a greater reactivity addition than one at Beginning of cycle because the moderator temperature coefficient is more negative at EOC.

13

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 011 Given the following plant conditions:

The plant is operating at 20% power A high radiation alarm has been received on RM-054B Counts are rising on RM-057 but are below the alarm setpoint Charging flow is 40 gpm Letdown flow is 30 gpm Pressurizer pressure is 2080 psia and steady Pressurizer level is 48% and steady The are no containment radiation alarms Containment pressure is steady Containment sump level is steady Which one of the following actions will occur automatically with these plant conditions?

A. Blowdown flow from Steam Generator, RC-2B will be isolated. Blowdown flow will continue from RC-2A.

B. Blowdown flow from both Steam Generators will be isolated.

C. RCV-978, 6th Stage Extraction Isolation Valve, will close.

D. A Ventilation Isolation Actuation Signal (VIAS) will be generated Thursday, August 02, 2012 12:32:21 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 11 Rev 0 KA #: 000038 2.4.31 Tier 1 Group 1: Steam Generator Tube Rupture Knowledge of annunciator alarms, indications, or response procedures.

Importance 4.2 / 4.1 CFR Number: 55.41(b)(11)

Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Fort Calhoun Objective:

EXPLAIN the response of secondary system parameters.

Question Pedigree Bank question that was used on the 2009 NRC exam.

K/A Fit:

Question addresses expected system response to a high blowdown radiation alarm during a steam generator tube rupture.

Choice A:

Distractor: Plausible because high radiation on RM-054B will isolate blowdown from RC-2B. Incorrect because blowdown flow from RC-2A will also be isolated.

Choice B:

Correct answer: A high radiation alarm on RM-054B will close one containment isolation valve on each steam generator.

Choice C:

Distractor: Plausible because RCV-978 will close on high radiation. Incorrect because RCV-978 closes for a high radiation alarm on RM-057, condenser offgas, and RM-057 is below it's alarm setpoint Choice D:

Plausible because several high radiation alarms result in a VIAS. Incorrect because RM-054B is not one of them.

KA#: 000038 2.4.31 Bank Ref #: 07-15-33 025 LP# / Objective: 0715-33 01.02 Exam Level: RO-11 Cognitive Level: HIGH Source: BANK

Reference:

STM-33 Handout: NONE Thursday, August 02, 2012 12:32:21 2

SYSTEM TRAINING MANUAL VOL. 33 RADIATION MONITORING SYSTEM DEMINERALIZED WATER FLUSH RM-054A SL-347 SAMPLE COOLER SL-147 SL-751 FE-RE-SL-8A 054A 2510 HCV -2506A HCV-2506B SL-346 SL-348 CCW TO RAW HCV-2508 W ATER INSIDE OUTSI DE SL-231 CONTAINMENT CONTAINMENT DEMINERALIZED WATER FLUSH SAMPLE SL-146 COOLER SL-752 RE- SL-349 FE- SL-351 SL-8B 054B 2511 HCV -2507A HCV-2507B CCW TO RM-054B SL-350 WASTE HCV-2509 DISPOSAL FIGURE 2-41: STEAM GENERATOR BLOWDOWN SAMPLING FLOW

• • Location 2.246 The local ratemeters for both channels (ADM-610) are located just outside of Room 60 in Corridor 26.

• • Power Supplies 2.247 Instrument power for RM-054A ratemeters is provided by 120 VAC instrument bus AI-40C.

2.248 Instrument power for RM-054B ratemeters is provided from 120 VAC instrument bus AI-40D.

• • Instrumentation and Control 2.249 The local ratemeters for both channels (ADM-610), located just outside of Room 60 in Corridor 26, output to the Control Room ratemeters on AI-33A.

2.250 The Control Room ratemeters (ADM-600) provide an output sig-nal to the ERF computer, radiation monitor recorder RR-049A, Control Room annunciator panel A33C and the four containment blowdown isolation valves.

2.251 A high alarm on RM-054A closes the inside containment blow-down isolation valves (HCV-1387A and HCV-1388A) for both steam generators. A high alarm on RM-054B closes the outside containment blowdown isolation valves (HCV-1387B and HCV-1388B) for both steam generators.

Level: Basic > Intermediate > Advanced Revision - 18

• 65

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 012 Given the following plant conditions:

A large steam line break occurred inside containment PPLS, CPHS, SGLS and CSAS actuation all occurred Two containment spray pumps are running Containment pressure is 41 psig and lowering All containment cooling units and cooling and filtering units are operating What actions are directed by EOP-05, "Uncontrolled Heat Extraction," and EOP/AOP Floating Step F, Containment Spray Termination, and why are these actions directed?

A. Stop one containment spray pump to conserve water in the SIRWT.

B. Stop both containment spray pumps to minimize water damage to equipment inside containment.

C. Start a third containment spray pump to reduce containment pressure faster.

D. Continue to operate two containment spray pumps because containment pressure is still above 3 psig.

Wednesday, August 08, 2012 15:58:48 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 12 Rev 0 KA #: CE-E05 EK2.02 Tier 1 Group 1 Excess Steam Demand: Knowledge of the interrelations between the (Excess Steam Demand) 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.

Importance 3.7 / 4.2 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

STATE from memory the condition(s) that must be met prior to termination of Containment Spray.

Question Pedigree New question.

K/A Fit:

Question addresses operation of a heat removal system (containment spray) during an excessive steam demand event.

Choice A:

Correct answer: With containment pressure above 30 psig and below 60 psig, EOP-05 and Floating Step F direct the operators to ensure only one containment spray pump is operating. TDB-EOP-05 states that one of the objectives of this action is to reduce the demand on the SIRWT.

Choice B:

Plausible because both of the referenced procedures contain instructions for termination of all containment spray and a caution in floating step F states that containment spray may affect equipment inside containment. Incorrect because containment pressure must be less than 30 psig to stop all containment spray.

Choice C:

Plausible because floating step F directs starting a containment spray pump after spray termination if containment pressure can not be maintained below 40 psig. Incorrect because EOP-05 contains a caution that states "Do NOT run SI-3B and SI-3C at the same time."

Choice D:

Plausible because floating step H directs resetting the CPHS lockout relays when containment pressure is below 3 psig. Incorrect because floating step F directs stopping all containment spray when containment pressure is less than 30 psig.

Wednesday, August 08, 2012 15:58:48 2

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 KA#: CE-E05 EK2.02 Bank Ref #: N/A LP# / Objective: 0718-15 03.09 Exam Level: RO-10 Cognitive Level: HIGH Source: NEW

Reference:

FLOATING STEP F Handout: NONE Wednesday, August 08, 2012 15:58:48 3

TBD-EOP-05 Page 44 of 85 INSTRUCTIONS CONTINGENCY ACTIONS Throughout the cooldown and depressurization, the operator should verify that the pressurizer pressure is being maintained within the maximum P-T limits. If the maximum Pressure-Temperature limits are being violated, then the operators should take actions to attempt to restore the RCS to within the P-T limits. The operator should perform the following actions as in such a situation:

Operate main or auxiliary spray and letdown Throttle or stop HPSI pumps Throttle or stop charging flow and manually control letdown flow.

NOTES Stopping SI-3A or SI-3B will result in closure of one spray valve, HCV-344 or HCV-345 by interlock which will extend the time to RAS.

24. IFAND Containment Spray has been initiated 24.1 IF Containment pressure can not be maintained less than 60 psig, ALL of the following conditions are THEN restore full CS flow by starting satisfied: available CS Pumps, either SI-3A/B or SI-3A/C.

Two CS pumps are operating Containment pressure is less than 60 psig and NOT rising At least one VA-3A/B, Containment Air Cooler Filter System in service At least one VA-7C/D, Containment Air Cooler in service THEN perform the following:

a. Ensure only ONE CS pump is operating.

b. Ensure only ONE of the following valves is open:

HCV-344 HCV-345

c. Ensure total CS flow is at least 2300 gpm.

Continuously Applicable or Non-Sequential Step R25

TBD-EOP-05 Page 45 of 85 INSTRUCTIONS CONTINGENCY ACTIONS

25. IF CS pump(s) are operating, AND ALL of the following conditions are 25.1 IF containment pressure can not be maintained less than 40 psig, satisfied: THEN initiate Containment Spray by performing the following:

Containment pressure is less than 30 psig and stable or lowering a. Start ONE CS Pump:

Containment Spray is NOT required for Containment cooling SI-3A At least on VA-3A/B, Containment SI-3B Air Cooler filter System in service At least one VA-7C/D, Containment b. Open ONE Containment Spray Air Cooler Valve:

HCV-344 THEN stop Containment Spray by HCV-345 performing Floating Step F, Containment Spray Termination.

c. Ensure total CS flow is at least 2300 gpm.

EPG Step: 23 Deviation 1) The EOP specifies containment design pressure is met. The EPG uses CSAS initiation pressure.

2) The EOP uses a two step approach to securing Containment Spray. The EPG utilizes a single step.

3) The EOP utilizes Floating Steps to perform Containment Spray Termination.

4) A note is used to warn operators of pump run out, The EPG does not have this note.

The note prior to step 24 is plant specific and is included to ensure the actuation of the interlock is not unexpected by the operators. This justifies deviation 3.

The intent of this step is to secure unneeded Containment Spray Pumps as early as practical after it has been confirmed that they have performed their safety function of pressure control.

The overall objective is to:

Reduce the demand on the SIRWT Reduce the flow rate to the sump during Containment recirculation Reduce the pressure differential across the sump screens if there is a buildup of debris Continuously Applicable or Non-Sequential Step R25

TBD-EOP-05 Page 46 of 85 INSTRUCTIONS CONTINGENCY ACTIONS The Containment Spray System is actuated automatically to mitigate an Uncontrolled Heat Extraction by a combination of CSAS and SGLS. This automatic action is addressed earlier in this section, along with other steps being taken to mitigate a potential UHE. Once minimum safeguards are verified and Containment pressure is verified within design limits, one train of Spray is secured. The design basis for this even only requires one train of CACFS/CAC in-service with containment spray to mitigate the event. Analysis shows that the Containment Vent Fans will maintain Containment pressure and temperature control. Failure of half of the Containment Vent Fan trains does not challenge the Containment pressure and temperature safety functions. This step allows operations to control the spray system to conserve SIRWT water. This meets the intent of the EPG. This justifies deviation 1.

Containment Spray is used as a backup to the Containment Air Cooling and Filtering System for beyond-design basis failure of the CACFS during a LOCA. The steps to secure Containment Spray ensure the energy levels in containment are low enough not to cause equipment damage.

The securing of Containment Spray will allow less wetting of containment components, and debris transport. Early termination is important to prevent RAS from initiating unnecessarily.

The use of two steps for Containment Spray termination provides closer control of the Spray System during restoration. Containment Spray is secured after the affected Steam Generator has blown down and containment pressure has lowered below 30 psig in order to conserve SIRWT water and preclude a RAS condition. Containment pressure and temperature is expected to rise after Containment Spray termination; therefore, spray may have to be cycled on and off to control Containment pressure (R10). This justifies deviation 2.

Floating steps are used to perform complicated actions that can be accomplished at any time during the event. The detail for restoration is placed in the floating steps to allow this. This justifies deviation 3.

Analysis shows that the Containment Vent Fans will maintain Containment pressure and temperature control. Failure of half of the Containment Vent Fan trains does not challenge the Containment pressure and temperature safety functions.

26. IF offsite power has been lost AND resources permit, THEN restore power to plant distribution and station loads by performing ANY or all of the following attachments:

Attachments 17, Restoring Off-Site Power to Bus 1A3 Attachment 18, Restoring Off-Site Power to Bus 1A4 Attachment 21, Energizing Buses 1A1 and 1A2 from Off-Site Power Continuously Applicable or Non-Sequential Step R25

EOP-05 Page 8 of 58 INSTRUCTIONS CONTINGENCY ACTIONS

5. (continued) 5.1 (continued)

3) IF two Containment Spray Pumps are operating, THEN open BOTH of the following Containment Spray Valves:

HCV-344 HCV-345

4) Ensure that total Containment Spray flow is greater than or equal to 2300 gpm.

(continue)

CAUTION Do NOT run SI-3B and SI-3C at the same time.

a.2 IF no Containment Spray Pumps are running OR total Containment Spray flow is less than 2300 gpm, THEN restore Containment Spray flow by performing the following:

(continue)

Continuously Applicable or Non-Sequential Step R25

TBD-EOP/AOP FLOATING STEPS Page 42 of 126 2.0 FLOATING STEPS F. CONTAINMENT SPRAY TERMINATION INSTRUCTIONS CONTINGENCY ACTIONS NOTES

1. Stopping SI-3A or SI-3B will result in closure of one spray valve, HCV-344 or HCV-345 by interlock which will extend the time to RAS.

2. A small rise in Containment temperature and pressure is an expected response during the initiation of RAS.

CAUTION Containment Spray may affect proper operation of RCPs, non-qualified equipment, Containment Sump, and instrumentation inside the Containment. When the termination criterion is satisfied, Containment Spray should be promptly secured.

1. IF Containment Spray has been 1.1 IF Containment pressure can not be initiated AND maintained less than 60 psig, ALL of the following conditions are THEN restore full CS flow by starting satisfied: available CS Pumps, either SI-3A/B or SI-3A/C Two CS pumps are operating Containment pressure is less than 60 psig and NOT rising At least one VA-3A/B inservice At least one VA-7C/D inservice THEN perform the following:

a. Ensure only ONE CS pump is operating.

b. Ensure only ONE of the following valves is open:

HCV-344 HCV-345 R3

TBD-EOP/AOP FLOATING STEPS Page 43 of 126 2.0 FLOATING STEPS F. CONTAINMENT SPRAY TERMINATION INSTRUCTIONS CONTINGENCY ACTIONS

c. Ensure total CS flow is at least 2300 gpm.

NOTE Terminating Containment Spray prior to resetting actuation relays will require increased monitoring of containment parameters.

2. IF CS pump(s) are operating, 2.1 IF containment pressure can not be AND ALL of the following conditions maintained less than 40 psig, are satisfied: THEN initiate Containment Spray by performing the following:

Containment pressure is less than 30 psig and stable or lowering a. Start ONE CS Pump:

Containment Spray is NOT required for Containment cooling SI-3A At least one VA-3A/B inservice SI-3B At least one VA-7C/D inservice

b. Open ONE Containment Spray THEN terminate Containment Spray Valve:

flow by performing the following:

HCV-344

a. Place the hand controller(s) for HCV-345 the open valve(s) in "OPEN":

c. Ensure total CS flow is at least HIC-344 2300 gpm.

HIC-345

b. Place all CS pumps, SI-3A/B/C in "PULL TO LOCK"

c. Close BOTH Containment c.1 IF either Containment Spray Spray Valves: Valve fails to close, THEN manually close the open HCV-344 valve(s) using the handwheel HCV-345 (Room 59).

R3

EOP/AOP FLOATING STEPS Page 49 of 115 2.0 FLOATING STEPS H. RESET OF ENGINEERED SAFEGUARDS INSTRUCTIONS CONTINGENCY ACTIONS CAUTION Do not perform this floating step if CSAS relays are tripped to prevent damage to VIAS relays.

2. IF CPHS has actuated AND Containment pressure is less than or equal to 3.0 psig, THEN reset CPHS by performing the following steps:

NOTE Resetting CPHS Lockout Relays may reset SGIS. HCV-1105 and HCV-1106 may reopen.

a. Reset BOTH of the following relays:

86A/CPHS 86B/CPHS

b. Reset BOTH of the following relays:

86A1/CPHS 86B1/CPHS (continue)

R3

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 013 Given the following plant conditions:

A loss of all feedwater event has occurred The conditions requiring Once-through-cooling have been met.

All 4160 volt buses are energized All Reactor Coolant pumps have been tripped HCV-150 and HCV-151," PORV Block Valves," are open All pressurizer heaters have been deenergized PPLS has been initiated using the test switches All HPSI loop injection valves are open Which one of the following describes the remaining steps required to establish once-through cooling per EOP-20, HR-4?

A. Ensure two HPSI pumps are running, open one PORV B. Ensure three HPSI pumps are running, open one PORV C. Ensure two HPSI pumps are running, open both PORVs D. Ensure three HPSI pumps are running, open both PORVs Thursday, August 09, 2012 14:07:39 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 13 Rev 0 KA #: 000054 2.1.20 Tier 1 Group 1: Loss of Main Feedwater Ability to interpret and execute procedure steps.

Importance 4.6 / 4.6 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

EXPLAIN the process of once through cooling.

Question Pedigree New question.

K/A Fit:

Question addresses executing procedure steps required to initiate once-through-cooling following a total loss of feedwater.

Choice A:

Distractor: Plausible because using only one PORV will conserve RCS inventory.

Incorrect because two PORVs are needed for adequate heat removal.

Choice B:

Distractor: Plausible because using only one PORV will conserve RCS inventory.

Incorrect because two PORVs are needed for adequate heat removal.

Choice C:

Correct answer. Two PORVs are needed for adequate heat removal and the procedures direct using two HPSI pumps.

Choice D:

Distractor: Plausible because two PORVs are needed for adequate heat removal and three HPSI pumps would increase injection flow. Incorrect because the procedures direct using two HPSI pumps.

KA#: 000054 2.1.20 Bank Ref #: N/A LP# / Objective: 0715-17 02.04 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

EOP-20. HR-4 Handout: NONE Thursday, August 02, 2012 13:19:32 2

EOP-20 Page 391 of 527 16.0 RCS AND CORE HEAT REMOVAL SAFETY FUNCTION: RCS and Core Heat Removal SUCCESS PATH: Once-Through-Cooling: HR-4 RESOURCE TREE: Tree E HR-4 INSTRUCTIONS CONTINGENCY ACTIONS NOTE The action in this success path are intended to establish once-through-cooling in which both PORVs are open and at least two HPSI pumps are running. This is the desired mode of operation, however, other configurations may also be successful.

CAUTIONS

1. This is a heat removal path of last resort, it should only be utilized when S/G heat removal is not possible.

2. Opening PORVs may cause RCS pressure to drop.

3. Do not allow Diesel Generator loads to exceed power and current rating limits.

1. IF both Vital 4160 V Buses are 1.1 IF once-through-cooling can NOT be energized, established because a Vital 4160 V THEN establish once-through-cooling bus is deenergized, by performing the following: THEN determine the appropriate method of recovery by performing

a. Stop all RCPs. step a, b, c or d:

(continue) (continue)

Continuously Applicable or Non-Sequential Step R25

EOP-20 Page 392 of 527 16.0 RCS AND CORE HEAT REMOVAL HR-4 INSTRUCTIONS CONTINGENCY ACTIONS

1. (continued) 1.1 (continued)

b. Deenergize all PZR Heaters. a. IF 1A3 is deenergized, AND BOTH of the PORV Block

c. Initiate PPLS by placing the Valves are open:

following switches in TEST:

HCV-150 HCV-151 86A/PPLS TEST SWITCH 86B/PPLS TEST SWITCH THEN GO TO Step 2.

d. Ensure ALL of the available pumps have started: b. IF 1A4 is deenergized, AND BOTH of the PORV Block Either HPSI Pumps, SI-2A/B or SI-2B/C Valves are open:

Charging Pumps, CH-1A/B/C HCV-150

e. Ensure all of the HPSI Loop HCV-151 Injection Valves are open.

THEN GO TO Step 12.

f. Ensure BOTH of the PORV Block Valves are open:

HCV-150 HCV-151 (continue)

(continue)

Continuously Applicable or Non-Sequential Step R25

EOP-20 Page 393 of 527 16.0 RCS AND CORE HEAT REMOVAL HR-4 INSTRUCTIONS CONTINGENCY ACTIONS

1. (continued) 1.1 (continued)

g. Open the PORVs. c. IF 1A3 is deenergized, AND ANY of the PORV Block Time: Valves are closed:

HCV-150

h. GO TO Step 60. HCV-151 THEN GO TO Step 21.

d. IF 1A4 is deenergized, AND ANY of the PORV Block Valves are closed:

HCV-150 HCV-151 THEN GO TO Step 42.

Continuously Applicable or Non-Sequential Step R25

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 014 AC-3B is the running CCW pump when a Loss of Offsite Power occurs.

Diesel Generator D-1 fails to start.

Diesel Generator D-2 operates as designed.

The event does not result in SIAS actuation.

What operator action is required to restore CCW?

A. Verify AC-3B automatically starts.

B. Manually start AC-3B C. Verify AC-3A and AC-3C automatically start D. Manually start AC-3A or AC-3C Thursday, August 02, 2012 14:06:14 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 14 Rev 0 KA #: 000056 AA1.09 Tier 1 Group 1: Loss of Off-Site Power Ability to operate and / or monitor the following as they apply to the Loss of Offsite Power: CCW pump Importance 3.3 / 3.3 CFR Number: 55.41(b)(8)

Components, capacity, and functions of emergency systems.

Fort Calhoun Objective:

EXPLAIN standby operation of CCW pumps in terms of switch positions and automatic actions.

Question Pedigree Modified question. The source question was used on the 2002 NRC exam.

K/A Fit:

The question addresses required operation to restore a CCW pump following a loss of offsite power.

Choice A:

Distractor: Plausible because AC-3B was the running pump. Incorrect because the control switch will be in after start and will not receive an autostart signal.

Choice B:

Correct Answer: AC-3B will not receive an autostart signal and must be manually started. Note: Fort Calhoun sequencers will not operate without a SIAS actuation.

Choice C:

Distractor: Plausible because AC-3A and AC-3C would automatically start if AC-3B trips. Incorrect because there will not be any power to them with bus 1A3 deenergized due to the D-1 failure, unless 480 volt buses were manually crosstied.

Choice D:

Plausible because they could be started if there were power to bus 1A3. Incorrect because the stem states the D-1 failed to start.

KA#: 000056 AA1.09 Bank Ref #: 07-11-06 061 LP# / Objective: 0711-06 01.04 Exam Level: RO-8 Cognitive Level: HIGH Source: MODIFIED

Reference:

STM 8 Handout: NONE Thursday, August 02, 2012 14:06:14 2

EOP-02 Page 7 of 43 INSTRUCTIONS CONTINGENCY ACTIONS

3. IMPLEMENT the Emergency Plan.

Time:

NOTE Floating Step BB, Minimizing DC Loads, requires operator action within 15 minutes of loss of either battery charger.

4. Monitor the Floating Steps.

5. Verify either Vital 4160 V Bus is 5.1 IF both Vital 4160 V Buses are energized. deenergized, THEN GO TO EOP-07, Station Blackout.

6. Attempt to start ALL of the following equipment:

• ONE CCW Pump, AC-3A/B/C

• ONE Raw Water Pump, AC-10A/B/C/D

• ONE Bearing Water Pump, AC-9A/B

• ONE Air Compressor, CA-1A/B/C

• ONE Charging Pump, CH-1A/B/C

• ONE AFW Pump, FW-6, FW-10, or FW-54 Continuously Applicable or Non-Sequential Step R18

SYSTEM TRAINING MANUAL VOL. 8 COMPONENT COOLING WATER SYSTEM

• • Design/Specifications 2.4 CCW pumps are single-stage, horizontal centrifugal pumps. CCW pumps are driven by 250 hp motors powered from the 480 VAC buses. Flowrate with one pump running during normal operation is 5400 gpm. After a design basis accident, with two pumps assumed running, each pump has a design flow of 3750 gpm. In reality, all three pumps will start if available.

2.5 During normal operation, one pump provides approximately 150 ft.

of total developed head (about 65 psid). With a nominal 40 psig nitrogen overpressure in the CCW surge tank, this results in a nor-mal system pressure of approximately 90 psig. After a design basis accident, the three pumps provide approximately 200 ft. total developed head (about 86 psid).

2.6 CCW temperature is maintained between 70¬xF and 90¬xF. If raw water temperature is low, a portion of the CCW is bypassed around the heat exchangers, or CCW flow is established through a CCW heat exchanger with the raw water secured. After a design basis accident, CCW temperatures may be as high as 120¬xF.

2.7 Design pressure of the CCW pumps is 150 psig with a design tem-perature of 200¬xF.

• • Location 2.8 The CCW pumps AC-3A/B/C are on the 1025 foot elevation of the Auxiliary Building at the east end of Room 69, just west of the surge tank.

• • Power Supplies 2.9 The 480 VAC power supplies for the CCW pumps are as follows:

• AC-3A from bus 1B3B

• AC-3B from bus 1B4A

• AC-3C from bus 1B3C-4C

• • Instrumentation and Control 2.10 The following is a detailed description of the Instrumentation and Controls for the Component Cooling Water Pumps.

• • • Local 2.11 There are three breaker status lights (green, amber and red) at the respective motor control center (MCC) in the switchgear room for each CCW pump breaker. The status lights are located just above the 69-permissive control switch. The green and red lights indicate breaker position (green-open and red-closed), while the amber Level: Basic > Intermediate > Advanced Revision - 39

• 8

SYSTEM TRAINING MANUAL VOL. 8 COMPONENT COOLING WATER SYSTEM light indicates breaker availability. The amber light will be lit if control power is available to the closing circuit and the breaker is racked in.

2.12 All three breaker status lights will turn off if the 69-permissive switch at the MCC is placed in PULL OUT or if the breaker is racked out.

• • • Remote 2.13 CCW pumps are operated from the control room on CB-1/2/3.

2.14 Each CCW pump has an ammeter, pistol-grip control switch and breaker status lights (green, white and red) on CB-1/2/3. Control switch positions are PULL TO LOCK, STOP and START. After operation, the control switch spring returns to the center position (AFTER STOP/START). The breaker status lights on CB-1/2/3 provide the following indications:

• Green light - is lit if the breaker is open and control power is available. The green light turns off if the 69-premissive switch is in PULL TO LOCK because it removes control power from the closing circuit. NOTE: control power will be energized at the breaker until the control power fuses are pulled.

• White light - is lit if there is disagreement between the control switch position and the local breaker position. For example, if the breaker trips open and the switch is in AFTER CLOSE.

• Red light - is lit if the breaker is closed. The red light is also used to verify trip circuit continuity during calibration testing of the breaker in the TEST position. This ensures that the trip cir-cuit, if needed, will function properly.

2.15 The CCW pump is disabled by turning its control switch to the PULL-TO-LOCK position and pulling out (indenting) the handle.

The PULL-TO-LOCK position actuates an alarm and turns off the breaker status lights at CB-1/2/3 and the respective MCC.

2.16 The CCW pump is aligned for standby operation (automatic start) by ensuring that the breaker is racked in, the 69-permissive switch is in AFTER CLOSE (red flag), the control switch on CB-1/2/3 is in AFTER STOP (green flag), and the green status light is lit.

• • • Interlocks 2.17 The standby pumps automatically start if the running pump trips.

Safeguards load sequencers start all CCW pumps.

2.18 Pressure switches PCS-412 and PCS-413 are located on the CCW pump discharge header in Room 69 and provide an interlock with containment isolation valves HCV-438A/B/C/D that supply CCW to the CEDM coolers and the Reactor Coolant Pump seal and lube oil coolers.

Level: Basic > Intermediate > Advanced Revision - 39

• 9

STM-ED - Volume 14 Figures Information Use Page 11 of 66 Electrical Distribution System Rev. 42 Figure 1 480 VAC Distribution

STM-ED - Volume 14 Figures Information Use Page 8 of 66 Electrical Distribution System Rev. 42 Figure 1 Station Supply/4160 Distribution

Question 14 Source Question QUESTIONS REPORT for ILO EXAM BANK 07-11-06 061 AC-3B is the running CCW pump when a loss of offsite power occurs. Both Diesel generators operate as designed. The event does not result in SIAS actuation. Assuming no operator action, what will be the status of the CCW pumps 60 seconds after the diesels start?

A. Only AC-3B will be running B. AC-3A and AC-3C will be running C. All three CCW pumps will be running D. No CCW pumps will be running Thursday, August 02, 2012 14:04:14 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 015 Given the following plant conditions:

The plant was operating at full power Power was lost to Instrument bus AI-40A AOP-16, LOSS OF INSTRUMENT BUS POWER, was entered AOP-16,Section II, "Loss of Instrument Bus AI-40A", Step 14 states "Consider closing BOTH of the PORV (Power-Operated Relief Valves) Block Valves" What is the purpose of this step?

A. An additional vital instrument bus failure will result in opening ONE of the PORVs.

B. An additional vital instrument bus failure will result in opening BOTH of the PORVs.

C. An additional vital instrument bus failure will result in the inability to close ONE of the PORV Block Valves.

D. An additional vital instrument bus failure will result in the inability to close BOTH of the PORV Block Valves.

Wednesday, August 08, 2012 15:59:26 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 15 Rev 0 KA #: 000057 AK3.01 Tier 1 Group 1: Loss of Vital AC Electrical Instrument Bus Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus: Actions contained in EOP for loss of vital ac electrical instrument bus Importance 4.1 / 4.4 CFR Number 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective Describe how the plant responds to a loss of instrument bus power in terms of how specific equipment is affected and how it affects overall plant operation and reliability.

Question Pedigree Bank question used on the 2009 NRC exam.

K/A Fit:

Question addresses reason for procedural action taken due to a loss of an instrument bus.

Choice A:

Distractor: Plausible because Applicant could confuse with a failure of a pressure instrument while LTOP is in service which would result in opening one PORV. Incorrect because 2/4 logic opens both PORVs.

Choice B:

Correct answer:Deenergizing 2 vital instrument buses results in a 2/4 logic that opens both PORVs.

Choice C:

Distractor: Plausible if the Applicant does not fully understand the PORV logic. Incorrect because a loss of two vital instrument buses will not affect operation of the PORV block valves.

Choice D:

Distractor: Plausible if the Applicant does not fully understand the PORV logic. Incorrect because a loss of two vital instrument buses will not affect operation of the PORV block valves.

KA#: 000057 AK3.01 Bank Ref #: 07-17-16 LP# / Objective: 0717-16 01.02 Exam Level: RO-10 Cognitive Level: HIGH Source: BANK

Reference:

TBD-AOP-16 Handout: NONE Wednesday, August 08, 2012 15:59:26 2

TBD-AOP-16 Page 17 of 143 Section II - Loss of Instrument Bus AI-40A INSTRUCTIONS CONTINGENCY ACTIONS

12. Consider closing BOTH of the PORV Block Valves:

HCV-150 HCV-151 One additional channel trip is needed to actuate the PORVs. The operator is given a note explaining this fact and an instruction to consider closing the PORV Block Valves. This will alert the operator to the status of the PORV system.

NOTE Upon loss of Instrument Bus A, ALL of the following instrumentation or equipment associated with the Core Heat Removal Safety Function are inoperable:

"SUBCOOLED MARGIN MONITOR A-168" "RC LOOP TEMPERATURES LOOP 1A "T-COLD" A/TI-112C" "RC LOOP TEMPERATURES LOOP 1 "T-HOT" A/TI-112H" "RC LOOP TEMPERATURES LOOP 2A "T-COLD" A/TI-122C" "RC LOOP TEMPERATURES LOOP 2 "T-HOT" A/TI-122H" "SHTDN HT EXCH AC-4A OUTLET VALVE CNTRLR HCV-484"

13. Verify at least one RCP is running. 14.1 IF all RCPs are stopped AND SDC is in service, THEN perform the following:

a. IF AC-4A, Shutdown Heat Exchanger is in service, THEN perform the following:

1) Ensure HCV-485 is closed.

2) Place HCV-481 control switch in "OPEN".

3) Place HCV-480 control switch in "CLOSE".

4) Throttle HCV-485 to maintain RCS temperature.

R18

SYSTEM TRAINING MANUAL VOL. 38 REACTOR PROTECTIVE SYSTEM AND DIVERSE SCRAM SYSTEM

• When the pretrip setpoint is reached, an annunciator will alarm DETAILED SYSTEM DESCRIPTION on the main control board and a white light will be energized on the respective trip unit.

• When the trip setpoint is reached, a another annunciator will alarm on the main control board and a red light will be ener-gized on the respective trip unit. The three red trip relay lights will be energized.

PRETRIP VARIABLE ALARM TO PRETRIP ALARM SETPOINT RELAY MANUAL PRETRIP SETPOINT INDICATOR WHITE LOW PRETRIP LEVEL TRIP TRIP PRETRIP RED TRIP COMPARATOR INDICATOR RED TRIP RELAY LITES (2 OUT OF 4 NSSS PARAMETER TRIP LOGIC MATRIX)

INPUT SIGNAL RELAY NO. 1 TRIP SETPNT PRETRIP TRIP TRIP RELAY NO. 2 ALARM COMPARATOR SETPOINT CONTACT CLOSES LITE TEST CABLE RECEPTACLE MANUAL WHEN TU TRIP (FOR TESTING)

SETPOINT BYPASSED RELAY NO. 3

+15 VDC TRIP VARIABLE ALARM SETPOINT RELAY TO TRIP ALARM AND SEQUENCE OF EVENTS RECORDER

• Used for Operational Bypass of trip units 2, 3, 6 & 9 only FIGURE 2-3: TYPICAL BISTABLE TRIP UNIT BLOCK DIAGRAM

• • Failure Modes 2.27 The trip units are reliable, qualified components that are required to generate a reactor trip signal when parameters are exceeded, and prevent a reactor trip signal when spurious faults are experienced.

The removal of any trip unit from the cubicle or loss of power to any trip unit will place that module in a tripped condition. Any one channel of each trip parameter may be placed in bypass for testing or maintenance.

High Power Level Trip Unit TU-1 2.28 The following is a detailed description of the high power level trip unit, TU-1.

• • Function 2.29 The high power level trip protects the core against fast reactivity excursions that are too fast for the high pressure or TM/LP trips.

Level: Basic > Intermediate > Advanced Revision - 20

• 16

SYSTEM TRAINING MANUAL VOL. 38 REACTOR PROTECTIVE SYSTEM AND DIVERSE SCRAM SYSTEM If a channel is put in BYPASS, it also bypasses PORV opening.

DETAILED SYSTEM DESCRIPTION 2.133 For example, if A channel high pressure is bypassed, it would take 2 of the other 3 (B, C, D) high pressure channels going into trip to open the PORVs.

2.134 The high pressure bistable comparators have a fixed pretrip at 2300 psia and trip at 2350 psia.

• • • Interlocks 2.135 Two out of four channels at 2350 psia initiates a reactor trip and opens the two pressurizer power operated relief valves.

• • • Alarms and Indications 2.136 The HIGH PRESSURIZER PRESSURE CHANNEL PRETRIP alarm on CB-4 actuates at 2300 psia (1 of 4 channels). The HIGH PRESSURIZER PRESSURE CHANNEL TRIP alarm actuates at 2350 psia on CB-4. The REACTOR TRIP alarm on CB-4 will also actuate at 2350 psia (2 of 4 channels).

• • Failure Modes 2.137 Loss of power to a channel produces a PORV open signal that cannot be bypassed, which results in a one-of-three logic for the remaining channels.

Thermal Margin/Low Pressure (TM/LP) Trip Unit TU-9 2.138 The following is a detailed description of the thermal margin/low pressure (TM/LP) trip unit, TU-9.

• • Function 2.139 The TM/LP trip ensures operation is within the DNBR safety limit stated in the Technical Specifications. This limit states that reactor power level shall not exceed the allowable limit for a given pressurizer pressure and cold leg temperature as shown on Figure 1-1 in the Technical Specifications (repeated below as the Thermal Margin/Low Pressure Safety curves, Figure 2-17). This figure rep-resents the values of reactor thermal power, RCS pressure, and cold leg temperature, where the value of DNBR is at the limiting value for the fuel.

2.140 To prevent exceeding the limits of Figure 2-17, a more restrictive set of curves is used for the TM/LP trip, as shown in the Thermal Margin/Low Pressure Limited Safety System Setting (LSSS) curve, shown in Figure 2-18. NOTE: TM/LP LSSS limits are sub-ject to change after each refueling cycle. The TM/LP limits of Fig-ure 2-18 are therefore only for illustration purposes. Refer to the COLR (TDB VI), Figure 1 for current values.

Level: Basic > Intermediate > Advanced Revision - 20

• 37

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 016 Why does AOP-16,Section VIII, "Loss of DC Bus 1," direct you to use HCV-1040, Atmospheric Dump Valve, for heat removal?

A. Control Power has been lost to MSIVs and MSIV bypass valves on both steam generators and they have have failed closed.

B. Control Power is lost to MS-291, MS-292 and the steam dump and bypass valves.

C. Control Power is lost to MS-291 and the steam dump and bypass valves. MSIV, HCV-1042A, has failed closed.

D. Control Power is lost to MS-292 and the steam dump and bypass valves. MSIV, HCV-1041A, has failed closed.

Wednesday, August 08, 2012 15:59:59 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 16 Rev 0 KA #: 000058 AK3.02 Tier 1 Group 1:Loss of DC Power Knowledge of the reasons for the following responses as they apply to the Loss of DC Power: Actions contained in EOP for loss of dc power Importance 4.0 / 4.2 CFR Number 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective Describe the major recovery actions of this AOP.

Question Pedigree New question.

K/A Fit:

Question addresses reason for steaming RC-2A using HCV-1040 following a loss of DC bus number 1 per AOP-16,Section VIII.

Choice A:

Distractor: Plausible if the Applicant believes both MSIVs will close on loss of DC bus 1 as they will on a loss of DC bus #2 and does not realize HCV-1040 is downstream of the MSIVs. Incorrect because MSIV HCV-1041A does not fail closed with a loss of DC bus #1.

Choice B:

Distractor: Plausible because Control power has been lost to MS-291 and the steam dump and bypass valves. Incorrect because control power is still available to MS-292.

Choice C:

Correct answer: Control power has been lost to MS-291 and the steam dump and bypass valves. S/G RC-2A can be steamed through HCV-1040 because MSIV HCV-1041A is still open.

Choice D:

Distractor: Plausible because control power will be lost to one S/G air operated safety valve, one MSIV and all steam dump and bypass valves. Incorrect because MS-292 and HCV-1041A still have control power.

KA#: 000058 AK3.02 Bank Ref #: N/A LP# / Objective: 0717-16 01.03 Exam Level: RO-10 Cognitive Level: HIGH Source: NEW

Reference:

TBD-AOP-16 Handout: NONE Wednesday, August 08, 2012 15:59:59 2

TBD-AOP-16 Page 105 of 143 Section VIII - Loss of DC Bus 1 INSTRUCTIONS CONTINGENCY ACTIONS NOTE Upon loss of DC Bus 1, ALL of the following instrumentation or equipment associated with the RCS Heat Removal Safety Function is affected as follows:

Control power to FW-4A, Feed Pump, is lost Control power to FW-5A, Heater Drain Pump, is lost Control power to FW-2A, Condensate Pump, is lost Control power to MS-291, Air Assisted Main Steam Safety Valve, is lost Power to Steam Dump and Bypass Valves is lost Control power is lost and SGIS closes HCV-1105, FCV-1101 Bypass Valve Control power to AFW Isolation Valves, HCV-1107A and HCV-1108A, is lost SGIS closes HCV-1042A, MSIV SGIS closes HCV-1103, Feed Reg Block Valves SGIS closes, HCV-1385, Feed Header Isolation Valve

21. Feed S/Gs with AFW to the AFW 21.1 IF a feedpath using AFW is NOT Nozzles to maintain S/G levels available, 35-85% NR (73-94% WR): THEN feed S/Gs to maintain S/G levels 35-85% NR (73-94% WR) by performing

a. a. Place the 43/FW Switch in the following:

"OFF".

a. Verify at least one of the following are running:

FW-4A/B/C FW-54

b. IF a SGIS has actuated, THEN override SGIS by placing the following affected valves SGIS Override Switches in "OVERRIDE":

OR/HC-1386 OR/HC-1385 OR/HC-1105 OR/HC-1106

c. Open the Feed Header Isolation Valves:

HCV-1386 HCV-1385 R18

TBD-AOP-16 Page 106 of 143 Section VIII - Loss of DC Bus 1 INSTRUCTIONS CONTINGENCY ACTIONS

b. Start at least one of the following d. Manually control EITHER of the AFW Pumps: following valves:

FW-10 HCV-1105 FW-6 HCV-1106

c. Verify HCV-1384, FW/AFW Header e. (LOCAL) Manually control EITHER Cross-Connect Valve is closed. of the following Valves (Room 81):

HCV-1105 HCV-1106

d. Open BOTH of the following AFW Isolation Valves:

HCV-1107A HCV-1108A

e. Manually control BOTH of the following AFW Isolation Valves:

HCV-1107B HCV-1108B With the loss of control power to normal feed paths to RC-2A the option to feed the Steam Generators through the Auxiliary Feedwater Nozzles is a preferred flow path. The valves HCV-1107A and HCV-1108A lose control power and fail open, this will require open verification using flow indication. Feedwater addition to both steam generators using the normal feed path is provided as a contingency, dependent on equipment availability. Local and remote operation of both valves is provided to allow flexibility in flow path selection.

The Steam Generator design and subsequent testing has removed the water hammer issue while feeding via the AFW nozzles. However, normal precautions should be used while feeding.

22. Steam RC-2A to control RCS 22.1 IF HCV-1040 is inoperable, temperature by performing the following: THEN control RCS temperature using MS-292, Air Assisted Main Steam Safety

a. Open MS-164, "MAIN STEAM Valve.

LINE "A" STEAM DUMP TO ATMOSPHERE ISOLATION VALVE" (Room 81).

R18

TBD-AOP-16 Page 107 of 143 Section VIII - Loss of DC Bus 1 INSTRUCTIONS CONTINGENCY ACTIONS

b. Ensure MS-407, "ATMOSPHERIC DUMP VALVE HCV-1040 2nd ISOLATION", is open (Room 81).

c. Open HCV-1040, Atmospheric Dump Valve.

Loss of control power to MS-291, PCV-910, and HIC-909 requires the use of HCV-1040 or MS-292 to control RCS Temperature.

NOTE Upon loss of DC Bus 1, ALL of the following equipment associated with the Containment Integrity Safety Function is affected as follows:

PCV-742E and PCV-742G, RM-050/051 Sample Valves, fail closed RM-050 and RM-051, Containment Atmosphere Radiation Monitors, are inoperable

23. Terminate all radioactive releases.

All radioactive releases are stopped until the status of effluent monitoring can be determined.

24. Ensure the RM-050/RM-051 Sample Pump is stopped.

RM-050/51 Sample Pump is verified stopped to prevent possible damage, since its sample lines are isolated.

25. Confirm Containment integrity by performing the following:

a. Check for no unexpected rise in Containment Sump level.

b. Check for no Containment Area Radiation Monitor alarms.

R18

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 017 What effect will a total loss of instrument air header pressure have on SIRWT level indication?

Assume that there is no actual change in SIRWT level and that the loss of pressure is of short enough duration that local air accumulators maintain pressure.

A. Low level will not be indicated, low level alarms will not be received, STLS actuation will not occur.

B. Low level will be indicated, low level alarms will be received, STLS actuation will not occur.

C. Low level will be indicated, low level alarms will not be received, STLS actuation will not occur.

D. Low level will be indicated, low level alarms will be received, STLS actuation will occur.

Wednesday, August 08, 2012 16:00:44 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 17 Rev 0 KA #: 000065 AA2.08 Tier 1 Group 1: Loss of Instrument Air Ability to determine and interpret the following as they apply to the Loss of Instrument Air: Failure modes of air-operated equipment Importance 2.9 / 3.3 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Describe how the plant responds to a loss of instrument air in terms of how specific equipment is affected and how it affects overall plant operation and reliability.

Question Pedigree Bank question used on 2005 NRC exam.

K/A Fit:

Question addresses failure modes of SIRWT air operated instrumentation (bubblers) following a loss of instrument air.

Choice A:

Distractor: Plausible if Applicant believes that all SIRWT level instruments have accumulators. Incorrect because only LC-383A/B/C/D, used for STLS actuation, have accumulators.

Choice B:

Correct answer. LT-381 and LT-382 provide level indication and alarm and do not have accumulators. LC-383A/B/C/D provide the STLS signal and they do have accumulators.

Choice C:

Distractor: Plausible if Applicant believes level indication and STLS actuation come from the same channels that are provided with accumulators. Incorrect because the level indication channels do not use accumulators.

Choice D:

Distractor: Plausible if the Applicant does not realize that the STLS instruments have accumulators. Incorrect because they do.

KA#: 000065 AA2.08 Bank Ref #: 07-17-17 002 LP# / Objective: 0717-17 01.02 Exam Level: RO-7 Cognitive Level: HIGH Source: BANK

Reference:

AOP-17 Handout: NONE Wednesday, August 08, 2012 16:00:44 2

AOP-17 Page 24 of 46 Attachment B Failure Position of Valves in Auxiliary Building - Controlled Area NOTE This attachment is not a complete listing of all air-operated valves in the controlled area of the Auxiliary Building.

1. CS, HPSI and LPSI Pumps suction and discharge valves all open.

2. SIRWT level indication (LT-381 and LT-382) and alarm (LIC-381 and LIC-382), in the Control Room are lost. An erroneous low level indication and alarm will result.

3. The SIRWT level controllers (LC-383A/B/C/D) which provide input to the RAS logic, are equipped with Air Accumulators and will continue to function for at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4. Spent Fuel Pool level indication is lost.

5. Concentrated Boric Acid Tanks level indication is lost.

6. Blowdown Tank level control and indication are lost.

7. All Auxiliary Building ventilation dampers close with the exceptions listed below. These valves may be manually closed.

SI Pump Rooms Inlet HCV-804A HCV-805A Charcoal Filter Outlets HCV-804C HCV-805C Spent Regenerant Room Inlet HCV-803A Charcoal Filter Outlets HCV-803C

8. The Vacuum Deaerator will overflow if DW-41 Pump is running.

R14

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 018 Given the following plant conditions:

The plant is operating at full power A grid disturbance has resulted in a loss of 161 KV to the plant Buses 1A3 and 1A4 fast transferred to 345 KV Main generator terminal voltage indicates 22,500 volts All Circulating Water Pumps are operating Which one of the following actions is directed by AOP-31, 161 KV GRID MALFUNCTIONS, section II, All 4160 V Buses Fed From 22 KV?

A. Lower the Main generator terminal voltage using the voltage regulator B. Raise the Main generator terminal voltage using the voltage regulator C. Contact System Operations and request grid voltage be lowered D. Contact System Operations and request grid voltage be raised Thursday, August 02, 2012 15:34:45 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 18 Rev 0 KA #: 000077 AA1.03 Tier 1 Group 1: Generator Voltage and Electric Grid Disturbances Ability to operate and/or monitor the following as they apply to Generator Voltage and Electric Grid Disturbances: Voltage regulator controls Importance 3.8 / 3.7 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Describe the major recovery actions of this AOP.

Question Pedigree New question.

K/A Fit:

Addresses operation of the main generator voltage regulator following a loss of 161 KV due to a grid disturbance.

Choice A:

Correct answer: AOP-31,Section II directs adjusting main generator terminal voltage less than 22,000 volts.

Choice B:

Distractor: Plausible because most electrical buses run at slightly higher voltages than designated. Incorrect because procedure directs lowering voltage below 22,000 volts.

Choice C:

Distractor: Plausible because AOP-31 section II contains direction to notify system operations and AOP-31,Section I does contain direction to request system operations adjust grid voltage for a 161 KV grid instability. Incorrect because procedure directs lowering main generator terminal voltage using the voltage regulator and 161 KV has been lost.

Choice D:

Distractor: Plausible because most electrical buses run at slightly higher voltages than designated, AOP-31 section II contains direction to notify system operations and AOP-31,Section I does contain direction to request system operations adjust grid voltage for a 161 KV grid instability. Incorrect because procedure directs lowering main generator terminal voltage using the voltage regulator and 161 KV has been lost.

KA#: 000077 AA1.03 Bank Ref #: N/A LP# / Objective: 0717-31 01.03 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

TBD-AOP-31 Handout: NONE Thursday, August 02, 2012 15:34:45 2

AOP-31 Page 15 of 28 Section II - All 4160 V Buses Fed From 22 KV 4.0 INSTRUCTIONS/CONTINGENCY ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS

1. IF Reactor power is greater than or 1.1 IF Reactor power is less than 50%,

equal to 50%, THEN align the Feedwater System by THEN ensure ALL of the following performing the following steps:

conditions are satisfied:

At least two Condensate Pumps are operating CAUTION At least two Feed Pumps are operating To protect Bus 1A1 in the event of a fault, At least two Heater Drain Pumps are FW-2A and FW-4A should not both be left operating running when the Feedwater System is realigned.

a. Ensure ALL of the following conditions are satisfied:

ONE Condensate Pump, FW-2A/B/C, is operating ONE Feed Pump, FW-4A/B/C, is operating ONE Heater Drain Pump, FW-5A/B/C, is operating

b. GO TO Step 4.

R11

AOP-31 Page 16 of 28 Section II - All 4160 V Buses Fed From 22 KV INSTRUCTIONS CONTINGENCY ACTIONS

2. Adjust Main Generator terminal voltage 2.1 IF terminal voltage is greater than or less than 22,000 volts by performing equal to 22,000 volts, the following steps: AND plant conditions permit operation of only two Circ Water Pumps,

a. Notify System Operations of the THEN stop ONE of the following Circ need to adjust voltage. Water Pumps:

b. Adjust the Voltage Regulator. CW-1B CW-1C

c. Verify terminal voltage is less than 22,000 volts.

2.2 IF terminal voltage is greater than 22,500 volts, THEN commence a power reduction to achieve less than 50% power within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> PER OP-4, Load Change and Normal Power Operation.

R11

AOP-31 Page 11 of 28 Section I - 161 KV Grid Instability INSTRUCTIONS CONTINGENCY ACTIONS

10. IF 161 KV Grid voltage exceeds 10.1 IF ANY Bus voltages are greater than 168.6 KV OR a Urgent High voltage 4390 V for greater than one hour, ERFCS alarm is received on ANY THEN reduce voltage by performing 4160 V bus, any of the following steps:

THEN verify voltages on 4160 V Buses are less than 4390 V. a. Contact System Operations and request Grid voltage be lowered.

b. Start additional plant equipment powered from the affected 4160 V Buses.

NOTE With high bus voltage, the operation of the LPSI and RCP motors above 4400 V for an extended period of time must be evaluated.

11. IF ANY 4160 V Bus voltage exceeds 11.1 IF a plant shutdown is required, 4390 V for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, THEN commence a plant shutdown THEN contact System Engineering for PER OP-4, Load Changes and Normal bus load evaluation. Power Operation.

11.2 IF plant conditions allow, THEN secure plant equipment as recommended by System Engineering.

12. GO TO Section 5, EXIT CONDITIONS.

End of Section 4.0 R11

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 019 Given the following initial conditions:

The plant was taken to hot shutdown during the middle of an operating cycle to repair feedwater heater tube leaks.

The repair took 4 days The reactor was restarted and power raised to 8%

The mode selector switch was in Manual Sequential The turbine was being warmed up Current plant conditions:

Reactor power level is 10% and increasing steadily RCS T-ave is increasing slowly Pressurizer level is following its programmed level Pressurizer pressure is slowly rising Containment pressure and temperature are normal Which one of the following events will result in the current plant conditions?

A. A condenser bypass valve has failed closed B. A continuous CEA withdrawal is occurring C. A turbine control valve is failing open D. Reactor Xenon-135 concentration is changing Thursday, August 02, 2012 15:55:16 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 19 Rev 0 KA #: 000001 AA2.04 Tier 1 Group 2: Continuous Rod Withdrawal Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal: Reactor power and its trend Importance 4.2 / 4.3 CFR Number: 55.41(b)(1)

Fundamentals of Reactor Theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

Fort Calhoun Objective:

A CEA withdrawal from full power Question Pedigree New question.

K/A Fit:

Question requires Applicant to diagnose a continuous rod withdrawal from reactor power and its trend and primary system response.

Choice A:

Distractor: Plausible because RCS temperature and pressurizer level and pressure will increase for a reduction in secondary heat removal such as a condenser bypass valve failing closed. Incorrect because a reduction in secondary heat removal will cause reactor power to lower.

Choice B:

Correct answer: A continuous rod withdrawal will cause reactor power, RCS temperature, pressurizer level and pressure to all increase.

Choice C:

Distractor: Plausible because a turbine control valve failing open would produce an increase in reactor power. Incorrect because it would cause RCS temperature, pressurizer pressure and pressurizer level to all decrease.

Choice D:

Distractor: Plausible because changes in xenon concentration can result in positive or negative reactivity changes depending on time after shutdown. Incorrect because the stem states that the repair took 4 days and the reactor will be xenon free.

KA#: 000001 AA2.04 Bank Ref #: N/A LP# / Objective: 0715-32 02.01 Exam Level: RO-1 Cognitive Level: HIGH Source: NEW

Reference:

LP 07-15-32 Handout: NONE Thursday, August 02, 2012 15:55:16 2

IHB: 07-15-32 Revision: 3 OUTLINE OF INSTRUCTION COMMENTS II. C.

6. To make the analysis conservative, the event Used 112% setpoint was assumed to occur at BOC with a slightly positive moderator temperature coefficient. The rods were assumed to be inserted to the PDIL.

Initial power was assumed to be 102%. Credit was taken for only the variable overpower trip.

7. The power response is shown in figure 2. As we discussed in the theory section, the power increases in a linear manner until the overpower reactor trip setpoint is reached. It then falls off rapidly due to the reactor trip. The RCS temperature, pressurizer level and pressurizer pressure increase in response to the power increase, and then decrease rapidly due to the trip.

8. As shown in the USAR analysis; the DNBR stays above it's limit and the peak LHR and the peak RCS pressure stay below their limits.

9. At Fort Calhoun, the required operator response to a rod withdrawal event is covered by AOP-02. The operator is instructed to place the mode selector switch in the off position. If the rods continue to withdraw, the operator is then instructed to manually trip the reactor.

10. Once the reactor trip occurs, the event is over and EOP-00 is entered. If there are no further events and the control systems function as designed, all of the safety functions will be satisfied.

D. CEA Bank Withdrawal from a Zero Power Condition (Startup Accident)

1. This event is also discussed in section 14.2 of the USAR.

12

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 020 During refueling operations, a new fuel assembly was inserted into the core. The following counts were recorded prior to insertion of the assembly: (Channels A, B & C are connected to scaler-timers)

A = 118 cps B = 132 cps C = 112 cps D = 107 cps After insertion of the assembly, the following counts were observed:

A = 240 cps B = 0 cps C = 225 cps D = 189 cps Which one of the following actions should be taken next?

A. Initiate Emergency Boration B. Move the "B" scaler-timer to channel "D" C. Withdraw the fuel assembly that was just loaded into the core D. Continue loading fuel into the core Thursday, August 02, 2012 16:14:48 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 20 Rev 0 KA #: 000024 2.1.07 Tier 1 Group 2: Emergency Boration Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Importance 4.4 / 4.7 CFR Number: 55.41(b)(1)

Fundamentals of Reactor Theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

Fort Calhoun Objective:

Describe the major recovery actions of this AOP.

Question Pedigree Modified question. The source question was used in NRC exam 2001-2.

K/A Fit:

The question addresses the need to initiate emergency boration based on the response of nuclear instrumentation during refueling.

Choice A:

Correct answer: Channels A and C have doubled meeting the entry condition for AOP-03, Emergency Boration.

Choice B:

Distractor: Plausible because one of the channels with a scaler-timer has failed.

Incorrect because emergency boration is required.

Choice C:

Distractor: Plausible because it would be reversing the reactivity change. Incorrect because emergency boration is required.

Choice D:

Distractor: Plausible If Applicant doesn't realize that counts have doubled on two channels. Incorrect because emergency boration is required.

KA#: 000024 2.1.07 Bank Ref #: 07-11-13 005 LP# / Objective: 0717-03 01.03 Exam Level: RO-1 Cognitive Level: HIGH Source: MODIFIED

Reference:

AOP-03 Handout: NONE Thursday, August 02, 2012 16:14:48 2

AOP-03 Page 2 of 9 1.0 PURPOSE This procedure provides guidance in the event Emergency Boration of the RCS is required.

2.0 ENTRY CONDITIONS The following conditions require the initiation of Emergency Boration using this procedure:

A. Uncontrolled Plant Cooldown.

B. Two or more CEAs inserted below the Emergency Boration Limit of the COLR Power Dependent Insertion Limit Curve.

C. Reactivity anomaly in a positive direction with all CEA's inserted.

D. Unexpected rise in count rate or unexpected doubling of count rate on two or more channels during refueling.

E. Failure of more than one Regulating or Shutdown CEA to insert following a Reactor trip.*

• Guidance for Emergency Boration with these conditions is provided in more appropriate procedures.

R5

Question 20 source question QUESTIONS REPORT for ILO EXAM BANK 07-11-13 005 During refueling operations, a new fuel assembly was inserted into the core. The following counts were recorded prior to insertion of the assembly: (Channels A, B & C are connected to scaler-timers A = 262 cps B = 290 cps C = 308 cps D = 228 cps After insertion of the assembly, the following counts were observed:

A = 270 cps B = 0 cps C = 312 cps D = 232 cps Can the next fuel assembly be inserted into the core?

A. Yes, only two channels of counts are required to proceed.

B. Yes, but only with written permission from the Plant Manager.

C. No, the bundle should be withdrawn to observe the affect on countrate.

D. No, since the base counts were taken on three channels, a new base count should be taken on two channels.

Thursday, August 02, 2012 16:13:15 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 021 Given the following conditions:

• Both level control channels (LC-101X and LC-101Y) are in cascade

• Level control channel selected to Channel Y

• Heater low level cutout selected to X/Y

• Pressurizer pressure is 2100 psia.

With no operator action, which of the following describes the response of the Pressurizer Level Control System if the reference leg of LI-101Y develops a leak?

A. Backup charging pumps (CH-1B and CH-1C) start, heaters energize, letdown flow decreases.

B. Backup charging pumps (CH-1B and CH-1C) start, heaters cutout, letdown flow decreases.

C. Normal running charging pump continues to run, any running backup charging pumps stop, backup heaters energize, letdown flow increases.

D. Normal running charging pump stops, backup heaters do not change state, letdown flow increases.

Thursday, August 02, 2012 16:44:05 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 21 Rev 0 KA #: 000028 AK1.01 Tier 1 Group 2: Pressurizer Level Control Malfunction Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Malfunctions: PZR reference leak abnormalities Importance 2.8 / 3.1 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective EXPLAIN the interlocks and control functions associated with RCS Instrumentation.

Question Pedigree Bank question used on 1999 NRC exam.

K/A Fit:

Question addresses how pressurizer level control responds to a reference leg leak Choice A:

Distractor: Plausible if the Applicant believes level indication fails low and doesn't fully understand pressurizer heater operation. Incorrect because level fails high.

Choice B:

Distractor: Plausible if the Applicant believes level indication fails low. Incorrect because level fails high.

Choice C:

Correct answer. Pressurizer level indication fails high causing backup charging pumps to start and letdown flow to increase. Backup heaters energize due to high level deviation.

Choice D:

Distractor: Plausible because pressurizer level indication does fail high causing letdown flow to increase. Incorrect because one charging pump will continue to operate and heater will energize due to high level deviation.

KA#: 000028 AK1.01 Bank Ref #: 07-11-20 025 LP# / Objective: 0711-20 04.04 Exam Level: RO-7 Cognitive Level: HIGH Source: BANK

Reference:

GFE C07 Handout: NONE Thursday, August 02, 2012 16:44:05 2

DEPRESSURIZED FILL CONNECTION the detector cell, causing the cell to lose its elasticity and altering its characteristics, as well as degrade sensitive electronics.

CONDENSING MAXIMUM POT DETECTABLE LEVEL "WET" REFERENCE LIQUID FAILURE INDICATIONS LEG MINIMUM PRESSURIZED HP LP DETECTABLE LEVEL For level detectors with a wet reference leg connected to the high pressure side of a D/P FILL CONNECTION DRAIN VALVE cell, the following failure modes exist. A break in the variable leg of the D/P cell creates a higher D/P being sensed by the D/P cell, resulting in the Figure 7-22 Closed Tank with Reference Leg level instrument indicating a low level.

Flashing Conversely, a reference leg break creates a lower D/P sensed across the D/P cell, resulting in an ENVIRONMENTAL EFFECTS ON indicated level higher than actual level. A OPERATION ruptured diaphragm or an open equalizing valve in a D/P cell allows the high pressure and low As previously discussed, variations in pressure chambers to equalize and produces a temperature and pressure can affect the reference minimum D/P signal. The minimum D/P signal leg density and the accuracy of the level will result in a maximum level indication.

measurement. These same density variations can Reference leg inventory losses usually occur be caused by the ambient conditions to which the during a rapid pressure decrease, resulting in level instruments are exposed, specifically the flashing part of the reference leg water to steam.

containment building. When this occurs, the pressure on the reference leg side of the D/P cell decreases, causing the Level measuring differential pressure cells are D/P to decrease and indicated level to increase.

not directly affected by changes in ambient For applications where the dry reference leg is pressure. Any pressure variations in the connected to the low pressure side of the D/P environment are felt on both sides of the open cell, the failure modes will be directly opposite vessel D/P level instrument and cancel each of those described above for P. See Table 7-3 other. The dry and wet reference leg level for failure indications.

detectors are only exposed to system pressure and therefore are not affected.

Temperature variations can significantly affect the level measurement accuracies. An increase in ambient temperature will cause the density of the wet reference leg to decrease. This will result in a lower D/P sensed by the D/P cell, and indicated level will be greater than actual level.

The opposite effect produces a lower indicated level when the ambient temperature decreases.

As previously discussed, radiation levels near the D/P cell affect the detector integrity. A high radiation environment can permanently embrittle PWR / COMPONENTS / CHAPTER 7 18 of 43 © 1999 GENERAL PHYSICS CORPORATION

/ SENSORS / DETECTORS REV 2

SYSTEM TRAINING MANUAL VOL. 36 RCS INSTRUMENTATION & REACTOR REGULATING SYSTEMS

• • • Fail Position on Loss of Air or Power DETAILED SYSTEM DESCRIPTION 2.94 Subcooled margin monitor SMM-A (B) is rendered inoperable on a loss of instrument bus AI-40A (B) or DC bus No.1 (No. 2).

Pressurizer Level, Temperature, and Pressure 2.95 The following is a detailed description of the pressurizer level, temperature, and pressure instrumentation.

• • Function 2.96 Pressurizer level instrumentation functions to maintain pressurizer level at the programmed setpoint by regulating the letdown flow rate and starting or stopping the backup charging pumps, as required.

2.97 Pressurizer pressure instrumentation is used to:

• Maintain pressure at 2100 psia during normal operation.

• Energize heaters during a decreasing pressure transient.

• De-energize heaters and open spray valves during an increasing pressure transient.

• Generate alarms to warn of abnormal pressures.

• Supply pressure signals to the RPS, ESC, SMM, RRS, DSS, QSPDS, and the ERF computer.

• Provide overpressure protection during cooldown operation.

• • Design/Specification 2.98 The Pressurizer Level Control System is designed to maintain pressurizer level between 48 percent at 535°F to 60 percent at 560°F. The Pressurizer Pressure Control System is designed to maintain pressurizer pressure at 2100 psia.

Level: Basic > Intermediate > Advanced Revision - 23

• 19

SYSTEM TRAINING MANUAL VOL. 36 RCS INSTRUMENTATION & REACTOR REGULATING SYSTEMS DETAILED SYSTEM DESCRIPTION 60 560oF, 60%

PRESSURIZER WATER LEVEL SETPOINT %

535oF, 48%

530 540 550 560 570 580 MEASURED TAVG - oF FIGURE 2-6: PRESSURIZER LEVEL SETPOINT 2.99 The pressurizer level sensors are differential pressure (P) cells.

The sensors measure the pressure difference between a reference leg and the variable leg, which corresponds to pressurizer level.

The reference leg is maintained full by a condensing pot exposed to containment ambient temperature.

Level: Basic > Intermediate > Advanced Revision - 23

• 20

SYSTEM TRAINING MANUAL VOL. 36 RCS INSTRUMENTATION & REACTOR REGULATING SYSTEMS DETAILED SYSTEM DESCRIPTION CONDENSING POT REFERENCE LEG 0 100%

VARIABLE LEG METAL DIAPHRAGM FIGURE 2-7: DIFFERENTIAL PRESSURE LEVEL SENSOR 2.100 Pressurizer level control channels LI-101X and LI-101Y are cali-brated for 643°F. LI-106 is calibrated for 120°F. Temperature cor-rection curves are used to determine actual level if the pressurizer is not at the calibration temperature. Reference leg temperature (which is determined by containment atmosphere temperature) also affects indicated pressurizer level. Therefore, a family of curves is provided, based on containment atmosphere temperature, to correlate indicated pressurizer level to actual level. Figure 2-8 shows a typical pressurizer level correction curve. If pressurizer temperature is below the calibration temperature, indicated level is higher than actual level.

Level: Basic > Intermediate > Advanced Revision - 23

• 21

SYSTEM TRAINING MANUAL VOL. 36 RCS INSTRUMENTATION & REACTOR REGULATING SYSTEMS DETAILED SYSTEM DESCRIPTION FIGURE 2-8: LI-101 X & Y TEMPERATURE CORRECTION (TDB.111.1.A) 2.101 The control channels function to maintain pressurizer level at the setpoint. The setpoint is developed by the RRS. RRS A provides the setpoint for channel X, and RRS B provides the setpoint for channel Y. The level control functions respond to deviation from setpoint. NOTE: The Tave signal from the RRS is hard wired and, therefore, not selectable with the A/B selector switch on CB-4.

2.102 There are six control channels of pressurizer pressure: four are safety channels and two are control channels. The control channels provide input to the ERF computer and pressurizer heater controls (automatic control rod positioning has been disabled). The safety channels provide input for RPS trips (high-pressure and TM/LP) and for PPLS. The only instrumentation on the pressurizer that is safety related is the pressurizer pressure instrumentation.

2.103 The four pressurizer pressure safety channels, A/B/C/D/PT-102, each have a transmitter and amplifier in a current loop with three voltage dropping resistors. The pressurizer pressure sigma meters have an internal voltage dropping resistor. The voltage dropped across the resistors is used for the high pressurizer pressure trip, TM/LP trip and PPLS.

Level: Basic > Intermediate > Advanced Revision - 23

• 22

SYSTEM TRAINING MANUAL VOL. 36 RCS INSTRUMENTATION & REACTOR REGULATING SYSTEMS The safety channels, A/B/C/D/PT-120, provide input for the DETAILED SYSTEM DESCRIPTION 2.104 Diverse Scram System (DSS). The pretrip is set at 2400 psia, and the trip is set at 2450 psia. The instrument range is from 1900 to 2900 psia. A/PI-120 provides pressure indication on AI-66A, and B/PI-120 provides pressure indication on AI-66B.

14.0% 116GPM LETDOWN EA CR DE EL V

LE G

SIN

+8.0% HI LEVEL ALARM +7.8% CLEAR HI LEVEL ALARM DEVIATION FROM PRESSURIZERLEVEL SETPOINT VARIABLE LETDOWN +5.0% BU HEATERS ON FLOW 2nd and 3rd CH PUMP OFF +3.3% CLEAR B.U.

HEATER SIGNAL PERMIT START STOP 2nd CH PUMP

+2.6% 2nd and 3rd CH PUMPS

+1.7% STOP 3rd CH PUMP 36 GPM AT 0.0 DEVIATION 0

-2.0% MIN. LETDOWN 26 GPM

-2.0%

START 2nd

-2.9% CH PUMP START 3rd

-3.3% CH PUMP

-5.0% CLEAR LOW LEVEL ERROR ALARM

-6.0%

INC LOW LEVEL ERROR RE ALARM BACKUP LE VE ASIN SIGNAL TO START L

ALL CH PUMPS G

FIGURE 2-9: LEVEL CONTROL CHRISTMAS TREE 2.105 Safety channels A/B/C/D/PT-102 provide the following inter-locks/controls. At 1700 psia decreasing, the pressurizer pressure safety channels allow a block of the PPLS signal. A PPLS is gener-ated by safety channels at 1600 psia decreasing if not previously bypassed and actuates the LO-LO PRESSURE alarm. Safety chan-nels also provide a TM/LP trip at a variable setpoint, a HIGH PRESSURE PRETRIP alarm at 2300 psia, and a high pressure trip and PORV actuation at 2350 psia.

2.106 Each safety channel pressure (A/B/C/D/PIA-102) is displayed on CB-1/2/3 from 1500 to 2500 psia. A signal from each channel is supplied to the computer. Meters A/B/C/D/PIA-102X display the TM/LP setpoints generated by the RPS.

Level: Basic > Intermediate > Advanced Revision - 23

• 23

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 022 During a large steam line break event, the CVCS is designed to initially supply _____

borated water to the RCS from the _____ for reactivity control.

Assume all systems work as designed.

A. 80 gpm, Boric Acid Storage Tanks B. 120 gpm, Boric Acid Storage Tanks C. 80 gpm, Safety Injection and Refueling Water Storage Tank D. 120 gpm, Safety Injection and Refueling Water Storage Tank Wednesday, August 08, 2012 16:01:25 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 22 Rev 0 KA #: CE-A11 AK1.01 Tier 1 Group 2: RCS Overcooling Knowledge of the operational implications of the following concepts as they apply to the (RCS Overcooling) Components, capacity, and function of emergency systems.

Importance 3.1 / 3.3 CFR Number: 55.41(b)(8)

Components, capacity, and functions of emergency systems.

Fort Calhoun Objective:

EXPLAIN the response of the CVCS to signals from the Engineered Safeguards Control System.

Question Pedigree New question K/A Fit:

Question addresses the charging flow capacity and function of the Boric Acid Storage Tanks following safeguards initiation due to an overcooling event (steam line break).

Choice A:

Distractor: Plausible if the Applicant believes that 2 of 3 charging pumps start on a SIAS just like 2 of 3 HPSI pumps start. Incorrect because all three charging pumps start.

Choice B:

Correct answer. SIAS starts all 3 charging pumps and opens valves from the suction of the Boric Acid Storage tanks to the suction of the charging pumps.

Choice C:

Distractor: Plausible because the SIRWT can be aligned to the suction of the charging pumps and the Applicant believes that 2 of 3 charging pumps start on a SIAS just like 2 of 3 HPSI pumps start. Incorrect because the initial supply of borated water is from the Boric Acid Storage Tank.

Choice D:

Distractor: Plausible because the SIRWT can be aligned to the suction of the charging pumps. Incorrect because the initial supply of borated water is from the Boric Acid Storage Tank.

KA#: CE-A11 AK1.01 Bank Ref #: N/A LP# / Objective: 0711-02 01.04 Exam Level: RO-8 Cognitive Level: LOW Source: NEW

Reference:

EOP-05 Handout: NONE Wednesday, August 08, 2012 16:01:25 2

TBD-EOP-05 Page 17 of 85 INSTRUCTIONS CONTINGENCY ACTIONS

6. IF RCS pressure is less than or equal to 1600 psia, THEN verify Engineered Safeguards are actuated by performing the following steps:

a. Ensure Emergency Boration is in progress.

b. Ensure SI flow is acceptable PER Attachment 3, Safety Injection Flow vs. Pressurizer Pressure.

EPG Step: 4, 5, 21a Deviation: 1) The EOP step anticipates automatic containment isolation based on a low RCS pressure while the EPG step bases it on CIAS.

2) The EOP requires Emergency Boration in progress, the EPG references the starting of the Charging Pumps.

3) The EPG provides specific guidance in the step for restoration, the EOP utilizes a floating step.

Plant specific Engineered Safeguards initiating signals are checked. Plant specific SIAS is actuated by either a Pressurizer Pressure Low Signal (PPLS) at 1600 psia or a Containment Pressure High Signal (CPHS) at 5 psig. Many functions performed by the EPG SIAS are directly or indirectly performed by the plant specific PPLS or CPHS signals. EPG SIAS may be referred to in the EOP as PPLS or SIAS.

The plant specific automatic Containment Isolation circuitry is designed to anticipate that a low primary pressure is the result of a primary system break inside containment (i.e., a primary break that will result in a direct release to the containment atmosphere). Hence, low RCS pressure due to a UHE will result in a Containment Isolation. This justifies deviation 1.

Emergency Boration provides the admission of concentrated Boric Acid directly to the suction of the charging pumps via forced flow or gravity feed from the concentrated Boric Acid Storage Tanks. This addition is prudent to counteract the positive reactivity inserted during a UHE due to the lowering moderator temperature. It is therefore prudent to ensure Emergency Boration is taking place. This justifies deviation 2.

Floating steps are used to perform complicated actions that can be accomplished at any time during the event. The detail for restoration is placed in the floating steps to allow this. This justifies deviation 3.

Continuously Applicable or Non-Sequential Step R25

SYSTEM TRAINING MANUAL VOL. 12 CHEMICAL AND VOLUME CONTROL SYSTEM The Second standby charging pump will auto start when pres-DETAILED SYSTEM DESCRIPTION 2.351 surizer level decreases to 3.3% below programmed level as long as the following conditions are met:

• The charging pump has been selected as Standby via the mode selector switch.

• The charging pump control switch is in the AFTER-STOP posi-tion.

2.352 The Second standby charging pump will auto stop when pres-surizer level increases to 1.7% above programmed level.

2.353 The First standby charging pump will auto stop when pressur-izer level increases to 2.6% above programmed level.

2.354 In the event pressurizer level decreases to 6% below programmed level, all charging pumps that are not currently running will receive a start signal regardless of their standby status. Once pres-surizer level increases to 5% above programmed level, the standby charging pumps will stop. The charging pump that was previously running will be unaffected.

2.355 The charging pumps will also start on a Pressurizer Pressure Low Signal (PPLS) or Containment Pressure High Signal (CPHS). This will cause the auto starts due to pressurizer level to be blocked.

• • • Alarms and Indications 2.356 The following charging pump alarms are located on CB-1/2/3:

• The CHARGING PUMP LUBE OIL PRESSURE LO alarm actuates via pressure control switch PCS-234 (-237, -240) if lube oil pump discharge pressure is less than 9 psig and the respective charging pump motor supply breaker is closed.

• The CHARGING FLOW LO alarm actuates via FIA-236 if the charging flow is less than 30 gpm, which could be indicative of a charging pump trip.

• The CHARGING PUMPS TRIP alarm is a common alarm for all three pumps that actuates via a low lube oil pressure control switch (9 psig), a low suction pressure control switch (10 psia),

a bus undervoltage, or a breaker overcurrent.

2.357 The following charging pump alarms are located on AI-30A(B):

• The CHARGING PUMP CH-1A (-1B or -1C) OFF NORMAL alarm actuates if the respective control switch on CB-1/2/3 is placed in the PULL-TO-LOCK position, the 69 switch is not in the AFTER-CLOSE position, or the pump breaker is not in the racked in position.

• The CHARGING PUMP CH-1A (-1B or -1C) AUTO START FAIL alarm actuates if the respective charging pump breaker fails to close within one second of receiving the signal to close Level: Basic > Intermediate > Advanced Revision - 45 *

• 72

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 023 The following plant conditions exist:

• A plant startup is in progress

• The plant is in MODE 2

• Steam Dumps/Turbine Bypass Valves are controlling RCS Temperature

• A maintenance error results in all running Circulating Water Pumps being stopped Which of the following describes the expected response of the Steam Dumps and Turbine Bypass Valves with no operator actions?

A. Any open valves will close immediately.

B. Any open valves will close when Condenser Vacuum degrades to 19 inches.

C. The Steam Dumps will close immediately; the Turbine Bypass Valve will close when Condenser Vacuum degrades to 19 inches.

D. The Turbine Bypass Valve will close immediately; the Steam Dumps will close when Condenser Vacuum degrades to 19 inches.

Wednesday, August 08, 2012 16:02:06 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 23 Rev 0 KA #: 000051 AK3.01 Tier 1 Group 2: Loss of Condenser Vacuum Knowledge of the reasons for the following responses as they apply to the Loss of Condenser Vacuum: Loss of steam dump capability upon loss of condenser vacuum Importance 2.8 / 3.1 CFR Number: 55.41(b)(4)

Secondary coolant and auxiliary systems that affect the facility.

Fort Calhoun Objective EXPLAIN the actions necessary to control main steam pressure using the steam dump and turbine bypass valves if the automatic pressure and temperature control outputs from the RRS are not available.

Question Pedigree Bank question used on 1995 NRC exam.

K/A Fit:

Question address steam dump and bypass valve response to a loss of condenser vacuum.

Choice A:

Distractor: Plausible if Applicant believes there is an interlock between circulating water pumps and steam dump and bypass valves. Incorrect because there is no interlock.

Choice B:

Correct answer. Steam dump and bypass valves will close when condenser vacuum degrades to 19 inches.

Choice C:

Distractor:Plausible combination of choices A and B. Incorrect because any open valve will close when condenser vacuum degrades to 19 inches.

Choice D:

Distractor:Plausible combination of choices A and B. Incorrect because any open valve will close when condenser vacuum degrades to 19 inches.

KA#: 000051 AK3.01 Bank Ref #: 07-12-31 001 LP# / Objective: 0712.-31 02.02 Exam Level: RO-4 Cognitive Level: LOW Source: BANK

Reference:

LP 0712.31 Handout: NONE Wednesday, August 08, 2012 16:02:06 2

STM-MS&SG - Volume 25 Information Use Page 69 of 103 Main Steam and Steam Generator System Rev. 35 Figure 2 Turbine Dump and Bypass Valve Control 2.5.5Ab (continued)

9) Turbine bypass valve signal auctioneering unit pressure control setpoint may be adjusted on PIC-910 using the setpoint controller and is variable from 800 psia to 1000 psia. It generates an electrical pressure error signal that is proportional to the difference between the main steam pressure and the established pressure setpoint.

The pressure error signal is supplied as one input to the signal auctioneering unit. The second input (programmed Tave from the RRS) is applied to the auctioneering unit through the main turbine trip contacts when the main turbine is tripped.

B. Interlocks

1. The auto signal from the RRS is overridden by a signal from the condenser on low vacuum. This signal prevents dump valves from opening in either MANUAL or AUTO if condenser vacuum drops below 19 inches of mercury.

2. The dump valves will not open on a reactor trip if RCS average temperature is less than 535°F. The opening signal from the RRS is overridden if auto-inhibit switch HC-909 is placed in INHIBIT.

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 024 With the plant in Mode 3, the following conditions exist:

• Circulating Water Pump, CW-1A is running

• Circulating Water Pumps, CW-1B & 1C are secured

• Raw water Pump, AC-10A is operating

• Waste Monitor Tank, WD-22B, is being released

• RM-055 is being used to monitor the release If CW-1A trips, which of the following actions should be taken immediately to ensure the offsite doses determined by the release permit are not exceeded?

A. An additional Raw Water Pump should be started.

B. The RM-055 setpoint should be lowered per the release permit.

C. The Waste Monitor Tank Release should be terminated.

D. The Waste Monitor Tank should be re-sampled.

Wednesday, August 08, 2012 16:02:45 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 24 Rev 0 KA #: 000059 AK3.04 Tier 1 Group 2: Accidental Liquid Radwaste Release Knowledge of the reasons for the following responses as they apply to the Accidental Liquid Radwaste Release: Actions contained in EOP for accidental liquid radioactive-waste release Importance 3.8 / 4.3 CFR Number: 55.41(b)(13)

Procedures and equipment available for handling and disposal of radioactive materials and effluents.

Fort Calhoun Objective:

Using the Offsite Dose Calculation Manual (ODCM) and OI's as reference, explain the minimum circulating water requirements during radioactive liquid effluent releases.

Question Pedigree Bank question used on the 2002 NRC exam. Slight rewording of the stem to better fit the K/A. Replaced choice B and D, reordered choices. Not counted as modified question.

K/A Fit:

Addresses procedural actions to be taken if a circulating water trip could result in a liquid release to the river that exceeds the release permit.

Choice A:

Distractor: Plausible because two Raw water pumps are required for a release with no circulating pumps operating. Incorrect because AOP-10 directs terminating a liquid release if a circulating water pump trips.

Choice B:

Distractor: Plausible because a lower setpoint would be needed with less dilution flow.

Incorrect because a release is not allowed with only one Raw Water pump operating and AOP-10 directs terminating a liquid release if a circulating water pump trips.

Choice C:

Correct answer. The release must be terminated with only one Raw Water Pump operating. AOP-10 directs terminating a liquid release if a circulating water pump trips.

Choice D:

Distracter. Plausible because this is an action that is taken before resuming a release if the radiation monitor failed. Incorrect because a release is not allowed with only one Raw Water pump operating and AOP-10 directs terminating a liquid release if a circulating water pump trips.

Wednesday, August 08, 2012 16:02:45 2

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 KA#: 000059 AK3.04 Bank Ref #: 07-11-03 017 LP# / Objective: 0711-03 01.06 Exam Level: RO-13 Cognitive Level: LOW Source: BANK

Reference:

AOP-10 Handout: NONE Wednesday, August 08, 2012 16:02:45 3

Page 1 of 11 Fort Calhoun Station Unit No. 1 AOP-10 LOSS OF CIRCULATING WATER Change No.: Convert Reason for Change: Convert document from WordPerfect to Word.

Initiator: N/A Preparer: K. Horn Issued: 05-02-06 3:00 pm R2

AOP-10 Page 2 of 11 1.0 PURPOSE This procedure provides guidance in the event of degraded operation or rupture of the Circulating Water System.

2.0 ENTRY CONDITIONS System operation has become degraded or a system rupture has occurred which may be indicated by any of the following:

A. Decreasing pressure indications on "CONDENSER CIRC WATER PRESSURE INLET PI-1913A".

B. White light(s) on CW Pumps.

C. Circulating Water Pump discharge pressure is low.

D. Decreasing Condenser vacuum.

E. "TURBINE BLDG SUMP LEVEL HI" alarm (CB-10,11; A11).

F. "INTAKE BLDG SUMP LEVEL HI" alarm (CB-10,11; A11).

R2

AOP-10 Page 3 of 11 3.0 PRECAUTIONS The following specific cautions and notes apply prior to or throughout this procedure.

A. CAUTIONS None B. NOTES None R2

AOP-10 Page 4 of 11 4.0 INSTRUCTIONS/CONTINGENCY ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS

1. Terminate radioactive liquid waste releases.

2. Ensure all CW Pumps are running.

3. IF all CW flow has been lost, THEN GO TO Step 11.

4. IF a system rupture has occurred, THEN GO TO Step 9.

R2

FORT CALHOUN STATION OI-WDL-3 OPERATING INSTRUCTION PAGE 2 OF 48 PRECAUTIONS (continued)

5. The Liquid Waste Disposal System is radioactively contaminated. Extra precautions should be exercised to avoid possible personnel or equipment contamination.

6. Personnel Corridor 4 may not be accessible during Post-Accident conditions due to High Radiation fields. An emergency hatch, a hoist to remove the hatch and a ladder have been installed in Corridor 52 for access to Panel AI-100 during Post-Accident conditions.

7. Only one Monitor Tank can be released at any one time.

8. A minimum of one Circulating or two Raw Water Pumps must be in operation to release a Monitor Tank.

REFERENCES/COMMITMENT DOCUMENTS

1. Procedures:

! Fort Calhoun Station Off-Site Dose Calculation Manual (ODCM), Section 2.1.1

! Chemistry Form FC-211, Waste Liquid Tank Release Permit

2. USAR:

! 9.13, Sampling

! 11.1.2, Liquid Wastes

3. Drawings File Description 11405-M-7 44338 Waste Disposal System 11405-M-9 44339 Waste Disposal System 627-D-8053 10474 Aquachem Piping and Instrument Diagram Model RW 900 SP 11405-M-257 44336 Circulating Water APPENDICES OI-WDL-3-CL-A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 OI-WDL-3-CL-B . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 OI-WDL-3-CL-C . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 OI-WDL-3-CL-D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 OI-WDL-3-CL-E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 OI-WDL-3-CL-F . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 OI-WDL-3-CL-G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 OI-WDL-3-CL-H . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 OI-WDL-3-CL-I . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 R21

Source Question for Question 24 QUESTIONS REPORT for ILO EXAM BANK 07-11-03 017 With the plant in Mode 3, the following conditions exist:

• Circulating Water Pump, CW-1A is running

• Circulating Water Pumps, CW-1B & 1C are secured

• Raw water Pump, AC-10A is operating

• Waste Monitor Tank, WD-22B, is being released Which of the following actions should be taken if CW-1A trips?

A. One of the other Circulating Water Pumps should be started immediately.

B. The Waste Monitor Tank Release should be terminated immediately.

C. An additional Raw water Pump should be started immediately.

D. No immediate action is required as long as one Raw Water Pump remains in service.

Friday, August 03, 2012 11:07:21 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 025 The following plant conditions exist:

The plant is operating at full power RM-052 is aligned to the auxiliary building stack The RM-052 STACK/CNTMT NOBLE GAS HIGH RADIATION annunciator is in alarm on panel A33C VIAS has actuated.

Aux Building Supply Fan, VA-35A. is in operation Aux Building Exhaust Fan, VA-40A, is in operation Aux Building Exhaust Fan, VA-40B, is out of service What actions should be taken to align Aux Building Ventilation per AOP-09, HIGH RADIATION?

A. Stop VA-35A B. Start VA-35B C. Stop VA-40A D. Start VA-40C Wednesday, August 08, 2012 16:03:26 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 25 Rev 0 KA #: 000060 AA1.02 Tier 1 Group 2: Accidental Gaseous Radwaste Release Ability to operate and / or monitor the following as they apply to the Accidental Gaseous Release: Ventilation system Importance 2.9 / 3.1 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Given the caution statements and/or notes listed in this AOP, explain the reason for each.

Question Pedigree Modified bank question.

K/A Fit:

Question addresses realigning the auxiliary building ventilation system following a high radiation alarm due to an accidental gaseous radwaste release.

Choice A:

Correct answer. AOP-09 directs that all supply fans be stopped and only one exhaust fan should be running.

Choice B:

Distractor: Plausible if Applicant believes two supply fans should be running. Incorrect because this is not in accordance with AOP-09.

Choice C:

Distractor: Plausible if Applicant believes all exhaust fans should be tripped. Incorrect because this is not in accordance with AOP-09.

Choice D:

Distractor: Plausible if Applicant believes two exhaust fans should be running to make the auxiliary building pressure more negative as is directed in AOP-08, "Fuel Handling Incident." Incorrect because this is not in accordance with AOP-09.

KA#: 000060 AA1.02 Bank Ref #: 07-17-09 019 LP# / Objective: 0717-09 01.05 Exam Level: RO-10 Cognitive Level: HIGH Source: MODIFIED

Reference:

AOP-09 Handout: NONE Wednesday, August 08, 2012 16:03:26 2

AOP-09 Page 5 of 15 INSTRUCTIONS CONTINGENCY ACTIONS

8. Verify at least one of the Stack 8.1 IF Stack Radiation Monitors are not Radiation Monitors has power and the powered or pumps are not energized, pump is energized (AI-33C). THEN align a power source and energize the associated pump:

• RM-062 and RM-063 (MCC-4C2)

• RM-052 (MCC-3B1) (normal)

• RM-052 (MCC-4C2), Attachment 2 of OI-RM-1, Radiation Monitoring

9. IF any of the Stack Radiation Monitor 9.1 IF NONE of the Stack Radiation Monitor radiation levels are rising, radiation levels are rising, THEN minimize Auxiliary Building THEN GO TO AOP-21, Reactor Coolant ventilation flow by performing the System High Activity.

following steps (AI-44):

a. Shutdown both Auxiliary Building Supply Fans, VA-35A/B.

(continue)

R10

AOP-09 Page 6 of 15 INSTRUCTIONS CONTINGENCY ACTIONS

9. (continued)

b. Ensure ONE of the following b.1 IF NO Auxiliary Building Exhaust Auxiliary Building Exhaust Fans is Fans are operating AND, in operation: a Radwaste Building Fan is in service,

• VA-40A THEN perform the following:

• VA-40B

• VA-40C

1) Request Security unlock doors 1007-1A and 1007-1B.

2) Open Auxiliary building to Radwaste Building door and West Radwaste Building Rollup door (Corr 26).

3) Monitor RM-043, RWP BLDG VENT STACK GAS RADIATION MONITOR.

10. Ensure the railroad siding rollup door is closed (Room 25).

11. IF VIAS has NOT occurred, 11.1 IF VIAS has occurred, THEN close ALL of the following supply THEN place ALL of the following and exhaust dampers (AI-44): switches in "PULL-TO-FILT" (AI-44):

• HCV-805A/B, East SI Pump Room

• VA-26A

• HCV-804A/B, West SI Pump Room

• VA-26B

• HCV-803A/B, Spent Regen Chem

• VA-27 Holdup Room R10

Source Question for Question 25 QUESTIONS REPORT for ILO EXAM BANK 07-17-09 019 In AOP-09, High Radiation, negative pressure in the Auxiliary Building is maintained by:

A. Shutting of all supply fans and all but one exhaust fan.

B. Shutting off all but one supply fan and all exhaust fans.

C. Shutting off all but one supply fan and all but one exhaust fan.

D. shutting off all supply and exhaust fans.

Friday, August 03, 2012 11:37:17 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 026 The following RM-078 setpoints are provided in in the technical data book, TDB-IV.8.

Normal <1.0 mrem Warn/Alert 10 mrem High 30 mrem A radiation source is moved in Corridor 4 which causes the measured radiation level to rise to 45 mrem and then go back down to normal. What is the status of the alarm status lights on the RM-078 panel at AI-33 after the radiation levels have returned to normal? Assume no operator action has been taken.

A. The Alert and Alarm status lights are both lit and will stay lit until the Alarm ACK button is pushed.

B. The Alert status light has automatically reset and is off. The Alarm status light is lit and will stay lit until the Alarm ACK button is pushed.

C. The Alarm status light has automatically reset and is off. The Alert status light is lit and will stay lit until the Alarm ACK button is pushed.

D. The Alert and Alarm status lights have automatically reset and are off.

Wednesday, August 08, 2012 16:10:24 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 26 Rev 0 KA #: 000061 AK2.01 Tier 1 Group 2: Area Radiation Monitoring (ARM) System Alarms Knowledge of the interrelations between the Area Radiation Monitoring (ARM) System Alarms and the following: Detectors at each ARM system location Importance 2.5 2.6 CFR Number: 55.41(b)(11)

Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Fort Calhoun Objective:

Front panel controls/ indications Question Pedigree Bank question used on the 2004 NRC exam.

K/A Fit:

The question addresses the interrelationship between RM-078 located in Auxiliary Building Corridor 4 and alarms at the Control Room radiation monitoring panel.

Choice A:

Correct answer: The radiation level at RM-078 went above both the alert and high alarms. Both the Alert and Alarm status lights will stay lit until the Alarm ACK button is pushed.

Choice B:

Distractor: Plausible if the Applicant believes that the alert status light automatically resets, but not the alarm status light. Incorrect because neither auto resets.

Choice C:

Distractor: Plausible if the Applicant believes that the alarm status light automatically resets, but not the alert status light. Incorrect because neither auto resets.

Choice D:

Distractor: Plausible if the Applicant believes that the alarm and alert both auto reset like the local monitor horn and flashing light. Incorrect because neither auto resets on the control room panel.

KA#: 000061 AK2.01 Bank Ref #: 07-12-03 057 LP# / Objective: 0712-03 03.03C Exam Level: RO-11 Cognitive Level: LOW Source: BANK

Reference:

STM 33 Handout: NONE Wednesday, August 08, 2012 16:10:24 2

SYSTEM TRAINING MANUAL VOL. 33 RADIATION MONITORING SYSTEM RE-070 THRU 089 RE AND RE-095 THRU 098 LOCATED RATEMETER RIA AND ALARM LOCATED ADJACENT TO DETECTOR TO ERF COMPUTER CONTROL ROOM RATE INPUT METERS ON AI-33B TO RR-099 ANN. A33C ANN. A33C AT AI-33C A AT AI-33C A AREA RADIATION AREA RADIATION HIGH TROUBLE FIGURE 2-53: NORMAL RANGE AREA RADIATION MONITOR CHANNEL ARRANGEMENT 2.338 Each detector is provided with three alarms:

• A downscale alarm to indicate failure of the monitor, loss of power, or loss of detectable background radiation.

• An alert-level alarm to indicate increasing radiation.

• A high-level alarm to alert personnel for evacuation.

• • Location 2.339 The read-out modules, independent channel power supplies, and multipoint recorder are installed on Control Room panel AI-33B.

• • Power Supplies 2.340 Instrument bus AI-40D supplies 120 VAC power to the normal-range area monitors and recorder RR-099 on panel AI-33B.

• • Instrumentation and Control 2.341 Each channel has two independently adjusted setpoints. The set-points are established based on the individual detector local back-ground radiation level. The ALERT alarm level indicates the dose rates in the area have increased to an abnormal level. The HIGH RADIATION alarm level indicates the area dose rate has reached the permissible limit for continued occupancy set forth in 10 CFR Level: Basic > Intermediate > Advanced Revision - 18

• 83

SYSTEM TRAINING MANUAL VOL. 33 RADIATION MONITORING SYSTEM

20. The alarm setpoints are changed on certain monitors for pro-longed shutdowns and outages due to changes in background radi-ation levels.

2.342 The Control Room ratemeters (Victoreen Model 946A) include the following indications and controls on their meter face:

• Main LED display with a five-digit capacity.

• A LED bargraph display with a range of 10-1 to 107 mR/hr.

• Status Lamps for HIGH ALARM (red), WARN ALARM (amber), FAIL ALARM (red) and RANGE ALARM (red).

• WARN and HIGH pushbuttons used to display the high and warn level setpoints.

• CHECK SOURCE pushbutton used to actuate the electronic check source and its associated green LED.

• ALARM ACK. pushbutton that resets relay outputs and causes alarm lights to stop flashing after acknowledgement.

• An alternate action ON/OFF pushbutton for power control.

mR/h CHECK SOURCE ALARM ACKNOWLEGE HIGH RATE CPM 10 102 10 3 10 4 105 106 107 ON ALARMS HIGH WARN FAIL RANGE RATE OFF NORMAL RANGE VICTOREEN FIGURE 2-54: AREA RADIATION MONITOR NORMAL RANGE RATEME-TER Level: Basic > Intermediate > Advanced Revision - 18

• 84

SYSTEM TRAINING MANUAL VOL. 33 RADIATION MONITORING SYSTEM

• • • Local 2.343 The local ratemeters provide personnel with continuous alarm and indication via a horn and light on each monitor. The horn and light are automatically reset when the radiation level lowers to less than the high alarm setpoint. NOTE: High alarm conditions cannot be canceled locally.

• • • Remote 2.344 The Control Room ratemeters (Victoreen Model 946A) include the following controls on the meter face:

• WARN and HIGH pushbuttons used to display the high and warn level setpoints.

• CHECK SOURCE pushbutton used to actuate the electronic check source and its associated green LED.

• ALARM ACK. pushbutton that resets relay outputs and causes alarm lights to stop flashing after acknowledgement.

• An alternate action ON/OFF pushbutton for power control.

• • • Interlocks 2.345 Not applicable.

• • • Alarms and Indications 2.346 The Control Room ratemeters (Victoreen Model 946A) include the following indications on the meter face:

• Main LED display with a five-digit capacity.

• A LED bargraph display with a range of 10-1 to 107 mR/hr.

• Status Lamps for HIGH ALARM (red), WARN ALARM (amber), FAIL ALARM (red) and RANGE ALARM (red).

mR/h CHECK SOURCE ALARM ACKNOWLEGE HIGH RATE CPM 10 10 2 103 104 105 106 107 ON ALARMS HIGH WARN FAIL RANGE RATE OFF NORMAL RANGE VICTOREEN Level: Basic > Intermediate > Advanced Revision - 18

• 85

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 027 An ATWS event has occurred. EOP-00 standard post-trip actions have been completed and the Reactivity Control Safety Function is still not satisfied.

After entering EOP-20, which one of the following actions will be performed that was NOT attempted using EOP-00?

A. Borating from the safety injection and refueling water storage tank B. Manually opening the clutch power supply breakers C. Borating from the boric acid storage tanks D. Manually actuating the DSS trip switches Friday, August 03, 2012 12:32:20 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 27 Rev 0 KA #: CE-E09 EA2.02 Tier 1 Group 2:Functional Recovery Ability to determine and interpret the following as they apply to the (Functional Recovery) Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

Importance 3.5 / 4.0 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Given the Resource Assessment Trees, basically DESCRIBE the Method, Path and Acceptance Criteria for each success path.

Question Pedigree Bank question used on the 2005 NRC exam K/A Fit:

Question addresses procedural response to loss of a safety function.

Choice A:

Correct answer: Borating from the SIRWT is addressed in EOP-20 but not in EOP-00.

Choice B:

Distractor: Plausible because it is a method to attempt to satisfy the safety function.

Incorrect becasue it is directed in both EOP-00 and EOP-20.

Choice C:

Distractor: Plausible because it is a method to attempt to satisfy the safety function.

Incorrect becasue it is directed in both EOP-00 and EOP-20.

Choice D:

Distractor: Plausible because it is a method to attempt to satisfy the safety function.

Incorrect becasue it is directed in both EOP-00 and EOP-20.

KA#: CE-E09 EA2.02 Bank Ref #: 07-18-18 020 LP# / Objective: 0718-18 01.05 Exam Level: RO-10 Cognitive Level: HIGH Source: BANK

Reference:

EOP-20 Handout: NONE Friday, August 03, 2012 12:32:20 2

EOP-00 Page 5 of 44 5.0 INSTRUCTIONS/CONTINGENCY ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS Time of RX Trip: ________

1. Verify Reactivity Control is 1.1 IF the reactor did NOT trip, established by performing steps THEN establish Reactivity Control by a and b: performing step a, b, c or d:

a. Verify ALL of the following:

a. Manually trip the Reactor (CB-4).

No more than one Regulating or Shutdown CEA is NOT b. Manually trip the Reactor (AI-31).

inserted Reactor power is lowering c. Place the DSS Manual Trip Startup rate is negative Switches in "TRIP" (AI-66A/B).

b. Monitor for an uncontrolled RCS d. Manually open the CEDM Clutch cooldown. Power Supply Breakers (AI-57).

(continue)

(continue)

R29

EOP-00 Page 6 of 44 INSTRUCTIONS CONTINGENCY ACTIONS

1. (continued) 1.2 IF more than one CEA is NOT fully inserted, OR an uncontrolled RCS cooldown is in progress, THEN emergency borate by performing the following:

a. Ensure BOTH of the following valves are closed:

FCV-269X, Demin Water Makeup Valve FCV-269Y, Boric Acid Makeup Valve

b. Open ALL of the following valves:

(continue) HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve HCV-265, CH-11A Gravity Feed Valve HCV-258, CH-11B Gravity Feed Valve

c. Start ALL of the following pumps:

Boric Acid Pumps, CH-4A/B Charging Pumps, CH-1A/B/C (continue)

R29

EOP-00 Page 7 of 44 INSTRUCTIONS CONTINGENCY ACTIONS

1. (continued) 1.2 (continued)

d. Close LCV-218-2, VCT Outlet Valve.

e. Ensure ALL of the following valves are closed:

LCV-218-3, Charging Pump Suction SIRWT Isolation Valve HCV-257, CH-4B Recirc Valve HCV-264, CH-4A Recirc Valve

f. Borate until adequate shutdown margin is established.

R29

EOP-20 Page 39 of 527 9.0 REACTIVITY CONTROL SAFETY FUNCTION: Reactivity Control SUCCESS PATH: Boration Using CVCS: RC-2 RESOURCE TREE: Tree A RC-2 INSTRUCTIONS CONTINGENCY ACTIONS

1. Maintain RCS temperature constant by 1.1 IF the Steam Dump and Bypass controlled steaming using the Steam Valves are NOT available, Dump and Bypass Valves. THEN control RCS temperature by performing step a or b:

a. Steam the S/Gs as follows:

1) Ensure MS-164, "MAIN STEAM LINE "A" STEAM DUMP TO ATMOSPHERE (continue) ISOLATION VALVE", is open (Room 81).

2) Ensure MS-407, "ATMOSPHERIC DUMP VALVE HCV-1040 2nd ISOLATION ", is open (Room 81).

(continue)

Continuously Applicable or Non-Sequential Step R25

EOP-20 Page 40 of 527 9.0 REACTIVITY CONTROL RC-2 INSTRUCTIONS CONTINGENCY ACTIONS

1. (continued) 1.1.a (continued)

3) Control HCV-1040, Atmospheric Dump Valve.

b. Operate at least one of the following Air Assisted Main Steam Safety Valves:

MS-291 MS-292

2. Verify a BAST is available for boration. 2.1 IF both BASTs are unavailable, THEN GO TO Step 4 to borate the RCS.

Continuously Applicable or Non-Sequential Step R25

EOP-20 Page 41 of 527 9.0 REACTIVITY CONTROL RC-2 INSTRUCTIONS CONTINGENCY ACTIONS NOTE If 1A3 is not energized, manual action may be needed to establish boric acid flow.

3. Commence emergency boration using BASTs to achieve adequate Shutdown Margin by performing the following:

a. Ensure BOTH of the following valves are closed:

FCV-269X, Demin Water Makeup Valve FCV-269Y, Boric Acid Makeup Valve

b. Open ALL of the following valves:

HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve HCV-265, CH-11A Gravity Feed Valve HCV-258, CH-11B Gravity Feed Valve (continue)

Continuously Applicable or Non-Sequential Step R25

EOP-20 Page 42 of 527 9.0 REACTIVITY CONTROL RC-2 INSTRUCTIONS CONTINGENCY ACTIONS

3. (continued)

c. Ensure ALL of the following c.1 IF the Charging Header is Charging Isolation Valves are inoperable, open: THEN open ALL of the following valves to charge via the HPSI HCV-247 Header:

HCV-238 HCV-248 HCV-239 HCV-308, Charging Pump HPSI Header Isolation Valve HCV-2987, HPSI Header Isolation Valve HPSI Loop Injection Valves

d. Start ALL of the following pumps:

Boric Acid Pumps, CH-4A/B Charging Pumps, CH-1A/B/C

e. Ensure ALL of the following valves are closed:

LCV-218-2, VCT Outlet Valve LCV-218-3, Charging Pump Suction SIRWT Isolation Valve HCV-257, CH-4B Recirc Valve HCV-264, CH-4A Recirc Valve Time:

f. GO TO Step 5.

Continuously Applicable or Non-Sequential Step R25

EOP-20 Page 43 of 527 9.0 REACTIVITY CONTROL RC-2 INSTRUCTIONS CONTINGENCY ACTIONS

4. Commence boration using the SIRWT to achieve adequate Shutdown Margin by performing the following:

a. Open LCV-218-3, Charging Pump Suction SIRWT Isolation Valve.

b. Stop both of the Boric Acid Pumps, CH-4A/B.

c. Ensure ALL of the following valves are closed:

HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve HCV-265, CH-11A Gravity Feed Valve HCV-258, CH-11B Gravity Feed Valve LCV-218-2, VCT Outlet Valve (continue)

Continuously Applicable or Non-Sequential Step R25

EOP-20 Page 44 of 527 9.0 REACTIVITY CONTROL RC-2 INSTRUCTIONS CONTINGENCY ACTIONS

4. (continued)

d. Ensure ALL of the following d.1 IF the Charging Header is Charging Isolation Valves are inoperable, open: THEN open ALL of the following valves to charge via the HPSI HCV-247 Header:

HCV-238 HCV-248 HCV-239 HCV-308, Charging Pump HPSI Header Isolation Valve HCV-2987, HPSI Header Isolation Valve HPSI Loop Injection Valves

e. Start all of the Charging Pumps, CH-1A/B/C.

Time:

Continuously Applicable or Non-Sequential Step R25

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 028 The Reactor was operating at full power with a normal electrical alignment when a reactor trip occurred. Following the trip, fast transfer to 161 KV failed.

Which Reactor Coolant Pumps would continue to operate in this situation?

A. RC-3A and RC-3B B. RC-3A and RC-3C C. RC-3C and RC-3D D. RC-3B and RC-3D Wednesday, August 08, 2012 16:11:05 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 28 Rev 0 KA #: 003000 K2.01 Tier 2 Group 1: Reactor Coolant Pump System Knowledge of bus power supplies to the following: RCPs Importance 3.1 3.1 CFR Number: 55.41(b)(3)

Mechanical components and design features of the reactor primary system.

Fort Calhoun Objective:

LIST the power supplies for the RCPs.

Question Pedigree Bank question used on the 1997 NRC exam.

K/A Fit:

Question addresses the power supplies to the RCPs.

Choice A:

Distractor:Plausible if Applicant does not understand RCP power supplies or the normal electrical alignment. Incorrect because RC-3A and RCP-3B would lose power if fast transfer failed.

Choice B:

Distractor:Plausible if Applicant does not understand RCP power supplies or the normal electrical alignment. Incorrect because RC-3A would lose power if fast transfer failed.

Choice C:

Correct answer. RC-3C and RC-3D are normally supplied by 161 KV and will continue to run.

Choice D:

Distractor:Plausible if Applicant does not understand RCP power supplies or the normal electrical alignment. Incorrect because RCP-3B would lose power if fast transfer failed.

KA#: 003000 K2.01 Bank Ref #: 07-11-20 024 LP# / Objective: 0711-20 01.07C Exam Level: RO-3 Cognitive Level: HIGH Source: BANK

Reference:

STM 37 Handout: NONE Wednesday, August 08, 2012 16:11:05 2

STM-RCS - Volume 37 Information Use Page 37 of 149 Reactor Coolant System Rev. 44 2.3.2 (continued)

W. Water from the Component Cooling Water System cools the reactor coolant pump shaft seal and pump motor. Both the upper and lower oil reservoirs in the pump motor have oil-to-water heat exchangers for cooling the oil.

X. The seal area cooling water is further subdivided into two streams:

Cooling for the thermal barrier area, the labyrinth passageway through which the controlled bleedoff enters the seal cavity area Cooling for the integral heat exchanger tubes through which the recirculating water passes Y. Component cooling water enters the integral heat exchanger and flows through the annular region formed by the heat exchanger's double pipe tube arrangement. Recirculating water, contained in the inner pipe of the heat exchanger tubing, transfers heat energy to the cooler component cooling water.

Z. Each pump has its own high-pressure oil lift pump to lubricate the thrust bearings and upper guide bearings during reactor coolant pump startup and shutdown. This oil pressure separates the thrust shoes and lubricates the thrust bearing for pump startup. While the RCP is running, the thrust runner acts as the oil pump. It is sometimes referred to as the low-pressure oil pump. The oil lift pump should be in operation for at least two minutes prior to starting the RCP.

2.3.3 Location A. The reactor coolant pumps are located on the 1013-foot elevation of the Containment Structure: RC pumps RC-3A and RC-3B are on the south side of the reactor vessel along with steam generator RC-2A, and RC pumps RC-3C and RC-3D are on the north side of the reactor vessel along with steam generator RC-2B.

2.3.4 Power Supplies A. The reactor coolant pumps are powered from the 4.16 kV buses. They receive power from buses 1A1, 1A2, 1A3 or 1A4 for RCPs A, B, C and D, respectively. The oil lift pumps and motor heaters for the RCPs are powered from motor control centers 3B1, 4A1, 3A1 and 4C1 for RCPs A, B, C and D, respectively.

STM-ED - Volume 14 Information Use Page 55 of 126 Electrical Distribution System Rev. 42 2.13.3 Location A. The 4160V switchgear is located in the main switchgear rooms separated by a fire barrier.

B. The manual transfer switches for breaker control power are located at the front of the auxiliary power compartment of the switchgear.

2.13.4 Power Supplies A. The normal power supply for buses 1A1 and 1A2 is the 22 kV System through unit auxiliary transformers T1A-1 and T1A-2. The normal power supply for buses 1A3 and 1A4 is the 161 kV System through engineered safeguards transformers T1A-3 and T1A-4. Buses 1A3 and 1A4 provide a limited source of emergency power from the diesel generators.

B. Control power for breaker operation is supplied from the 125V DC Distribution System. Buses 1A1 and 1A3 are normally supplied from DC Bus no. 1 and 1A2 and 1A4 are supplied from DC Bus No. 2.

Alternate or emergency power is available from the opposite bus.

2.13.5 Instrumentation and Control A. Each 4160V bus has voltmeters for each phase on the vertical section of CB-20.

B. Two of two loss-of-voltage signals will initiate a load shed. A load shed opens breakers on the affected bus to prevent overloading an incoming supply and to allow the operator to realign the system. A backup load shed signal is generated if the circuit breakers feeding the bus are open.

Load shed can also be caused by the lockout relays on the breakers feeding the 4160V buses using the auxiliary relay contacts.

C. The 4160V supply breakers 1A11, 1A13, 1A22, and 1A24 will trip on any of the following conditions:

Undervoltage on the supply transformer secondary OPLS (1A13, 1A24 only)

Main Generator Lockout (86/G1)

Main Transformer Lockout (86/GT1)

Main Turbine Lockout (86/SVG1) 345 kV Breaker Failure Lockouts (86/BF-4 or 86/BF-5)

D. Each of the breakers will trip and lockout if a signal is received from its inverse time overcurrent (51) relay or lock-out (86) relay.

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 029 Given the following plant conditions:

The reactor tripped from full power due to low steam generator level following the loss of all Main Feedwater Pumps Water Level in Steam Generator A is lower than the water level in Steam Generator B The water level in both steam generators is lowering at the same rate Assuming no operator action is taken, how will the Auxiliary Feedwater System respond in this situation?

A. AFW will be supplied to S/G "A" only when narrow range level in S/G "A" falls to 32%

B. AFW will be supplied to S/G "A" only when wide range level in S/G "A" falls to 32%

C. AFW will be supplied to both S/G's when narrow range level in S/G "A" falls to 32%

D. AFW will be supplied to both S/G's when wide range level in S/G "A" falls to 32%

Friday, August 03, 2012 13:03:31 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 29 Rev 0 KA #: 059000 K3.02 Tier 2 Group 1: Main Feedwater System Knowledge of the effect that a loss or malfunction of the MFW will have on the following: AFW system Importance 3.6 / 3.7 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN the operation of the AFW System following an Engineered Safeguards AFAS.

Question Pedigree New question.

K/A Fit:

Question addresses AFW system response to a loss of main feedwater.

Choice A:

Distractor: Plausible if Applicant believes AFAS is initiated by narrow range S/G level.

Incorrect because it is initiated by wide range level.

Choice B:

Correct answer. Since the stem states that water level in the "A" S/G is lower than the level in the "B" S/G and the level in both S/G's is lowering at the same rate, 32% wide range level in S/G "A" will be reached first and will initiate flow to S/G "A" only.

Choice C:

Distractor: Plausible if Applicant believes AFAS is actuated on narrow range level and also believes that the logic initiates flow to both S/G's. Incorrect because it is initiated on wide range level and only feeds the S/G with wide range level below 32%.

Choice D:

Distractor: Plausible if Applicant believes that the logic initiates flow to both S/G's.

Incorrect because it only feeds the S/G with wide range level below 32%.

KA#: 059000 K3.02 Bank Ref #: N/A LP# / Objective: 0711-01 01.04 Exam Level: RO-7 Cognitive Level: HIGH Source: NEW

Reference:

STM 04 Handout: NONE Friday, August 03, 2012 13:03:31 2

SYSTEM TRAINING MANUAL VOL. 4 AUXILIARY FEEDWATER SYSTEM During plant shutdown, the main feedwater pumps are used to feed GENERAL SYSTEM DESCRIPTION 1.17 the steam generators until the plant heat load is within the capacity of the motor-driven or engine-driven AFW pump. AFW is aligned to the main feed header for plant shutdown and the main feed pump is shut down.

1.18 The turbine-driven AFW pump is not normally used for routine plant operation. During heatup, the turbine is a heat loss that slows the heatup rate. The turbine exhaust is to atmosphere, so the inven-tory is lost. The turbine is designated the emergency pump and is periodically started for surveillance testing to verify operability.

1.19 If the steam generator level instrumentation senses a low level (32 percent wide range) in either steam generator, the AFW System is designed to start both AFW pumps and supply emergency feedwa-ter to the intact steam generator that has low level. Feedwater is automatically directed through the auxiliary feedwater nozzles.

After the situation stabilizes, the operator can realign the system for AFW flow through the feed regulating bypass valves or take manual control to avoid thermal shock to the steam generators.

Pa>500 AUTOMATIC PSIA INITIATION RC-2A LOW LEVEL P>85

(<32% WR) PSID Pa-Pb P<85 Pa<500 NO PSID PSIA INITIATION Pa-Pb FIGURE 1-3: AUTOMATIC INITIATION AUXILIARY FEEDWATER 1.20 When both AFW pumps are operating, the turbine-driven pump will be providing a slightly greater proportion of the total flow because of its head versus flow curve being higher and flatter. The steam header pressure is normally maintained by the steam dump and bypass system at 900 psig, during startup and shutdown. The motor-driven pump is a constant speed pump and its discharge pressure is controlled by the feedwater header pressure. The shut-off head of the motor-driven pump is about 1200 psi.

Level: Basic > Intermediate > Advanced Revision - 48

• 5

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 030 When placing a second charging pump in service, what action is necessary to match charging and letdown flows and minimize the pressurizer level transient?

A. The pressure setpoint on Pressure Controlled, "PIC-210," is adjusted until charging and letdown flows are matched.

B. The level bias potentiometer on Letdown Flow Controller, HIC-101-1/101-2, is adjusted until charging and letdown flows are matched.

C. The selected Letdown Control Valve, LCV-101-1 or LCV-101-2, is manually adjusted until charging and letdown flows are matched D. The setpoint on the selected Pressurizer Level Controller, LRC-101X or LRC-101Y, is adjusted until charging and letdown flows are matched.

Wednesday, August 08, 2012 16:11:40 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 30 Rev 0 KA #: 004000 A4.08 Tier 2 Group 1:Chemical and Volume Control System Ability to manually operate and/or monitor in the control room: Charging Importance 3.8 / 3.4 CFR Number: 55.41(b)(5)

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.

Fort Calhoun Objective:

EXPLAIN, the manual and automatic functions of control valves in the CVCS.

Question Pedigree Bank Question used on 2001-2 NRC exam. Revised two distractors but left the stem as is. Not counted as modified question.

K/A Fit:

Question addresses operations for placing an additional charging pump in service.

Choice A:

Distractor:Plausible because PIC-210 is adjusted to maintain letdown pressure at 300 psig. Incorrect because the bias pot is used to match charging and letdown flows.

Choice B:

Correct answer:The bias pot is used to match charging and letdown flows.

Choice C:

Distractor: Plausible because letdown flow could be manually adjusted until charging and letdown flow are matched. Incorrect because OI-CH-1 directs using the bias pot.

Choice D:

Distractor: Plausible because the pressurizer level controller setpoint could be manually adjusted until charging and letdown flow are matched. Incorrect because OI-CH-1 directs using the bias pot.

KA#: 004000 A4.08 Bank Ref #: 07-11-02 010 LP# / Objective: 0711-02 01.02 Exam Level: RO-5 Cognitive Level: LOW Source: BANK

Reference:

OI-CH-1 Handout: NONE Wednesday, August 08, 2012 16:11:40 2

FORT CALHOUN STATION OI-CH-1 OPERATING INSTRUCTION CONTINUOUS USE PAGE 15 OF 83 Attachment 3 - Raising Charging and Letdown Flows PROCEDURE (continued) () INITIALS 2

f. Open and backseat the selected charging pump Discharge Valve:

CH-193 (for CH-1A)

CH-192 (for CH-1B)

CH-190 (for CH-1C)

Ind Verif

g. Place control switch for selected pump in AFTER STOP:

CH-1A CH-1B CH-1C

h. Repeat this step as necessary for each affected charging pump.

3. Start the selected Charging Pump:

CH-1A, Charging Pump CH-1B, Charging Pump CH-1C, Charging Pump NOTES 1 PIC-210 Letdown Press Cntrlr should be continuously monitored while adjusting letdown flow.

2. Steps 4 and 5 may be performed concurrently without the procedure in hand. Sign-offs may be completed after these steps are performed.

4. Raise the bias on HIC-101-1/101-2, Letdn Throttle Valves Controller, and observe an increase in Letdown flow.

5. Adjust PIC-210 as necessary to maintain Letdown pressure approximately 300 psig.

6. Continue to adjust the bias on HIC-101-1/101-2 until Pressurizer level is stabilized at the programmed setpoint with two Charging Pumps in operation.

Completed by Date/Time /

R89

Source question for question 30 QUESTIONS REPORT for ILO EXAM BANK 07-11-02 010 When placing a second Charging Pump in service, what action is necessary to match Charging and Letdown flows and minimize the Pressurizer level transient?

A. The Letdown Control Valves, LCV-101-1 or LCV-101-2, are controlled manually when more than one (1) Charging Pump is in operation.

B. The Level Bias Potentiometer on Letdown Flow Controller, HIC-101-1/101-2, is manually adjusted to match flows.

C. The pressure setpoint on PIC-210 is changed until Charging and Letdown flows are matched.

D. No manual adjustments are required. Charging and Letdown are matched automatically with no change in Pressurizer level.

Friday, August 03, 2012 13:18:53 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 031 Given the following plant conditions:

A plant cooldown was in progress Shutdown cooling is in service using LPSI Pump, SI-1A The CRS has entered AOP-19, "Loss of Shutdown Cooling," due to lowering RCS water level RVLMS, LI-197, LI-199 and LIS- 119 all indicate that RCS water level is below the centerline of the hot legs.

What action should be taken immediately by the RO per AOP-19?

A. Manually initiate PPLS B. Start LPSI Pump, SI-1B C. Start all charging pumps D. Stop LPSI Pump, SI-1A Wednesday, August 08, 2012 16:12:19 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 31 Rev 0 KA #: 005000 2.4.01 Tier 2 Group 1: Residual Heat Removal System Knowledge of EOP entry conditions and immediate action steps.

Importance 4.6 / 4.8 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Use the Loss of Shutdown Cooling Procedure to mitigate the consequences of a loss of cooling to the Reactor Coolant System.

Question Pedigree New question.

K/A Fit:

Question addresses immediate action to be taken if LPSI pump is about to lose suction while on shutdown cooling.

Choice A:

Distractor:Plausible because PPLS is used for a loss of RCS inventory during normal operation. Incorrect because ECCS is not aligned for normal operation and AOP-19 directs stopping the running LPSI pump.

Choice B:

Distractor: Plausible if Applicant believes this will increase RCS inventory. Incorrect because the result may be two cavitating LPSI pumps and AOP-19 directs stopping the running LPSI pump.

Choice C:

Distractor: Plausible as this will increase RCS inventory. Incorrect because AOP-19 directs stopping the running LPSI pump.

Choice D:

Correct answer: Aop-19 directs stopping the running LPSI pump to prevent suction voiding.

KA#: 005000 2.4.01 Bank Ref #: N/A LP# / Objective: 0717-19 01.00 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

AOP-19 Handout: NONE Wednesday, August 08, 2012 16:12:19 2

AOP-19 Page 5 of 110 4.0 INSTRUCTIONS/CONTINGENCY ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS

1. Ensure ALL refueling operations are stopped.

2. (STA) IMPLEMENT Attachment C, Time to Boil Determination.

3. IMPLEMENT the Emergency Plan.

4. Verify RCS Water Level is above the 4.1 IF RCS Water Level is NOT above the centerline of the Hot Leg using at least centerline of the Hot Leg, two of the following level indications: THEN stop the operating LPSI or CS Pump.

RVLMS (29%)

LI-197 (1006.5 feet)

LI-199 (1006.5 feet, Containment)

LIS-119 (1006.5 feet)

R17

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 032 Which of the following will result in a Containment Isolation Actuation Signal (CIAS)?

A. Containment Radiation Monitor, RM-051, fails high and alarms.

B. Pressure in both steam generators drops below 500 psia due to a steam line break downstream of the MSIVs.

C. Pressurizer pressure channel B/PIA-102Y fails low.

D. Inadvertent actuation of ESF relay, 86A/CPHS.

Question # 32 Rev 0 KA #: 006000 K4.30 Tier 2 Group 1: Emergency Core Cooling System Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following: Containment isolation Importance 3.6 / 3.9 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN how each prime and backup actuation signal is developed.

Question Pedigree Bank question used on the 2005 NRC exam.

K/A Fit:

Question addresses failure of a relay that will initiate containment isolation.

Choice A:

Distractor: Plausible because this will cause a VIAS actuation. Incorrect because it will not initiate CIAS.

Choice B:

Distractor: Plausible because this will cause a SGLS actuation. Incorrect because it will not initiate CIAS.

Choice C:

Distractor: Plausible because this will provide a partial logic to a PPLS signal that would also initiate CIAS. Incorrect because only one pressure channel failed and 2/4 logic is required.

Choice D:

Correct answer: Actuation of relay 86A/CPHS will result in CPHS.

Wednesday, August 08, 2012 16:13:01 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 KA#: 006000 K4.30 Bank Ref #: 07-12-14 017 LP# / Objective: 0712-14 01.04 Exam Level: RO-7 Cognitive Level: LOW Source: BANK

Reference:

STM 19 Handout: NONE Wednesday, August 08, 2012 16:13:01 2

SYSTEM TRAINING MANUAL VOL. 19 ENGINEERED SAFEGUARDS CONTROL SYSTEM GENERAL SYSTEM DESCRIPTION OC OU S 30 DERIVED SIGNAL TO 86B1 DERIVED INITIATION LOCKOUT RELAYS ON AI-30A CUTOFF SWITCH 86B/CRHS 86B/STLS 86B/CPHS 86B/PPLS TO SGIS LOCKOUT RELAYS OR OR AND AND 86B/VIAS 86B/RAS 86B/SIAS 86B/CSAS SEQUENCER DIESEL START BREAKER AND PROTECTION OVERRIDE 86B/OPLS 86BX/RAS 86BX/SIAS 86B/CIAS ESLOAI30B FIGURE 1-4: LOCKOUT RELAYS AI-30B 1.34 Most of the lockout relays are arranged in four groups on AI-30A and AI-30B. Each group comprises twelve lockout relays. The 24 lockout relays on AI-30B are functionally redundant to those on AI-30A and are arranged to form a mirror image of the lockout relays on AI-30A. The prime initiation relays and prime actuation relays form the upper group of 12 lockout relays on each panel and are designated 86A on AI-30A and 86B on AI-30B. The derived initiation relays and backup actuation relays form the lower group of 12 lockout relays on each panel and are designated 86B1 on AI-30A and 86A1 on AI-30B. Each derived signal lockout relay is tripped by its associated prime initiation relay. For example, 86A1/

CPHS will trip when 86A/CPHS trips. Each backup actuation lockout relay is tripped by the appropriate logical combination of derived lockout relays located in the same group. For example, 86A1/CSAS will trip when 86A1/CPHS and 86A1/PPLS trip.

Level: Basic > Intermediate > Advanced Revision - 37

• 9

SYSTEM TRAINING MANUAL VOL. 19 ENGINEERED SAFEGUARDS CONTROL SYSTEM CIAS lockout relay actuation:

DETAILED SYSTEM DESCRIPTION 2.129

• De-energizes 22 containment isolation relays and 6 ventilation isolation relays.

• Trips lockout relays on AI-43A and AI-43B.

• Prime lockout relays provide signals to the ERF computer.

• Provides annunciation.

• • Location 2.130 Containment isolation panels AI-43A and AI-43B are in the Con-trol Room and contain lockout relays, the containment isolation relays, position indicating lights for all containment isolation valves, control switches for all isolation valves which do not have control switches elsewhere in the Control Room, and annunciator panels.

• • Power Supplies 2.131 The self-reset alarm relays are powered from the 125V DC buses.

• • Instrumentation and Control 2.132 Containment isolation panels AI-43A and AI-43B are in the Con-trol Room and contain lockout relays, the containment isolation relays, position indicating lights for all containment isolation valves, control switches for all isolation valves which do not have control switches elsewhere in the Control Room, and annunciator panels. If a given line has two isolation valves, normally one valve receives an isolation signal from each containment isolation panel.

If a flowpath has only one isolation valve, it receives an isolation signal from both containment isolation panels.

Level: Basic > Intermediate > Advanced Revision - 37

• 49

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 033 Given the following:

E CCS actuation has occurred due to a PPLS E

OP-03, LOSS OF COOLANT ACCIDENT, has been entered S

tep 33 has been reached S

tep 33 says, "Verify that the Containment Sump level rises as the SIRWT level lowers" S

IRWT level is 95 inches and lowering C

ontainment sump level is low and not rising What action should be taken per EOP-03?

A. Minimize ECCS flow to the minimum required to remove decay heat.

B. Initiate containment spray to ensure adequate containment sump level.

C. Begin blended makeup to the SIRWT to increase SIRWT inventory.

D. Trip the LPSI pumps to reduce SIRWT depletion rate.

Friday, August 03, 2012 14:06:32 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 33 Rev 0 KA #: 006000 A1.15 Tier 2 Group 1: Emergency Core Cooling System Ability to predict and/or monitor changes in parameters RWST Level and temperature Importance 3.3 / 3.9 CFR Number 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective GIVEN a copy of the Technical Basis Documents (TBDs), EXPLAIN the bases behind the major operator actions contained in EOP-03, LOCA.

Question Pedigree Bank question used on the 2009 NRC exam.

K/A Fit:

Question addresses monitoring level in the SIRWT during an interfacing system LOCA.

Choice A:

Distractor: Plausible because this is an action directed for indications of containment sump strainer clogging .Incorrect because the ECCS pumps are still drawing suction from the SIRWT.

Choice B:

Distractor: Plausible because this is an action that may be recommended by SAMG evaluators during an interfacing system LOCA. Incorrect because it is not directed by the EOPs.

Choice C:

Correct answer: EOP-03 directs a blended makeup to the SIRWT.

Choice D:

Distractor: Plausible because action would reduce SIRWT depletion rate. Incorrect because the LPSI pumps may still be required to operate at this point in a LOCA.

KA#: 006000 A1.15 Bank Ref #: 07-18-13 076 LP# / Objective: 0718-13 01.04 Exam Level: RO-10 Cognitive Level: HIGH Source: BANK

Reference:

EOP-03 Handout: NONE Friday, August 03, 2012 14:06:32 2

EOP-03 Page 40 of 82 INSTRUCTIONS CONTINGENCY ACTIONS

32. (continued)

c. Close ALL of the following valves:

• HCV-268, Boric Acid Pump Header to Charging Pumps Isolation Valve

• HCV-265, CH-11A Gravity Feed Valve

• HCV-258, CH-11B Gravity Feed Valve Time:

33. Verify that the Containment Sump level 33.1 IF Containment Sump level is NOT rises as the SIRWT level lowers. rising, THEN commence blending to the SIRWT by performing the following:

a. Open CH-152, "CHARGING PUMPS CH-1A,B&C SUCT HDR SI AND REFUELING WATER (continue)

TANK SI-5 BLENDED BORIC ACID SUPPLY VALVE" (Corridor 26).

b. Start at least one Boric Acid Pump, CH-4A/B.

(continue)

Continuously Applicable or Non-Sequential Step R36

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 034 Given the following plant conditions:

The Pressurizer Quench Tank is being filled to clear a low level alarm During the fill, the QUENCH TANK PRESS HI annunciator went into alarm on CB-1,2,3/A4.

Quench Tank pressure is 10.5 psig Which one of the following actions should be taken to restore pressure in the Pressurizer Quench Tank per OI-RC-6, PRESSURIZER QUENCH TANK NORMAL OPERATION?

A. Open HCV-155, "Quench Tank Vent Valve."

B. Lower the setpoint on PCV-2624, "Nitrogen Pressure Regulator."

C. Open HCV-2637, "N2 Hdr to PQT/RCDT Regulator Bypass valve."

D. Open HCV-153, "Quench Tank Drain Valve."

Friday, August 03, 2012 14:19:38 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 34 Rev 0 KA #: 007000 A2.05 Tier 2 Group 1: Pressurizer Relief Tank / Quench Tank System Ability to (a) predict the impacts of the following malfunctions or operations on the P S; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Exceeding PRT high-pressure limits Importance 3.2 / 3.6 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

EXPLAIN the purpose of the quench tank.

Question Pedigree New question.

K/A Fit:

Question addresses mitigating a high pressure in the quench tank during filling operations.

Choice A:

Correct answer: OI-RC-6 directs opening the vent valve to maintain pressure below 10 psig.

Choice B:

Distractor: Plausible because nitrogen is provided to the quench tank through this regulator. Incorrect because nitrogen to the quench tank would be isolated during this operation.

Choice C:

Distractor: Plausible if Applicant believes opening this bypass valve will allow nitrogen to flow from the quench tank to the RCDT. Incorrect because it won't.

Choice D:

Distractor: Plausible because draining the quench tank to the RCDT will lower pressure.

Incorrect because it will also lower level and the objective of this operation is to raise level in the quench tank.

KA#: 007000 A2.05 Bank Ref #: N/A LP# / Objective: 0711-20 01.09 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

OI-RC-6 Handout: NONE Friday, August 03, 2012 14:19:38 2

FORT CALHOUN STATION OI-RC-6 OPERATING INSTRUCTION C PAGE 4 OF 19 Continuous Use Attachment 1 - Filling the Quench Tank PREREQUISITES () INITIALS

1. Procedure Revision Verification Revision No. Date:

2. Demineralized Water System is operable to HCV-1560A and HCV-1560B per OI-DW-4.

3. Checklist OI-RC-6-CL-A has been completed.

PROCEDURE NOTE Quench Tank level and pressure may also be read on ERF Computer Points L132, P130 and P131.

1. Open the following valves (CB-11):

HCV-1560A, Deaerated Water Header Isolation Valve HCV-1560B, Deaerated Water Header Isolation Valve CAUTION During normal Reactor operation, Quench Tank pressure shall not exceed 10 psig on PIA-131.

2. Monitor PIA-131, Pressurizer Quench Tank, pressure (CB-1/2/3).

3. Open HCV-155, Quench Tank Vent Valve, as necessary, to maintain pressure less than 10 psig (CB-1/2/3).

4. WHEN filled to the desired level on LIA-132, Pressurizer Quench Tank level (normal 73%),

THEN close the following valves:

HCV-1560A HCV-1560B Completed by Date/Time /

R12

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 035 A SIAS signal will automatically ________ the CCW inlet and outlet valves to CCW Heat Exchanger AC-1C, HCV-492A/B, and _________ the CCW Heat Exchanger Bypass Valve, HCV-497.

A. Open, Open B. Open, Close C. Close, Open D. Close, Close Wednesday, August 08, 2012 16:13:50 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 35 Rev 0 KA #: 008000 A3.08 Tier 2 Group 1: Component Cooling Water System Ability to monitor automatic operation of the CCWS, including: Automatic actions associated with the CCWS that occur as a result of a safety injection signal Importance 3.6 / 3.7 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective EXPLAIN the response of the CCW System to signals from the Engineered Safeguards Control System.

Question Pedigree Modified question. Source question was used on the 1995 NRC exam.

K/A Fit:

Question addresses automatic actions for the CCW system following a SIAS.

Choice A:

Distractor: Plausible if Applicant believes all valves open to maximize system flow.

Incorrect because HCV-497 closes.

Choice B:

Correct answer: CCW heat exchanger inlet and outlet valves all open and HCV-497 closes to maximize flow through the heat exchangers. Note: HCV-497 was modified. In the original plant design, HCV-497 did not change position.

Choice C:

Distractor: Plausible if Applicant does not fully understand system design. Incorrect because heat exchanger valves open and HCV-497 closes.

Choice D:

Distractor: Plausible if Applicant does not fully understand system design. Incorrect because heat exchanger valves open.

KA#: 008000 A3.08 Bank Ref #: 07-11-06 012 LP# / Objective: 0711-06 01.05 Exam Level: RO-7 Cognitive Level: LOW Source: MODIFIED

Reference:

LP 0711,06 Handout: NONE Wednesday, August 08, 2012 16:13:50 2

SYSTEM TRAINING MANUAL VOL. 8 COMPONENT COOLING WATER SYSTEM

• • Location 2.80 The CCW inlet and outlet isolation valves are located near their respective heat exchangers. HCV-489A/B and HCV-490A/B are at the east end of Corridor 4 by heat exchangers AC-1A and AC-1B.

HCV-491A/B and HCV-492A/B are located in Room 18 with heat exchangers AC-1C and AC-1D.

• • Power Supplies 2.81 Not applicable

• • Instrumentation and Controls 2.82 The following is a detailed description of the Instrumentation and Controls for the Component Cooling Water Heat Exchanger Isola-tion valves and Bypass valve.

• • • Local 2.83 The heat exchanger isolation valves are equipped with a hand jack that is used to prevent automatic opening of the valves during heat exchanger maintenance. Section 3.0 has details on their operation.

• • • Remote 2.84 The isolation valves HCV-489A/B through HCV-492A/B (eight valves) for the CCW heat exchangers are operated in pairs (inlet and outlet valves) from CB-1/2/3. Each pair of valves has a control switch with three positions (CLOSED/NORMAL/OPEN).

2.85 The CCW heat exchangers can be bypassed by remotely operated valve HCV-497 with a hand controller on panel AI-45. The valve can be positioned from zero to 100% open.

• • • Interlocks 2.86 A safety injection actuation signal (SIAS) automatically opens the CCW heat exchanger inlet/outlet valves and closes the bypass valve. However, if CCW is lost during the accident, the signal may be overridden to a selected set of CCW isolation valves (four) and heat exchangers (two), so that the RW interface valves can be man-ually opened to supply backup cooling. The signal may also be overridden to the bypass valve.

2.87 The control switch HC-4350 overrides SIAS to the CCW isolation valves for AC-1A and AC-1C. The control switch HC-4351 over-rides SIAS to the valves for AC-1B and AC-1D. Both override switches are on CB-1/2/3 and have two positions (NORMAL/

OVERRIDE).

2.88 The control switch HC-497 can override SIAS to the CCW Heat Exchanger Bypass Valve or can close the bypass valve. This switch is located on CB-1, 2, 3 and has three positions (OVER-RIDE, NORMAL/CLOSE).

Level: Basic > Intermediate > Advanced Revision - 39

• 18

Source question for question 35 QUESTIONS REPORT for ILO EXAM BANK 07-11-06 012 Which one of the following describes the expected response of the CCW Heat Exchanger Bypass Valve (HCV-497) to an SIAS?

A. Goes fully open.

B. Goes fully closed.

C. Closes to minimum flow (15% open).

D. Does not change position.

Friday, August 03, 2012 14:25:51 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 036 Given the following plant conditions:

The plant is operating at full power Channel PRC-103X is selected as the controlling channel and is in automatic RPS Channel B TMLP Trip unit is in bypass RPS Channel "C" High Pressurizer Pressure Trip Unit is in bypass Assuming NO operator action, which of the following combinations of pressurizer pressure instrument failures would result in a reactor trip?

A. Pressurizer Control Channel, PT-103X, and Pressurizer Safety Channel, A/PT-102, both fail high.

B. Pressurizer Safety Channel, A/PT-102, and DSS Pressure Channel, B/PT-120, both fail high.

C. DSS Pressure Channels, C/PT-120 and D/PT-120, both fail low.

D. Pressurizer Control Channel, PT-103Y, and Pressurizer Safety Channel, D/PT-102, both fail low.

Wednesday, August 08, 2012 16:14:21 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 36 Rev 0 KA #: 010000 K1.01 Tier 2 Group 1: Pressurizer Pressure Control System Knowledge of the physical connections and/or cause-effect relationships between the PZR PCS and the following systems: RPS Importance 3.9 / 4.1 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective When given specific plant conditions, EXPLAIN operating principles to predict response of Reactor Coolant System (RCS) Instrumentation.

Question Pedigree New question.

K/A Fit:

Question address pressurizer control channel failure that results in a RPS reactor trip.

Choice A:

Correct answer: Since PRC-103X is selected as the controlling channel, if PT-103X fails high, pressurizer spray valves will open and lower pressure to the RPS TMLP setpoint.

With B TMLP channel bypassed and A pressure channel failed high, TMLP channels C

& D will still provide the 2/4 logic required for a reactor trip.

Choice B:

Distractor: Plausible because 2 RPS channels have failed high. Incorrect because only normal RPS channel has failed and only 1 DSS channel has failed.

Choice C:

Distractor: Plausible because if these two channels fail high a diverse scram trip will result. Incorrect because both channels failed low and there is no DSS low pressure trip.

Choice D:

Distractor: Plausible because the answer would be correct if PRC-103Y was selected as the controlling channel. Incorrect because PRC-103X is the controlling channel.

KA#: 010000 K1.01 Bank Ref #: N/A LP# / Objective: 0712-20 04.00 Exam Level: RO-7 Cognitive Level: HIGH Source: NEW

Reference:

STM 36 Handout: NONE Thursday, August 09, 2012 14:23:51 2

SYSTEM TRAINING MANUAL VOL. 36 RCS INSTRUMENTATION & REACTOR REGULATING SYSTEMS DETAILED SYSTEM DESCRIPTION FIGURE 2-8: LI-101 X & Y TEMPERATURE CORRECTION (TDB.111.1.A) 2.101 The control channels function to maintain pressurizer level at the setpoint. The setpoint is developed by the RRS. RRS A provides the setpoint for channel X, and RRS B provides the setpoint for channel Y. The level control functions respond to deviation from setpoint. NOTE: The Tave signal from the RRS is hard wired and, therefore, not selectable with the A/B selector switch on CB-4.

2.102 There are six control channels of pressurizer pressure: four are safety channels and two are control channels. The control channels provide input to the ERF computer and pressurizer heater controls (automatic control rod positioning has been disabled). The safety channels provide input for RPS trips (high-pressure and TM/LP) and for PPLS. The only instrumentation on the pressurizer that is safety related is the pressurizer pressure instrumentation.

2.103 The four pressurizer pressure safety channels, A/B/C/D/PT-102, each have a transmitter and amplifier in a current loop with three voltage dropping resistors. The pressurizer pressure sigma meters have an internal voltage dropping resistor. The voltage dropped across the resistors is used for the high pressurizer pressure trip, TM/LP trip and PPLS.

Level: Basic > Intermediate > Advanced Revision - 23

• 22

SYSTEM TRAINING MANUAL VOL. 36 RCS INSTRUMENTATION & REACTOR REGULATING SYSTEMS The safety channels, A/B/C/D/PT-120, provide input for the DETAILED SYSTEM DESCRIPTION 2.104 Diverse Scram System (DSS). The pretrip is set at 2400 psia, and the trip is set at 2450 psia. The instrument range is from 1900 to 2900 psia. A/PI-120 provides pressure indication on AI-66A, and B/PI-120 provides pressure indication on AI-66B.

14.0% 116GPM LETDOWN EA CR DE EL V

LE G

SIN

+8.0% HI LEVEL ALARM +7.8% CLEAR HI LEVEL ALARM DEVIATION FROM PRESSURIZERLEVEL SETPOINT VARIABLE LETDOWN +5.0% BU HEATERS ON FLOW 2nd and 3rd CH PUMP OFF +3.3% CLEAR B.U.

HEATER SIGNAL PERMIT START STOP 2nd CH PUMP

+2.6% 2nd and 3rd CH PUMPS

+1.7% STOP 3rd CH PUMP 36 GPM AT 0.0 DEVIATION 0

-2.0% MIN. LETDOWN 26 GPM

-2.0%

START 2nd

-2.9% CH PUMP START 3rd

-3.3% CH PUMP

-5.0% CLEAR LOW LEVEL ERROR ALARM

-6.0%

INC LOW LEVEL ERROR RE ALARM BACKUP LE VE ASIN SIGNAL TO START L

ALL CH PUMPS G

FIGURE 2-9: LEVEL CONTROL CHRISTMAS TREE 2.105 Safety channels A/B/C/D/PT-102 provide the following inter-locks/controls. At 1700 psia decreasing, the pressurizer pressure safety channels allow a block of the PPLS signal. A PPLS is gener-ated by safety channels at 1600 psia decreasing if not previously bypassed and actuates the LO-LO PRESSURE alarm. Safety chan-nels also provide a TM/LP trip at a variable setpoint, a HIGH PRESSURE PRETRIP alarm at 2300 psia, and a high pressure trip and PORV actuation at 2350 psia.

2.106 Each safety channel pressure (A/B/C/D/PIA-102) is displayed on CB-1/2/3 from 1500 to 2500 psia. A signal from each channel is supplied to the computer. Meters A/B/C/D/PIA-102X display the TM/LP setpoints generated by the RPS.

Level: Basic > Intermediate > Advanced Revision - 23

• 23

SYSTEM TRAINING MANUAL VOL. 36 RCS INSTRUMENTATION & REACTOR REGULATING SYSTEMS Two single-pen pressure recorder controllers, PRC-103X and DETAILED SYSTEM DESCRIPTION 2.134 PRC-103Y, are located on CB-1/2/3. Channel selector switch HC-103 selects which channel controls the heaters and spray valves.

The setpoint, nominally 2100 psia, is adjustable by depressing the respective setpoint pushbutton (up/down) on each pressure con-troller. The selected controller generates an analog signal that modulates the proportional heater power and spray valve opening.

A HI-LO PRESSURE alarm actuates at 2145 psia (HI) or 2080 psia (LO).

2545 2nd CODE SAFETY RE SSU PRE 2500 1st CODE SAFETY 2450 DIVERSE SCRAM TRIP G

SIN REA 2400 DIVERSE SCRAM DEC PRE-TRIP 2350 TRIP & PORV'S OPEN 2300 HI PRESS PRETRIP 2225 FULLY OPEN SPRAY VALVE MODULATION BAND 2175 FULLY CLOSED 2145 HI PRESS ALARM 2115 PROP HTRS OFF 2100 SETPOINT 2080 LO PRESS ALARM 2075 BACKUP 2075 PROP HTRS FULLY ON HEATERS OFF ALL HEATERS 2060 BACKUP HTRS ON PRES FULLY ON SUR E INC REA TM/LP TRIP 1750 SING PPLS BLOCK PERMISSIVE 1700 LO LO ALARM and PPLS 1600 FIGURE 2-12: RCS PRESSURE CONTROL WITH 2100 PSIA SETPOINT 2.135 Low-range pressure channel P-118 provides Control Room indi-cation on CB-1/2/3 from zero to 1600 psia. The low-range pressure channel actuates a close signal to the shutdown cooling suction valves if pressure exceeds 300 psia. P-118 must be energized to allow opening of HCV-347 and HCV-348. A loss of power over-ride switch is provided to allow shutdown cooling operation during a loss of power from AI-40B. The redundant shutdown cooling valve interlock is provided by P-115, which requires power to pro-vide the interlock, so no override switch is needed.

Level: Basic > Intermediate > Advanced Revision - 23

• 29

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 037 Given the following plant conditions:

A Reactor Coolant System (RCS) cooldown is in progress for a Refueling Outage S

hutdown Cooling is in operation per OI-SC-1 R

CS temperature is 120°F R

CS pressure is 235 psia R

eactor Coolant Pumps, RC-3A and RC-3B, are operating.

P ressure is being controlled manually using Main pressurizer spray and heaters.

What actions should be taken before shutting down Reactor Coolant Pumps, RC-3A and RC-3B, in accordance with OI-RC-9?

A. RCS pressure should be lowered to less than 215 psia.

B. RCS temperature should be lowered to less than 110°F.

C. The RCS should be borated to the refueling boron concentration.

D. RCS pressure control using Auxiliary spray should be established.

Wednesday, August 08, 2012 16:14:54 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 37 Rev 0 KA #: 010000 K6.03 Tier 2 Group 1: Pressurizer Pressure Control System Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: PZR sprays and heaters Importance 3.2 / 3.6 CFR Number 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective DISCUSS the prerequisites and precautions for shutdown of the RCS.

Question Pedigree Bank question used on the 2009 NRC exam.

K/A Fit:

Question addresses loss of RCS pressure control following a loss of pressurizer spray due to tripping RCP's.

Choice A:

Distractor:Plausible because there are RCS pressure limitations for operation of the RCP's. Incorrect because because pressure should be maintained above 225 psia with RCPs running.

Choice B:

Distractor: Plausible if Applicant believes that are temperature restrictions on shutting down RCP's. Incorrect because there is no requirement to lower RCS temperature before tripping RCPs.

Choice C:

Distractor: Plausible if the Applicant believes RCP's are needed for RCS mixing until the RCS is at the refueling boron concentration. Incorrect because refueling boron concentration is not yet required.

Choice D:

Fort Calhoun suffered a loss of shutdown cooling from these conditions when the reactor coolant pumps were secured. Securing the pumps caused main pressurizer spray flow to be lost and RCS pressure increased to the setpoint where shutdown cooling loop isolation valves automatically closed. The caution in the provided reference (OI-RC-9) about establishing auxiliary pressurizer spray was added after this event.

KA#: 010000 K6.03 Bank Ref #: 07-11-20 187 LP# / Objective: 0711-20 03.06A Exam Level: RO-10 Cognitive Level: HIGH Source: BANK

Reference:

OI-RC-9 Handout: NONE Wednesday, August 08, 2012 16:14:54 2

FORT CALHOUN STATION OI-RC-9 OPERATING INSTRUCTION PAGE 14 OF 67 C

Continuous Use Attachment 2 - Shutdown Reactor Coolant Pumps (Coupled)

PREREQUISITES (T) INITIALS

1. Procedure Revision Verification Revision No. Date:

2. The Reactor is Shutdown (Mode 3, Mode 4 or Mode 5).

3. The Zero Power Mode Bypass Switches on AI-31A/B/C/D are in bypass to disable the RPS Low Flow Trip before the first RCP is stopped.

PROCEDURE NOTE When a RCP is shutdown its seal Bleedoff temperature will rise.

CAUTION Failure to establish pressure control using Auxiliary PZR Spray prior to securing all Reactor Coolant Pumps may result in Reactor Coolant System pressure increasing and potentially result in a Loss of Shutdown Cooling if pressure exceeds the HCV-347/348, SHUTDOWN COOLING LOOP 2 ISOLATION VALVES interlock setpoint value of 300 psia.

1. Secure the selected RCP by placing its Control Switch to AFTER STOP:

! RC-3A, RC Pump

! RC-3B, RC Pump

! RC-3C, RC Pump

! RC-3D, RC Pump R73

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 038 Given the following plant conditions:

B us Power Failure DC Distribution Panel 1 Light is off

" DC BUS#1 LOW VOLT" annunciator is in alarm D C Bus #1 Voltage indicates 0 volts P ower has been lost to the "A" DSS 86-Relay How will operation of the Diverse Scram System (DSS) be affected?

A. The DSS will initiate a Reactor Trip.

B. The DSS will go to a "1 of 3" Trip Logic.

C. The "A" DSS 86-Relay will not be capable of generating a Reactor Trip.

D. The DSS will not be capable of generating a Reactor Trip.

Friday, August 03, 2012 15:42:21 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 38 Rev 0 KA #: 012000 K2.01 Tier 2 Group 1: Reactor Protection System Knowledge of bus power supplies to the following: RPS channels, components, and interconnections Importance 3.3 / 3.7 CFR Number 55.41(b)(6)

Design, components,and functions of reactivity control mechanisms and instrumentation.

Fort Calhoun Objective DESCRIBE the effects of a loss of power on the Diverse Scram System.

Question Pedigree Bank question used on the 2009 NRC exam.

K/A Fit:

Question addresses a loss of a DC power supply to the DSS portion of the RPS.

Choice A:

Distractor: Plausible if Applicant believes a loss of DC bus will initiate a 2/4 reactor trip.

Incorrect because DSS is an energize to trip system Choice B:

Distractor: Plausible if the Applicant believes that 1 of the 4 inputs are tripped. Incorrect because the power for the 2/4 logic comes from the AC instrument buses.

Choice C:

Correct answer: DSS is an energize to trip system. There are two independent trains, each powered by a different DC bus. With power to the "A" train lost, it would not be capable of generating a reactor trip, but the "B" train would be.

Choice D:

Distractor: Plausible if the Applicant believes a loss of DC bus #1 affects both the "A" and "B" DSS trains. Incorrect because the "B" DSS 86 Relay could still produce a trip.

KA#: 012000 K2.01 Bank Ref #: 07-12-25 LP# / Objective: 0712-25 05.06 Exam Level: RO-6 Cognitive Level: HIGH Source: BANK

Reference:

STM-38 Handout: NONE Friday, August 03, 2012 15:42:21 2

SYSTEM TRAINING MANUAL VOL. 38 REACTOR PROTECTIVE SYSTEM AND DIVERSE SCRAM SYSTEM DETAILED SYSTEM DESCRIPTION SEE PREVIOUS DRAWING

+ -

CH A MATRIX A/TS-DSS MANUAL CLOSED IN TRIP #2 TRIP OPEN IN A1/TS-DSS BYPASS 86A DSS TRIP CH B 94-A1 DSS W

94-A2 DSS W

A1/TS-DSS 3-A1 DSS CLOSED IN TEST 3-A2 DSS CB-AB NE UV CB-CD NE COIL UV 125 VDC BUS #1 COIL FIGURE 2-43: CHANNEL A TRIP CIRCUIT 2.298 The 86A/DSS relay is normally de-energized. If a trip occurs, 86A/DSS energizes to close a contact that energizes 94-A1/DSS and 94-A2/DSS, opening contacts to remove power to the under-voltage coil of both CB-AB and CB-CD. This removes power to the CEDM clutches, tripping the reactor.

2.299 Without 125 VDC power, a reactor trip from the DSS is not pos-sible.

2.300 Either matrix A or B of the DSS is capable of tripping the reactor.

2.301 The following conditions would be indicative of a 125 VDC power failure:

• Loss of matrix lights

• Loss of the 86A/DSSS (or B) LOCKOUT RELAY light

• The DSS 86A/DSS (or B) TROUBLE alarm actuates.

Level: Basic > Intermediate > Advanced Revision - 20

• 76

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 039 With the plant operating at 100% power, pressurizer pressure instrument A/PIA-102Y failed low.

What effect does this have on the Engineered Safeguards Actuation System?

A. PPLS Matrix A is in 1 out of 3 logic, PPLS Matrix B is unaffected.

B. PPLS Matrix A is in 2 out of 3 logic, PPLS Matrix B is unaffected.

C. Both PPLS Matrix A and PPLS Matrix B are in 1 out of 3 logic.

D. Both PPLS Matrix A and PPLS Matrix B are in 2 out of 3 logic.

Friday, August 03, 2012 15:57:29 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 39 Rev 0 KA #: 013000 K6.01 Tier 2 Group 1: Engineered Safety Features Actuation System Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS: Sensors and detectors Importance 2.7 / 3.1 CFR Number: 55.41(b)(8)

Components, capacity, and functions of emergency systems.

Fort Calhoun Objective:

EXPLAIN how ESC signals respond to loss of power or sensor failures.

Question Pedigree Bank question.

K/A Fit:

Question addresses how a loss of a pressure detector would affect the engineered safeguards features actuation system logic.

Choice A:

Distractor: Plausible if Applicant believes A channel detector only affects the "A" matrix.

Incorrect because it affects both the A matrix and the B matrix.

Choice B:

Distractor: Plausible if Applicant believes A channel detector only affects the "A" matrix and the channel is bypassed instead of tripped. Incorrect because it affects both the A matrix and the B matrix and the channel is tripped.

Choice C:

Correct answer: Failure of the detector trips the channel placing both the A matrix and the B matrix in 1/3 logic.

Choice D:

Distractor: Plausible if the Applicant believes the channel is bypassed. Incorrect because the channel is tripped.

KA#: 013000 K6.01 Bank Ref #: 07-12-14 055 LP# / Objective: 0712-14 02.04 Exam Level: RO-8 Cognitive Level: HIGH Source: BANK

Reference:

STM-19 Handout: NONE Friday, August 03, 2012 15:57:29 2

SYSTEM TRAINING MANUAL VOL. 19 ENGINEERED SAFEGUARDS CONTROL SYSTEM The following components are located on AI-30A/B:

DETAILED SYSTEM DESCRIPTION 2.8

• CPHS lockout relays

• CPHS test switches

• PPLS test switches

• SIAS lockout relays and backup SIAS lockout relays

• CPHS/SIAS test switch keys

• RAS test switches 2.9 The following auxiliary SIAS relays are located inside main control board CB-4:

• A/94-1/SIAS

• A/94-2/SIAS

• B/94-1/SIAS

• B/94-2/SIAS

• B/94-5/SIAS 2.10 43-SIAS/FW2 and 43-SIAS/FW4 are located on CB-10/11.

2.11 The load shed test switch CS-A(B)/LST is located in AI-109A(B).

• • Power Supplies 2.12 Power supplies for engineered safeguards components such as an indicator relay or lockout relay are fed from their respective AC or DC engineered safeguards power supply. For example, pressurizer meter relay A/PIA-102Y is fed from 120 VAC instrument bus AI-40A, and the B, C and D relays are fed from their respective bus (AI-40B/C/D). Similarly, lockout relay A/94-3/SIAS is fed from 125 VDC bus AI-41A, and the B relay is fed from AI-41B.

• • Instrumentation and Control 2.13 The following describes the instrumentation and controls for the safety injection actuation signal (SIAS).

2.14 Four pressurizer pressure meter relays, A/B/C/D/PIA-102Y pro-vide signals for PPLS. These pressurizer pressure safety channel current loops are the same instrumentation loops that serve the RPS thermal margin/low pressure (TM/LP) trip and RPS high pressurizer pressure reactor trip and PORV control circuits. Each pressure meter relay in CB-1,2,3 provides pressurizer pressure indication, and actuates 2 auxiliary relays, one for the PPLS actua-tion circuitry and one for the PPLS block circuity. The 2 auxiliary relays provide a PRESSURIZER SAFETY INJECTION SIGNAL LOW-LOW PRESS alarm at 1600 psia, one contact to the PPLS matrix relays, and two contacts to PPLS block circuitry.

Level: Basic > Intermediate > Advanced Revision - 37

• 18

SYSTEM TRAINING MANUAL VOL. 19 ENGINEERED SAFEGUARDS CONTROL SYSTEM DETAILED SYSTEM DESCRIPTION PRESSURIZER PRESSURE TRANSMITTERS A/PT-102 B/PT-102 C/PT-102 D/PT-102 AI-31A AI-31B AI-31D (RPS) (RPS) (RPS)

TEST TEST B/PIA-102Y C/PIA-102Y D/PIA-102Y SIGNAL JACK SIGMA METER RELAY A/PIA-102Y A/PIA- A/PIA-102Y-B 102Y-A B/PIA-102Y D/PIA-102Y 1700 PSIA A LO-LO PRESS 120VAC D/PIA-102Y C/PIA-POWER DECREASING 1600 PSIA 102Y C/PIA-102Y DECREASING 120VAC CS-PPLS POWER BLOCK B/PIA-102Y CS/PPLS OR OR OR OR BLOCK AND AND PPLS BLOCK A PPLS BLOCK B PPLS A/A1 PPLS B/B1 A A BLOCKED 120VAC 120VAC BLOCKED POWER POWER AND AND D/PIA- C/PIA- B/PIA- MATRIX RELAY MATRIX RELAY B/PIA- C/PIA- D/PIA-102Y-1 102Y-1 102Y-1 A/PIA-102Y-1 A/PIA-102Y-2 102Y-2 102Y-2 102Y-2 PPLS 2/4 MATRIX A PPLS 2/4 MATRIX B TEST TEST OR OR PPLS-A PPLS-B FIGURE 2-2: PPLS LOGIC 2.15 Each pressurizer pressure channel provides one contact to the PPLS Block-A circuit and one contact to the PPLS Block-B circuit (via the B auxiliary block relay). PPLS is manually blocked dur-ing shutdown to allow cooldown and depressurization of the RCS.

Closure of contacts from channels A or B and channels C or D allows the operator to block PPLS when pressure decreases below 1700 psia. When pressure increases above 1700 psia, the PPLS block is removed automatically, or may be manually removed at any time by the operator.

2.16 The key for the PPLS block switch is normally captive in the key holder. When in the REMOVE position, the key holder provides a PPLS BLOCK CIRCUIT KEY REMOVED alarm on CB-1,2,3.

The CS-PPLS block switch is a three-position key-operated switch with EMERGENCY RESET/NORMAL/BLOCK positions with spring return to NORMAL.

2.17 The operator blocks PPLS by momentarily turning the key switch to the BLOCK position. Operating the switch energizes the PPLS Block A and PPLS Block B relays if pressurizer pressure is less than 1700 psia. Turning the switch to the EMERGENCY RESET position unblocks PPLS.

Level: Basic > Intermediate > Advanced Revision - 37

• 19

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 040 The plant is operating at full power Detector well cooling fan VA-12A is out of service due to maintenance on its breaker.

The "480 VOLT OVERLOAD TRIP" annunciator is in alarm Detector well cooling fan VA-12B tripped and cannot be restarted.

Using the attached references and the following data, determine the time available until the plant must be in Hot Shutdown.

TI-732A 130° TI-735A 155° TI-733A 132° TI-736A 146° TI-733B 139° TI-736B 142° TI-734A 155° TI-737A 139° A. 9-11 minutes B. 12-14 minutes C. 15-17 minutes D. 18-20 minutes Friday, August 03, 2012 16:15:44 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 40 Rev 0 KA #: 022000 K3.02 Tier 2 Group 1: Containment Cooling System Knowledge of the effect that a loss or malfunction of the CCS will have on the following:

Containment instrumentation readings Importance 3.0 / 3.3 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Given the Operating Instruction OI-VA-1, explain how to calculate the length of time allowed to operate without Nuclear Detector Well Cooling.

Question Pedigree Bank question.

K/A Fit:

Question addresses containment temperature readings following a loss of detector well cooling fans that are a part of the containment cooling system.

Choice A:

Distractor: Plausible if Applicant uses the references incorrectly. Incorrect because the time is too short.

Choice B:

Distractor: Plausible if Applicant uses the references incorrectly. Incorrect because the time is too short.

Choice C:

Correct answer: the time available is approximately 16 minutes.

Choice D:

Distractor: Plausible if Applicant uses the references incorrectly. Incorrect because the time is too long.

Handouts: OI-VA-1 Attachment 3D and TDB III.24 KA#: 022000 K3.02 Bank Ref #: 07-14-05 LP# / Objective: 0714-05 03.00 Exam Level: RO-10 Cognitive Level: HIGH Source: BANK

Reference:

OI-VA-1, TDB III 26 Handout: OI-VA-1, TDB III 24 Friday, August 03, 2012 16:15:44 2

FORT CALHOUN STATION OI-VA-1 OPERATING PROCEDURE C PAGE 17 OF 183 Continuous Use Attachment 3B - Concrete H/U Determination All NDWC Lost PREREQUISITES () INITIALS

1. Procedure Revision Verification Revision No. Date:

PROCEDURE NOTE TIC-732A is the most representative of bulk concrete temperature and should be used if operable. TIC-734A and TIC-736A are conservatively high and should be used only if TIC-732A is inoperable. TIC-733B and TIC-735A read more conservatively high than TIC-734A and TIC-736A and should be used only if TIC-732A, TIC-734A, and TIC-736A are inoperable.

CAUTION Concrete Temperature Heatup determination should be performed as soon as possible upon loss of NDWC.

1. Record the following temperatures: (AI-44)

TIC-732A TIC-733B TIC-734A TIC-735A TIC-736A

2. Determine the concrete temperature using TIC-732A per TDB Figure III.24.

3. IF TIC-732A is inoperable, THEN determine the concrete temperature using TIC-734A and/or TIC-736A.

4. IF TIC-732A, TIC-734A, and TIC-736A are inoperable, THEN determine the concrete temperature using TIC-733B and/or TIC-735A.

R79

FORT CALHOUN STATION OI-VA-1 OPERATING PROCEDURE C PAGE 18 OF 183 Continuous Use Attachment 3B - Concrete H/U Determination All NDWC Lost PROCEDURE (continued) () INITIALS

5. Determine the time remaining before the Plant must be shutdown by performing the following:

148°F - °F = °F ÷ 1°F/min = Minutes (Conc Temp)

6. Notify the Shift Manager of the time before Plant Shutdown.

Completed by Date/Time /

R79

PAGE 1 OF 2 Fort Calhoun Station Unit 1 TDB-III.24 TECHNICAL DATA BOOK NUCLEAR DETECTOR WELL TEMPERATURES Change No. EC 36060 Reason for Change Converted from WordPerfect to Word.

Requestor N/A Preparer J. Collier Editorial Correction N/A Issue Date 04-13-05 3:00 pm R7

FORT CALHOUN STATION TDB-III.24 TECHNICAL DATA BOOK PAGE 2 OF 2 Detector Well Temperature vs Highest Concrete Temperature NOTE: TIC-732A is the most representative of bulk concrete temperature and should be used if operable. TIC-734A and TIC-736A are conservatively high and should be used only if TIC-732A is inoperable. TIC-733B and TIC-735A read more conservatively high than TIC-734A and TIC-736A and should be used only if TIC-732A, TIC-734A, and TIC-736A are inoperable.

R7

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 041 Technical Specification Limiting Conditions for Operations 2.4, "Containment Cooling" allows the reactor to be taken critical for power operation with one _________

inoperable.

A. Containment Air Cooling Unit B. CCW Heat Exchanger C. CCW Pump D. RW Pump Friday, August 03, 2012 16:28:09 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 41 Rev 0 KA #: 022000 2.2.22 Tier 2 Group 1: Containment Cooling System Knowledge of limiting conditions for operations and safety limits.

Importance 4.0 / 4.7 CFR Number: 55.41(b)(5)

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.

Fort Calhoun Objective:

Given a copy of Technical Specifications, apply the applicable Limiting Conditions for Operation (LCO).

Question Pedigree New question.

K/A Fit:

Question addresses LCO for containment cooling Choice A:

Distractor: Plausible because there are 2 air cooling and filtering units and 2 air cooling units. Incorrect because all must be operable before taking the reactor critical.

Choice B:

Distractor: Plausible because there are 4 CCW heat exchangers. Incorrect because all must be operable before taking the reactor critical.

Choice C:

Distractor: Plausible because there are 3 CCW pumps. Incorrect because all must be operable before taking the reactor critical.

Choice D:

Correct answer: T.S. 2.4 does allow the reactor to be taken critical with 1 Raw Water pump inoperable.

KA#: 022000 2.2.22 Bank Ref #: N/A LP# / Objective: 0711-22 01.12 Exam Level: RO-5 Cognitive Level: LOW Source: NEW

Reference:

TS 2.4 Handout: NONE Friday, August 03, 2012 16:28:09 2

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.4 Containment Cooling Applicability Applies to the operating status of the containment cooling systems.

Objective To assure operability of equipment required to remove heat from the containment during normal operating and emergency situations.

Specifications (1) Minimum Requirements

• a. The reactor shall not be made critical, except for low-temperature physics tests, unless all the following are met:

i. The following equipment normally associated with diesel-generator DG-1 (4.16-kV bus 1A3 and associated non-automatically transferring 480-Volt bus sections) is operable, except as noted:(1)

Raw water pump AC-10A Raw water pump AC-10C Component cooling water pump AC-3A Component cooling water pump AC-3C Containment spray pump SI-3A Containment air cooling and filtering unit VA-3A Containment air cooling unit VA-7C ii. The following equipment normally associated with diesel-generator DG-2 (4.16-kV 1A4 and associated non-automatically transferable 480 Volt bus sections) is operable, except as noted.(1)

Raw water pump AC-10B Raw water pump AC-10D Component cooling water pump AC-3B Containment spray pump SI-3B Containment air cooling and filtering unit VA-3B Containment air cooling unit VA-7D iii. Four component cooling heat exchangers shall be operable.

iv. All valves, piping and interlocks associated with the above components and required to function during accident conditions are operable.

(1)

Reactor may be made critical with one inoperable raw water pump. LCO action statements shall apply.

• SEE TDB-VIII 2.4 - Page 1 Amendment No. 120,175, 244, 249

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.4 Containment Cooling (Continued)

b. During power operation one of the components listed in (1)a.i. or ii. may be inoperable. If the inoperable component is not restored to operability within seven days, the reactor shall be placed in hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the inoperable component is not restored to operability within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. For cases involving Raw Water pump inoperability, if the river water temperature is below 60 degrees Fahrenheit, one Raw Water pump may be inoperable indefinitely without applying any LCO action statement. When the river water temperature is greater than 60 degrees Fahrenheit, an inoperable Raw Water pump shall be restored to operability within 7 days or the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the inoperable Raw Water pump is not restored to operability within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(2) Modification of Minimum Requirements

a. During power operation, the minimum requirements may be modified to allow a total of two of the components listed in (1)a.i. and ii. to be inoperable at any one time.

(This does not include: 1) One Raw Water pump which may be inoperable as described above if the river water temperature is below 60 degrees Fahrenheit or,

2) SI-3A and SI-3B being simultaneously inoperable; or 3) VA-3A and VA-3B, or VA-7C and VA-7D, being simultaneously inoperable. Only two raw water pumps may be out of service during power operations. Either containment spray pump, SI-3A or SI-3B, must be operable during power operations. One train of the containment air cooling and filtering systems (VA-3A and VA-7C), or (VA-3B and VA-7D), must be operable during power operations). If the operability of one of the two components is not restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. LCO 2.4(1)b. shall be applied if one of the inoperable components is restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the operability of both components is not restored within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. During power operation one component cooling heat exchanger may be inoperable.

If the operability of the heat exchanger is not restored within 14 days, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If two component cooling heat exchangers are inoperable, the reactor shall be placed in hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the inoperable heat exchanger(s) is not restored to operability within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. Any valves, interlocks and piping directly associated with one of the above components and required to function during accident conditions shall be deemed to be part of that component and shall meet the same requirements as for that component.

• d. Any valve, interlock or piping associated with the containment cooling system which is not included in the above paragraph and which is required

• SEE TDB-VIII 2.4 - Page 2 Amendment No. 49,111,120,175, 244,249, 255

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 042 A large Steam Line Break in containment coincident with numerous tube ruptures in the faulted S/G has caused PPLS, CPHS and SGLS.

Following RAS, What would be the expected trends in SIRWT level (LIC-381 and LIC-382) and Containment Sump levels (LI-387 and LI-388) if the seals on one of the Containment Spray Pumps failed?

SIRWT Containment Sump Level Level A. Steady Steady B. Steady Lowering C. Rising Lowering D. Lowering Rising Friday, August 03, 2012 17:01:42 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 42 Rev 0 KA #: 026000 A1.03 Tier 2 Group 1: Containment Spray System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including: Containment sump level Importance 3.5 / 3.5 CFR Number: 55.41(b)(5)

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.

Fort Calhoun Objective:

EXPLAIN what is meant by an "Interfacing system LOCA."

Question Pedigree Bank question used on 2004 exam.

K/A Fit:

Question addresses monitoring containment sump level during containment spray operation.

Choice A:

Distractor: Plausible if Applicant does not understand sump level response due to CS pump seal failure. Incorrect because containment sump level will lower.

Choice B:

Correct answer: SIRWT level will be steady because RAS has occured. Containment sump level will be lowering because recirculating water is flowing out the CS pump seal.

Choice C:

Distractor: Plausible if Applicant does not understand that ECCS recirculation valves to the SIRWT close on RAS and does not understand the results of the CS pump seal leak.

Choice D:

Distractor: Plausible because these are the expected trends before RAS occurs.

Incorrect because containment sump level will be lowering.

KA#: 026000 A1.03 Bank Ref #: 026000 017 LP# / Objective: 0715-23 01.08 Exam Level: RO-5 Cognitive Level: HIGH Source: BANK

Reference:

STM-15 Handout: NONE Friday, August 03, 2012 17:01:42 2

STM-ECC Volume 15 Information Use Page 6 of 93 Emergency Core Cooling System Rev. 53 1.2 Functions The following is a summary of the functions associated with the Emergency Core Cooling System.

1.2.1 Safety-Related Functions A. The ECCS is required to provide emergency core cooling following a loss of primary or secondary coolant. Portions of the ECCS equipment are used to provide shutdown cooling. Auxiliary functions of the Emergency Core Cooling System equipment include:

Fill and drain of the refueling cavity Provide a backup cooling system for the Spent Fuel Pool Cooling System Provide a means of cooling containment spray water following a Recirculation Actuation Signal (RAS)

Provide a means to fill and drain the safety injection tanks B. Provide a supply of water for initial fill and flushing of the reactor coolant pump mechanical seals.

1.2.2 Non-Safety-Related Functions None 1.3 Basic System Overview The following is a brief overview of the Emergency Core Cooling System.

1.3.1 Flowpaths A. During normal plant operation at power, the Emergency Core Cooling System is maintained in a standby mode with all of its components aligned for emergency operation. Upon receiving a safety injection actuation signal (SIAS), two of the three high-pressure safety injection pumps and the low-pressure safety injection pumps automatically start and the motor-operated safety injection loop isolation valves automatically open.

B. During the injection mode of operation, the safety injection pumps take a suction from the SIRWT and inject borated water into the Reactor Coolant System. The safety injection tanks are pressurized to inject borated water when the Reactor Coolant System pressure falls below their set pressure.

STM-ECC Volume 15 Information Use Page 7 of 93 Emergency Core Cooling System Rev. 53 1.3.1 (continued)

Figure 1 Safety Injection and Containment Spray Block Diagram C. SI-3A and SI-3B containment spray pumps are started by the containment spray actuation signal (CSAS) coincident with a steam generator low pressure signal (SGLS). The CSAS is actuated by high containment pressure coincident with low pressurizer pressure. The combination of these two signals CSAS and SGLS function to limit the containment pressure rise following a MSLB.

D. Low water level in the SIRWT automatically initiates recirculation. The recirculation actuation signal (RAS) shuts down the low-pressure safety injection pumps because they are no longer needed, opens both recirculation line isolation valves to provide suction for HPSI and containment spray, and closes the suction header isolation valves and the minimum flow line isolation valves to isolate the SIRWT from the containment sump water. The HPSI pumps continue to operate with suction from the containment sump to provide cooling for the complete spectrum of break sizes. A portion of the cooled water from the Containment Spray System can be diverted to the suction of the HPSI pumps.

E. A provision is made for maintaining core cooling and preventing boric acid buildup in the core by simultaneous hot and cold leg injection of the HPSI flow in the event of a large break LOCA. The Reactor Coolant System can also be cooled down and depressurized by opening the power-operated relief valves to accelerate refilling of the RCS in the event of a small-break LOCA.

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 043 With the plant operating at full power, inadvertent isolation of extraction steam to FW Heater, FW-16B will result in decreased ______________.

A. FW Temperature to S/G-2B only B. FW Temperature to both S/G's C. FW Flow to S/G-2B only D. FW Flow to both S/G's Saturday, August 04, 2012 12:46:56 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 43 Rev 0 KA #: 039000 K1.08 Tier 2 Group 1:

Main and Reheat Steam System: Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems: MFW Importance 2.7 / 2.9 CFR Number: 55.41(b)(4)

Secondary coolant and auxiliary systems that affect the facility.

Fort Calhoun Objective:

EXPLAIN the maximum allowable consequences for the Excess Load and Loss of Feedwater Heating events.

Question Pedigree New question K/A Fit:

Question addresses the affect of a loss of MRSS extraction steam to a feedwater heater.

Choice A:

Distractor: Plausible if Applicant believes that B train of feedwater heaters supplies water to the B S/G. Incorrect because FW-16A and FW-16B are in parallel Choice B:

Correct answer: FW-16A and FW-16B are in parallel Choice C:

Distractor: Plausible if Applicant believes that B train of feedwater heaters supplies water to the B S/G and doesn't understand that FW flow is controlled downstream of where the heater drain pumps discharge joins the feedwater lines. Incorrect because FW-16A and FW-16B are in parallel Choice D:

Distractor: Plausible if the Applicant doesn't understand that FW flow is controlled downstream of where the heater drain pumps discharge joins the feedwater lines.

Incorrect because FW flow is controlled downstream of heater drain connection.

KA#: 039000 K1.08 Bank Ref #: N/A LP# / Objective: 0715-20 03.02 Exam Level: RO-4 Cognitive Level: HIGH Source: NEW

Reference:

STM-20 Handout: NONE Saturday, August 04, 2012 12:46:56 2

SYSTEM TRAINING MANUAL VOL. 20 FEEDWATER AND CONDENSATE SYSTEM GENERAL SYSTEM DESCRIPTION AUXILIARY CONTAINMENT BUILDING TURBINE BUILDING BUILDING HCV 1106 S/G F.C. FW-16A FW-16B 2B HCV HCV FCV 1385 1104 1102 F.A.I. AFW M M M PI RECIRC PI RECIRC PI RECIRC HCV TO TO TO 1105 COND. COND. COND.

FW-4C FW-4B FW-4A S/G 2A F.C.

HCV HCV FCV M M M 1386 1103 1101 FT FE FT FE FT FE F.A.I.

FW FW FW FW FW FW 317 752 753 318 754 319 FIGURE 1-7: FEEDWATER SYSTEM 1.17 The outlet of each feed pump connects to a common header that splits to supply high-pressure feedwater heaters FW-16A and -16B in parallel. Normally each heater carries about 50 percent of the feedwater flow. Each heater is isolable and capable of carrying 80 percent of full power feedwater flow. The outlets of the high-pres-sure heaters combine and again split to supply the two feedwater headers, one for each steam generator.

1.18 The flow through each feedwater header passes through a flow nozzle and a feedwater regulating valve that controls the amount of feedwater to the steam generators above 15 percent power. A feed-water bypass valve around each feedwater regulating valve is used to control steam generator water level at low power levels. The feedwater then flows through two motor-operated isolation valves in each header and into the steam generators.

1.19 The heater drains are cascaded from the highest pressure to the lowest pressure heaters. One drain flow path (or train) starts with high-pressure feedwater heater FW-16A (B) draining to low-pres-sure heater FW-15A (B), then to heater FW-14A (B) to the heater drain tank. Four moisture separator drain tanks also discharge to Level: Basic > Intermediate > Advanced Revision - 46

• 8

SYSTEM TRAINING MANUAL VOL. 20 FEEDWATER AND CONDENSATE SYSTEM GENERAL SYSTEM DESCRIPTION the heater drain tank. The contents of the heater drain tank are pumped by the heater drain pumps back to the condensate header between low-pressure heaters FW-14A/B and FW-15A/B.

H. P. HEATER 6A H. P. HEATER 6B L. P. HEATER 3B L.P. HEATER 3A LCV LCV 1202B 1202A LCV LCV LCV LCV 1201A 1196A 1196B 1201B L. P. HEATER 5A L.P. HEATER 5B L. P. HEATER 2B L. P. HEATER 2A L. P. HEATER 1B L. P. HEATER 1A LCV LCV LE V EL LEV E L 1197A 1197B CO NTRO L CO NTRO L TANK B TA NK A M OI S TURE L. P. HEATER 4A L. P. HEATER 4B S E P ARATOR DRAI N TA NKS P CV DRAI N COOLER B DRAI N COOLER A 1300 LCV LCV 1203D 1203C LCV TO LCV LCV LCV 1199 CONDE NS ATE 1302 1303 13 04 HE ATE RS LCV LCV LCV 4 &5 1203B 1203A 1301 HEATER DRAI N TANK LCV FCV LCV LCV 1202C 1216A 1202D 1201D LCV LCV DRAI N DRAI N 1196D 1201C FCV LCV PUMP A CONDENSER CONDENS ER 1216B 1196C

'B' 'A' LCV 1197D LCV DRAI N LCV FCV 1197C PUM P B 1198 1216C DRAI N PUM P C CF WHTRDR FIGURE 1-8: FEEDWATER HEATER DRAIN SYSTEM BLOCK DIAGRAM 1.20 The second drain flow path starts with low-pressure heater FW-13A (B) draining to heater FW-12A (B) then to heater FW-11A (B) and its level control tank. The level control tanks drain through the drain coolers to the main condenser. Each heater (except FW-14A/B and FW-11A/B) also has a drain path to the main condenser to dump excess water if the associated level controller fails to maintain the proper level.

1.21 The shell side of each heater is vented to the main condenser. Each heater is provided with connections that continuously vent through orifices sized to remove noncondensables while restricting steam losses to a minimum. The orifices on FW-14A/B, FW-15A/B, and FW-16A/B have a bypass valve that can be opened during plant startup to speed up the removal of noncondensible gases.

• • Interfaces 1.22 The following systems interface with the Feedwater and Conden-sate System:

Level: Basic > Intermediate > Advanced Revision - 46

• 9

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 044 Which of the following conditions would result in the greatest reactivity addition due to a steam line break assuming all systems operate as designed.

A. Beginning of cycle, zero power B. Beginning of cycle, full power C. End of cycle, zero power D. End of cycle, full power Wednesday, August 08, 2012 16:15:30 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 44 Rev 0 KA #: 039000 K5.08 Tier 2 Group 1: Main and Reheat Steam System Knowledge of the operational implications of the following concepts as the apply to the MRSS: Effect of steam removal on reactivity Importance 3.6 / 3.6 CFR Number: 55.41(b)(1)

Fundamentals of Reactor Theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

Fort Calhoun Objective:

EXPLAIN the response of primary system parameters.

Question Pedigree Bank question.

K/A Fit:

The question addresses reactivity addition due to a steam line break.

Choice A:

Distractor: Plausible because S/G inventory is greatest at zero power. Incorrect because the MTC is less negative at BOC.

Choice B:

Distractor: Plausible because most reactivity accidents analyzed in the USAR are more severe at BOC. Incorrect because the greatest reactivity addition is EOC, zero power.

Choice C:

Correct answer: The reactivity addition is greatest at EOC, zero power because the MTC is the most negative and the S/G inventory is the greatest.

Choice D:

Distractor: Plausible because the MTC is most negative at EOC. Incorrect because the S/G water inventory is lower at full power.

KA#: 039000 K5.08 Bank Ref #: 07-15-20 006 LP# / Objective: 0715-20 01.01 Exam Level: RO-1 Cognitive Level: LOW Source: BANK

Reference:

LP 07-15-20 Handout: NONE Wednesday, August 08, 2012 16:15:30 2

IHB: 07-15-20 Revision: 6 OUTLINE OF INSTRUCTION COMMENTS II. A. 4.

c. It can be seen that an excessive heat removal event has the potential to meet two of the three conditions required to produce a brittle fracture (At least in the very conservative calculations). Proper operator action will reduce the severity of the cooldown and minimize the total tensile stress during an excessive heat removal event.

5. Reactivity added by an excessive heat removal EO 1.5 event

a. The reactivity addition that occurs during an excessive heat removal event depends on the extent of the cooldown and on the moderator temperature coefficient. The extent of the cooldown depends on the type, severity and duration of the cooldown event.

b. In the case of a major steam line break, the extent of the cooldown will be much more severe if the break is not isolated by SGIS since that will allow one steam generator to blow down completely. In that case, the steam generator inventory will affect the extent of the cooldown. If the steam generator water inventory is increased, more heat will be removed from the primary side and the extent of the cooldown will be greater. Since the steam generator water inventory is greater at zero power than it is at full power, a steam line break initiated at zero power will result in a greater reactivity addition than one initiated at full power. Additional feedwater flow to the steam generator will have the same effect as increased water inventory; it will add additional reactivity.

II. A. 5.

14

IHB: 07-15-20 Revision: 6 OUTLINE OF INSTRUCTION COMMENTS

c. An excessive heat removal event occurring at End of Cycle will result in a greater reactivity addition than one at Beginning of cycle because the moderator temperature coefficient is more negative at EOC.

B. MITIGATION

1. Potential Causes of an Excessive Heat Removal EO 2.1 Event

a. There are many different potential causes of an excessive heat removal event. We will discuss several of them.

A steam line break is the classic cooldown event from an analysis point of view. The location of the break can have a significant affect on the plant response and indications. The break locations could be:

(1) Inside containment could result in a rapid cooldown and an increase in containment pressure.

The affected steam generator could not be isolated.

15

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 045 The TDSF (Time Delayed Start Failure) relay on diesel generator, DG-1, will initiate a "DIESEL START FAIL" alarm on AI-30A/A30 B-2 if:

A. DG-1's speed is less than 100 rpm 10 seconds after an autostart signal B. DG-1's speed is less than 750 rpm 10 seconds after an autostart signal C. DG-1's speed is less than 100 rpm 15 seconds after an autostart signal D. DG-1's speed is less than 750 rpm 15 seconds after an autostart signal Wednesday, August 08, 2012 16:16:05 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 45 Rev 0 KA #: 064000 K4.05 Tier 2 Group 1: Emergency Diesel Generators Knowledge of ED/G system design feature(s) and/or interlock(s) which provide for the following: Incomplete-start relay Importance 2.8 / 3.2 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Explain abnormal operation of the EDG.

Question Pedigree New question K/A Fit:

The question addresses the TDSF relay that initiates the Diesel Start Failure alarm.

Choice A:

Correct answer: The TDSF relay initiates the diesel start failure alarm if the D/G speed is less than 100 rpm, 10 seconds after an autostart signal.

Choice B:

Distractor Plausible because the time is correct and the D/G's field flashes at 750 rpm.

Incorrect because the TDSF speed setpoint is 100 rpm.

Choice C:

Distractor: Plausible because the speed is correct and the diesel lube oil pressure low alarm has a time delay of 15 seconds after an autostart. Incorrect because the TDSF relay has a 10 second time delay.

Choice D:

Distractor: Plausible because the D/G's fiel flashes at 750 rpm and the diesel lube oil pressure low alarm has a time delay of 15 seconds after an autostart. Incorrect because the values for the TDSF relay are 10 seconds and 100 rpm.

KA#: 064000 K4.05 Bank Ref #: N/A LP# / Objective: 0713-05 01.15 Exam Level: RO-7 Cognitive Level: HIGH Source: NEW

Reference:

ARP-AI-30A/A30 Handout: NONE Wednesday, August 08, 2012 16:16:05 2

FORT CALHOUN STATION CONTINUOUS USE ARP-AI-30A/A30 ANNUNCIATOR RESPONSE PROCEDURE PAGE 11 OF 66 Panel: AI-30A Annunciator: A30 Window: B-2 DIESEL GENERATOR #1 FAILED TO START DIESEL START SAFETY RELATED FAIL Tech Spec

References:

2.7 Initiating Device TDSF Relay Setpoint <100 rpm after 10 sec Power DP2-D1__

OPERATOR ACTIONS

1. Verify DG-1 not running.

1.1 IF DG1 is not running, THEN check the following:

DG-1 Starting Air Pressures DG-1 Control Switches for proper position Any DG-1 Trip Signals 1.2 IF DG-1 is required for emergency operations, THEN attempt to start DG-1 with the Manual Diesel Emergency Start pushbutton.

1.3 IF DG-1 is declared inoperable, THEN test DG-2 according to Technical Specifications 2.7.

1.4 Notify System Engineer of failure to start on demand.

PROBABLE CAUSES Low Starting Air Pressure DG-1 Switch(es) not aligned for Automatic operation DG-1 Trip Signal present if not an Emergency Start REFERENCES B120F14501 Sh 1 17396 B120F14501 Sh 2 17397 161F597 Sh 9 09809 R34

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 046 According to Tech Spec 2.5, "Emergency Feedwater Storage Tank," FW-19, must contain a minimum of ________ gallons of water to ensure that reactor decay heat can be removed for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

A. 50,000 B. 55,000 C. 60,000 D. 65,000 Saturday, August 04, 2012 13:42:25 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 46 Rev 0 KA #: 061000 K5.02 Tier 2 Group 1: Auxiliary / Emergency Feedwater System Knowledge of the operational implications of the following concepts as the apply to the AFW: Decay heat sources and magnitude Importance 3.2 / 3.6 CFR Number: 55.41(b)(4)

Secondary coolant and auxiliary systems that affect the facility.

Fort Calhoun Objective:

Given a copy of the Technical Specifications, INTERPRET the requirements for the AFW System.

Question Pedigree New question K/A Fit:

Question discusses the amount of AFW needed to remove decay heat for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Choice A:

Distractor: Plausible because 50,000 gallons is a reasonable value. Incorrect because tech spec 2.5 requires 55,000 gallons.

Choice B:

Correct answer: Tech spec 2.5 requires 55,000 gallons based on removing decay heat for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Choice C:

Distractor: Plausible because 60,000 gallons is a reasonable value. Incorrect because tech spec 2.5 requires 55,000 gallons.

Choice D:

Distractor: Plausible because 65,000 gallons is a reasonable value. Incorrect because tech spec 2.5 requires 55,000 gallons.

KA#: 061000 K5.02 Bank Ref #: N/A LP# / Objective: 0711-01 01.05 Exam Level: RO-4 Cognitive Level: LOW Source: NEW

Reference:

TS 2.5 Handout: NONE Saturday, August 04, 2012 13:42:25 2

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.5 Steam and Feedwater Systems Applicability When steam generators are relied upon for reactor coolant system heat removal.

NOTE: When heating the reactor coolant above 300°F the steam driven auxiliary feedwater (AFW) pump is only required to be OPERABLE prior to making the reactor critical.

Objective To define certain conditions for the steam and feedwater system necessary to assure adequate decay heat removal.

Specifications

• (1) Two AFW trains shall be OPERABLE when Tcold is above 300°F.

A. With one steam supply to the turbine driven AFW pump inoperable, restore the steam supply to OPERABLE status within 7 days and within 8 days from discovery of failure to meet the LCO.

B. With one AFW train inoperable for reasons other than condition A, restore the AFW train to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. If the required action and associated completion times of condition A or B are not met, then the unit shall be placed in MODE 2 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, in MODE 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and less than 300°F without reliance on the steam generators for decay heat removal within the next 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

D. With both AFW trains inoperable, then initiate actions to restore one AFW train to OPERABLE status immediately. Technical Specification (TS) 2.0.1 and all TS actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.

(2) The motor driven train is required to be OPERABLE when Tcold is below 300°F and the steam generators are relied upon for heat removal. With the motor driven AFW train inoperable, then initiate actions to restore one AFW train to OPERABLE status immediately. Technical Specifications (TS) 2.0.1 and all TS actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.

(3) A minimum of 55,000 gallons of water in the emergency feedwater storage tank (EFWST) and a backup water supply to the emergency feedwater storage tank shall be available. With the EFWST inoperable, verify operability of the backup water supply within four hours and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter, and restore the EFWST

• TDB-VIII 2.5 - Page 1 Amendment No. 49,127,169,212

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.5 Steam and Feedwater Systems to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If these action requirements cannot be satisfied, then the unit shall be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and less than 300°F without reliance on the steam generators for decay heat removal within the next 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

(4) The main steam stop valves are OPERABLE when Tcold is above 300°F and capable of closing in four seconds or less under no-flow conditions.

Basis A reactor shutdown from power requires a removal of core decay heat. Immediate decay heat removal requirements are normally satisfied by the steam bypass to the condenser.

Therefore, core decay heat can be continuously dissipated via the steam bypass to the condenser as long as feedwater to the steam generator is available. Normally, the capability to supply feedwater to the steam generators is provided by operation of the turbine cycle feedwater system. In the unlikely event of complete loss of electrical power to the station, decay heat removal is by steam discharge to the atmosphere via the main steam safety and atmospheric dump valves. Either auxiliary feed pump can supply sufficient feedwater for removal of decay heat from the plant. Technical Specification 2.1.1 establishes when the steam generators are required for heat removal. Each train includes the pump, piping, instruments, and controls to ensure the availability of an OPERABLE flow path capable of taking suction from the EFWST and delivering water to the steam generators. The eight day completion time for 2.5(1)A provides a limit in the maximum time allowed for any combination to be inoperable during any continuous failure to meet the LCO. With one of the required AFW trains inoperable, actions must be taken to restore OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With no AFW trains OPERABLE the unit is in a seriously degraded condition with no safety related means for conducting a cooldown, and only limited means for conducting cooldown with nonsafety grade equipment. In such a condition the unit should not be perturbed by any action, including a power change, that might result in a trip.

The minimum amount of water in the emergency feedwater storage tank is the amount needed for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of such operation. The tank can be resupplied with water from the raw water system.(1)

A closure time of 4 seconds for the main steam stop valves is considered adequate time and was selected as being consistent with expected response time for instrumentation as detailed in the steam line break analysis.(2)(3)

References (1) USAR, Section 9.4.6 (2) USAR, Section 10.3 (3) USAR, Section 14.12 2.5 - Page 2 Amendment No. 127,212

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 047 Following a loss of DC bus #2, Auxiliary Feedwater Pump _________ cannot be started from the control room until _______________.

A. FW-6, the alternate DC power supply is selected.

B. FW-10, the alternate DC power supply is selected.

C. FW-6, power is restored to DC bus #2 D. FW-10, power is restored to DC bus #2 Wednesday, August 08, 2012 16:16:45 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 47 Rev 0 KA #: 061000 A2.03 Tier 2 Group 1: Auxiliary / Emergency Feedwater System 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 Importance 3.1 / 3.4 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective List the primary (preferred) and alternate (if any) power supplies to each bus/component.

Question Pedigree New question K/A Fit:

Question addresses impact of a loss of a DC bus on the AFW system and mitigating actions.

Choice A:

Distractor: Plausible if Applicant believes that FW-6 breaker control power comes from DC bus 2. Incorrect because breaker control power of FW-6 comes from DC bus number 1.

Choice B:

Correct answer. DC bus 2 supplies DC power to FW-10. A switch is provided to transfer FW-10 DC power to DC bus 1.

Choice C:

Distractor: Plausible if Applicant believes that FW-6 breaker control power comes from DC bus 2 and doesn't realize breaker control power can be transfered to the other DC bus.. Incorrect because breaker control power for FW-6 comes from DC bus number 1.

Choice D:

Distractor: Plausible if Applicant does not realize the DC power to FW-10 can be transfered to DC-bus 1. Incorrect because it can be transfered.

KA#: 061000 A2.03 Bank Ref #: N/A LP# / Objective: 0713-04 01.03 Exam Level: RO-7 Cognitive Level: LOW Source: NEW

Reference:

AFW STM Handout: NONE Wednesday, August 08, 2012 16:16:45 2

SYSTEM TRAINING MANUAL VOL. 4 AUXILIARY FEEDWATER SYSTEM

• FW-10 TURBINE DRIVEN FEEDWATER PUMP FAILED DETAILED SYSTEM DESCRIPTION TO START alarm actuates if valve YCV-1045 fails to fully open via time delay relay 74-3/1045.

• FW-10 TURBINE DRIVEN FEEDWATER PUMP IN TEST OR CONTROLS OFF NORMAL alarm actuates via HC-1045 if the hand control is not in AUTO or via auto start relay test switch TS-1045 if the switch in not in the NORMAL position.

• The FW-10 INLET SYSTEM LINE LOW PRESSURE alarm actuates via pressure switch PS-923 if the pressure in the steam line lowers to 400 psig.

2.36 The 40-psig pressure switch that de-energizes the auxiliary lube oil pump also illuminates a red light (PUMP FW-10 RUNNING) on panel AI-179.

2.37 A white light on panel AI-179 is illuminated if an overpressure condition occurs on the discharge header for AFW pump FW-10.

2.38 PI-5062 shaft driven oil pump LO-54 discharge pressure indicator (located on the southeast side of the machine near the mounting base). This gauge has no isolation valve and indicates pressure whenever FW-10 is running.

2.39 PI-5064 DC powered lube oil pump LO-39 discharge pressure gauge. This gauge is located on the front of the machine and is normally isolated.

• • • Local Instrumentation 2.40 PI-5062 Shaft driven oil pump LO-54 discharge pressure indicator (Located on the wall side of the machine near the mounting base.)

This gauge has no isolation valve and indicates pressure whenever FW-10 is running.

2.41 PI-5064 DC powered lube oil pump LO-39 discharge pressure gauge. This gauge is located on the front of the machine and is nor-mally isolated.

• • Failure Modes 2.42 The turbine driver for FW-10 is a reliable unit that operates with-out AC power with a wide range of steam pressures. During nor-mal operation, DC power is used to start the oil pump which pressurizes the oil system to open the governor steam admission valve.

2.43 If the preferred source of DC power were to fail, the alternate bus can be selected from a switch in the inside back of AI-179 in the penetration area. In the unlikely event that both DC sources were to fail and FW-10 did not start when YCV-1045 and YCV-1045A/

B fail open, a pry bar may be used to locally jack up the steam Level: Basic > Intermediate > Advanced Revision - 48

• 18

SYSTEM TRAINING MANUAL VOL. 4 AUXILIARY FEEDWATER SYSTEM DETAILED SYSTEM DESCRIPTION OIL RELAY 8 STARTING LEVER MAIN CONTROLS LEVER CATCH BASIN OIL RELAY PISTON SPINDLE CLEVIS PIN OIL PILOT VALVE UPPER PILOT VALVE BUSHING OIL RELAY PISTON OIL RELAY PISTON RING OIL INLET PIPE OIL PRESSURE FROM OIL PUMP OIL RELAY CYLINDER COVER GASKET OIL LEAKOFF TO PUMP OIL SUMP LOWER PILOT VALVE BUSHING FIGURE 2-7: OIL RELAY ASSEMBLY (FW-10) 2.18 Closing air for YCV-1045 can be supplied from the Instrument Air System or from a permanently installed air accumulator (3-hour rating) to ensure closing air is available.

• • Location 2.19 Turbine-driven AFW pump FW-10 is located in Room 19 on the 991-foot elevation of the Auxiliary Building, south of the motor-driven AFW pump, FW-6, and north of the plant air compressors.

2.20 Discharge header overpressure control panel AI-279 is mounted near AFW pump FW-10.

• • Power Supplies 2.21 The DC motor for the auxiliary oil pump on FW-10 is powered from 125 VDC bus No. 2.

• • Instrumentation and Controls 2.22 The following is a detailed description of the Instrumentation and Controls of the Turbine-Driven AFW Pump FW-10.

• • • Local 2.23 FW-10 is manually controlled from the Control Room or locally from AFW shutdown panel AI-179. The controlling station is selected from REMOTE/LOCAL transfer switches on AI-179. The switches are normally positioned to REMOTE, which selects the Control Room as the controlling station. It is also possible to man-ually open the steam supply valves locally, which should start the turbine. If the balanced governor steam valve is closed, it can be opened with a pry bar to commence rolling the turbine.

Level: Basic > Intermediate > Advanced Revision - 48

• 15

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 048 With a normal electrical system lineup, how would the plant respond to the DC input breaker to inverter "C" failing open?

A. Power would be lost to instrument bus "C" until manually restored.

B. The supply for instrument bus "C" would automatically switch to the bypass transformer for inverter "C".

C. The supply for instrument bus "C" would automatically switch to swing inverter, EE-8T D. The cross tie breakers between instrument buses "A" and "C" would automatically close to supply instrument bus "C".

Wednesday, August 08, 2012 16:17:22 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 48 Rev 0 KA #: 062000 K4.10 Tier 2 Group 1:A.C. Electrical Distribution Knowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following: Uninterruptable ac power sources Importance 3.1 / 3.5 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

List the primary (preferred) and alternate (if any) power supplies to each bus/component.

Question Pedigree Bank question used on the 2001-1 NRC exam. Replaced one of the distractors, changed stem to have DC input breaker fail open. Not counted as modified question.

K/A Fit:

Question addresses automatic transfer of power to an instrument bus to provide uninterruptable AC power.

Choice A:

Distractor: Plausible if Applicant does not understand auto transfer to bypass transformer. Incorrect because instrument bus power will auto transfer.

Choice B:

Correct answer: The supply for instrument bus "C" would automatically switch to the bypass transformer for inverter "C".

Choice C:

Distractor: Plausible because swing inverter EE-8T can be aligned to supply instrument bus "C." Incorrect because it requires manual alignment.

Choice D:

Distractor: Plausible because instrument bus "A" can be cross-tied to supply instrument bus "C." Incorrect because it requires manual alignment.

KA#: 062000 K4.10 Bank Ref #: 07-13-04 007 LP# / Objective: 0713-04 01.03 Exam Level: RO-7 Cognitive Level: LOW Source: BANK

Reference:

ED STM Handout: NONE Wednesday, August 08, 2012 16:17:22 2

STM-ED - Volume 14 Information Use Page 96 of 126 Electrical Distribution System Rev. 42 2.16.2 Design/Specification A. The 120V Instrument AC Distribution System comprises six buses, each supplied by a solid state inverter fed from a 125V DC bus with a backup source of power via a 480/120V voltage regulating transformer. There are two inverters that act as installed spares, one for busses A and C and one for busses B and D. An inverter functions to electronically convert DC to a reliable supply of regulated AC power. Each inverter is equipped with a static switch which monitors the output of the inverter and automatically switches the load to the backup source without a loss of power to the load if the inverter output is lost. The static switches can be manually controlled to bypass an inverter for maintenance.

B. Inverters 1 and 2 each have an additional 480/120V bypass transformer which provides a source of power to instrument bus No. 1 and No. 2 during maintenance and testing of the inverter. This transformer is selected for operation via a fast acting rotary snap switch installed between the inverter output and the instrument bus. This switch is a make before break switch to provide continuity of power to the instrument bus. This switch, along with a voltmeter to indicate the transformer output voltage, are provided in a junction box mounted near Inverters 1 and 2.

Figure 2 EE-102 Inverter No. 1 Testing Transformer C. The six safety-related instrument inverters A, B, C, D, EE-8T and EE-8U are 7.5 kVA 125V DC/120V AC single phase units with +2% voltage regulation. The non-safety related instrument inverters 1 and 2 are similar with 10 kVA capacity.

STM-ED - Volume 14 Information Use Page 97 of 126 Electrical Distribution System Rev. 42 2.16.3 Location A. The additional snap switch provided for inverters 1 and 2, along with a voltmeter to indicate the transformer output voltage, are provided in a junction box mounted near Inverters 1 and 2.

2.16.4 Power Supplies A. Each 120VAC Instrument bus is supplied by a solid state inverter fed from a 125V DC bus with a backup source of power via a 480/120V voltage regulating transformer. There are two swing inverters that act as installed spares, one for busses A and C and one for busses B and D.

The swing inverters also have a backup source of power via a 480/120V voltage regulating transformer.

2.16.5 Instrumentation and Control A. The inverters are controlled from the switchgear rooms. Instruments on the front of the cabinets monitor the voltage and load. The output voltage and current of the inverters are also indicated in the control room. The safety-related inverters are indicated on panels AI-40A, AI-40B, AI-40C and AI-40D and the output of the non-safety related inverters is displayed at panels AI-42A and AI-42B. Ground detection indication for each inverter appear at each inverter distribution panel in the control room.

Figure 2 Vital Inverters (EE-8K; EE-8L)

Source question for question 48 QUESTIONS REPORT for ILO EXAM BANK 07-13-04 007 With a normal electrical system lineup, which one of the following statements would be true if Instrument Inverter "C" failed?

A. Power would be lost to instrument bus "C" until manually restored.

B. The supply for instrument bus "C" would automatically switch to the bypass transformer for inverter "C".

C. The cross tie breakers between instrument buses "A" and "C" would automatically close to supply instrument bus "C".

D. The cross tie breakers between instrument buses "1" and "C" would automatically close to supply instrument bus "C".

Saturday, August 04, 2012 14:17:06 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 049 Given the following plant conditions:

A station blackout occurred 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago The Equipment Operator completed step 1 of EOP/AOP Attachment 6, Minimizing DC loads, 50 minutes ago The Equipment Operator opened G-03, Generator ST-2 Vent Header Isolation Valve, 30 minutes ago The turbine has just stopped rolling What action should be taken from the control room at this time and why should it be taken?

A. Stop LO-4, the DC Oil Pump, to ensure the batteries can supply power for the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

B. Stop LO-12B, the DC Seal Oil Pump, to ensure the batteries can supply power for the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

C. Stop LO-4, the DC Oil Pump, to ensure the batteries can supply power for the next 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

D. Stop LO-12B, the DC Seal Oil Pump, to ensure the batteries can supply power for the next 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

Saturday, August 04, 2012 14:46:26 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 49 Rev 0 KA #: 063000 A1.01 Tier 2 Group 1: D.C. Electrical Distribution 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 Importance 2.5 / 3.3 CFR Number: 55.41(b)(8)

Components, capacity, and functions of emergency systems.

Fort Calhoun Objective:

GIVEN a copy of Attachment 6, EXPLAIN the steps necessary to minimize DC loads.

Question Pedigree Modified question. The source question was used on the 2007 NRC exam.

K/A Fit:

The question addresses actions taken per procedures to reduce the battery discharge rate.

Choice A:

Distractor: Plausible because the action is correct and Fort Calhoun's blackout rule coping time is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Incorrect because the objective of minimizing DC loads is the extend the battery life to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Choice B:

Distractor: Plausible because LO-12B is stopped as part of minimizing DC loads and Fort Calhoun's blackout rule coping time is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Incorrect because LO-12B is tripped 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the blackout and the objective of minimizing DC loads is the extend the battery life to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Choice C:

Correct answer. LO-4 is stopped when the turbine stops rolling and the objective of minimizing DC loads is the extend the battery life to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Choice D:

Distractor: Plausible because LO-12B is stopped as part of minimizing DC loads.

Incorrect because it is stopped 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the blackout.

KA#: 063000 A1.01 Bank Ref #: 07-18-17 021 LP# / Objective: 0718-17 02.03 Exam Level: RO-8 Cognitive Level: HIGH Source: MODIFIED

Reference:

EOP-AOP ATT 6 Handout: NONE Saturday, August 04, 2012 14:46:26 2

EOP/AOP ATTACHMENTS Page 28 of 173 Attachment 6 Minimizing DC Loads INSTRUCTIONS CONTINGENCY ACTIONS NOTE Performing the following step will allow up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> operation of the control and instrumentation devices required for Reactor shutdown without Battery Charger operation.

1. Reduce DC loads by performing the following steps within 15 minutes of the loss of Battery Chargers:

a. Place BOTH of the following DC Bus 2 breakers in "OFF" (West Switchgear Room):

EE-8G-CB12, "400 CYCLE INVERTER EE-21" EE-8G-CB8, "EMERGENCY LIGHTING PNL ELP-2 TRANSFER SWITCH"

b. Place DC Bus 1 breaker EE-8F-CB11, "AUX BLDG EMGY LIGHTING PANEL ELP-1", in "OFF" (East Switchgear Room).

(continue)

R32

EOP/AOP ATTACHMENTS Page 29 of 173 Attachment 6 Minimizing DC Loads INSTRUCTIONS CONTINGENCY ACTIONS

1. (continued)

c. Place 125 VDC Panel DC-PNL-1, Breaker 15, "ELP-5 EMERGENCY. LIGHTING PANEL" in "OFF" (West wall of Turbine Room, upper level).

2. Vent hydrogen from the Generator using G-03, "GENERATOR ST-2 VENT HEADER ISOLATION VALVE" (behind AI-134).

NOTE To ensure adequate battery capacity, the DC Oil Pump should be stopped as soon as the turbine stops rolling, which will occur in approximately one hour.

3. WHEN the turbine has stopped rolling, THEN stop LO-4, DC Oil Pump.

R32

EOP/AOP ATTACHMENTS Page 30 of 173 Attachment 6 Minimizing DC Loads INSTRUCTIONS CONTINGENCY ACTIONS

4. WHEN two hours has elapsed since the loss of Battery Chargers, THEN reduce DC loads by performing the following steps:

a. Ensure BOTH of the following breakers are closed (AI-42A):

I-BUS-I1-1, " INSTRUMENT BUS 1 MAIN BREAKER" "CIRCUIT #1 AI-53 NORM FEED" (continue)

R32

EOP/AOP ATTACHMENTS Page 31 of 173 Attachment 6 Minimizing DC Loads INSTRUCTIONS CONTINGENCY ACTIONS

4. (continued)

b. Place ALL of the following breakers in "OFF" (AI-42A):

"CIRCUIT #2 AI-56 FEED" "CIRCUIT #3 AI-100 FEED" "CIRCUIT #4 IB-1A FEED" "CIRCUIT #5 CB-10,11 FEED" "CIRCUIT #7 CB-1,2,3 FEED" "CIRCUIT #8 AI-195 FEED" "CIRCUIT #9 AI-44 FEED" "CIRCUIT #10 AI-58 & AI-59 FEED" "CIRCUIT #11 CB-20 FEED" "CIRCUIT #12 AI-42 & AI-60 FEED" "CIRCUIT #13 CB-4 FEED" "CIRCUIT #18 AI-43A &

AI-33C FEED"

c. Ensure BOTH of the following breakers are closed (AI-42B):

I-BUS-I2-1, " INSTRUMENT BUS 2 MAIN BREAKER" "CIRCUIT #1 AI-53 EMERG FEED" (continue)

R32

EOP/AOP ATTACHMENTS Page 32 of 173 Attachment 6 Minimizing DC Loads INSTRUCTIONS CONTINGENCY ACTIONS

4. (continued)

d. Place ALL of the following breakers in "OFF" (AI-42B):

"CIRCUIT #2 AI-50 FEED" "CIRCUIT #3 AI-105, 107, 107-1, 181" "CIRCUIT #4 AI-55 FEED" "CIRCUIT #5 AI-101B FEED" "CIRCUIT #6 IB-2A FEED" "CIRCUIT #7 CB-10,11 FEED" "CIRCUIT #8 EE-32 ALARM PNL (EMGY)"

"CIRCUIT #9 CB-4 FEED" "CIRCUIT #10 EE-32 AUX ALARM PNL (EMGY)"

"CIRCUIT #11 AI-44 FEED" "CIRCUIT #12 AI-187 FEED" "CIRCUIT #13 CB-1,2,3 FEED" "CIRCUIT #14 AI-195 FEED" "CIRCUIT #15 AI-270 FEED" "CIRCUIT #16 AI-43B &

AI-65B FEED" "CIRCUIT #17 CB-20 FEED" "CIRCUIT #18 AI-292 FEED" R32

EOP/AOP ATTACHMENTS Page 33 of 173 Attachment 6 Minimizing DC Loads INSTRUCTIONS CONTINGENCY ACTIONS NOTE Securing LO-12B, DC Seal Oil Pump takes precedence over complete venting of hydrogen from the Main Generator.

e. Stop LO-12B, Emergency Seal Oil Pump.

End of Attachment 6 R32

Source question for question 49.

QUESTIONS REPORT for ILO EXAM BANK 07-18-17 021 A station blackout has occurred due to a loss of 161 kv and 345 kv to the switchyard.

The Equipment Operator reports that step one of "Minimizing DC loads has been completed. What additional action must be taken in control room.

A. Stop LO-4, the DC Oil Pump, once the turbine stops turning.

B. Stop LO-12B, the DC Seal Oil Pump, when the generator stops turning.

C. Transfer AI-41A to its emergency DC power source.

D. Transfer AI-41B to its emergency DC Power source.

Saturday, August 04, 2012 14:45:05 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 050 Given the following plant conditions:

OI-DG-1, Attachment 1, "Idle Speed Start and Loading," is being performed Diesel Generator, D-1, is operating at 900 RPM The synchroscope is rotating slowly in the fast direction Incoming and outgoing voltages are matched.

What action should be taken after breaker 1AD1 is closed?

A. Increase voltage using voltage regulator, CS-90/D1 B. Decrease voltage using voltage regulator, CS-90/D1 C. Increase load using governor control, CS-65/D1 D. Decrease load using governor control, CS-65/D1 Saturday, August 04, 2012 15:17:20 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 50 Rev 0 KA #: 064000 A3.05 Tier 2 Group 1: Emergency Diesel Generators Ability to monitor automatic operation of the ED/G system, including: Operation of the governor control of frequency and voltage control in parallel operation Importance 2.8 / 2.9 CFR Number: 55.41(b)(5)

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.

Fort Calhoun Objective:

Explain the interrelationships between the EDG and the Electrical Distribution System.

Question Pedigree New question.

K/A Fit:

Question addresses operation of the governor control switch in parallel operation.

Choice A:

Distractor: Plausible if Applicant confuses voltage control with speed control. Incorrect because voltages are matched.

Choice B:

Distractor: Plausible if Applicant confuses voltage control with speed control and does not understand the relationship of the governor and load. Incorrect because voltages are matched.

Choice C:

Correct answer: Increasing the governor control switch setting will pick up load in parallel operation.

Choice D:

Distractor: Plausible if the Applicant does not understand the relationship of the governor and load. Incorrect because load must be increased.

KA#: 064000 A3.05 Bank Ref #: N/A LP# / Objective: 0713-05 01.02 Exam Level: RO-5 Cognitive Level: HIGH Source: NEW

Reference:

OI-DG-1 Handout: NONE Saturday, August 04, 2012 15:17:20 2

FORT CALHOUN STATION OI-DG-1 OPERATING INSTRUCTION C PAGE 9 OF 44 Continuous Use Attachment 1 - Idle Speed Start and Loading PROCEDURE (continued) () INITIALS 3

h. Verify the following dampers are open:

YCV-871G, Inlet YCV-871H, Inlet YCV-871E, Exhaust

i. Notify the Operator to perform the following:

Visual inspection Verify oil visible in the Upper Sightglass

j. Record the Start Time in the Control Room Log and FC-1046, Diesel Generator Demand Record.

4. IF Loading DG-1, THEN perform the following:

a. Notify the Operator to verify Jacket Water Inlet Temperature is greater than 120°F.

b. Place CS-65/D1, Diesel Generator D1 Governor, to Raise until the Diesel Speed is 900 rpm.

c. Verify the Generator Field flashed.

d. Place D1/BUS 1A3 Sync Switch to ON.

e. Adjust CS-90/D1, Diesel Generator D1 Voltage Regulator, until the RUNNING VOLTS is approximately matched to the INCOMING VOLTS on the Synchroscope or the ERF DGD Display.

NOTE Recommended synchroscope speed is less than 1 revolution per 10 seconds.

f. Adjust CS-65/D1 until the Synchroscope is rotating slowly in the FAST direction.

R54

FORT CALHOUN STATION OI-DG-1 OPERATING INSTRUCTION C PAGE 10 OF 44 Continuous Use Attachment 1 - Idle Speed Start and Loading PROCEDURE (continued) () INITIALS 4

NOTE Steps 4.g and 4.h may be performed without the procedure in hand.

Sign-offs may be completed after these steps are performed.

CAUTIONS

1. Load must be immediately picked up following closure of 1AD1 to prevent motorizing the Diesel Generator.

2. Governor controls are extremely sensitive.

g. WHEN the Synchroscope is between 11 and 12 O'CLOCK, THEN close 1AD1 BREAKER.

h. Place CS-65/D1 to Raise to pick up 250-350 KW.

i. Place D1/BUS 1A3 Sync Switch to OFF.

j. IF the Diesel is loaded and Y3287A, ERF 1A3 Bus Voltage, is greater than 4375 VAC, THEN immediately notify the System Engineer.

NOTES

1. Load should be maintained below the 2000 hr Rating vs Ambient Temp curve per TDB-III.26A Figure 1, DG-1 Output Power Rate.

2. Power factor may be determined by using TDB-III.26, Diesel Generator Capability Curve.

3. Current is normally limited to 400 amps at 2500 KW.

4. Diesel Generator manual loading and unloading rates should be maintained at less than 500 KW per minute.

5. Steps 4.k and 4.l may repeated as necessary while the diesel is loaded.

Sign-offs may be completed after these steps are performed.

k. Place CS-65/D1 to RAISE picking up the desired DG-1 Load.

R54

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 051 Given the following plant conditions:

The Radiation Monitoring system is in normal alignment with all components operable RM-052 is aligned to the Auxiliary Building stack A waste gas decay tank has ruptured Radiation levels on RM-052 and RM-062 are reading 5x104 CPM and rising RM-063 is reading low and is not changing Why is RM-063 not responding?

A. RM-063 must be manually placed into operation.

B. RM-062 countrate is not high enough to initiate sample flow to RM-063 C. RM-063 is not sensitive to noble gases contained in the waste gas decay tank D. RM-063 sample flow is isolated when RM-052 is aligned to the stack Wednesday, August 08, 2012 16:26:01 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 51 Rev 0 KA #: 073000 A4.02 Tier 2 Group 1: Process Radiation Monitoring System Ability to manually operate and/or monitor in the control room: Radiation monitoring system control panel Importance 3.7 / 3.7 CFR Number: 55.41(b)(11)

Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Fort Calhoun Objective:

LIST radiation monitors with automatic actuations and STATE the automatic actuations that occur.

Question Pedigree New question.

K/A Fit:

Question addresses monitoring RM-063 response during an accident.

Choice A:

Distractor: Plausible if Applicant believes RM-063 is operated like RM-064 and RM-065.

Incorrect because RM-063 will begin to operate when the countrate on RM-062 is high enough.

Choice B:

Correct answer. Sample flow to RM-063 will begin when countrate is higher on RM-062.

Choice C:

Distractor: Plausible if Applicant does not understand that RM-063 is a noble gas monitor with the capability for manual grab sampling for particulate and iodine. Incorrect because RM-063 is sensitive to noble gases.

Choice D:

Distractor: Plausible if Applicant believes RM-063 is connected to RM-052. Incorrect because it is connected to RM-062.

KA#: 073000 A4.02 Bank Ref #: N/A LP# / Objective: 0712-03 04.01 Exam Level: RO-11 Cognitive Level: HIGH Source: NEW

Reference:

STM 33 Handout: NONE Wednesday, August 08, 2012 16:26:02 2

SYSTEM TRAINING MANUAL VOL. 33 RADIATION MONITORING SYSTEM SAMPLE CONTROL ADM-600 PUMP CONTROL R A A HIGH ALE RT NORM FAIL G A R VENT ACCIDENT PURGE 1.38 E3 CPM PUMP OFF SAMPLE FLOW PUMP ON SKID HIGH/LOW 101 102 103 104 105 106 107 LOCAL OFF CR VENT PURGE KEYPAD START OFF CR CHECK RESET TEST SOURCE OFF ON MODE SET LOCAL CONTROL (AI-83)

RM-062 CONTROL PANELS CONTROL ROOM (AI-33A)

SAMPLE CONTROL ADM-600 PUMP CONTROL R A A HIGH ALE RT NORM FAIL G A R 1.38 E3 CPM PUMP OFF SAMPLE FLOW PUMP ON VENT ACCIDENT PURGE SKID HIGH/LOW 101 102 103 104 105 106 107 VENT PURGE KEYPAD STOP START CHECK RESET SOURCE TEST OFF ON G G MODE SET CONTROL CONTROL ROOM ROOM FIGURE 2-26: RM-062 CONTROL PANELS

• • Failure Modes 2.136 RM-062 fails on loss of AC power. There are no air-powered actuation valves in the system. Compensatory actions are required whenever radiation monitor RM-062 is removed from service or fails. Details are in the ARP, ODCM and OI-RM-1.

• • • Fail Position on Loss of Air or Power 2.137 RM-062 LED display becomes blank on loss of AC power. The CRHS signal is not actuated. The RM-062 AUX BLDG STACK TROUBLE alarm on AI-33C will actuate.

Accident Range Vent Stack Skid Monitor RM-063 2.138 The following is a detailed description of the accident range vent stack skid monitor, RM-063.

• • Function 2.139 The accident range vent stack monitor skid (RM-063) is an off-line accident range noble gas sample skid designed to work in con-junction with RM-062 to sample the Auxiliary Building vent stack during accident conditions when RM-062 reaches full scale.

Level: Basic > Intermediate > Advanced Revision - 18

• 41

SYSTEM TRAINING MANUAL VOL. 33 RADIATION MONITORING SYSTEM FIT-052-1 RM-050/51 RM-052 PCV-742 H CONTAINMENT CONTAINMENT F.C.

ONLY OR STACK CONTAINMENT AI-80 ROOM 69 AI-81 ROOM 69 PCV-742 F F.C. RM-051 RM-052 AUX.

BLDG.

RM-062 VENT OR STACK FE-CRHS 758 RM-063 RM-062 POST ACCIDENT NORMAL STACK STACK AI-84 AI-83 CORRIDOR 26 CORRIDOR 26 FIGURE 2-27: CONTAINMENT/AUX BLDG MONITOR ARRANGEMENT

• • Design/Specification 2.140 The skid assembly (AI-84) is an open-frame construction and located adjacent to RM-062 in the east end of Auxiliary Building Corridor 26 and is comprised of the following equipment:

• RM-063 receives sample flow through a 3/4-inch sample line from a splitter block located on RM-062.

• Motor-operated inlet isolation valve MV1 opens automatically if RM-062 exceeds 5.0 E6 CPM or RM-063 exceeds 5.0 E-3 uCi/cc and closes when RM-063 decreases to less than 5.0 E-3 uCi/cc.

• Grab sample connections are located on the skid inlet and outlet piping.

• Two trains of fixed filter particulate/Iodine filters.

2.141 NOTE: The original design was to have active detector channels for each particulate/Iodine filter; however, due to calibration prob-lems, it was decided to not use the channels and abandon the detec-tors (RE-044A/B) in place while still using the samplers for manual grab sampling.

• Solenoid valves SV1 and SV2 work together to close off sample flow from the particulate and Iodine filters and provide filtered room air for purging the flow through ion chamber.

Level: Basic > Intermediate > Advanced Revision - 18

• 42

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 052 During liquid releases to the river, what radiation monitors are used to ensure that the releases do not result in a dose that exceeds 1.5 mrem to the total body or 5 mrem to any organ during any given calendar quarter.

A. RM-054A/B and RM-055 B. RM-055 and RM-057 C. RM-057 and RM-064 D. RM-054A/B and RM-064 Wednesday, August 08, 2012 16:26:38 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 52 Rev 0 KA #: 073000 K5.03 Tier 2 Group 1: Process Radiation Monitoring System Knowledge of the operational implications as they apply to concepts as they apply to the PRM system: Relationship between radiation intensity and exposure limits Importance 2.9 / 3.4 CFR Number: 55.41(b)(12)

Radiological safety principles and procedures.

Fort Calhoun Objective Without references, state the purpose, applicability and content of Offsite Dose Calculation Manual.

Question Pedigree New question.

K/A Fit:

Question addresses radiation monitors used to ensure radionuclide concentrations maintain public doses within limits during liquid releases.

Choice A:

Correct answer: RM-054A/B monitor blowdown releases and RM-055 monitors monitor tank releases.

Choice B:

Distractor: Plausible because both monitor releases. Incorrect because RM-057 (condenser offgas) monitors gaseous releases.

Choice C:

Distractor: Plausible because both monitor releases. Incorrect because RM-057 and RM-064 (steam line) monitor gaseous releases.

Choice D:

Distractor. Plausible because both monitor releases. Incorrect because RM-064 monitors gaseous releases.

KA#: 073000 K5.03 Bank Ref #: N/A LP# / Objective: 0762-02 5.01.00 Exam Level: RO-12 Cognitive Level: LOW Source: NEW

Reference:

ODCM 3.1.2 Handout: NONE Wednesday, August 08, 2012 16:26:38 2

FORT CALHOUN STATION CH-ODCM-0001 OFF-SITE DOSE CALCULATION MANUAL REFERENCE USE PAGE 12 OF 130 Table 2.1.1 - Radioactive Liquid Effluent Monitoring Instrumentation MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release.

a. Liquid Radwaste Effluent Line (RM-055) 1 1, 5

b. Steam Generator Blowdown Effluent Line (RM-054 A and 11 2, 5 B)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line 1 3

b. Steam Generator Blowdown Effluent Line 1 3

3. Radioactivity Recorders

a. Liquid Radwaste Effluent Line 1 4

b. Steam Generator Blowdown Effluent Line 1 4 1 If one of the two radiation monitors is inoperable, the activity of both blowdown lines shall be monitored by the operable monitor within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the declaration of inoperability by the Shift Manager, or the action steps of ACTION 2, Table 2.1.1 should be performed on the Steam Generator that is not being monitored.

R21

FORT CALHOUN STATION CH-ODCM-0001 OFF-SITE DOSE CALCULATION MANUAL REFERENCE USE PAGE 24 OF 130 3.1.2 Dose from Radioactive Liquid Effluents A. Limiting Condition for Operation

1) The dose or dose commitment to an individual in unrestricted areas from radioactive materials in liquid effluents shall be limited to the following:

a) During any calendar quarter: Less than or equal to 1.5 mrem to the total body and 5 mrem to any organ; and b) During any calendar year: Less than or equal to 3 mrem to the total body and 10 mrem to any organ.

APPLICABILITY: At all times ACTION:

a) If the dose contribution, due to the cumulative release of radioactive materials in liquid effluents, exceeds the annual or quarterly dose objectives, submit a Special Report to the NRC, per Section 5.2.3, within 30 days.

B. Surveillance Requirements

1) Cumulative dose contributions from liquid effluents for the current calender quarter and the current calendar year shall be determined in accordance with the methodology and parameters in Part II of the Off-Site Dose Calculation Manual at least once per quarter.

R21

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 053 Which of the following buses directly provides power to Raw Water Pump, AC-10C A. 4160V Bus 1A1 B. 4160V Bus 1A3 C. 480V Bus 1B3C D. 480V Bus 1B3C-4C Wednesday, August 08, 2012 16:27:15 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 53 Rev 0 KA #: 076000 K2.01 Tier 2 Group 1:Service Water System Knowledge of bus power supplies to the following: Service water Importance 2.7 / 2.7 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation,signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

When given specific plant conditions APPLY operating principles to diagnose Raw Water System response.

Question Pedigree Bank question used on 2005 NRC exam K/A Fit:

Question addresses power supply to a Raw water pump.

Choice A:

Distractor: Plausible if Applicant does not know power supply to AC-10C. Incorrect because power supply is bus 1A3.

Choice B:

Correct answer: Bus 1A3 is the power supply for AC-10C.

Choice C:

Distractor: Plausible if Applicant does not know power supply to AC-10C. Incorrect because power supply is bus 1A3.

Choice D:

Distractor: Plausible if Applicant does not know power supply to AC-10C. Incorrect because power supply is bus 1A3.

KA#: 076000 K2.01 Bank Ref #: 07-11-19 037 LP# / Objective: 0711-19 01.00 Exam Level: RO-7 Cognitive Level: LOW Source: BANK

Reference:

STM-35 Handout: NONE Wednesday, August 08, 2012 16:27:15 2

STM-RW - Volume 35 Information Use Page 15 of 55 Raw Water System Rev. 27 2.1.3 Location A. The raw water pumps are located inside the Intake Structure, 35 feet back from the river. Two of the pumps are located inside their individual cells while the remaining two pumps share a cell.

Figure 2 Raw Water Pump Floor Plan Elevation 993'6" B. The raw water pump breaker control switches are on CB-1,2,3.

2.1.4 Power Supplies A. Each raw water pump is driven by a 3 phase, 200 hp, 1200 rpm, 4160 VAC motor that is powered from one-of-two vital electrical buses.

Bus 1A3 supplies pumps AC-10A/AC-10C, and bus 1A4 supplies pumps AC-10B/AC-10D. During normal operation, off-site power from the 161 kV System supplies vital buses 1A3 and 1A4, and backup power is provided by its respective emergency diesel generator.

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 054 The plant was in hot shutdown when the following Annunciators came into alarm in the control room:

"COOLING WATER PRESSURE LOW" "INSTRUMENT AIR PRESS LO" "PLANT AIR PRESS LO" Air pressure indication on both PI-1750 and PI-1700 was lowering.

The operator noted that the running bearing water pump and the running air compressors have tripped. The operators were able to start a backup air compressor, but it ran for only a few minutes. They were unable to start either bearing water pump.

What action is required to restore instrument air pressure?

A. Close the cross-tie valve between service air and instrument air.

B. Bypass the Instrument air dryers.

C. Align raw water backup cooling to an air compressor and then restart it.

D. Align potable water backup cooling to an air compressor and then restart it.

Wednesday, August 08, 2012 16:27:56 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 54 Rev 0 KA #: 078000 K1.04 Tier 2 Group 1: Instrument Air System Knowledge of the physical connections and/or cause-effect relationships between the IAS and the following systems: Cooling water to compressor Importance 2.6 / 2.9 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Explain the principles of Abnormal operation of the Compressed Air System in terms of flow paths, major paramaters, (temperature, pressure, flow, etc.), alarms and control devices.

Question Pedigree Bank question used on 2007 NRC exam.

K/A Fit:

Question addresses backup cooling water to the air compressors.

Choice A:

Distractor: Plausible because this is the correct action for a service air leak. Incorrect because a loss of cooling water has occurred.

Choice B:

Distractor: Plausible because this is the correct action for a clogged dryer. Incorrect because a loss of cooling water has occurred.

Choice C:

Distractor: Plausible because Raw water backup cooling is used for many plant components. Incorrect because it is not used for the air compressors.

Choice D:

Correct answer: The air compressors have tripped due to a loss of normal cooling (bearing water). Backup cooling (potable water) should be aligned to the air compressors KA#: 078000 K1.04 Bank Ref #: 07-11-07 022 LP# / Objective: 0711-07 01.05 Exam Level: RO-10 Cognitive Level: HIGH Source: BANK

Reference:

OI-CA-5 Handout: NONE Thursday, August 09, 2012 14:33:16 2

FORT CALHOUN STATION OI-CA-5 OPERATING INSTRUCTION C PAGE 3 OF 12 Continuous Use Attachment 1 - Initiation of Backup Cooling to Air Compressor(s)

PREREQUISITES () INITIALS

1. Procedure Revision Verification Revision No. Date:

2. Air Compressor(s) are operating per OI-CA-1.

3. Potable Water System is in service per OI-PW-1.

4. Verify Blair water supply pressure is greater than 45 psig on PI-1601-2, Blair Water Pressure gauge (Water Plant Office).

5. If this will be the only in service Air Compressor, protected equipment labels should be placed at the component and its associated breaker.

PROCEDURE

1. IF CA-1A is to be lined up to backup cooling, THEN complete the following:

a. Ensure compressor is not running.

b. Close the following valves:

AC-586, Air Compressor Aftercooler CA-2A Inlet Valve AC-588, Air Compressor CA-1A Outlet Valve AC-584, Air Compressor CA-1A Intercooler Inlet Valve AC-589, Air Compressor CA-1A Aftercooler CA-2A Outlet Valve NOTE AC-1043, Potable Water Outlet Valve AND AC-1045, Aftercooler Potable Water Outlet Valve may need to be throttled to prevent Air Compressor trip due to low cooling water pressure.

c. Open the following valves:

AC-1042, Air Compressor CA-1A Intercooler Potable Water Inlet Valve AC-1044, Air Compressor CA-1A Aftercooler CA-2A Potable Water Inlet Valve AC-1028, Comp Cooling Wtr Valve FCV-1990A Bypass Valve R15

FORT CALHOUN STATION OI-CA-5 OPERATING INSTRUCTION C PAGE 4 OF 12 Continuous Use Attachment 1 - Initiation of Backup Cooling to Air Compressor(s)

PROCEDURE (continued) () INITIALS 1

NOTE The air compressors will automatically trip on low jacket cooling water pressure at 30 psig or a cooling water temperature of 130°F.

d. Throttle open the following valves to maintain a pressure of 40 to 50 psig on PI-1988A:

AC-1045, Air Compressor CA-1A Aftercooler CA-2A Potable Water Outlet Valve AC-1043, Air Compressor CA-1A Potable Water Outlet Valve _______

e. Start CA-1A per OI-CA-1. _______

f. IF air compressor trips, THEN raise cooling water pressure by throttling closed on the Aftercooler and/or air compressor cooling water outlet valve(s).

1) Adjust valves to desired pressure:

AC-1045 ___

AC-1043

2) Go to Step 1.e.

g. Close AC-1028, Comp Cooling Wtr Valve FCV-1990A Bypass Valve.

h. Verify potable water cooling flow by checking the following:

FI-1955A, Air Compressor CA-1A Cooling Water Flow Indicator TI-1702A, Air Compressor Aftercooler CA-2A Temperature Indicator

i. Allow temperatures to stabilize and if necessary, adjust flow in small increments using Aftercooler/Potable Water Outlet valves to maintain temperature and pressure.

AC-1045 AC-1043 R15

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 055 The plant is operating at full power. A containment pressure reduction is in progress.

Which of the following conditions is considered to be a Loss of Containment Integrity?

A. HCV-746A and HCV-746B, "Containment Pressure Relief Valves," are both open.

B. The outer Personnel Air Lock door has failed it's seal leakage test C. An automatic CIAS valve failed a surveillance test due to excessive stroke time.

D. The Limitorque operator for a locked closed containment isolation MOV is inoperable.

Wednesday, August 08, 2012 16:28:31 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 55 Rev 0 KA #: 103000 K3.02 Tier 2 Group 1: Containment System Knowledge of the effect that a loss or malfunction of the containment system will have on the following: Loss of containment integrity under normal operations Importance 3.8 / 4.2 CFR Number: 55.41(b)(5)

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.

Fort Calhoun Objective:

DESCRIBE how containment integrity is monitored.

Question Pedigree Bank question used on 1995 NRC exam. Minor rewording, not counted as modified question.

K/A Fit:

Question addresses loss of containment integrity during normal operation.

Choice A:

Distractor: Plausible because these are two containment isolation valves in series.

Incorrect because they may both be opened during a containment pressure reduction with an approved release permit.

Choice B:

Distractor: Plausible because the PAL doors function to maintain containment integrity.

Incorrect because the inner PAL door is still intact.

Choice C:

Correct answer per AOP-12 Choice D:

Distractor: Plausible because an inoperable containment isolation MOV could be a loss of containment integrity. Incorrect because it is locked closed.

KA#: 103000 K3.02 Bank Ref #: 07-11-08 023 LP# / Objective: 0711-08 02.01 Exam Level: RO-5 Cognitive Level: LOW Source: BANK

Reference:

AOP-12 Handout: NONE Wednesday, August 08, 2012 16:28:31 2

AOP-12 Page 2 of 8 1.0 PURPOSE This procedure provides guidance in the event of the loss of required Containment Integrity.

2.0 ENTRY CONDITIONS A loss of Containment Integrity has occurred which may be indicated by any of the following:

A. Non-automatic Containment Isolation Valves are open or blind flanges are not sealed as required for Containment Integrity.

B. The Equipment Hatch is not properly sealed.

C. Neither Personnel Air Lock Door is properly sealed.

D. Automatically operated Containment Isolation Valves are inoperable and not locked closed.

E. Containment building leakage rates have exceeded the allowable limits of Technical Specification 3.5, Containment Test.

F. Noticeable air leakage from Containment.

R7

Source question for question 55 QUESTIONS REPORT for ILO EXAM BANK 07-11-08 023 With the plant at power and no evolutions in progress, which of the following conditions is a Loss of Containment Integrity?

A. HCV-746A (Pressure Relief) is opened.

B. One of the Personnel Air Lock doors is open.

C. An automatic CIAS valve fails a surveillance test due to not cycling full open.

D. A containment isolation MOV is inoperable but, is locked closed.

Monday, August 06, 2012 09:36:49 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 056 Excessive leakage from the RCS through the Control Element Drive Mechanism shaft seals will result in rising level in _______________.

A. The Reactor Coolant Drain Tank B. The Pressurizer Quench Tank C. The Spent Regenerant Tank D. The Volume Control Tank Monday, August 06, 2012 09:50:45 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 56 Rev 0 KA #: 001000 K3.02 Tier 2 Group 2: Control Rod Drive System Knowledge of the effect that a loss or malfunction of the CRDS will have on the following: RCS Importance 3.4 / 3.5 CFR Number: 55.41(b)(3)

Mechanical components and design features of the reactor primary system.

Fort Calhoun Objective:

Describe the interface/interaction between the CRDS and the following systems/components:

Question Pedigree Bank Question.

K/A Fit:

Question addresses Control Rod Drive system leakage to a RCS component.

Choice A:

Correct answer.

Choice B:

Distractor: Plausible if Applicant does not understand where CRDS seal leakage goes.

Incorrect because it goes to the RCDT.

Choice C:

Distractor: Plausible if Applicant does not understand where CRDS seal leakage goes.

Incorrect because it goes to the RCDT.

Choice D:

Distractor: Plausible if Applicant does not understand where CRDS seal leakage goes.

Incorrect because it goes to the RCDT.

KA#: 001000 K3.02 Bank Ref #: 07-12-26 01.02 LP# / Objective: 0712-26 01.02 Exam Level: RO-3 Cognitive Level: LOW Source: BANK

Reference:

STM 11 Handout: NONE Monday, August 06, 2012 09:50:45 2

SYSTEM TRAINING MANUAL VOL. 11 CONTROL ROD DRIVE SYSTEM DETAILED SYSTEM DESCRIPTION 1/4" NPT PLUG REF AUTOCLAVE NUT CEDM HOUSING WASHER AUTOCLAVE PLATE GASKET FIGURE 2-19: CEDM SEALING 2.21 The tool access tube is sealed by a bolted closure with a flexitalic gasket. The drive shaft from the clutch penetrates the autoclave and is sealed with a rotating mechanical seal. The mechanical seal uses a chrome oxide and graphitar mating surface. Approximately 1.5 gpm component cooling water (CCW) flow is supplied to a jacket around the seal to ensure the temperature does not exceed 250°F. Leakage through the seal is piped to the reactor coolant drain tank, and excessive RCS leakage is sensed by thermocouples installed in the leakoff lines which alarm at 200°F on the plant computer.

2.22 There are two clutches installed in each shutdown and regulating group CEDM, a trip clutch and an anti-reverse rotation clutch. The trip clutch transmits torque from the rod drive motor to the drive shaft. The trip clutch has two toothed sections. When the 33V DC power is applied, the lower toothed section rises and engages the teeth of the driving section. A reactor trip interrupts clutch power and allows gravity to pull the lower section away and disengage the clutch teeth. The pinion, drive shaft, and lower portion of the clutch rotate as the control rod falls into the core.

Level: Basic > Intermediate > Advanced Revision - 19

• 19

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 057 Following a control room evacuation, AOP-07, "Control Room Evacuation," directs the Operators to maintain RCS Pressure 2050-2150 psia by controlling "Pressurizer Back-Up Heaters," Bank 4 Groups 10, 11 and 12.

What 480 volt MCC powers these heaters?

A. MCC-3A4 B. MCC-3B3 C. MCC-3C4C-2 D. MCC-4C1 Thursday, August 09, 2012 14:58:46 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 57 Rev 0 KA #: 011000 K2.02 Tier 2 Group 2: Pressurizer Level Control System Knowledge of bus power supplies to the following: PZR heaters Importance 3.1 / 3.2 CFR Number: 55.41(b)(3)

Mechanical components and design features of the reactor primary system.

Fort Calhoun Objective LIST the power supplies for the pressurizer heaters.

Question Pedigree New question K/A Fit:

Question addresses power supply to pressurizer heater bank 4, groups 10, 11 and 12.

Choice A:

Distractor: Plausible if Applicant does not know power supply to these heaters. Incorrect because the power supply is MCC-4C1 Choice B:

Distractor: Plausible if Applicant does not know power supply to these heaters. Incorrect because the power supply is MCC-4C1.

Choice C:

Distractor: Plausible if Applicant does not know power supply to these heaters. Incorrect because the power supply is MCC-4C1.

Choice D:

Correct answer. The power supply is MCC-4C1 KA#: 011000 K2.02 Bank Ref #: N/A LP# / Objective: 0711-20 01.06C Exam Level: RO-3 Cognitive Level: LOW Source: NEW

Reference:

AOP-07 Handout: NONE Thursday, August 09, 2012 14:58:46 2

AOP-07 Page 12 of 56 Section I - Plant to Hot Shutdown INSTRUCTIONS CONTINGENCY ACTIONS

5. Monitor Reactor power on NR-004 and NI-004, "WIDE RANGE NEUTRON FLUX CHAN "D" SIGNAL PROCESSOR" (AI-212).

CAUTION Charging to the RCS may cause overpressurization due to the isolation of Letdown and RCS Heatup.

6. Maintain PZR level (45%-60%) by operating CH-1B, Charging Pump (AI-185).

7. Maintain a record of Charging Pump run time to support estimates of RCS boron concentration.

8. Maintain RCS pressure 2050-2150 psia by controlling "PRESSURIZER BACK-UP HEATERS", Bank 4 Groups 10, 11 and 12 (MCC-4C1).

R14

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 058 Given the following plant conditions:

A power increase is in progress per OP-4, LOAD CHANGE AND NORMAL POWER OPERATION The reactor is at 80% power Group 4 CEAs are at 52 inches All other CEAs are fully withdrawn.

Using the attached PDIL curve, determine what procedure should be entered as a result of these conditions?

A. AOP-02, "CEA and Control System Malfunctions" B. AOP-03, "Emergency Boration" C. AOP-05, "Emergency Shutdown" D. EOP-00, "Standard Post-Trip Actions" Monday, August 06, 2012 10:11:53 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 58 Rev 0 KA #: 014000 2.4.04 Tier 2 Group 2: Rod Position Indication System Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Importance 4.5 / 4.7 CFR Number: 55.41(b)(5)

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.

Fort Calhoun Objective:

DISCUSS the power dependent insertion limit including: Describing required action when the Emergency Boration Limit is exceeded.

Question Pedigree Bank question used on 2009 NRC exam.

K/A Fit:

Question addresses rod position that requires entry into AOP-03.

Choice A:

Distractor Plausible because AOP-02 can be entered for CEA mispositioning. Incorrect because PDIL is violated and emergency boration is required.

Choice B:

Correct answer: PDIL is violated and AOP-03, emergency boration is required.

Choice C:

Distractor Plausible because AOP-05 can be entered for CEA mispositioning. Incorrect because PDIL is violated and emergency boration is required.

Choice D:

Distractor Plausible because EOP-00 can be entered for CEA mispositioning. Incorrect because PDIL is violated and emergency boration is required.

KA#: 014000 2.4.04 Bank Ref #: 07-05-09 LP# / Objective: 0705-09 01.13E Exam Level: RO-5 Cognitive Level: HIGH Source: BANK

Reference:

AOP-03 Handout: PDIL CURVE Thursday, August 09, 2012 15:03:26 2

AOP-03 Page 2 of 9 1.0 PURPOSE This procedure provides guidance in the event Emergency Boration of the RCS is required.

2.0 ENTRY CONDITIONS The following conditions require the initiation of Emergency Boration using this procedure:

A. Uncontrolled Plant Cooldown.

B. Two or more CEAs inserted below the Emergency Boration Limit of the COLR Power Dependent Insertion Limit Curve.

C. Reactivity anomaly in a positive direction with all CEA's inserted.

D. Unexpected rise in count rate or unexpected doubling of count rate on two or more channels during refueling.

E. Failure of more than one Regulating or Shutdown CEA to insert following a Reactor trip.*

• Guidance for Emergency Boration with these conditions is provided in more appropriate procedures.

R5

FORT CALHOUN STATION TDB-VI TECHNICAL DATA BOOK PAGE 12 OF 19 R40

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 059 The wiring from one Core Exit Thermocouple (CET) has just failed open.

How would "Representative CET Temperature" as indicated on the ERF computer be affected?

A. The Representative CET Temperature would increase to 2300°F.

B. The Representative CET Temperature would decrease to containment temperature.

C. The Representative CET Temperature would decrease to 32°F.

D. The Representative CET Temperature would not change.

Monday, August 06, 2012 10:36:33 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 59 Rev 0 KA #: 017000 K6.01 Tier 2 Group 2: In-Core Temperature Monitor System Knowledge of the effect of a loss or malfunction of the following ITM system components: Sensors and detectors Importance 2.7 / 3.0 CFR Number: 55.41(b)(2)

General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Fort Calhoun Objective:

EXPLAIN the principles of operation (both normal and abnormal) of the QSPDS System in terms of signal flow paths, major parameters (temperature, pressure, flow, etc.)

alarms and control devices.

Question Pedigree New question.

K/A Fit:

Question addresses how a loss of a CET would affect displayed representative CET temperature.

Choice A:

Distractor: Plausible if Applicant believes CET will fail to maximum scale but does not realize how representative CET temperature is calculated.

Choice B:

Distractor: Plausible if Applicant believes CET will fail to reference junction temperature but does not realize how representative CET temperature is calculated.

Choice C:

Distractor: Plausible if Applicant believes CET will fail to minimum scale but does not realize how representative CET temperature is calculated.

Choice D:

Correct answer: Representative CET temperature is calculated by rejecting the outlier CET temperatures and averaging the rest.

KA#: 017000 K6.01 Bank Ref #: N/A LP# / Objective: 0712-23 01.06 Exam Level: RO-2 Cognitive Level: HIGH Source: NEW

Reference:

OI-QSP-1 Handout: NONE Thursday, August 09, 2012 15:09:08 2

FORT CALHOUN STATION OI-QSP-1 OPERATING INSTRUCTION I PAGE 13 OF 21 Information Use Attachment 7 - Core Exit Thermocouple (CET) Temperature Calculation PREREQUISITES () INITIALS

1. Procedure Revision Verification Revision Number Date:

2. At least one Qualified Safety Parameter Display System channel is operational.

PROCEDURE

1. The QSPDS displays individual Core Exit Temperatures with a Core Map, the highest and next highest Core Exit Temperature in each quadrant, and a representative Core Exit Temperature.

2. The Representative Core Exit Temperature is calculated based on statistical analysis with practical checks from other inputs.

a. Using the Representative CET Temperature the Pressure AND Temperature Saturation Margins are calculated and displayed.

However, the CET calculated variables will be flagged with a question mark during the time that the QSPDS is identifying Out-of-Range and Suspect CET inputs.

b. WHEN the QSPDS completes one (1) complete cycle of the CET algorithm without flagging or removing a flag from any CET input, THEN the inputs are considered stable AND accurate and the question mark is removed from the calculated variables.

c. The CET calculated variables will also be flagged with all question marks if the number of valid CET inputs is less than nine (9).

3. Inputs considered invalid are either failed or deviated from the mean of the CET inputs by a specific amount.

a. Failed inputs are displayed as all question marks.

b. Deviated inputs are displayed as a question mark in front of the displayed value indicating it is suspect.

R8

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 060 Given the following plant conditions:

The plant was operating at full power A large LOCA occurred in containment All safeguards equipment operated as designed The CNTMT VENT FAN VA-3A FILTER TEMP HI annunciator alarmed.

TIC-866 indicates that the charcoal bed temperature is 460°F What action should be taken regarding HCV-864, "Charcoal Filter Spray Valve," in response to these conditions?

A. Verify HCV-864 opened automatically.

B. Manually open HCV-864.

C. Manually initiate Containment Spray and open HCV-864.

D. HCV-864 is not operated per FCS procedures.

Wednesday, August 08, 2012 16:29:45 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 60 Rev 0 KA #: 027000 A2.01 Tier 2 Group 2: Containment Iodine Removal System Ability to (a) predict the impacts of the following malfunctions or operations on the CIRS; and (b) based on those predictions, use Procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High temperature in the filter system Importance 3.0 / 3.3 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Explain automatic controls associated with the Containment Air Cooling and Filtering System.

Question Pedigree New question.

K/A Fit:

Question addresses procedural response for a high charcoal filter temperature.

Choice A:

Distractor: Plausible if Applicant believes the valve opens automatically. Incorrect because it doesn't.

Choice B:

Distractor: Plausible if Applicant believes there is procedural direction to open the valve and doesn't realize containment spray is also required for filter spray. Incorrect because the is no procedural direction to open this valve.

Choice C:

Distractor: Plausible if Applicant believes there is procedural direction and realizes that containment spray flow is required filter spray. Incorrect because there is no procedural direction.

Choice D:

Correct answer. HCV-864 was designed to mitigate a charcoal filter fire in the original plant design. However, opening HCV-864 was removed from FCS procedures following an engineering analysis.

KA#: 027000 A2.01 Bank Ref #: N/A LP# / Objective: 0714-02 01.04 Exam Level: RO-10 Cognitive Level: HIGH Source: NEW

Reference:

ARP-AI-30A/A33-1 Handout: NONE Wednesday, August 08, 2012 16:29:45 2

FORT CALHOUN STATION ARP-AI-30A/A33-1 ANNUNCIATOR RESPONSE PROCEDURE CONTINUOUS USE PAGE 56 OF 68 Panel: AI-30A Annunciator: A33-1 Window: H-4 CONTAINMENT FILTER HIGH TEMPERATURE CNTMT VENT FAN SAFETY RELATED VA-3A FILTER TEMP HI Tech Spec

References:

2.4 Initiating Device TIC-866 Setpoint >450ºF Power AI-41A__

OPERATOR ACTIONS

1. Verify TIC-866, Containment Filter Temperature High (AI-30A).

2. Verify TE-715, Containment Temperature High.

CAUTION Water vapor may cause a significant temperature rise (90-140ºF) when air flow through the filter is stopped.

2.1 Ensure VA-3A, Cntmt Vent Fan is in operation.

2.2 Ensure CCW is supplied to VA-15A.

2.3 Verify filter temperature is reduced.

3. Ensure compliance with requirements of Technical Specification 2.4.

PROBABLE CAUSES Water vapor in VA-15A housing with no flow thru VA-3A REFERENCES 161F597 Sh 10 09810 136B2331 Sh 139 07251 IC-CP-01-0866 11405-M-1 Sh 1 10431 R24

IHB: 07-14-02 Revision: 12 OUTLINE OF INSTRUCTION COMMENTS II. C. 6.

e. Each cooling and filtering unit contains Adsorption - assimilation 288 charcoal filter cells. of gas, vapor, or dissolved matter by the surface of a solid or liquid.

f. Rated flow through each cell is 383 cfm.

g. Elemental iodine adsorption efficiency is 99.9%.

h. Overheating of the Charcoal is not probable, due to the following:

(1) Charcoal filter ignition temperature is 644F.

(2) Maximum Containment temperature after a DBA is 417F (EA-93-022),

Containment temperature after a Super Heated Condition or a 300 PSIA spike is reduced to less than 288F in 65 seconds.

(3) Maximum Charcoal temperature rise due to Iodine decay is less than 6F (5.88) approximately 500 minutes after an accident which is well after DBA or MSLB peak temperatures.

i. FCS no longer uses Containment Spray to douse the Charcoal filters, due to the following:

(1) It has been found that temperature based fire detection methods in air cleaning units are unreliable.

20

IHB: 07-14-02 Revision: 12 OUTLINE OF INSTRUCTION COMMENTS II. C. 6. i.

(2) Water sprays were found to be ineffective, and spraying the hot carbon beds results in an initial temperature rise due to the heat of adsorption of water (90-140F).

(3) U.S. NRC Regulatory Guides 1.52, Rev. 2, and 1.140, do not require water sprays for adsorber beds.

j. The EOPs/AOPs no longer reference the use of HCV-864 and 865 for dousing the Charcoal filters, but the following is a description of the equipment locations and controls since they are still present in the system.

(1) Inside containment are two air-operated isolation valves one from each spray header leading to the filter.

(2) Valve HCV-864 operated from EO 1.2 panel AI-30A provides dousing water to filter VA-6A.

(3) Valve HCV-865 operated from EO 1.2 panel AI-30B provides dousing water to filter VA-6B.

(4) Handswitch positions CLOSE/

NORMAL/OPEN.

(5) Green shut/red open indicating EO 1.1 lights above handswitch.

k. Temperature indication Visual Aid 1.4 SHB page 6 EO 1.1 21

IHB: 07-14-02 Revision: 12 OUTLINE OF INSTRUCTION COMMENTS II. C. 6. kl.

(1) Thermistor strip temperature detectors pass through 24 uni-formly separated beds within each cooling and filtering unit.

(2) TE-866 (TE-867) provides temperature indication on panel AI-30A (AI-30B) for vent fan VA-3A (VA-3B) filter VA-6A (VA-6B).

l. Alarms NOTE: Refer to current revision of ARP.

EO 1.3 (1) Containment ventilation fan VA-3A filter temperature HI on panel AI-30A annunciator A33-1.

(a) Setpoint 450F (b) Trip device TIC-866 NOTE: Ask what actions are required.

(2) Containment ventilation fan VA-3B filter temperature HI on panel AI-30A annunciator A34-1.

(a) Setpoint 450F (b) Trip device TIC-867

7. Cooling coils (VA-1A and VA-1B) (VA-8A and VA-8B)

a. Located within containment cooling and Visual Aid 1.3 filtering and containment cooling units SHB page 5 VA-3A, VA-3B, VA-7C and VA-7D.

b. Cool containment air.

22

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 061 Given the following plant conditions:

A LOCA occurred Containment Pressure is 8.2 psig and lowering You have been directed to place the containment hydrogen analyzers in service What action must be taken to open the H2 Analyzer Containment Isolation Valves?

A. Place the valve control switches in "OPEN" B. Place the valve control switches in "O'RIDE" C. Reset CIAS, then place the valve control switches in "OPEN" D. Reset CIAS, then place the valve control switches in "O'RIDE" Monday, August 06, 2012 11:13:55 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 61 Rev 0 KA #: 028000 A1.02 Tier 2 Group 2: Hydrogen Recombiner and Purge Control System Ability to predict and/or monitor changes in parameter (to prevent exceeding design limits) associated with operating the HRPS controls including: Containment pressure Importance 3.4 / 3.7 CFR Number: 55.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Using the operating instructions DESCRIBE how to operate the hydrogen analyzer as required by plant conditions.

Question Pedigree New question K/A Fit:

Question addresses opening the hydrogen analyzer containment isolation valves with containment pressure above setpoint that allows reset of containment isolation.

Choice A:

Distractor: Plausible if Applicant does not recognize that the valves will not open in the open position with CIAS. Incorrect because then will not open with control switches in the open position.

Choice B:

Correct answer. Override position allows opening these valves with a CIAS.

Choice C:

Distractor: Plausible if Applicant believes CIAS can be reset. Incorrect because containment pressure is too high to reset CIAS.

Choice D:

Distractor: Plausible if Applicant does not realize that CIAS does not have to be reset to open valves in override position. Incorrect because CIAS can not be reset at this containment pressure.

KA#: 028000 A1.02 Bank Ref #: N/A LP# / Objective: 0714-03 02.00 Exam Level: RO-7 Cognitive Level: HIGH Source: NEW

Reference:

EOP/AOP ATT 16 Handout: NONE Monday, August 06, 2012 11:13:55 2

EOP/AOP ATTACHMENTS Page 83 of 173 Attachment 16 Containment Hydrogen Analyzer Startup INSTRUCTIONS CONTINGENCY ACTIONS

1. Start the Hydrogen Analyzers, by performing the following steps:

a. Open ONE of the Containment Hydrogen Sampling Valves, HCV-820C/D/E/F/G/H.

b. Open ONE of the Containment Hydrogen Sampling Valves, HCV-883C/D/E/F/G/H.

c. Place ALL of the following switches in "O'RIDE":

"H2 ANALYZER VA-81A ISOLATION VALVES OUTBD HCV-820A/821A" "H2 ANALYZER VA-81B ISOLATION VALVES INBD HCV-883A/884A" "H2 ANALYZER VA-81A ISOLATION VALVES INBD HCV-820B/821B" H2 ANALYZER VA-81B ISOLATION VALVES OUTBD HCV-883B/884B" (continue)

R32

SYSTEM TRAINING MANUAL VOL. 23 CONTAINMENT H2 ANALYZER SYSTEM

• • Instrumentation and Controls DETAILED SYSTEM DESCRIPTION 2.13 The following is a detailed description of the instrumentation and controls of the hydrogen analyzers, VA-81A/B.

• • • Local 2.14 The local panels and remote panels each have a function selector switch and local-remote selector switch. However, the analyzers are normally operated from the remote panels in the control room.

System operation from the local panels is normally performed by I&C technicians during calibration and surveillance testing. Each local panel has a breaker and motor-starter controller with a reset push button in the rear of the panels in Room 59.

INTERNAL JUNCTION CUSTOMER JUNCTION TERMINAL BLOCKS CONTROL TERMINAL BLOCKS CUSTOMER JUNCTION SIGNAL TERMINAL BLOCK CIRCUIT TB1 TB5 SAMPLE EXIT BREAKER VALVE TB3 SAMPLE ENTRY TB6 PUMP VALVE TB2 STARTER TB4 SAMPLE PRESSURE SWITCH (1 REQ'D)

REGENT & CALIB. REGENT & CALIB.

GASES PRESSURE GASES SOLENOID SWITCH (2 REQD.) VALVES (3 REQ'D) 275'F HEATED COMPARTMENT REGENT & CALIB. BROOKS FLOW GASES PRESSURE METER (1 REQD.)

REGULATORS (2 REQ'D) ANALYZER BYPASS FLOW (FIC-3)

SAMPLE PUMP z 1 HP, 460 VAC AIR-COOLED 3 0, 60 H HEAT EXCHANGER GROUND BUS BAR 1x26-1/2x1/4" COPPER REAR VIEW PUMP MOUNTING (DOOR OMITTED FOR CLARITY) BOLTS FIGURE 2-3: LOCAL CONTROL PANEL - REAR VIEW

• • • Remote 2.15 The inboard containment isolation valves and outboard isolation valves are operated in pairs via a three-position control switch on the respective containment isolation panel, AI-43A/B. The switch positions are CLOSE, OPEN and OVERRIDE. Control switches HC-820A/821A and HC-883A/884A are on AI-43A, and control switches HC-820B/821B and HC-883B/884B are on AI-43B.

Level: Basic > Intermediate > Advanced Revision - 9

• 8

SYSTEM TRAINING MANUAL VOL. 23 CONTAINMENT H2 ANALYZER SYSTEM The control switches for the sample isolation valves are grouped in DETAILED SYSTEM DESCRIPTION 2.16 pairs, but only one sample valve may be opened at any one time by an individual switch. The control switch positions are OPEN, BOTH VALVES CLOSED and OPEN. For example, HCV-820C is opened by placing the control switch in the first (left-hand side)

OPEN position, and HCV-820D is opened by placing the control switch in the second (right-hand side) OPEN position on AI-65A.

The mid-position sends a signal to close both valves. Each valve has two status lights (green-closed and red-open) on AI-65A/B for position indication.

• • • Interlocks 2.17 Actuation of a containment isolation actuation signal (CIAS) will close the inboard and outboard sample isolation valves. However, the signal may be overridden with its respective control switch on panel AI-43A/B, as allowed by Technical Specifications 2.12 and 2.15.

• • • Alarms and Indications 2.18 The following alarms are located on control room panel AI-65A (similar alarms are located on AI-65B):

• CONTAINMENT H2 LEVEL HI alarm actuates if hydrogen concentration exceeds 4% on VA-81A.

• CONTAINMENT H2 SAMPLING SYSTEM COMMON FAILURE alarm actuates for any of the following: analyzer low temperature, sample flow low, reagent or calibration gas low pressure, hydrogen analyzer cell malfunction, hydrogen ana-lyzer power supply failure.

• CONTAINMENT H2 SAMPLING SYSTEM REMOTE /

LOCAL OFF NORMAL alarm actuates whenever the LOCAL position is selected on the push button control switch at remote panel AI-65A or local panel VA-81A.

• CONTAINMENT H2 SAMPLE VALVES LOSS OF POWER alarm actuates if power is lost to any of the six sample valves:

HCV-820C/D/E/F/G/H.

• CONTAINMENT H2 SAMPLE VALVES OFF NORMAL alarm actuates if the control switch on AI-65A is in the OPEN position for any of the six sample isolation valves.

2.19 The following alarm is on control room panel AI-43A (a similar alarm is on panel AI-43B) - H2 PURGE/ANALYZER ISOLA-TION VALVES OVERRIDE alarm actuates if the control switch for any hydrogen analyzer (or hydrogen purge) isolation valve associated with AI-43A is in the OVERRIDE position.

Level: Basic > Intermediate > Advanced Revision - 9

• 9

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 062 Due to a Loss of Instrument Air, both MSIVs (Main Steam Isolation Valves) failed closed.

Assuming NO operator action, which of the following Control Room indications would indicate that MS-291 and MS-292 (Air Operated Safety Valves) are open?

A. The red position indication lights on CB-10 B. The valve position icons on the DCS display C. Steam Generator pressure indicating approximately 900 psia D. Steam Generator pressure indicating approximately 1000 psia Wednesday, August 08, 2012 16:30:20 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 62 Rev 0 KA #: 035000 K4.06 Tier 2 Group 2: Steam Generator System Knowledge of S/GS design feature(s) and/or interlock(s) which provide for the following:

S/G pressure Importance 3.1 / 3.4 CFR Number: 55.41(b)(5)

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.

Fort Calhoun Objective:

EXPLAIN how changes in plant conditions will effect the Main Steam System.

Question Pedigree Bank Question used on the 2004 NRC exam. Two distractors changed. Not counted as modified question.

K/A Fit:

Question addresses automatic operation of S/G safety valves and setpoints.

Choice A:

Distractor: Plausible because red light will light if MS-291 and MS-292 are opened from the control room. Incorrect because red light will not light if the open due to high pressure.

Choice B:

Distractor: Plausible because DCS display added many new indications. Incorrect because the positions of MS-291 and MS-292 are not indicated on DCS.

Choice C:

Distractor: Plausible because the steam dump and bypass system normally controls S/G pressure at 900 psia. Incorrect because both MSIVs are closed.

Choice D:

Correct answer. The opening setpoint for MS-291 and MS-292 is 1000 psia.

KA#: 035000 K4.06 Bank Ref #: 07-11-17 032 LP# / Objective: 0711-17 01.06 Exam Level: RO-5 Cognitive Level: HIGH Source: BANK

Reference:

STM 25 Handout: NONE Wednesday, August 08, 2012 16:30:20 2

STM-MS&SG - Volume 25 Information Use Page 49 of 103 Main Steam and Steam Generator System Rev. 34 Figure 2 Main Steam System Diagram 2.3.2 (continued)

C. Each valve starts to open when the steam pressure reaches the initial setpoint and will be fully open when the steam pressure reaches 103% of the initial setpoint value (3% accumulation). The safety valves have a blowdown of 4% which means the valve is fully closed when the steam pressure drops below 96% of the initial relief setpoint. When all five main steam safety valves on one main steam header are simultaneously wide open, they are capable of relieving 3,302,500 lbm/hr. The actual setpoints are shown below:

Table 1 - Main Steam Safety Valve Setpoints Valves Setpoints MS-291/292 1000 psia MS-275/279 1015 psia MS-276/280 1025 psia MS-277/281 1040 psia MS-278/282 1050 psia

STM-MS&SG - Volume 25 Information Use Page 50 of 103 Main Steam and Steam Generator System Rev. 34 2.3.2 (continued)

D. Safety valves MS-275 through MS-282 (total of eight) are each 6-inch valves, capable of discharging 794,062 pounds mass of steam per hour (lbm/hr). The remaining two 2.5-inch valves (MS-291 and MS-292) each have a discharge capacity of 126,299 lbm/hr. Valves MS-291/292 have a pneumatic operator to allow opening the valves from the Control Room.

This allows the steam generators to be used for plant cooldown or RCS temperature control if the MSIVs are closed or the main condenser is otherwise unavailable.

E. The handswitches for MS-291/292 are located on CB-10/11 and have three positions: OPEN, NORMAL, or CLOSE. On loss of air or control power, the pneumatic operator will fail shut. The power for solenoid operation comes from the 125V DC buses. The solenoid is energized to admit control air to the pneumatic operating cylinder to open the valve.

2.3.3 Location A. All ten main steam safety valves are located in Room 81 of the Auxiliary Building on the 1039-foot elevation. Each steam header has five safety valves.

2.3.4 Power Supplies None 2.3.5 Instrumentation and Control None A. Local

1. Main steam safety valves MS-291 and MS-292 are provided with non-CQE air accumulators to allow valve operation after a reactor trip for cooldown purposes. The time limit the pneumatic operator will remain operable is not specified. Manual Lift Levers are provided for manual operation to continue a controlled cooldown.

B. Remote

1. The steam safety valves MS-291/292 each have a three-position (CLOSE, NORM and OPEN) control switch and valve position status lights (green-closed and red-open) on the vertical section of CB-10/11 in the Control Room.

Source question for question 62 QUESTIONS REPORT for ILO EXAM BANK 07-11-17 032 Due to a Loss of Instrument Air, both MSIVs (Main Steam Isolation Valves) failed closed. Assuming no operator action, which of the following Control Room indications would indicate that MS-291 and MS-292 (Pneumatically Operated Safety Valves) are open?

A. The red position indication lights on CB-10 B. Safety tail pipe temperature indicating 300°F C. Red LEDs on the Safety Valve Acoustic Monitors D. Steam Generator pressure indicating approximately 1000 psia Monday, August 06, 2012 11:27:54 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 063 20 minutes after an uncomplicated reactor and turbine trip from full power, RCS T-cold stabilizes at 532°F.

Which of the following are the expected positions of the steam dump valves (TCV-909A/B/C/D) and the turbine bypass valve (PCV-910)?

A. The steam dump valves will be closed and the turbine bypass valve will modulate to maintain RCS T-cold at approximately 532°F B. The steam dump valves and the turbine bypass valves will all modulate to maintain the RCS T-cold at approximately 532°F C. The turbine bypass valve will be closed and one of the steam dump valves will modulate to maintain RCS T-cold at approximately 532°F D. The turbine bypass valve will be closed and all of the steam dump valves will modulate to maintain RCS T-cold at approximately 532°F Monday, August 06, 2012 12:00:17 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 63 Rev 0 KA #: 041000 K1.05 Tier 2 Group 2: Steam Dump System and Turbine Bypass Control Knowledge of the Physical connections and/or cause-effect relationships between the SDS and the following systems: RCS Importance 3.5 / 3.6 CFR Number: 55.41(b)(4)

Secondary coolant and auxiliary systems that affect the facility.

Fort Calhoun Objective:

EXPLAIN the operation of the Reactor Regulating System.

Question Pedigree Modified bank question.

K/A Fit:

Question addresses how steam dump and bypass valves operate to control RCS temperature.

Choice A:

Correct answer: 20 minutes after a reactor trip, the turbine bypass valve, PCV-910, will be the only valve open.

Choice B:

Distractor: Plausible if the Applicant does not realize how the steam dump and bypass valve positions are staged by DCS. Incorrect because only PCV-910 will be open.

Choice C:

Distractor: Plausible if the Applicant does not realize how the steam dump and bypass valve positions are staged by DCS. Incorrect because only PCV-910 will be open.

Choice D:

Distractor: Plausible if the Applicant does not realize how the steam dump and bypass valve positions are staged by DCS. Incorrect because only PCV-910 will be open.

KA#: 041000 K1.05 Bank Ref #: 07-12-31 015 LP# / Objective: 0712-31 02.00 Exam Level: RO-4 Cognitive Level: HIGH Source: MODIFIED

Reference:

SYSTEM LESSON PLAN 3 Handout: NONE Thursday, August 09, 2012 15:12:37 2

EO *1.7 Major Component Description Steam Dump and Turbine Bypass Valves (TCV-909-1/2/3/4 and PCV-910)

The valves are operated in automatic or manual mode as selected at DCS workstations in the Control Room.

The control system utilizes RCS Tave and main steam pressure to control the steam dump and bypass valves via staging to maintain steam header pressure/RCS temperature at the required setpoint.

Normal valve staging is as follows:

(1) PCV-910 opens first (2) When needed TCV-909-1 opens (3) Followed by TCV-909-2 (4) Then TCV-909-3 (5) Finally TCV-909-4.

88

When less steaming is required, the valves will be staged closed in reverse order.

88

Source question for question 63.

QUESTIONS REPORT for ILO EXAM BANK 07-12-31 015 Following a reactor and turbine trip from full power, main steam pressure stabilizes at 900 psia. With regard to the Steam Dump and Turbine Bypass valves, which of the following statements is correct?

A. The dumps will be closed and the Bypass valve will modulate to hold 900 psia.

B. The dumps will be partially open and the Bypass valve will modulate to hold 900 psia.

C. The dumps and the Bypass valve will modulate to hold 900 psia.

D. The dumps will be closed and the Bypass valve will remain wide open until pressure is < 750 psia.

Monday, August 06, 2012 11:55:48 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 064 An Operator was involved in the performance of OI-WDL-5, Attachment 2, "Spent Resin Transfer and Liner Dewatering in RadWaste Bldg."

During this operation he spent 45 minutes in an area with a general area dose rate of 12 mrem/hr before he inadvertently spent 1 minute above the level of the process shield where the dose rate was 75 R/hr.

What is the Operator's total estimated dose for this operation?

A. 9 mrem B. 10 mrem C. 1250 mrem D. 1259 mrem Wednesday, August 08, 2012 16:30:59 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 64 Rev 0 KA #: 068000 K5.03 Tier 2 Group 2: Liquid Radwaste System Knowledge of the operational implication of the following concepts as they apply to the Liquid Radwaste System: Units of radiation, dose, and dose rate Importance 2.6 / 2.6 CFR Number: 55.41(b)(12)

Radiological safety principles and procedures.

Fort Calhoun Objective:

SOLVE stay-time and dose problems.

Question Pedigree New question.

K/A Fit:

Question addresses dose due to liquid radwaste operation and requires knowledge of dose and dose rate units.

Choice A:

Distractor: Plausible because this is the dose assuming only the first 45 minutes.

Incorrect because it is not the correct dose.

Choice B:

Distractor: Plausible if the Applicant uses 75 mr/hr instead of R/hr. Incorrect becasue it is not the correct dose.

Choice C:

Distractor: Plausible if the Applicant only considers dose from the 75 R/hr exposure.

Incorrect because it is not the total dose.

Choice D:

Correct answer: The total dose is (12 mr/hr X 0.75 hrs) +(75000 mr/hr x 01667 hrs)

KA#: 068000 K5.03 Bank Ref #: N/A LP# / Objective: 1924-03A 06.02 Exam Level: RO-12 Cognitive Level: HIGH Source: NEW

Reference:

OI-WDL-5, ATT 2 Handout: NONE Wednesday, August 08, 2012 16:30:59 2

FORT CALHOUN STATION OI-WDL-5 OPERATING INSTRUCTION C PAGE 13 OF 32 Continuous Use Attachment 2 - Spent Resin Transfer and Liner Dewatering in RadWaste Bldg PROCEDURE () INITIALS

1. IF desired, THEN perform Spent Resin Transfer and Liner Dewatering as follows:

NOTE The following steps are categorized for easy reference. Complete the steps in any order and frequency to keep resin flowing. Maintain correct levels in the SRST and the disposal container. Maintain frequent dose rate surveys in accessible areas during resin transfer.

CAUTIONS

1. Minimize time spent above the level of the process shield. During resin transfer radiation levels could exceed 100 R/HR.

2. Closely monitor all tank and container levels. Do not overfill.

a. Close WD-533, Spent Resin Storage Tank WD-33 Gas Analyzer Valve (Room 28).

CAUTION Do not let Vent Header pressure exceed 2.25 psig.

b. Open WD-539, Spent Resin Storage Tank WD-33 Nitrogen Sparging Valve to break up hard packed resin in the SRST while ensuring flow through FI-644.

c. Monitor Vent Header pressure and Operate Waste Gas Compressors as necessary to maintain pressure less than 2 psig.

d. Continue to nitrogen sparge the SRST for a minimum of 10 minutes.

e. When N2 sparging is complete, close WD-539.

f. Throttle Open HCV-693, Spent Resin Tk Recycle Valve (AI-101B, Drumming Panel) as necessary to provide sufficient backpressure, but not to exceed 15 psig on PT-635 during the following flush and resin transfer steps.

R19

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 065 PCV-1753, Service Air System Automatic Isolation Valve, automatically closes when

_____ air pressure is ________ psig or less.

A. service, 70 B. service, 80 C. instrument, 70 D. instrument, 80 Monday, August 06, 2012 12:28:22 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 65 Rev 0 KA #: 079000 A4.01 Tier 2 Group 2: Station Air System Ability to manually operate and/or monitor in the control room: Cross-tie valves with IAS Importance 2.7 / 2.7 CFR Number: 55.41(b)(4)

Secondary coolant and auxiliary systems that affect the facility.

Fort Calhoun Objective:

Given the caution statements and/or notes listed in this AOP, explain the reason for each.

Question Pedigree Bank Question.

K/A Fit:

Question addresses monitoring automatic operation of Service air isolation valve.

Choice A:

Distractor: Plausible if Applicant believes the valve closes on service air pressure and does not know the setpoint. Incorrect because the valve closes when instrument air pressure drops to 80 psig.

Choice B:

Distractor: Plausible if Applicant believes the valve closes on service air pressure and knows the setpoint. Incorrect because the valve closes when instrument air pressure drops to 80 psig.

Choice C:

Distractor: Plausible if Applicant knows the valve closes on instrument air pressure but does not know the setpoint. Incorrect because the valve closes when instrument air pressure drops to 80 psig.

Choice D:

Correct answer: The valve closes when instrument air pressure drops to 80 psig.

KA#: 079000 A4.01 Bank Ref #: 07-11-07 016 LP# / Objective: 0717-10 01.05 Exam Level: RO-4 Cognitive Level: LOW Source: BANK

Reference:

AOP-17 Handout: NONE Monday, August 06, 2012 12:28:22 2

AOP-17 Page 6 of 46 INSTRUCTIONS CONTINGENCY ACTIONS

3. IF the "AIR DRYER TROUBLE" alarm (CB-10,11; A11) annunciates, THEN direct an operator to proceed to AI-79 to determine the cause of the alarm (Room 19).

4. Direct all available operators to search for the source of the air leakage.

5. IF Instrument Air pressure is less than 5.1 IF PCV-1753 is NOT closed, 80 psig, THEN close CA-121, "SERVICE AIR THEN verify PCV-1753, "SERVICE SUPPLY SYSTEM MANUAL AIR SYSTEM AUTOMATIC ISOLATION VALVE" (Room 19).

ISOLATION VALVE", is closed (Room 19).

6. IF Instrument Air pressure returns to a normal 98-108 psig after isolating Service Air, THEN GO TO Section 5.0, Exit Conditions.

7. IF Instrument Air pressure is less than 7.1 IF Instrument Air pressure is less than 78 psig, 78 psig THEN verify that PCV-1752, "AIR AND PCV-1752 is closed, DRYERS CA-31 & CA-12 BYPASS THEN open CA-197, "BYPASS VALVE", is open (Room 19). CONTROL VALVE PCV-1752 BYPASS VALVE" (Room 19).

R14

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 066 Which one of the following meets the requirements for personnel within the "Control Room Boundary" whenever fuel is in the reactor?

A. One RO B. One CRS C. One RO and one Shift Manager.

D. Two RO's and one unlicensed STA.

Question # 66 Rev 0 KA #: 000000 2.1.01 Tier 3 Group 4:

Generic Knowledges and Abilities Knowledge of conduct of operations requirements.

Importance 3.8 / 4.2 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective DISCUSS Personnel Presence and Conduct in the Control Room.

Question Pedigree Modified question. Source question was used on the 2001-1 NRC exam.

K/A Fit:

Question addresses required control room staffing.

Choice A:

Distractor: Plausible if Applicant confuses with requirements for the "At the controls area." Incorrect because at least one RO and one SRO (CRS or Shift Manager) is required.

Choice B:

Distractor: Plausible if Applicant confuses with requirements for the "At the controls area." Incorrect because at least one RO and one SRO (CRS or Shift Manager) is required.

Choice C:

Correct answer: One RO and one SRO are required.

Choice D:

Distractor: Plausible if Applicant does not understand that a SRO is also required.

Monday, August 06, 2012 12:55:02 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 KA#: 000000 2.1.01 Bank Ref #: ADM-OPS 018 LP# / Objective: 0767-05 04.00B Exam Level: RO-10 Cognitive Level: LOW Source: MODIFIED

Reference:

SO-O-1 Handout: NONE Monday, August 06, 2012 12:55:02 2

FORT CALHOUN STATION SO-O-1 STANDING ORDER INFORMATION USE PAGE 35 OF 74 5.9.2.D ERF Computer - See OI-ERFCS-1 5.9.2.E Operations Memorandums - See SO-O-13.

5.9.2.F Technical Data Book 5.9.2.G Temporary Labels, Curves, Notes, and Instructions - See SO-O-41.

5.10 Operations Shift Manning 5.10.1 Responsibility NOTE: The RERP Table B-1 requires one STA and Control Room Communicator in all modes of operation for ERO duties.

5.10.1.A It is the responsibility of the Shift Manager to ensure that the required number of Licensed and Equipment Operators are available.

5.10.1.B Additional Operations Department personnel may be required on shift because of unusual plant conditions or operational needs. The Shift Manager shall obtain the additional personnel as necessary. Activities requiring additional personnel will not be undertaken until the required personnel are available.

5.10.2 Staffing Requirements (Minimum) 5.10.2.A Shift manning will be as specified in Technical Specification 5.2.2.a, b, c, and e. When overtime is required for shift coverage, then the SM will use the operator overtime list as a guideline. When overtime is not required for on-shift needs, additional personnel shall be scheduled, as appropriate, on the weekly schedule.

5.10.2.B The Operations Call List (located in the Shift Manager's Office), the Duty Assignment Call List or the Operator overtime list will be used by the Shift Manager to facilitate the recall of personnel as required.

5.10.2.C In the event it is anticipated that these requirements cannot be satisfied, the Manager-Operations, Manager-Shift Operations, or the Manager-Fort Calhoun Station or designated alternate shall be notified immediately.

5.10.2.D Technical Specification 5.2 designates Licensed Operator placement as follows:

5.10.2.D.1) Two Licensed individuals (Shift Manager, Control Room Supervisor, or Reactor Operator) must be within the Control Room boundary, as defined in Attachment 7.1 at all times whenever fuel is in the reactor except as noted in Step 5.10.2.E.

R94

FORT CALHOUN STATION SO-O-1 STANDING ORDER INFORMATION USE PAGE 36 OF 74 5.10.2.D.2) The placement of the two Licensed Operators required above must also meet the following requirements; 5.10.2.D.2)a) Either the SM or the CRS must be within the Control Room boundary, as defined in Attachment 7.1 at all times whenever fuel is in the reactor except as noted in Step 5.10.2.E.

5.10.2.D.2)b) One RO or SRO should be within the at the controls area as defined in Attachment 7.1 at all times whenever fuel is in the reactor except as noted in Step 5.10.2.E. Further clarification is provided in the OPD (Procedure Maintenance and Ownership).

5.10.2.D.2)c) Occasionally, two SROs are present in the Control Room. It should be clearly understood that there is only one official CRS on watch at any one time and that the functional duties of the two SROs are not to be swapped back and forth on the same shift.

The official CRS is the one who accepted the turnover from his/her counterpart at the beginning of the shift and who signed into the shift turnover log as the oncoming CRS. This CRS shall remain as the official CRS for the duration of the shift unless properly relieved by the other CRS on-shift.

5.10.2.E The (2) Licensed members of the plant staff shall be in the Control Room at all times except during off-normal operation conditions having potential safety related impact which could jeopardize the health and safety of the public and which requires prompt operator actions outside the Control Room complex. Safety related impact is subject to Operator discretion at the time of the off-normal operating condition.

5.10.2.F Shift crew assignments during periods of core alterations shall include a Licensed Senior Reactor Operator to directly supervise the core alterations.

This Licensed Senior Reactor Operator shall not have other concurrent duties.

5.10.2.G The Manager-Shift Operations will determine the appropriate amount of management oversight during critical evolutions, and during Refueling Outages. This oversight will be documented in an Operations RFO Organizational Plan. This plan will normally designate a shutdown SRO, in case one is needed, and will provide resources for oversight in addition to the SM and CRS both in the Control Room and in the field as appropriate.

R94

Source question for question 66 QUESTIONS REPORT for ILO EXAM BANK ADM-OPS 018 What are the minimum requirements for personnel within the "at the controls" area whenever fuel is in the reactor?

A. One licensed individual (RO or SRO).

B. One SRO.

C. One RO and one SRO.

D. Two licensed individuals (RO or SRO).

Monday, August 06, 2012 12:45:17 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 067 The Shift Manager has implemented the two-person rule for vital areas per AOP-37, "Security Events."

Assuming you are one of two extra operators in the control room, which one of the following actions would require you to take an additional operator with you?

A. Acknowledging an annunciator alarm on AI-100, Alarm Panel A50 B. Acknowledging an annunciator alarm on AI-134, Alarm Panel A30-2 C. Verifying the position of Auxiliary Feedwater Pump FW-54 engine, FW-56, throttle lever.

D. Verifying the position of Quadvoter solenoid valves, PCV-3303A./B and PCV-3303C/D.

Monday, August 06, 2012 13:05:40 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 67 Rev 0 KA #: 000000 2.1.13 Tier 3 Group 4: Generic Knowledges and Abilities Knowledge of facility requirements for controlling vital / controlled access.

Importance 2.5 / 3.2 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective DESCRIBE how the plant operators respond to Security Events.

Question Pedigree New question K/A Fit:

Question addresses requirements for access to vital areas.

Choice A:

Correct answer: AI-100 is located in a vital area.

Choice B:

Distractor: Plausible if Applicant believes AI-134 is located in a vital area. Incorrect because AI-134 is located within a non-vital area of the turbine building.

Choice C:

Distractor: Plausible because FW-54 is important to safety per the PRA and is located in its own room. Incorrect because the FW-54 room is not a vital area.

Choice D:

Distractor: Plausible if Applicant believes the Quadvoter is located in a vital area.

Incorrect because the Quadvotor is located within a non-vital area of the turbine building.

Note: AOP-37 is contained on the "SUNSI" CD.

KA#: 000000 2.1.13 Bank Ref #: N/A LP# / Objective: 0717-37 01.04 Exam Level: RO-10 Cognitive Level: HIGH Source: NEW

Reference:

AOP-37 Handout: NONE Monday, August 06, 2012 13:05:40 2

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 068 The reactor has been operating at 100% power for three months. In accordance with OPD-4-19, "Reactivity Management," the "Shift Reactivity Briefing" is required to include discussion of:

A. Core xenon reactivity B. RCS boron concentration C. Moderator temperature coefficient D. Available shutdown margin Question # 68 Rev 0 KA #: 000000 2.1.37 Tier 3 Group 4: Generic Knowledges and Abilities Knowledge of procedures, guidelines, or limitations associated with reactivity management.

Importance 4.3 / 4.6 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

DESCRIBE the Performance Standards listed in the OPD Manual.

Question Pedigree New question.

K/A Fit:

The question addresses the required contents of a reactivity briefing.

Choice A:

Distractor:Plausible because this is a significant reactivity parameter. Incorrect because it is not required by OPD-4-19.

Choice B:

Correct answer. OPD-4-19 requires that RCS boron concentration be discussed during a reactivity briefing.

Choice C:

Distractor:Plausible because this is a significant reactivity parameter. Incorrect because it is not required by OPD-4-19.

Choice D:

Distractor:Plausible because this is a significant reactivity parameter. Incorrect because it is not required by OPD-4-19.

Wednesday, August 08, 2012 16:31:41 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 KA#: 000000 2.1.37 Bank Ref #: N/A LP# / Objective: 0767-05 02.00 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

OPD-4-19 Handout: NONE Wednesday, August 08, 2012 16:31:41 2

OPD-4-19 Information Use Page 15 of 15 Reactivity Management Rev. 15 Table 3 - Shift Reactivity Briefing Checklist Review Current Reactivity Status RCS boron concentration and neutral blend/batch Last shift's and expected reactivity additions Reactivity thumb-rules Discuss Potential External Impacts on Reactivity Weather/River conditions affecting Condenser operations Raw Water and CCW operations Electrical Grid conditions Are there any contingency actions required?

Discuss Planned Activities and Reactivity Impacts Reactor power changes for activities (e.g., Control Valve testing)

Feedwater/Condensate/Heater Drain system operations Main Steam/Steam Dump and Bypass/SG Blowdown operations Turbine EHC operations CVCS operations Other items (e.g., testing on critical components/systems)

Are there any contingency actions required?

Discuss Operator Challenges with Reactivity Impacts Review Operator Challenge List from Plan of the Day for reactivity impacts Reactivity control or feedwater controls system deficiencies Annunciator/Alarm deficiencies Are there any contingency actions required?

Input/Questions? Follow-up Actions?

Get input from all personnel in attendance Do any of the above items require RE, Reactivity SRO, or other support?

The purpose of the Shift Reactivity Briefing Checklist is to provide a consistent method for crew reactivity briefings in accordance with SO-G-117, Reactivity Management. These briefings should occur at the beginning of each shift and at the discretion of the SM/CRS should plant conditions warrant.

The briefing is led by the CRS, with input provided by the LOs, STA, and SM. The CRS, or an LO as designated by the CRS, should review the checklist in a section by section manner. Each section provides a list of items intended to stimulate discussion during the briefing.

The contingency actions discussion should include items such as:

Critical parameters to monitor when applicable, Post-trip actions that would be different due to plant/system/component status, and Reactivity-related AOP/EOP actions that would be different due to plant/system/component status.

The beginning-of-shift reactivity is not a substitute for a reactivity briefing for a specific activity (e.g., MTC testing).

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 069 According to Standing Order G-30, "Procedure Changes and Generation," who is responsible for ensuring that the 50.59 and/or 72.48 screening is completed for a proposed procedure change?

A. The Shift Manager or CRS.

B. The Qualified Reviewer.

C. The PRC Technical Secretary.

D. The Document Control Supervisor.

Question # 69 Rev 0 KA #: 000000 2.2.06 Tier 3 Group 4: Generic Knowledges and Abilities Knowledge of the process for making changes in procedures.

Importance 3.0 / 3.6 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective STATE some of the activities, covered by Standing Orders, which require written procedures per Regulatory Guide 1.33.

Question Pedigree Bank question used on the 2007 NRC exam.

K/A Fit:

Question addresses responsibilities for procedures changes.

Choice A:

Distractor: Plausible if Applicant does not understand this is the responsibility of the Qualified Reviewer. Incorrect because the Qualified Reviewer is responsible.

Choice B:

Correct answer: The Qualified Reviewer is responsible for this action.

Choice C:

Distractor: Plausible if Applicant does not understand this is the responsibility of the Qualified Reviewer. Incorrect because the Qualified Reviewer is responsible.

Choice D:

Distractor: Plausible if Applicant does not understand this is the responsibility of the Qualified Reviewer. Incorrect because the Qualified Reviewer is responsible.

Monday, August 06, 2012 13:24:51 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 KA#: 000000 2.2.06 Bank Ref #: ADM-CONTROL 031 LP# / Objective: 0762-01 02.00 Exam Level: RO-10 Cognitive Level: LOW Source: BANK

Reference:

SO G-30 Handout: NONE Monday, August 06, 2012 13:24:51 2

SO-G-30 Information Use Page 12 of 78 Procedure Changes and Generation Revision 124 3.3.9 Ensuring proposed procedure changes which are determined to involve a change in the Technical Specifications are submitted to the NRC for approval prior to implementation.

3.4 The Qualified Reviewer is responsible for:

3.4.1 Ensuring proper review of proposed procedure changes and documenting recommendations for approval/disapproval in accordance with guidance of Step 4.17.

3.4.2 Ensuring the 50.59 and/or 72.48 Screen and, if applicable, 50.59 and/or 72.48 Evaluation is completed as required by NOD-QP-3. The review may be performed by the QR or may be performed prior to the QR review process. (AR 19408) 3.4.3 Ensuring proposed procedure changes are reviewed in a timely manner.

3.4.4 Ensuring completeness and thoroughness of any cross-disciplinary/ cross-functional or other reviews.

3.5 Document Control is responsible for:

3.5.1 Distributing Operating Manual procedures in a timely manner.

3.5.2 Maintaining and issuing the current index of Operating Manual procedures.

3.5.3 Maintaining records required by this procedure.

3.5.4 Distributing a copy of Setpoint Change Review Form (PED-SEI-9.1) and 50.59 Screen/Evaluation (if applicable) to the Manager-Design Engineering Nuclear following approval of procedure changes involving any changes to instrument setpoint, tolerance, or acceptance criteria as delineated in PED-SEI-9.

3.5.5 Notifying on-shift Shift Manager if an OI Checklist has been revised and may require performance of the OI Checklist immediately after issuance (as specified by the QR). (CR 199701648) 3.6 The Plant Review Committee (PRC) is responsible for:

3.6.1 Reviewing procedure or procedure changes which require a 50.59 or 72.48 Evaluation.

3.6.2 Reviewing those procedure changes determined as a Very High Risk Activity.

3.6.3 Reviewing candidates to serve as Qualified Reviewers and recommending approval.

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 070 Safety Limit 1.1.1(a) states:

"The reactor power level shall not exceed the allowable limit for the pressurizer pressure and the cold leg temperatures as shown in Figure 1-1 for 4-pump operation.

The safety limit is exceeded if the point defined by the combination of reactor coolant cold leg temperature and power level is at any time above the appropriate pressurizer pressure line."

What is the basis for this Safety Limit?

A. To ensure that DNBR during normal operation is not greater than the DNBR limit.

B. To ensure that DNBR during normal operation is not lower than the DNBR limit.

C. To ensure that DNBR during a design basis LOCA is not greater than the DNBR limit.

D. To ensure that DNBR during a design basis LOCA is not lower than the DNBR limit.

Monday, August 06, 2012 13:34:05 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 70 Rev 0 KA #: 000000 2.2.25 Tier 3 Group 4: Generic Knowledges and Abilities Knowledge of bases in Technical Specifications for limiting conditions for operations and safety limits.

Importance 3.2 / 4.2 CFR Number: 55.41(b)(5)

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.

Fort Calhoun Objective:

STATE the two (2) safety limits and the basis for each.

Question Pedigree New question.

K/A Fit:

Question addresses knowledge of the basis for a safety limit.

Choice A:

Distractor: Plausible if Applicant does not realize that DNBR must stay above the limit.

Choice B:

Correct answer: DNBR must stay above the limit for normal operation.

Choice C:

Distractor: Plausible if Applicant does not realize that DNBR must stay above the limit and believes the limit is for a LOCA instead of normal operation.

Choice D:

Distractor: Plausible if Applicant believes the DNBR limit applies to a LOCA.

KA#: 000000 2.2.25 Bank Ref #: N/A LP# / Objective: 0762-08 03.00 Exam Level: RO-5 Cognitive Level: LOW Source: NEW

Reference:

TS 1.1.1(A) BASIS Handout: NONE Monday, August 06, 2012 13:34:06 2

TECHNICAL SPECIFICATIONS 1.0 SAFETY LIMITS 1.1 Safety Limits (SLs) 1.1.1 Reactor Core SLs Applicability This specification applies to the limiting combinations of reactor power and reactor coolant system flow, temperature and pressure during operation.

Objective To maintain the integrity of the fuel cladding and prevent the release of significant amounts of fission products to the reactor coolant.

Specifications (a) The reactor power level shall not exceed the allowable limit for the pressurizer pressure and the cold leg temperatures as shown in Figure 1-1 for 4-pump operation. The safety limit is exceeded if the point defined by the combination of reactor coolant cold leg temperature and power level is at any time above the appropriate pressurizer pressure line.

(b) Peak fuel centerline temperature shall be maintained at < 5081°F, decreasing by 58°F per 10,000 MWD/MTU and adjusted for burnable poison per XN-NF-79-56(P)(A), Revision 1, Supplement 1.

1.1.2 Reactor Coolant System Pressure SL Applicability Applies to the limit on reactor coolant system pressure.

Objective To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity to the containment.

Specification The reactor coolant system pressure shall not exceed 2750 psia when fuel assemblies are located within the reactor vessel.

1.0 - Page 1 Amendment No. 47,70,196, 252

TECHNICAL SPECIFICATIONS 1.0 SAFETY LIMITS 1.1 Safety Limits (SLs) (continued)

Basis To maintain the integrity of the fuel cladding and prevent the release of significant amounts of fission products to the reactor coolant, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant saturation temperature. The upper boundary of the nucleate boiling regime is termed "departure from nucleate boiling" (DNB).

At DNB there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperature and the possibility of clad failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of reactor thermal power and reactor coolant flow, temperature and pressure can be related to DNB through a correlation. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux at that location, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients corresponds to a 95% probability at a 95%

confidence level that DNB will not occur, which is considered an appropriate margin to DNB for all operating conditions.(1)

The curves of Figure 1-1 represent the loci of points for reactor thermal power (either neutron flux instruments or T instruments), reactor coolant system pressure, and cold leg temperature for which the minimum DNBR is not less than the minimum DNBR limit.

The area of safe operation is below these lines.

SL 1.1.1(b) ensures that fuel centerline temperature remains below the fuel melt temperature 5081°F during normal operating conditions or design anticipated operational occurrences (AOOs) with adjustments for burnup and burnable poison. An adjustment of 58°F per 10,000 MWD/MTU has been established in XN-NF-82-06(P)(A), Revision 1, Supplements 2, 4 and 5 (Ref. 8) and adjustments for burnable poisons are established based on XN-NF-79-56(P)(A), Revision 1, Supplement 1 (Ref. 9).

The reactor core safety limits are based on radial peaks limited by the CEA insertion limits in Section 2.10 and axial shapes within the axial power distribution trip limits in the COLR. The Thermal Margin/Low Pressure trip requirements shall be within the limits provided in the COLR. The Thermal Margin/Low Pressure trip is based on an unrodded integrated total radial peak (FRT) that is provided in the COLR.

1.0 - Page 2 Amendment No. 8,32,43,47, 70,77,92,117,126,141,170,196, 252 TSBC-08-002-0

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 071 How will placing the key switch on a Control Room process radiation monitor ratemeter in the "KEYPAD" position affect operation of the radiation monitor?

A. The ratemeter will display 0 cpm and equipment actuations will be blocked.

B. The ratemeter will display actual countrate but equipment actuations will be blocked.

C. The ratemeter and equipment actuations will still work but output to the ERF computer will be blocked.

D. Operation of the radiation monitor will not be affected.

Wednesday, August 08, 2012 16:32:18 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 71 Rev 0 KA #: 000000 2.3.05 Tier 3 Group 4: Generic Knowledges and Abilities Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Importance 2.9 / 2.9 CFR Number: 55.41(b)(11)

Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Fort Calhoun Objective:

EXPLAIN the operations, actuations and applications of the individual radiation monitors.

Question Pedigree Bank question used on the 2009 NRC exam. The wording of the choices has been changed but not counted as a modified question.

K/A Fit:

Question addresses the operation of installed radiation monitors control pads.

Choice A:

Distractor: Plausible if Applicant believes the display will go to zero. Incorrect because it will display the actual countrate.

Choice B:

Correct answer: The ratemeter will display actual countrate but equipment actuations will be blocked.

Choice C:

Distractor: Plausible if Applicant believes only the output to the ERF computer will be blocked. Incorrect because equipment actuations will be blocked.

Choice D:

Distractor: Plausible if Applicant believes placing ratemeter in keypad will have no effect. Incorrect because equipment actuations will be blocked.

KA#: 000000 2.3.05 Bank Ref #: 07-12-03 032 LP# / Objective: 0712-03 04.00 Exam Level: RO-11 Cognitive Level: LOW Source: BANK

Reference:

STM-33 Handout: NONE Wednesday, August 08, 2012 16:32:18 2

SYSTEM TRAINING MANUAL VOL. 33 RADIATION MONITORING SYSTEM 2.49 When the Control Room ratemeter keyswitch is in the OFF posi-tion, the local ratemeter will be unaffected and continue to provide a local read-out of activity and alarms only; however, the process signal will not be passed on to the ERF computer process radiation monitor recorder, high radiation and trouble alarms on AI-33C, and external relays for system actuation.

2.50 With the Control Room ratemeter keyswitch in the KEYPAD posi-tion, the alarms and system actuation relays will not operate. How-ever, activity will still be displayed and sent to the process radiation monitor recorder and the ERF computer.

RAM606 HIGH ALERT NORM FAIL 1.38 E3 CPM 101 102 103 104 105 106 107 CHECK KEYPAD RESET TEST SOURCE OFF ON MOD SET E

NOTE: ALSO USED AS LOCAL RATEMETERS ON SKID UNITS RM-050/51, RM-052, RM-062 AND RM-063 FIGURE 2-14: PROCESS RADIATION MONITOR CONTROL ROOM RATE-METER (RAM606) 2.51 The keypad contains the following switches:

2.52 RESET- switch is used to reset acknowledge alarm conditions (lights and actuation relays) when the count rate has returned to below the alert or alarm setpoints. The monitor must be reset before external alarms on AI-33C and external system actuation relays can be reset.

2.53 CHECK SOURCE- switch begins a pre-timed program for check source activation (30 to 120 seconds depending on type of moni-tor). The CHECK SOURCE program can be activated from the Control Room ratemeter with its keyswitch in the ON position.

However, when activating CHECK SOURCE from the local rate-meter, its keyswitch must be in the KEYPAD position. While in Level: Basic > Intermediate > Advanced Revision - 18

• 19

Source question for question 71 QUESTIONS REPORT for ILO EXAM BANK 07-12-03 032 How will placing the keyswitch on a Control Room process radiation monitor ratemeter in the "KEYPAD" position affect operation of the radiation monitor?

A. Annunciation and equipment actuations will be blocked.

B. Annunciation will still work but equipment actuations will be blocked.

C. Annunciation and equipment actuations will still work but output to the radiation monitor recorder will be blocked.

D. Annunciation and equipment actuations will still work but output to the ERF computer will be blocked.

Monday, August 06, 2012 13:53:09 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 072 Why does EOP-04, "Steam Generator Tube Rupture" direct the operators to ensure that FW-268, "Condensate Dump Valve LCV-1193 Outlet Isolation Valve" and FW-266, "Condensate Dump Valve LCV-1193 Bypass Valve" are closed?

A. To ensure that the water level in the condenser hotwell remains high enough to meet the NPSH requirements of the condensate pumps.

B. To ensure that the water level in the condenser hotwell remains high enough to adequately dilute radioactive water from the ruptured steam generator.

C. To isolate flow to the condensate storage tank to prevent it from overflowing.

D. To isolate flow to the condensate storage tank to prevent the spread of contamination.

Wednesday, August 08, 2012 16:32:52 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 72 Rev 0 KA #: 000000 2.3.14 Tier 3 Group 4: Generic Knowledges and Abilities Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Importance 3.4 / 3.8 CFR Number: 55.41(b)(12)

Radiological safety principles and procedures.

Fort Calhoun Objective:

DEMONSTRATE the knowledge required to use EOP-04, Steam Generator Tube Rupture (SGTR), to mitigate the consequences of a SGTR.

Question Pedigree New question.

K/A Fit:

The question addresses radiation contamination hazards during a S/G tube rupture.

Choice A:

Distractor: Plausible if the Applicant believes these valves are closed to retain water in the hotwell. Incorrect because they are closed to prevent contamination of the CST.

Choice B:

Distractor: Plausible if the Applicant believes these valves are closed to retain water in the hotwell. Incorrect because they are closed to prevent contamination of the CST.

Choice C:

Distractor: Plausible if Applicant believes the valves are closed to prevent overfilling the CST. Incorrect because they are closed to prevent contamination of the CST.

Choice D:

Correct answer: They are closed to prevent contamination of the CST.

KA#: 000000 2.3.14 Bank Ref #: N/A LP# / Objective: 0718-14 01.00 Exam Level: RO-12 Cognitive Level: HIGH Source: NEW

Reference:

TBD-EOP-04 Handout: NONE Wednesday, August 08, 2012 16:32:52 2

TBD-EOP-04 Page 24 of 68 INSTRUCTIONS CONTINGENCY ACTIONS

d. Monitor S/G levels.

EPG Step: 12 Deviation: None In order to minimize the release of radioactivity, the S/G with the tube leak must be identified so it can be isolated when plant conditions for S/G isolation are established. The S/G with the tube leak may be determined by some or all of the following:

• Sampling the S/Gs indicates higher activity in one of the S/Gs

• Monitoring the steam headers indicate higher activity in one of the steam headers

• Monitoring the S/G blowdown indicates higher activity in one of the blowdown lines

• For a large SGTR, a rise in the level of the affected S/G may be noticed or a constant level may be observed while a mismatch between feed rate and steaming rate exists.

13. Minimize the spread of contamination by performing the following steps:

a. Ensure HCV-2509, SAMPLE DRAIN TO DRAIN HEADER, is open (AI-107, Room 60).

b. Ensure HCV-2508, SAMPLE DRAIN TO CONDENSER C.W.

TUNNEL, is closed (AI-107, Room 60).

c. Ensure FW-268, CONDENSATE DUMP VALVE LCV-1193 OUTLET ISOLATION VALVE, is closed (Turbine Building Mezzanine).

d. Ensure FW-266, CONDENSATE DUMP VALVE LCV-1193 BYPASS VALVE, is closed (Turbine Building Mezzanine).

Continuously Applicable or Non-Sequential Step R26

TBD-EOP-04 Page 25 of 68 INSTRUCTIONS CONTINGENCY ACTIONS EPG Step: None Deviations: 1) The EOP provides direction for realigning blowdown sample discharge from the Raw Water system to the Waste Disposal system.

2) The EOP isolates Condensate dump to the storage tank.

Blowdown sample flow is normally discharged to the Raw Water system which discharges to the river. During a SGTR event the sample flow is redirected to the Spent Regenerant tank to maintain any radioactive sample within the plant's waste disposal system. This justifies deviation 1.

Condensate dump to the storage tank is isolated to prevent the spread of contamination. This justifies deviation 2.

14. WHEN RCS TH is less than or equal to 510°F, THEN isolate the most affected S/G by performing the following steps:

a. IF RC-2A is most affected, a.1 IF RC-2B is most affected, THEN isolate RC-2A by THEN isolate RC-2B by performing performing the following steps: the following steps:

Continuously Applicable or Non-Sequential Step R26

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 073 Given the following plant conditions:

A LOCA occurred 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> ago RCS pressure is 150 psia and lowering SIRWT Level is 32 inches and lowering Pressurizer Level is 50% and rising slowly RCS T-cold is 325°F What condition must change before EOP/AOP Attachment 7, SDC with RAS can be implemented?

A. 30 additional minutes must elapse B. RCS pressure must be 140 psia or less C. SIRWT level must be 16 inches or less D. RCS T-cold must be 300°F or less Monday, August 06, 2012 14:40:27 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 73 Rev 0 KA #: 000000 2.4.02 Tier 3 Group 4: Generic Knowledges and Abilities Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

Importance 4.5 / 4.6 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

GIVEN a copy of Attachment 7, EXPLAIN the steps necessary to initiate Shutdown Cooling Operations with RAS.

Question Pedigree New question.

K/A Fit:

Question addresses setpoint entry conditions for an EOP attachment that is common to different EOPs.

Choice A:

Distractor: Plausible because this is required for EOP/AOP Attachment 9. Incorrect because it is not required for Attachment 7.

Choice B:

Distractor: Plausible because this is required for EOP/AOP Attachment 9. Incorrect because it is not required for Attachment 7.

Choice C:

Correct answer. SIRWT level must be less than the STLS setpoint, 16 inches, before RAS can occur.

Choice D:

Distractor: Plausible because RCS pressure must be 300 psia or less. Incorrect Attachment 7 can be performed with T-cold at 325°F.

KA#: 000000 2.4.02 Bank Ref #: N/A LP# / Objective: 0718-13 02.04 Exam Level: RO-10 Cognitive Level: HIGH Source: NEW

Reference:

EOP-03 Handout: NONE Thursday, August 09, 2012 15:15:40 2

EOP-03 Page 42 of 82 INSTRUCTIONS CONTINGENCY ACTIONS

34. IF SIRWT level falls to 16 inches, 34.1 IF RAS is NOT actuated by STLS, THEN verify that STLS initiates RAS by THEN perform the following to performing the following: manually establish RAS flow paths:

a. Ensure BOTH of the following a. Open both SI Pump Suction valves are open: Containment Isolation Valves, HCV-383-3/4.

• HCV-383-3, SI Pump Suction Containment Isolation Valve

• HCV-383-4, SI Pump Suction b. Close both SI Pump Suction Containment Isolation Valve SIRWT Isolation Valves, LCV-383-1/2.

b. Ensure ALL of the following valves are closed: c. Ensure both LPSI Pumps stop,

• LCV-383-1, SI Pump Suction SI-1A/B.

SIRWT Isolation Valve

• LCV-383-2, SI Pump Suction d. Close BOTH of the following SIRWT Isolation Valve

• HCV-385, SIRWT Recirc Valve SIRWT Recirc Valves:

• HCV-386, SIRWT Recirc Valve

• HCV-480, AC-4A CCW Inlet

• HCV-385 Valve

• HCV-386

• HCV-481, AC-4B CCW Inlet Valve

• HCV-484, AC-4A CCW Outlet Valve Time:

• HCV-485, AC-4B CCW Outlet Valve

c. Ensure BOTH LPSI pumps stop. c.1 IF LPSI pumps are not stopped, THEN IMPLEMENT Floating Time:

Step B, LPSI Stop and Throttle.

Continuously Applicable or Non-Sequential Step R36

EOP-03 Page 54 of 82 INSTRUCTIONS CONTINGENCY ACTIONS

44. WHEN ALL of the following SDC entry 44.1 IF SDC can NOT be established, conditions are established: THEN GO TO Step 45.

• PZR level is greater than or equal to 45% and constant or rising

• RCS subcooling is greater than or equal to 20°F

• RCS pressure is less than or equal to 300 psia.

• RCS TC less than 350°F THEN initiate SDC operation PER ONE of the following Attachments:

• Attachment 4, SDC Without RAS

• Attachment 7, SDC with RAS

• Attachment 8, Cooled SI Flow with RAS Time:

CAUTION If RAS has occurred, area radiation levels may require access to AI-100 via the Emergency Access Plug.

45. Monitor both SI Pump Room Sumps for indication of excessive leakage (ERF Analog Data Points L565, L566).

Continuously Applicable or Non-Sequential Step R36

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 074 While performing the actions of an AOP, a step in the AOP directs you to "GO TO" an EOP.

How should the procedures be used?

A. Exit the AOP and enter the EOP.

B. Perform the EOP actions in parallel with the AOP actions.

C. Enter the EOP and use the reference material in the AOP.

D. Perform the actions in the EOP and then come back to the AOP.

Wednesday, August 08, 2012 16:33:32 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 74 Rev 0 KA #: 000000 2.4.08 Tier 3 Group 4: Generic Knowledges and Abilities Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Importance 3.8 / 4.5 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

DEMONSTRATE general knowledge about the structure, terminology and usage of the Emergency Operation Procedures (EOP's).

Question Pedigree New question.

K/A Fit:

Question generically addresses how AOP's are used in conjunction with EOP's.

Choice A:

Correct answer:OPD-4-09 states that "GO TO" means exit this procedure and use the procedure identified after the "GO TO."

Choice B:

Distractor: Plausible if Applicant confuses with "Implement." Incorrect because the AOP must be exited.

Choice C:

Distractor: Plausible if Applicant confuses with "Refer to." Incorrect because the AOP must be exited.

Choice D:

Distractor: Plausible if Applicant does not realize that the AOP must be exited. Incorrect because the AOP must be exited.

KA#: 000000 2.4.08 Bank Ref #: N/A LP# / Objective: 0718-10 01.00 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

OPD 4-09 Handout: NONE Wednesday, August 08, 2012 16:33:32 2

OPD-4-09 Information Use Page 20 of 58 EOP/AOP Users Guidelines Rev. 15 4.19 Declarative Steps 4.19.1 Some steps are merely statements, e.g., "a leak in the Aux Building is indicated". These are declarative steps and only meant as an operator aid.

4.20 Announcements 4.20.1 Announcements appear in quotes following a sentence beginning with the verb "ANNOUNCE".

4.21 Evaluation Order 4.21.1 Steps are to be performed in order of appearance unless otherwise noted. If steps are permitted to be performed out of order, then a note will specify that.

Additionally, steps within EOP's are designated as continuously applicable or non-sequential. These steps can be performed in any order as conditions are met, and are not considered deviations.

4.21.2 Time dependent steps will have the time limit in the step. When a set of steps is time dependent, there will be a note to state that ahead of the first step in the set.

4.21.3 Equally acceptable steps are a set of steps any of which may accomplish the goal of the overall step. The order of appearance is most-to-least preferred.

Operators should attempt to perform step a. If unable to perform step a, attempt b, etc.

4.22 References 4.22.1 When an EOP/AOP refers to another procedure of any type, it will direct the operator to the appropriate section. When references are made to tables and figures this is not applicable.

4.23 Referencing Terms 4.23.1 "REFER TO" indicates that a source of information is available to aid in performance of the current step, at the discretion of the operator.

4.23.2 "IMPLEMENT" indicates another set of instructions is to be performed in parallel with the concurrent procedure.

4.23.3 "PER" can replace both "IMPLEMENT" and "REFER TO" where they would be inappropriate, e.g., "Verify RCP NPSH requirements are met PER Attachment 2."

4.24 Branching Terms 4.24.1 "GO TO" means exit this procedure and use the procedure identified after the "GO TO".

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 075 Given the following:

The AI-149 FIRE DETECTION ALARM OR TROUBLE alarm was received The Control Room Fire Detection Computer, PC-66, indicates that multiple detectors in the Cable Spread Room are in alarm The EONT was dispatched and reports that Halon is being discharged into the Cable Spread Room In accordance with AOP-06, FIRE EMERGENCY, what action should be taken by the operating crew?

A. Trip the reactor, open clutch power supply breakers, close the PORV block valves and evacuate the Control Room.

B. Trip the reactor, close the PORV block valves, start both Fire pumps, FP-1A and FP-1B, and evacuate the Control Room.

C. Trip the reactor, open clutch power supply breakers, close the PORV block valves and place the Control Room HVAC in Recirculation Mode.

D. Close the PORV block valves, start both Fire pumps, FP-1A and FP-1B, and place the Control Room HVAC in Recirculation Mode.

Wednesday, August 08, 2012 16:34:08 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 75 Rev 0 KA #: 000000 2.4.25 Tier 3 Group 4: Generic Knowledges and Abilities Knowledge of fire protection procedures.

Importance 3.3 / 3.7 CFR Number: 55.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Describe how the plant may respond to a fire in the following locations: Upper Electrical Penetration Room.

Question Pedigree Bank question used on the 2009 NRC exam.

K/A Fit:

The question addresses actions to be taken per FCS Fire Protection Procedures.

Choice A:

Correct answer: AOP-06 requires that the Control Room be evacuated for a Cable Spread Room fire. The reactor is tripped, clutch power supply breakers are opened and PORV block valves are closed prior to evacuating the Control Room.

Choice B:

Distractor: Plausible because it contains some of the correct actions. Incorrect because fire pumps should not be started.

Choice C:

Distractor: Plausible because it includes some correct actions. Incorrect because the Control Room is not evacuated.

Choice D:

Distractor: Plausible because it includes some correct actions. Incorrect because the Control Room is not evacuated.

KA#: 000000 2.4.25 Bank Ref #: 07-17-06 013 LP# / Objective: 0717-06 01.02C Exam Level: RO-10 Cognitive Level: LOW Source: BANK

Reference:

AOP-06 Handout: NONE Wednesday, August 08, 2012 16:34:08 2

AOP-06 Page 16 of 130 Section II - Control Room Evacuation 4.0 INSTRUCTIONS/CONTINGENCY ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS

1. Direct the Shift Manager to perform Attachment M, Shift Manager Actions in Support of AOP-06 Section II.

2. Perform the following steps prior to evacuating the Control Room:

a. Manually trip the Reactor.

b. Open BOTH Clutch Power Supply Circuit Breakers (AI-57):

"CB-AB" "CB-CD"

c. Close BOTH of the following PORV Block Valves:

HCV-150 HCV-151 (continue)

R26

AOP-06 Page 17 of 130 Section II - Control Room Evacuation INSTRUCTIONS CONTINGENCY ACTIONS

2. (continued)

d. Place BOTH PORV handswitches in "CLOSE":

PCV-102-1 PCV-102-2

e. (CRS) Obtain the AOP-6 Keys and a transceiver.

f. Direct all operators in the Control Room to obtain AOP-06 Keys and a transceiver.

g. Ensure extra transceivers are taken to AI-185, "ALTERNATE SHUTDOWN PANEL" (West Upper Electrical Penetration Room).

R26

AOP-06 Page 5 of 130 Section I - Fire Emergency 4.0 INSTRUCTIONS/CONTINGENCY ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS

1. Direct an operator to locate and report the status of the fire.

2. IF an acrid odor is detected, THEN IMPLEMENT Attachment O, Acrid Odor.

3. IF a Control Room evacuation is required, THEN GO TO AOP-06 Section II, Control Room Evacuation.

4. IF the Cable Spread Room is affected, THEN GO TO AOP-06 Section II, Control Room Evacuation.

5. IF a Plant fire exists AND fire brigade activation is warranted, THEN sound the Fire Alarm for 10-15 seconds.

R26

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 076 A plant transient has led to a high pressure reactor trip. Following the trip, a pressurizer safety valve opened and did not reclose. The RCS pressure then lowered and PPLS actuation started ECCS pumps. The operators tripped all RCPs.

It is 10 minutes after the trip and current plant conditions are:

• RCS Pressure is 900 psia and rising

• Hot leg and CET temperatures indicate 530ºF

• Pressurizer level is 85% and rising

• 2 HPSI, 2 LPSI and all charging pumps are running

• EOP-00, "Standard Post Trip Actions" have been completed What action should be directed in response to the rising pressurizer level?

A. Enter EOP-03, "Loss of Coolant Accident," and allow HPSI and charging pumps to continue injecting at full capacity.

B. Enter EOP-03, "Loss of Coolant Accident," and reduce HPSI and charging pump flow per floating step A, "HPSI Stop and Throttle Criteria."

C. Enter EOP-20, "Functional Recovery Procedure," and allow HPSI and charging pumps to continue injecting at full capacity.

D. Enter EOP-20, "Functional Recovery Procedure," and reduce HPSI and charging pump flow per floating step A, "HPSI Stop and Throttle Criteria."

Wednesday, August 08, 2012 16:40:55 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 76 Rev 0 KA #: 000008 AA2.23 Tier 1 Group 1: Pressurizer Vapor Space Accident Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident: Criteria for throttling high-pressure injection after a small LOCA Importance 3.6 / 4.3 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

EXPLAIN how the "Stop and Throttle" criteria is used to prevent reducing HPSI flow when full HPSI flow is required.

Question Pedigree Modified question. Source question was used on 2005 NRC exam.

K/A Fit:

Question addresses reducing HPSI flow after a small LOCA Choice A:

Correct answer: HPSI stop and throttle criteria not net due to inadequate subcooling.

EOP-03 entry conditions met.

Choice B:

Distractor: Plausible because pressurizer level is high and rising. Incorrect because criteria are not met.

Choice C:

Distractor: Plausible because action is correct. Incorrect because EOP-20 entry conditions are not met.

Choice D:

Distractor: Plausible because pressurizer level is high and rising. Incorrect because criteria are not met and EOP-20 entry conditions are not met.

KA#: 000008 AA2.23 Bank Ref #: 07-15-23 029 LP# / Objective: 0715-23 02.06 Exam Level: SRO-5 Cognitive Level: HIGH Source: MODIFIED

Reference:

LP 07-15-23 Handout: NONE Wednesday, August 08, 2012 16:40:56 2

EOP-00 Page 34 of 44 NOTE B

Indication of primary to secondary leakage are steam plant activity or S/G level mismatch.

Is pressurizer No pressure >1800 PSIA and stable or rising?

Yes High Subcooled Indications of No No margin and low primary to S/G pressure? secondary leakage?

Yes Yes Consider Consider Consider Uncontrolled Heat Steam Generator Loss of Coolant Extraction Event Tube Rupture Event Event (EOP-03)

(EOP-05) (EOP-04)

GO TO C on next page R29

EOP-00 Page 35 of 44 Certain events (i.e., LOCA, SGTR, UHE and Loss of All Feedwater) do not require offsite power in order to adequately, mitigate the effects of the accident. For this reason, the LOCA, SGTR, UHE or Loss of All Feedwater procedure may be C implemented even if a Loss of Offsite Power has also occurred.

Are all acceptance No Has any event No criteria except IA been satisfied? diagnosed?

Yes Yes No GO TO Functional Is the diagnosis a Recover Procedure single event (EOP-20)

GO TO the Reactor only?

Trip Recovery Procedure (EOP-01)

Yes GO TO the appropriate Optimal Recovery Procedure R29

EOP/AOP FLOATING STEPS Page 21 of 115 2.0 FLOATING STEPS A. HPSI STOP AND THROTTLE CRITERIA INSTRUCTIONS CONTINGENCY ACTIONS CAUTIONS

1. If emergency boration is required then at least one charging pump must remain running.

2. As natural circulation develops, the expected rise in TH will reduce subcooling. This may jeopardize HPSI Stop and Throttle Criteria.

3. Reducing SI flow should be approached cautiously.

4. The purpose of HPSI stop and throttle is to prevent an over pressurization of the RCS and a solid PZR, however, maintaining RCS inventory is more important than pressure control.

1. Verify ALL of the following stop and throttle criteria are satisfied:

RCS subcooling is greater than or equal to 20F PZR level is greater than or equal to 10% and not lowering At least one S/G is available for RCS heat removal RVLMS indicates level is at or above the top of the Hot Leg (43%, ERF "I" display)

R3

Source Question for question 76 QUESTIONS REPORT for ILO EXAM BANK 07-15-23 029 A plant transient has led to a high pressure reactor trip. Following the trip, a pressurizer safety valve opened and did not reclose. The RCS pressure then lowered and PPLS actuation started all of the ECCS pumps. The operators tripped all RCPs.

It is 20 minutes after the trip and current plant conditions are:

• RCS Pressure is steady at 900 psia

• Hot leg and CET temperatures indicate 530ºF

• Pressurizer level is 85% and rising

• All HPSI, LPSI and charging pumps are running.

What action should be taken in response to the rising pressurizer level?

A. Continue to provide full injection flow. Allow all ECCS pumps to continue injecting.

B. Trip one HPSI pump and all LPSI pumps. Continue to provide full charging flow.

C. Trip all HPSI pumps. Trip charging pumps as required to control pressurizer level.

D. Trip HPSI pumps as needed to control pressurizer level. Continue to provide full charging flow.

Monday, August 06, 2012 15:50:35 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 077 The plant was operating at full power with the following conditions:

Containment pressure is 1.2 psig and slowly rising Containment temperature is 105°F and slowly rising Containment dew point is 125°F and slowly rising Pressurizer pressure and level are steady Charging flow is 40 gpm and steady Letdown flow is 32 gpm and lowering What action should be taken as a result of these conditions?

A. Enter OI-VA-1, "Containment Heating, Cooling and Ventilation Systems Normal Operation," and place an additional containment cooling unit in service.

B. Enter OI-VA-1, "Containment Heating, Cooling and Ventilation Systems Normal Operation," and initiate containment pressure relief.

C. Enter AOP-22, "Reactor Coolant Leak," isolate letdown and determine the leak rate.

D. Trip the reactor and enter EOP-00, "Standard Post Trip Actions. Then enter EOP-03, "Loss of Coolant Accident."

Wednesday, August 08, 2012 16:41:29 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 77 Rev 0 KA #: 000009 EA2.11 Tier 1 Group 1: Small Break LOCA Ability to determine or interpret the following as they apply to a small break LOCA:

Containment temperature, pressure, and humidity Importance 3.8 / 4.1 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Use the Reactor Coolant Leak Procedure to mitigate the consequences of a leak in the Reactor Coolant System.

Question Pedigree New question.

K/A Fit:

Question addresses using containment pressure, temperature and dew point to mitigate a small break in containment.

Choice A:

Distractor: Plausible because containment pressure and temperature are rising.

Incorrect because AOP-22 entry conditions have been reached.

Choice B:

Distractor: Plausible because containment pressure is rising. Incorrect because AOP-22 entry conditions have been reached.

Choice C:

Correct answer. Indications of a 4 gpm leak in containment. AOP-22 entry conditions have been reached. AOP-22 will direct letdown isolation and leakrate determination.

Choice D:

Distractor: Plausible because a reactor trip is directed for RCS leakage greater than charging pump capacity. Incorrect because leakage is much less than 40 gpm.

KA#: 000009 EA2.11 Bank Ref #: N/A LP# / Objective: 0717-22 01.00 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

AOP-22 Handout: NONE Wednesday, August 08, 2012 16:41:29 2

AOP-22 Page 3 of 49 Section I - Leak Rate Determination and Leak Isolation 1.0 PURPOSE This procedure provides direction in the event of any size RCS leak from any initial RCS condition.

2.0 ENTRY CONDITIONS RCS Leak Rate has increased as indicated by any one or more of the following conditions:

A. Lowering VCT level.

B. Unexplained mismatch between Charging and Letdown flow.

C. Abnormal change in PZR level.

D. Lowering RCS pressure.

E. Total RCS leak rate exceeds the limits of Technical Specification 2.1.4.

F. High temperature alarms on PORVs or PZR Code Safeties.

G. High Quench Tank level, pressure or temperature.

H. High RCDT level or pressure.

I. High Containment dew point, radiation, pressure or Sump level.

J. Rising S/G Blowdown or Condenser Off-Gas radiation.

K. High Auxiliary Building airborne activity.

L. Primary-to-Secondary leakage verified by Chemistry sampling.

R33

AOP-22 Page 7 of 49 Section I - Leak Rate Determination and Leak Isolation INSTRUCTIONS CONTINGENCY ACTIONS

5. (continued)

d. Verify PZR level is restored. d.1 IF the Reactor is critical AND PZR level continues to drop, THEN initiate a Reactor Trip by performing the following steps:

1) Trip the Reactor.

2) GO TO EOP-00, Standard Post Trip Actions.

6. Determine the RCS leakage rate.

7. IF RCS leakage rate is greater than 40 gpm AND the Reactor is critical, THEN initiate a Reactor Trip by performing the following steps:

a. Trip the Reactor.

b. GO TO EOP-00, Standard Post Trip Actions.

R33

AOP-22 Page 8 of 49 Section I - Leak Rate Determination and Leak Isolation INSTRUCTIONS CONTINGENCY ACTIONS

8. Direct the Shift Chemist to verify 8.1 IF primary to secondary leak rate is primary to secondary leak rate less greater than 1 gpd, than 1 gpd PER CH-AD-0007, Primary THEN IMPLEMENT Attachment B, to Secondary Leak Rate Primary to Secondary Leak Rate Determination. Actions.

NOTES

1. If all Charging Pump auto-start signals are defeated and the Reactor is critical, the Plant must be placed in Hot Shutdown within six hours.

2. If Pressurizer Heater control is lost and the Reactor is critical, the Plant must be placed in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

9. Determine the source of leakage by performing the following steps:

a. Isolate Letdown by closing BOTH Letdown Isolation Valves:

TCV-202 HCV-204

b. Isolate CVCS by performing the following steps:

1) Place the Charging Pump Control Switches in PULL-TO-LOCK.

(continue)

R33

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 078 Given the following plant conditions:

A RCS cooldown is in progress T-cold is 310°F The cooldown rate is 75°F/hr A transient caused RCS pressure to increase to 1900 psia RCS pressure has been reduced to 600 psia What actions are required as a result of these conditions?

A. No actions are required.

B. Enter Technical Specification LCO 2.1.2, "Heatup and Cooldown Rate," and direct engineering to perform a fracture toughness evaluation.

C. Enter TDB-IX, "RCS Pressure and Temperature Limits Report" and direct engineering to perform an Appendix "E" evaluation.

D. Enter EOP/AOP Attachment 27, "P-T Limit Restoration," and maintain current temperature for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Monday, August 06, 2012 16:33:39 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 78 Rev 0 KA #: 000027 2.1.25 Tier 1 Group 1: Pressurizer Pressure Control System Malfunction Ability to interpret reference materials such as graphs, curves, etc.

Importance 3.9 / 4.2 CFR Number: 55.43(b)(2)

Facility operating limitations in the technical specifications and their bases.

Fort Calhoun Objective:

EXPLAIN the basis for the RCS heatup and cooldown curves and STATE the limits.

Question Pedigree New question.

K/A Fit:

Question addresses using graphs and curves to determine action required by technical specifications following a malfunction of the LTOP system.

Choice A:

Distractor: Plausible because all limits are currently being met.

Choice B:

Correct answer. The PTLR limit was exceeded. Tech Spec 2.1.2 requires a fracture toughness evaluation.

Choice C:

Distractor: Plausible because PTLR limits were exceeded. Incorrect because the Appendix E limits were not exceeded.

Choice D:

Distractor: Plausible because this is an action that would be taken if PTLR limits were exceeded during a steam line break. Incorrect because EOP entry conditions have not been met.

HANDOUTS: EOP/AOP Attachment 2 curve and TDB-IX Figure 5-1 KA#: 000027 2.1.25 Bank Ref #: N/A LP# / Objective: 0711-20 02.02 Exam Level: SRO-2 Cognitive Level: HIGH Source: NEW

Reference:

TS 2.1.2 Handout: EOP/AOP ATT 2 Monday, August 06, 2012 16:33:39 2

EOP/AOP ATTACHMENTS Page 4 of 173 Attachment 2 RCS Pressure-Temperature Limits for 40 EFPY NOTES

1. During forced circulation, use TH for the 20° F subcooled curve and saturation curve. During natural circulation, use CETs for the 20° F subcooled curve and the saturation curve. Use TC for all other curves.

2. Curve 3 is the Maximum Pressure for First Start RCP curve.

Reference:

EA06-036 R32

FORT CALHOUN STATION TDB-IX TECHNICAL DATA BOOK REFERENCE USE PAGE 12 OF 69 Figure 5 Fort Calhoun Station Composite P-T Limits 40 EFPY R6

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate Applicability Applies to the temperature change rates and pressure of the Reactor Coolant System (RCS).

Objective To specify limiting conditions of the reactor coolant system heatup and cooldown rates.

Specification The combination of RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR and as designated below:

a. Allowable combinations of pressure and temperature (Tc) for a specific heatup rate shall be below and to the right of the applicable limit lines as shown on the pressure and temperature (P-T) limit Figure(s) in the PTLR.

b. Allowable combinations of pressure and temperature (Tc) for a specific cooldown rate shall be below and to the right of the applicable limit lines as shown on the P-T limit Figure(s) in the PTLR.

c. The heatup rate of the pressurizer shall not exceed 100°F in any one hour period.

d. The cooldown rate of the pressurizer shall not exceed 200EF in any one hour period.

Required Actions (1) When any of the above limits are exceeded, the following corrective actions shall be taken:

1. Immediately initiate action to restore the temperature or pressure to within the limit.

2. Perform an analysis to determine the effects of the out of limit condition on the fracture toughness properties of the reactor coolant system.

3. Determine that the reactor coolant system remains acceptable for continued operation or be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

(2) Before the radiation exposure of the reactor vessel exceeds the exposure for which they apply, the P-T limit Figure(s) shown in the PTLR shall be updated in accordance with the following criteria and procedures:

2.1 - Page 8 Amendment No. 22,74,161, 207,221, 259

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 079 Given the following plant conditions:

The plant is operating at full power The STEAM GEN FEED PUMP B OVERLOAD/TRIP annunciator on CB-10,11/A12 is in alarm The control switch status for Feedwater Pump FW-4A is red light on / green flag The control switch status for Feedwater Pump FW-4B is green light on / red flag The control switch status for Feedwater Pump FW-4C is red light on / red flag Level in both S/G's is 61% narrow range DCS display indicates that both LCV-1101 and LCV-1102 are in three element mode What action should be taken as a result of these conditions?

A. Trip the reactor, enter EOP-00, "Standard Post Trip Actions," and ensure an auxiliary FW pump is running.

B. Trip the reactor, complete EOP-00, "Standard Post Trip Actions," and enter EOP-06, "Loss of All Feedwater."

C. Enter ARP-DCS-FW and place control for LCV-1101 and LCV-1102 in single element mode D. Enter ARP-CB-10,11/A12 and ensure the 43-SIAS/FW4 switch is selected to FW-4C Wednesday, August 08, 2012 16:42:35 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 79 Rev 0 KA #: 000054 AA2.02 Tier 1 Group 1: Loss of Main Feedwater Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW): Differentiation between loss of all MFW and trip of one MFW pump Importance 4.1 / 4.4 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

INTERPRET Feedwater System alarms and actions in the Control Room using the Alarm Response Procedures as a guide.

Question Pedigree New question.

K/A Fit:

Question addresses differentiating between a pump trip and a loss of all feedwater and procedural action to be taken.

Choice A:

Distractor: Plausible if Applicant believes all Feedwater has been lost. Incorrect because only one pump tripped.

Choice B:

Distractor: Plausible if Applicant believes all main and auxiliary feedwater has been lost.

Incorrect because only one pump tripped.

Choice C:

Distractor: Plausible if Applicant believes feedwater transient is due to DCS control malfunction. Incorrect because a FW pump tripped.

Choice D:

Correct answer. FW-4B tripped and FW-4C auto started. ARP directs that 43-SIAS/FW4 switch be selected to FW-4C.

KA#: 000054 AA2.02 Bank Ref #: N/A LP# / Objective: 0711-11 02.07 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

ARP-CB-10,11/A12 Handout: NONE Wednesday, August 08, 2012 16:42:35 2

FORT CALHOUN STATION CONTINUOUS USE ARP-CB-10,11/A12 ANNUNCIATOR RESPONSE PROCEDURE PAGE 17 OF 32 Panel: CB-10 Annunciator: A12 Window: A-6L STEAM GENERATOR FEEDWATER PUMP FW-4B Page 1 of 2 OVERLOAD OR TRIP STEAM GEN SAFETY RELATED FEED PUMP B OVERLOAD/TRIP Tech Spec

References:

None Initiating Device 74 / FW-4B Setpoint Breaker OPEN Power DC BUS 2_

OPERATOR ACTIONS

1. Dispatch an Operator to check the status of FW-4B, Feedwater Pump.

1.1 IF FW-4B is operating, THEN check 83/FW-4B, Overload Relay at 1A2 (1A2-8).

2. IF FW-4B is tripped, THEN ensure that standby Feedwater Pump FW-4A or C has automatically started.

2.1 IF two pumps are required, THEN ensure Feedwater Pumps FW-4A and C are running.

2.2 Ensure 43-SIAS/FW4 Post SIAS Running Feedwater Pump is selected to FW-4C.

2.3 IF no Feedwater Pump can be started, THEN trip the Reactor and GO TO EOP-00.

(continue)

PROBABLE CAUSES Steam Generator Feedwater Pump low oil pressure Feedwater Pump motor overload or trips on phase or differential overcurrent Loadshed on 4160 Volt Bus 1A2 Feedwater Pump stopped from the local 69 permissive control switch REFERENCES 136B3083 Sh 23 06305 0223R0454 Sh 9 09928 R13

FORT CALHOUN STATION CONTINUOUS USE ARP-CB-10,11/A12 ANNUNCIATOR RESPONSE PROCEDURE PAGE 18 OF 32 Panel: CB-10 Annunciator: A12 Window: A-6L STEAM GENERATOR FEEDWATER PUMP FW-4B Page 2 of 2 OVERLOAD OR TRIP SAFETY RELATED OPERATOR ACTIONS (continued) 2.4 Check the following for the cause of the FW-4B trip:

49/FW-4B, Overcurrent Relay tripped at 1A2 (1A2-8)

Motor stopped from local stop pushbutton Motor stopped from the 69 permissive switch (1A2-8)

Low voltage on 4160V Bus 1A2 Low oil pressure on Feedwater Pump 86/FW-4B, Feedwater Pump Motor Differential Overcurrent (AI-12) 2.5 Notify Electrical Maintenance of Feedwater Pump trip.

2.6 Place the running (standby) pump in AFTER-START.

NOTE: Prevent a pump in PULL-OUT from auto starting by first matching flags for the pumps not in PULL-OUT.

3. IF it is desired to align another pump to Standby, THEN perform the following:

3.1 Ensure the Control Switch for the running pump is in AFTER-START (Red Flag).

3.2 Place the Control Switch for the tripped pump in AFTER-STOP (Green Flag).

3.3 Place the Control Switch for the pump in PULL-OUT to AFTER-STOP (Green Flag).

R13

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 080 The reactor was operating at full power, when the following annunciators came into alarm on CB-20, Panel A15:

DC BUS #1 GROUND DC BUS #1 LOW VOLTAGE INVERTER #1 TROUBLE INVERTER A TROUBLE INVERTER C TROUBLE A few seconds later, the reactor tripped. During the performance of EOP-00, Standard Post Trip Actions, the Balance of Plant Operator reported that DC Bus #1 is not energized.

What action should be taken following completion of EOP-00?

A. Enter AOP-16, "Loss of Instrument Bus Power," and transfer Instrument bus power to swing inverter EE-8T.

B. Enter AOP-16, "Loss of Instrument Bus Power," and transfer switchgear DC control power to DC bus #2.

C. Enter EOP-20, "Functional Recovery Procedure," and transfer Instrument bus power to swing inverter EE-8T.

D. Enter EOP-20, "Functional Recovery Procedure," and transfer switchgear DC control power to DC bus #2.

Thursday, August 09, 2012 15:27:39 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 80 Rev 0 KA #: 000058 2.1.07 Tier 1 Group 1: Loss of DC Power Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Importance 4.4 / 4.7 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Given the Resource Assessment Trees, basically DESCRIBE the Method, Path and Acceptance Criteria for each success path.

Question Pedigree New question.

K/A Fit:

Question addresses diagnosing symptoms of a loss of a DC bus and taking proper procedural actions.

Choice A:

Distractor: Plausible because instrument inverter alarms are in. Incorrect because AOP-16 is not entered from EOP-00.

Choice B:

Distractor: Plausible because DC bus number 1 has been lost and AOP-16 does contain directions for transferring DC control power. Incorrect because AOP-16 is not entered from EOP-00.

Choice C:

Plausible because of instrument inverter trouble alarms and EOP-20 is the correct procedure. Incorrect because a DC bus has been lost.

Choice D:

Correct answer: EOP-00 directs transfer to EOP-20 and EOP-20 directs transferring DC control power.

KA#: 000058 2.1.07 Bank Ref #: N/A LP# / Objective: 0718-18 01.05 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

EOP-00, EOP-20 Handout: NONE Thursday, August 09, 2012 15:27:39 2

EOP-00 Page 33 of 44 Diagnostic Actions No Are Reactivity Control Requirements met?

Yes No Are both DC Buses GO TO Functional Recovery energized? Procedure (EOP-20)

Yes No Is at least one vital Consider Station Blackout 4160V Bus energized? Event (EOP-07)

Yes Is at least one No non-vital 4160V Bus energized?

Yes No Consider a loss of offsite power / loss of forced Is at least one RCP circulation event (EOP-02) running?

Yes Does at least one No Consider a loss of S/G have adequate feedwater event (EOP-06) feed?

Yes GO TO B on next page R29

EOP-20 Page 110 of 527 11.0 MAINTENANCE OF VITAL AUXILIARIES - DC MVA-DC INSTRUCTIONS CONTINGENCY ACTIONS

2. IF both DC Buses are energized, 2.1 IF either DC Bus is deenergized, THEN GO TO Step 81. THEN determine the appropriate recovery method by performing the following:

a. IF DC Bus 1 is deenergized, THEN GO TO Step 3.

b. IF DC Bus 2 is deenergized, THEN GO TO Step 42.

CAUTION HCV-438A and C may go closed after control power is transferred on AI-41A.

3. Transfer DC control power to emergency by performing the following:

a. Press the "EMERG. SOURCE 125 VDC" pushbutton (AI-41A).

(continue)

Continuously Applicable or Non-Sequential Step R25

EOP-20 Page 111 of 527 11.0 MAINTENANCE OF VITAL AUXILIARIES - DC MVA-DC INSTRUCTIONS CONTINGENCY ACTIONS

3. (continued)

b. Press PB-2/1B3A-4A-MTS, MANUAL TRANSFER PUSHBUTTON 1B3A-4A-MTS EMERG. SOURCE" pushbutton (East Switchgear Room, 1B3A).

c. Press PB-2/1B3C-4C-MTS, MANUAL TRANSFER PUSHBUTTON 1B3C-4C-MTS EMERG. SOURCE" pushbutton (East Switchgear Room, 1B3C).

d. Press PB-2/1A1-1A3-MTS, MANUAL TRANSFER PUSHBUTTON 1A1-1A3-MTS EMERGENCY SOURCE pushbutton ("1A1-1A3 AUX POWER COMPARTMENT").

(continue)

Continuously Applicable or Non-Sequential Step R25

EOP-20 Page 112 of 527 11.0 MAINTENANCE OF VITAL AUXILIARIES - DC MVA-DC INSTRUCTIONS CONTINGENCY ACTIONS

3. (continued)

e. Press ATD-D1, DIESEL D1 125 VDC MANUAL TRANSFER SWITCH" "EMERGENCY" pushbutton, (D-1 Room, North Wall).

Time:

NOTE PCV-1849, Containment Instrument Air Inboard Isolation Valve may need to be held open while repressurizing the Containment Air header.

4. Restore Instrument Air to Containment by performing the following:

a. Open PCV-1849B, Containment Instrument Air Isolation Valve (Outboard).

a. b. Open PCV-1849A, Containment Instrument Air Isolation Valve (Inboard).

Continuously Applicable or Non-Sequential Step R25

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 081 Given the following plant conditions:

The plant was operating at full power The "PLANT AIR PRESS LO" annunciator alarmed 5 minutes ago The "INSTRUMENT AIR PRESS LO" annunciator alarmed 5 minutes ago The "FW REG SYS FCV-1101 TROUBLE" annunciator just alarmed The "FW REG SYS FCV-1102 TROUBLE" annunciator just alarmed The Balance of Plant Operator reports instrument air pressure is 48 psig and lowering What SEQUENCE of procedures and actions should be used to mitigate this event.?

A. Enter AOP-17, "Loss of Instrument Air," trip the reactor, stop all main FW pumps then implement EOP-00, "Standard Post Trip Actions."

B. Trip the Reactor, Enter EOP-00, "Standard Post Trip Actions," implement AOP-17, "Loss of Instrument Air," then stop all main FW pumps.

C. Enter AOP-17, "Loss of Instrument Air," trip the reactor, reset and close FCV-1101 and FCV-1102 then implement EOP-00, "Standard Post Trip Actions."

D. Trip the Reactor, Enter EOP-00, "Standard Post Trip Actions," implement AOP-17, "Loss of Instrument Air," then reset and close FCV-1101 and FCV-1102.

Wednesday, August 08, 2012 16:43:58 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 81 Rev 0 KA #: 000065 2.1.32 Tier 1 Group 1: Loss of Instrument Air Ability to explain and apply system limits and precautions.

Importance 3.8 / 4.0 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Use the Loss of Instrument Air Procedure to mitigate the consequences of a partial or complete loss of instrument air.

Question Pedigree New question.

K/A Fit:

Question addresses FW control limitations during a loss of instrument air.

Choice A:

Correct answer: AOP-17 is initially entered. When instrument air pressure drops below 50 psig, AOP-17 directs the operators to trip the reactor, stop all main FW pumps and then implement EOP-00, "Standard Post Trip Actions.

Choice B:

Distractor: Plausible because the reactor is tripped. Incorrect because AOP-17 should have been entered on the initial alarms and AOP-17 directs that the reactor be tripped and FW pumps tripped before performing EOP-00.

Choice C:

Distractor: Plausible because AOP-17 is initially entered and directs that the reactor be tripped. Incorrect because AOP-17 directs that the FW pumps be tripped because FCV-1101 and FCV-1102 can not be closed due to the loss of instrument air pressure.

Choice D:

Distractor: Plausible because the reactor is tripped. Incorrect because AOP-17 directs that the FW pumps be tripped because FCV-1101 and FCV-1102 can not be closed due to the loss of instrument air pressure KA#: 000065 2.1.32 Bank Ref #: N/A LP# / Objective: 0717-17 01.00 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

TBD-AOP-17 Handout: NONE Wednesday, August 08, 2012 16:43:58 2

TBD-AOP-17 Page 10 of 46 INSTRUCTIONS CONTINGENCY ACTIONS CAUTION Extended operation with Compressed Air pressure less than 80 psig will result in depletion of Air Accumulator reserves.

13. IF Instrument Air pressure is greater 13.1 IF Instrument Air pressure is less than than or equal to 50 psig, 50 psig, THEN restore Instrument Air by THEN initiate a Reactor Shutdown by performing the following steps: performing the following steps:

a. Determine the source of air a. Trip the Reactor.

leakage.

b. Stop all Main Feed Pumps to

b. Evaluate the need to shutdown prevent an RCS cooldown.

the Reactor PER ONE of the following procedures: c. IMPLEMENT EOP-00, Standard Post Trip Actions.

OP-4, Load Change and Normal Power Operations d. IMPLEMENT the Emergency AOP-05, Emergency Plan.

Shutdown

c. Direct Maintenance to repair the e. Close ONE of the following valves source of air leakage. to isolate LCV-1190 and prevent draining of Condensate Storage

d. Ensure Service Air and Air Dryers Tank to Condenser Hotwell have been returned to normal. (Turbine Mezzanine; West Side):

e. IF air pressure is greater than or FW-269, CONDENSATE equal to 98 psig, MAKEUP VALVE LCV-1190 THEN GO TO Section 5.0, Exit INLET VALVE Conditions. FW-270, CONDENSATE MAKEUP VALVE LCV-1190 OUTLET VALVE R14

TBD-AOP-17 Page 11 of 46 INSTRUCTIONS CONTINGENCY ACTIONS With Instrument Air pressure less than 50 psig, the pneumatic controllers will respond sporadically, if at all. Valves which are necessary for proper operation and control of the plant (i.e., Main Feedwater Reg and Bypass Valves) will not be operable at <50 psig. Therefore, the most conservative course of action would be to trip the plant to prevent more severe complications (R8).

The caution warns the operators that indefinite operation with abnormally low Instrument Air pressures is not possible. The contingency allows for short-term operation with low air pressure while the system is being restored to normal.

EPIP-OSC-1, Emergency Classification, provides instructions for emergency classification upon loss or degradation of the Instrument Air System.

Steps 13 through 25 provide instructions for actions to be taken in the event Instrument Air pressure is lost. This meets an NRC commitment (C1).

If instrument air can not be maintained above 50 psig, then either FW-269 or FW-270 is closed to isolate LCV-1190. LCV-1190 fails open on a loss of instrument air pressure. Backup N2 will keep LCV-1190 closed for a limited time. Manual closure of FW-269 or FW-270 will isolate this line and prevent draining of the condensate storage tank to the condenser hotwell, thereby preserving a suction source for engine-driven AFW pump FW-54.

The Main Feedwater Pumps will cooldown the RCS due to the large amounts of "cold" water added to the Steam Generators. Securing the Main Feedwater Pumps will prevent and/or greatly decrease the RCS cooldown rate by slowing down the rate of cold water addition.

NOTES

1. YCV-1045A and YCV-1045B may open as their air accumulators bleed down, causing FW-10, Steam AFW Pump, to start.

2. Open and close operation of HCV-1107B and HCV-1108B is possible for a minimum of three cycles.

14. Feed S/Gs with AFW to the AFW Nozzles to maintain S/G levels 35-85% NR (73-94% WR) by performing the following steps:

a. Ensure the "43/FW" Switch is in "OFF".

b. Start FW-6, Electric AFW Pump.

R14

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 082 The reactor tripped on low RCS flow following a loss of power to 4160V bus 1A3.

During the performance of EOP-00, Standard Post Trip Actions, the At The Controls Operator reported that two trippable CEAs failed to insert.

Assuming bus 1A3 remains deenergized, which one of the following actions should be taken?

A. Stay in EOP-00, "Standard Post Trip Actions," and dispatch the Aux Building Operator to manually open HCV-268, "Boric Acid Pump Discharge Header Isolation Valve."

B. Stay in EOP-00, "Standard Post Trip Actions," and dispatch the Aux Building Operator to manually open LCV-218-3, "Charging Pump Suction SIRWT Isolation Valve."

C. Enter AOP-03, "Emergency Boration," and dispatch the Aux Building Operator to manually open HCV-265, "CH-11A Gravity Feed Valve."

D. Enter AOP-03, "Emergency Boration," and dispatch the Aux Building Operator to manually close LCV-218-2, Charging Pump Suction VCT Isolation Valve."

Tuesday, August 07, 2012 14:48:44 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 82 Rev 0 KA #: 000024 2.2.37 Tier 1 Group 2: Emergency Boration Ability to determine operability and/or availability of safety related equipment.

Importance 3.6 / 4.6 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Use the Emergency Boration AOP to mitigate the consequences of an uncontrollable or unexplained positive reactivity addition.

Question Pedigree Bank question used on 2004 NRC exam. Reworded to SRO question. Not counted as a modified question.

K/A Fit:

Question addresses operability of equipment needed for emergency boration with a loss of bus 1A3.

Choice A:

Correct answer. HCV-268 is powered from 480 bus 1B3C which is powered from 4160 bus 1A3. When HCV-268 is manually opened, it will allow flow from the powered BA pump, CH-4B.

Choice B:

Distractor: Plausible because LCV-218-3 is used for emergency boration in EOP-20.

Incorrect because it is not used in EOP-00.

Choice C:

Distractor: Plausible because HCV-265 is opened as part of emergency boration.

Incorrect because AOP-03 is not entered and VCT pressure will prevent gravity feed unless LCV-218-2 is closed.

Choice D:

Distractor: Plausible because closing LCV-218-2 will allow emergency boration through the powered gravity feed valve. Incorrect because AOP-03 is not entered.

KA#: 000024 2.2.37 Bank Ref #: 07-17-03 003 LP# / Objective: 0717-03 01.00 Exam Level: SRO-5 Cognitive Level: HIGH Source: BANK

Reference:

STM 12 Handout: NONE Tuesday, August 07, 2012 14:48:44 2

SYSTEM TRAINING MANUAL VOL. 12 CHEMICAL AND VOLUME CONTROL SYSTEM ning, and the concentrated boric acid tanks were the highest pres-DETAILED SYSTEM DESCRIPTION sure suction source available to the charging pumps. In that case, the pumps would further borate the reactor.

• • • Fail Position on Loss of Air or Power 2.586 On a loss of power, HCV-258 and -265 will "fail as is", and the respective indicating lights will go dark.

Boric Acid Pump to Charging Pump Suction Valve HCV-268 2.587 The following is a detailed description of HCV-268.

• • Function 2.588 The boric acid pump to charging pump suction valve, HCV-268, is a normally closed, motor-operated valve that connects the com-bined discharges of the boric acid pumps directly to the suctions of the charging pumps via the VCT outlet line

• • Design/Specification 2.589 This piping arrangement bypasses the boric acid filter to prevent a clogged filter from impeding boric acid flow. In all aspects of control and indication, this valve is identical to HCV-258 and HCV-265. The motor receives power from MCC-3C2.

2.590 HCV-268 has a 1/8-inch hole drilled in the upstream disk to pre-vent thermal binding and pressure locking (MR-FC-96-001).

• • Location 2.591 HCV-268 is located on the 1016-foot elevation of Corridor 26 in the Auxiliary Building, by boric acid storage tanks CH-11A/B.

• • Power Supplies 2.592 HCV-268 receives 480 VAC power from MCC-3C2.

• • Instrumentation and Control 2.593 In all aspects of control and indication, (local, remote, interlocks, alarms and indications) HCV-268 is identical to HCV-258 and HCV-265.

• • Failure Modes 2.594 The following paragraphs describe system response in the event HCV-268 was to fail open or closed.

2.595 In terms of operational characteristics, this valve is identical to HCV-258 and HCV-265. Therefore, its failure modes are also the same.

Level: Basic > Intermediate > Advanced Revision - 45 *

• 108

SYSTEM TRAINING MANUAL VOL. 12 CHEMICAL AND VOLUME CONTROL SYSTEM SYSTEM OPERATION DW LETDOWN B A BAST BAST CH-112 LCV-218-1 WDL M M FCV-269X H2 VENT HCV- HCV-265 258 N2 LT 219 FCV-269Y FCV-LCS 269 VCT F 218 M

LCV-218-2 SI HCV-268 LCV-218-3 CH-152 M

SI-157 SIRWT FIGURE 3-3: EMERGENCY BORATION 3.30 When the RCS is depressurized, all makeup from the CVCS is supplied to the charging pump suction. The boric acid pump and/or demineralized water header supplies sufficient head to provide flow through the idle charging pump and into the RCS, via the high-pressure safety injection header.

3.31 The manual blend mode is normally used when makeup to the SIRWT is required in order to restore level following the use of the SIRWT for safety injection or refueling.

Abnormal System Operation 3.32 The following describes system operation during a loss of letdown and a loss of charging.

• • Loss of Letdown 3.33 Loss of letdown can be affected by the closing of any of the fol-lowing valves:

• Letdown stop valve (TCV-202)

• Letdown flow control valves (LCV-101-1 and 101-2)

Level: Basic > Intermediate > Advanced Revision - 45 *

• 130

SYSTEM TRAINING MANUAL VOL. 12 CHEMICAL AND VOLUME CONTROL SYSTEM normally exist during makeup to the VCT. A locally read pressure DETAILED SYSTEM DESCRIPTION indicating controller is connected to the outlet of each pump.

These controllers provide the low-pressure condition input to the BORIC ACID PUMP CH-4A (CH-4B) AUTO MAKE-UP SIG ON AND DISCH PRESS LO alarm whenever discharge pressure drops below 40 psig.

BORIC ACID BATCHING TANK STRAINER BORIC ACID BORIC ACID STORAGE STORAGE TANK TANK HCV-258 HCV-265 M M A BA B BA PUMP PUMP FILTER HCV M 268 FCV 269Y TO VCT TO CHARGING PUMPS DW FIGURE 2-31: BORIC ACID SECTION 2.524 Each pump is driven by a 480 VAC, 30 horsepower motor. Pump CH-4A is powered from MCC-3C2 and CH-4B powered from MCC-4A2. Design head for each pump is 208 ft. (90 psig). Con-trols and indications for the motors are identical, therefore the fol-lowing explanation will address the pumps generically, with differences being noted.

• • Location 2.525 The boric acid pumps are located in Corridor 26, on the ground floor near the northwest corner of the Auxiliary Building.

• • Power Supplies 2.526 Pump CH-4A is powered from MCC-3C2 and pump CH-4B is powered from MCC-4A2.

Level: Basic > Intermediate > Advanced Revision - 45 *

• 96

STM-ED - Volume 14 Information Use Page 19 of 126 Electrical Distribution System Rev. 42 Figure 1 480 VAC Distribution Figure 1 Instrument AC Distribution

Source question for question 82 QUESTIONS REPORT for ILO EXAM BANK 07-17-03 003 The plant was operating at 100% steady state power when the reactor tripped due to a loss of power to bus 1A3. All control room actions for Emergency Boration were taken.

Assuming bus 1A3 remains deenergized, which of the following local operations would result in emergency boration flow?

A. Opening HCV-258, "CH-11B Gravity Feed Valve" B. Opening HCV-265, "CH-11A Gravity Feed Valve" C. Opening HCV-268, "Boric Acid Pump Discharge Header Isolation Valve" D. Opening LCV-218-3, "Charging Pump Suction SIRWT Isolation Valve" Tuesday, August 07, 2012 14:47:24 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 083 The reactor is operating at full power with the following conditions:

Wide Range NI Channel D failed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ago.

Wide Range NI Channel C failed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> ago.

Wide Range NI Channels A and B just failed What action must be taken per Technical Specifications assuming all four Wide Range Channels remain inoperable?

A. Immediately, trip the reactor and enter EOP-00.

B. Immediately, begin a power reduction to less than 70% power.

C. 4 days from now, initiate a plant shutdown.

D. 7 days from now, initiate a plant shutdown.

Tuesday, August 07, 2012 15:01:50 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 83 Rev 0 KA #: 000033 AA2.10 Tier 1 Group 2: Loss of Intermediate Range Nuclear Instrumentation Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Tech-Spec limits if both intermediate-range channels have failed.

Importance 3.1 / 3.8 CFR Number: 55.43(b)(2)

Facility operating limitations in the technical specifications and their bases.

Fort Calhoun Objective:

Explain the indication provided on AI-212 provided by WR NIS.

Question Pedigree New question.

K/A Fit:

Question addresses Tech Spec actions if all 4 wide range NI channels become inoperable.

Choice A:

Distractor: Plausible because this would be the action to take if all 4 power range NI channels failed. Incorrect because a trip is not required.

Choice B:

Distractor: Plausible because this would be the required action if 2 power range NI channels failed. Incorrect because a power reduction is not required at this time.

Choice C:

Correct answer: With reactor power above 15%, Tech Spec 2.15 requires that a reactor shutdown be initiated 7 days after wide range channel "D" becomes inoperable. Since it became inoperable 3 days ago, 4 days remain until a shutdown must be initiated.

Choice D:

Distractor: Plausible because Tech Spec 2.15 does have a 7 day LCO. Incorrect because 3 days of the allowable 7 have already elapsed.

KA#: 000033 AA2.10 Bank Ref #: N/A LP# / Objective: 0712-18 02.13 Exam Level: SRO-2 Cognitive Level: HIGH Source: NEW

Reference:

TS 2.15 Handout: NONE Tuesday, August 07, 2012 15:01:50 2

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION

• 2.15 Instrumentation and Control Systems (Continued)

If after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of initiating a hot shutdown procedure at least one inoperable engineered safety features or isolation functions channel has not been restored to OPERABLE status, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This specification applied to the high rate trip-wide range log channel when the plant is at or above 10-4% power and is operating below 15% of rated power.

(3) In the event the number of channels on a particular engineered safety features (ESF) or isolation logic subsystem in service falls below the limits given in the columns entitled "Minimum Operable Channels" or "Minimum Degree of Redundancy," except as conditioned by the column entitled "Permissible Bypass Conditions," sufficient channels shall be restored to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> so as to meet the minimum limits or the reactor shall be placed in a hot shutdown condition within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; however, operation can continue without containment ventilation isolation signals available if the ventilation isolation valves are closed. If after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of initiating a hot shutdown procedure sufficient channels have not been restored to OPERABLE status, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(4) In the event the number of channels of those particular systems in service not described in (3) above falls below the limits given in the columns entitled "Minimum Operable Channels" or "Minimum Degree of Redundancy," except as conditioned by the column entitled "Permissible Bypass Conditions," the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If minimum conditions for engineered safety features or isolation functions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of discovering loss of operability, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the number of OPERABLE high rate trip-wide range log channels falls below that given in the column entitled "Minimum Operable Channels" in Table 2-2 and the reactor is at or above 10-4% power and at or below 15% of rated power, reactor critical operation shall be discontinued and the plant placed in an operational mode allowing repair of the inoperable channels before startup or reactor critical operation may proceed.

If during power operation, the rod block function of the secondary CEA position indication system and rod block circuit are inoperable for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or the plant computer PDIL alarm, CEA group deviation alarm and the CEA sequencing function are inoperable for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the CEAs shall be withdrawn and maintained at fully withdrawn and the control rod drive system mode switch shall be maintained in the off position except when manual motion of CEA Group 4 is required to control axial power distribution.

(5) In the event that the number of operable channels of the listed Alternate Shutdown Panels or the Auxiliary Feedwater Panel instrumentation or control circuits falls below the required number of channels, either restore the required number of channels to OPERABLE status within seven (7) days, or be in hot shutdown (Mode 3) within the next twelve hours. This specification is applicable in Modes 1 and 2.

• SEE TDB-VIII 2.15 - Page 2 Amendment No. 8, 20,54,65,88,125,157,194, 208, 249

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION

• 2.15 Instrumentation and Control Systems (Continued)

Function/Instrument Required Number or Control Parameter Location of Channels

1. Reactivity Control

a. Source Range Power AI-212 1

b. Reactor Wide Range AI-212 1 Logarithmic Power

2. Reactor Coolant System Pressure Control

a. Pressurizer Wide Range AI-179 1 Pressure (0-2500 psia)

3. Decay Heat Removal via Steam Generators

a. Reactor Coolant Hot Leg AI-185 1 (Note 1)

Temperature

b. Reactor Coolant Cold Leg AI-185 1 (Note 1)

Temperature

c. Steam Generator Pressure AI-179 1 per Steam Generator

d. Steam Generator Narrow AI-179 1 per Steam Range Level Generator

e. Steam Generator Wide AI-179 1 per Steam Range Level Generator

4. Reactor Coolant System Inventory Controls

a. Pressurizer Level AI-185 1

b. Volume Control Tank Level AI-185 1

c. Charging Pump CH-1B and AI-185 1 its associated controls

d. Charging Isolation Valve AI-185 1 Control

5. Transfer Functions

a. All Transfer Switches/Lockout AI-185 1 Relays

b. All Transfer Switches/Lockout AI-179 1 Relays Note 1: One reactor coolant hot leg temperature indication and one reactor coolant cold leg temperature indication channel must both be operable on the same steam generator (i.e., RC-2A or RC-2B).

• SEE TDB-VIII 2.15 - Page 3 Amendment No. 208, 249

TECHNICAL SPECIFICATIONS

• TABLE 2-2 Instrument Operating Requirements for Reactor Protective System Test, Maintenance Minimum Minimum Permissible and Functional Operable Degree of Bypass Inoperable No. Unit Channels Redundancy Condition Bypass 1 Manual (Trip Buttons) 1 None None N/A 2 High Power Level 2(b)(c) 1(c) Thermal Power (e)

Input Bypassed below 10-4% of Rated Power(a)(d) 3 Thermal Margin/Low 2(b) 1 Below 10-4% of (e)

Pressurizer Pressure Rated Power(a)(d) 4 High Pressurizer 2(b) 1 None (e)

Pressure 5 Low R.C. Flow 2(b) 1 Below 10-4% of (e)

Rated Power (a)(d) 6 Low Steam Generator 2/Steam 1/Steam None (e)

Water Level Gen(b) Gen 7 Low Steam Generator 2/Steam 1/Steam Below 600 (e)

Pressure Gen(b) Gen psia(a)(d) 8 Containment High 2(b) 1 During Leak Test (e)

Pressure 9 Axial Power 2(b)(c) 1(c) Below 15% of (e)

Distribution Rated Power(g) 10 High Rate Trip-wide 2(b) 1 Below 10-4% and (e)

Range Log Channels above 15% of Rated Power(a)(g) 11 Loss of Load 2(b) 1 Below 15% of (e)

Rated Power(g) 12 Steam Generator 2(b) 1 None (e)

Differential Pressure

a. Bypass automatically removed.

b. Specification 2.15(2) is applicable.

• SEE TDB-VIII 2.15 - Page 7 Amendment No. 60,77,88,153,194, 249

TECHNICAL SPECIFICATIONS

• TABLE 2-2 (Continued)

c. If two channels are inoperable, load shall be reduced to 70% or less of rated power.

d. For low power physics testing this trip may be bypassed up to 10-1% of rated power.

e. Specification 2.15(1) is applicable.

f. Deleted.

g. For each channel, the same bistable automatically activates the Loss of Load and Axial Power Distribution (APD) trips and automatically bypasses the high rate trip at 15% of rated power. Only the APD trip is a Limiting Safety System Setting. Therefore, the bistable is set to actuate within the APD tolerance band.

• SEE TDB-VIII 2.15 - Page 8 Amendment No. 60,77,88, 194, 249

SYSTEM TRAINING MANUAL VOL. 29 NUCLEAR INSTRUMENTATION SYSTEM DETAILED SYSTEM DESCRIPTION AI-31A RPS CHANNEL A CHANNEL A WR MONITOR DRAWER CB4 & COMPUTER DETECTORS AMPLIFIER ASSEMBLY AI-31B RPS CHANNEL B WR MONITOR CHANNEL B CB4 & COMPUTER DRAWER DETECTORS AMPLIFIER AI-31E ASSEMBLY AUDIO CONTAINMENT AI-31C COUNT RATE SPEAKER DRAWER RPS CHANNEL C CHANNEL C DETECTORS AMPLIFIER WR MONITOR ASSEMBLY DRAWER CB4 & COMPUTER AI-31D RPS CHANNEL D CHANNEL D OPTICAL DETECTORS AMPLIFIER WR MONITOR ISOLATOR ASSEMBLY DRAWER CB4 & COMPUTER AI-212 REMOTE SHUTDOWN MONITOR NIROOM57 FIGURE 2-1: WIDE RANGE NI SYSTEM OVERVIEW 2.5 In addition, an Appendix R channel (channel D) consists of an opti-cal isolator and wide range monitor. A single audible count rate chassis is provided selectable to any one of the four channels.

2.6 Wide Range NI System channel D (NI-004) provides input to an independent signal processor at alternate shutdown panel AI-212.

For this channel the amplifier signals are fed to an optical isolator NA-004 where the signals are split to two identical wide range monitors. This channel is designed so that a failure of either wide range monitor cannot affect the operation of the other.

• • Location 2.7 Four cabinets (AI-31A, B, C and D) each house one wide range NI channel along with the power range channels. The cabinets are located side by side in the Control Room.

2.8 Alternate shutdown panel AI-212 houses a display for channel D and is located in the upper electrical penetration area (Room 57) of the Auxiliary Building.

• • Power Supplies 2.9 Wide range NI channels A, B, C and D are powered from their respective 120 VAC instrument buses, AI-40A/B/C/D.

Level: Basic > Intermediate > Advanced Revision - 23

• 12

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 084 A fire was detected and confirmed in Room 19.

What procedure should be used by the Control Room Operators to identify safe shutdown equipment potentially threatened by the fire and operator actions to be taken?

A. AOP-06-01, "Fire Emergency Auxiliary Building Radiation Controlled Areas and Containment."

B. AOP-06-02, "Fire Emergency Uncontrolled Areas of Auxiliary Building."

C. Standing Order G-28, "Station Fire Plan."

D. Standing Order G-102, "Fire Protection Program Plan."

Tuesday, August 07, 2012 15:28:31 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 84 Rev 0 KA #: 000067 AA2.12 Tier 1 Group 2: Plant Fire on Site Ability to determine and interpret the following as they apply to the Plant Fire on Site:

Location of vital equipment within fire zone Importance 2.9 / 3.9 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Use the Fire Emergency AOP to mitigate the consequences of a fire on the plant site.

Question Pedigree New question.

K/A Fit:

Question addresses procedure used to identify safe shutdown equipment within a fire zone.

Choice A:

Distractor: Plausible because AOP-06-01 is used to identify safe shutdown equipment in the RCA and containment. Incorrect because Room 19 is in the uncontrolled area of the Auxiliary Building.

Choice B:

Correct answer: AOP-06-02 is used to identify safe shutdown equipment in Room 19.

Choice C:

Distractor. Plausible because SO-G-28 does contain information about safe shutdown equipment located in various areas of the plant. Incorrect because SO-G-28 provides information to the Fire Brigade Incident Commander and Fire Fighters.

Choice D:

Distractor: Plausible because SO-G-102 does address the Fire Protection Program.

Incorrect because it does not identify safe shutdown equipment in specific plant areas.

KA#: 000067 AA2.12 Bank Ref #: N/A LP# / Objective: 0717-06 01.00 Exam Level: SRO-5 Cognitive Level: LOW Source: NEW

Reference:

AOP-06-02 Handout: NONE Tuesday, August 07, 2012 15:28:31 2

AOP-06-02 Page 104 of 149 Section XII - Fire in Room 19 (AFW/Air Compressor Area) 1.0 PURPOSE This procedure provides steps to be followed in the event of a fire in Room 19 (AFW/Air Compressor Area). It provides a Control Room strategy that will protect credited Fire Safe Shutdown equipment. The following information is provided to assist in understanding the strategy employed by this procedure:

Area Name: Room 19 (AFW/Air Compressor Area)

Location: Auxiliary Building - 989' elev.

Fire Area No: 32 Credited Safe Shutdown Equipment Threatened by This Fire: 4160 Volt Electrical Distribution System 480 Volt Electrical Distribution System SI-1A/B/C, HPSI Pumps Charging System PCV-103-1, Pressurizer Spray Valve PCV-103-2, Pressurizer Spray Valve HCV-150, PORV Block Valve HCV-151, PORV Block Valve PCV-102-1, PORV PCV-102-2, PORV FW-6, Motor Driven AFW Pump FW-10, Steam Driven AFW Pump Electrical Supply Panels Located in This Area: None Instrumentation Available A/LT-911 B/LT-911 LT-101X for Fire Area: A/LT-912 B/LT-912 LT-101Y A/PT-902 B/PT-902 NE-003 A/PT-905 B/PT-905 NE-004 A/TE-112C B/TE-112C PT-105 A/TE-112H B/TE-112H PT-115 R3

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 085 Given the following plant conditions:

Fuel movement is in progress inside containment and at the spent fuel pool The containment equipment hatch is open Both PAL doors are closed All other containment penetrations open to the atmosphere are closed VIAS and CRHS actuations have occurred Radiation Monitors RM-050 and RM-051 are in alarm The refueling SRO reports that a fuel assembly has been dropped Which one of the following actions should be taken as a result of these conditions?

A. Enter AOP-08, "Fuel Handling Incident," and direct securing of the Room 66 openings to isolate any release to the atmosphere.

B. Enter AOP-08, "Fuel Handling Incident," and ensure that the AB supply and exhaust fans are operating as required to maintain a negative pressure.

C. Enter AOP-09, "High Radioactivity," and ensure that the AB supply and exhaust fans are operating as required to maintain a negative pressure.

D. Enter AOP-12, "Loss of Containment Integrity," and direct securing of the room 66 openings to isolate any release to the atmosphere.

Tuesday, August 07, 2012 15:54:26 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 85 Rev 0 KA #: 000036 2.4.18 Tier 1 Group 2: Fuel Handling Incidents Knowledge of the specific bases for EOPs.

Importance 3.3 / 4.0 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Use the Fuel Handling Incident Procedure to mitigate the consequences in the event an irradiated fuel assembly is dropped or damaged.

Question Pedigree New question.

K/A Fit:

Question addresses procedures, procedural action and their bases during a fuel handling incident.

Choice A:

Correct answer: AOP-08 is entered and it directs securing the Room 66 openings.

Choice B:

Distractor: Plausible because AOP-08 is entered and fuel movement is in progress in the Auxiliary Building. Incorrect because RM-050 and RM-051 monitor the containment atmosphere and AOP-08 only realigns AB ventilation if the incident is in the AB.

Choice C:

Distractor: Plausible because RM-050 and RM-051 alarming are entry conditions for AOP-09. Incorrect because AOP-09 contains a not that AOP-08 provides guidance for dropped fuel assembly.

Choice D:

Distractor: Plausible because the containment equipment hatch is open. Incorrect because this is not an AOP-12 action.

KA#: 000036 2.4.18 Bank Ref #: N/A LP# / Objective: 0717-08 01.00 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

TBD-AOP-08 Handout: NONE Tuesday, August 07, 2012 15:54:26 2

AOP-08 Page 5 of 12 INSTRUCTIONS CONTINGENCY ACTIONS NOTE It is required to close the following within one hour: Room 66 Roll-up Doors and Room 66 Construction Access Opening or the Equipment Hatch, all containment penetrations open to the outside atmosphere and one door in the PAL.

5. Secure Room 66 Openings by 5.1 IF unable to close Room 66 Roll-up Doors and Room 66 Construction performing the following:

Access Opening,

• Direct Security Shift Manager to THEN direct Shift Outage Manager to close Room 66 Roll-up Doors close the Equipment hatch.

1009-1, 1013-3, and 1013-4.

• Direct Maintenance to install cover for the Room 66 Construction Access Opening.

6. Direct Shift Outage Manager to close all containment penetrations open to the outside atmosphere.

7. WHEN Room 66 Roll-up Doors and Room 66 Construction Access Opening OR Equipment Hatch and all applicable containment penetrations have been closed, THEN direct the EONA to close at least one PAL door.

R9

AOP-08 Page 6 of 12 INSTRUCTIONS CONTINGENCY ACTIONS

8. Verify VIAS actuation PER 8.1 IF VIAS did NOT actuate, Attachment A, VIAS Actuation. THEN ensure VIAS actuates by performing the following steps:

a. Trip VIAS using CRHS test switches.

b. Verify VIAS actuation PER Attachment A, VIAS Actuation.

9. IF the incident was in the Auxiliary Building, THEN align ventilating equipment by performing the following steps (AI-44):

a. Ensure VA-66, Spent Fuel Area Charcoal Filter, is in the filtered mode.

(continue)

R9

AOP-08 Page 7 of 12 INSTRUCTIONS CONTINGENCY ACTIONS

9. (continued)

b. Verify at least one of the Stack b.1 IF Stack Radiation Monitors are Radiation Monitors has power not powered or pumps are not and the pump is energized energized, (AI-33C). THEN align a power source and energize the associated pump:

• RM-062 and RM-063 (MCC-4C2)

• RM-052 (MCC-4C2) (normal)

• RM-052 (MCC-3B1),

Attachment 2 of OI-RM-1, Radiation Monitoring

c. Ensure a negative pressure exists in the Auxiliary Building by performing the following steps:

1) Ensure ANY two of the following Auxiliary Building Exhaust Fans are in service:

• VA-40A

• VA-40B

• VA-40C

2) Ensure ONE Auxiliary Building Supply Fan is in service:

• VA-35A

• VA-35B R9

AOP-09 Page 3 of 15 3.0 PRECAUTIONS The following specific cautions and notes apply prior to or throughout this procedure.

A. CAUTIONS None B. NOTES

1. AOP-08, Fuel Handling Incident, provides guidance in the event an irradiated fuel assembly is dropped or otherwise damaged.

R10

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 086 The following plant conditions exist:

The reactor is at 100% power The RC-3A "SEAL LEAKAGE FLOW HI" annunciator is in alarm VCT pressure is 50 psia.

RC-3A upper seal inlet pressure is 110 psia RC-3A middle seal inlet pressure is 1095 psia Pressurizer pressure is 2100 psia What action should be taken as a result of these conditions per AOP-35, Reactor Coolant Pump Malfunctions?"

A. Stay in AOP-35 and continue to monitor seal parameters.

B. Perform a plant shutdown using OP-4, "Load Change and Normal Power Operation" and OP-3A, "Plant Shutdown."

C. Perform a plant shutdown using AOP-5, "Emergency Shutdown."

D. Trip the reactor and enter EOP-00, "Standard Post-Trip Actions" then trip RCP RC-3A.

Tuesday, August 07, 2012 16:17:35 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 86 Rev 0 KA #: 003000 2.1.28 Tier 2 Group 1: Reactor Coolant Pump System Knowledge of the purpose and function of major system components and controls.

Importance 4.1 / 4.1 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

EXPLAIN the operation of the RCP seal package.

Question Pedigree Modified question. Stem conditions modified to make another choice correct and selection of appropriate procedure added.

K/A Fit:

Question addresses loss of a reactor coolant pump seal.

Choice A:

Correct answer: The upper RCP seal has failed. AOP-35 directs the operators to continue to monitor seal parameters.

Choice B:

Distractor: Plausible because AOP-35 directs a plant shutdown if two seals have failed.

Incorrect because only one seal has failed and AOP-05 would be used for the shutdown if two seals failed.

Choice C:

Distractor: Plausible because this is the correct answer if two seals failed. Incorrect because only one seal failed.

Choice D:

Distractor: Plausible because this would be the correct answer if more than two seals failed. Incorrect because only one seal failed.

KA#: 003000 2.1.28 Bank Ref #: 07-11-20 154 LP# / Objective: 0711-20 01.07D Exam Level: SRO-5 Cognitive Level: HIGH Source: MODIFIED

Reference:

AOP-35 Handout: NONE Tuesday, August 07, 2012 16:17:35 2

AOP-35 Page 4 of 24 Section I - RCP Seal Failure 3.0 PRECAUTIONS The following specific cautions and notes apply prior to or throughout this procedure.

A. CAUTIONS

1. All indications shall be verified by independent instrumentation (where available).

B. NOTES

1. RCP operating parameters are available on the ERF computer on Page 342 in a summary format for comparison between pumps and on pages 440, 441, 442 and 443 for individual pumps.

2. A differential pressure across a seal stage of less than 100 psid indicates a seal stage failure.

3. Middle seal inlet pressure less than 1000 psig with Controlled Bleedoff less than 0.75 gpm, indicates blockage of the lower pressure breakdown device.

R5

AOP-35 Page 5 of 24 Section I - RCP Seal Failure 4.0 INSTRUCTIONS/CONTINGENCY ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS

1. Verify NONE of the following conditions 1.1 IF any condition exists exist for any RCP: AND the Reactor is critical, THEN stop the affected RCPs by

• Lower seal cavity temperature exceeds 200°F performing the following steps:

• Vapor seal pressure equals RCS pressure

• More than two seals have failed a. Trip the Reactor.

b. IMPLEMENT EOP-00, Standard Post Trip Actions.

c. Stop the affected RCPs.

d. GO TO Section 5.0, Exit Conditions.

1.2 IF any condition exists AND the Reactor is shutdown, THEN stop the affected RCPs.

R5

AOP-35 Page 6 of 24 Section I - RCP Seal Failure INSTRUCTIONS CONTINGENCY ACTIONS

2. Verify proper RCP seal operation by ALL 2.1 IF only one seal per RCP has failed, of the following conditions: THEN continue to monitor seal parameters for the affected RCPs.

• P across each RCP seal is greater than 100 psid 2.2 IF ANY of the following conditions exist:

• Middle seal inlet pressure is greater than 500 psig with Controlled Bleedoff

• Two seals have failed flow greater than 0.5 gpm

• Middle seal inlet pressure is less than

• Bleedoff temperature is less than or equal to 500 psig and Controlled 250°F Bleedoff flow is less than 0.5 gpm

• Bleedoff temperature is greater than or equal to 250°F AND the Reactor is critical, THEN stop the affected RCPs by performing the following steps:

a. Commence an immediate Plant Shutdown PER AOP-05, Emergency Shutdown.

(continue)

b. Continue to monitor seal parameters for the affected RCPs.

c. Monitor for entry conditions of AOP-22, Reactor Coolant Leak.

(continue)

R5

Source question for question 86 QUESTIONS REPORT for ILO EXAM BANK 07-11-20 154 The RO reports the following plant conditions:

The reactor is at 100% power RCS pressure is 2100 psia The RC-3A "SEAL LEAKAGE FLOW HI" annunciator is in alarm VCT pressure is 50 psia.

RC-3A middle seal inlet pressure is 110 psia RC-3A upper seal inlet pressure is 80 psia What actions should you direct your crew to take?

A. Monitor the RC-3A's seals on the ERF computer. Continue full power operation.

B. Perform a normal plant shutdown using OP-3A, then shutdown RC-3A.

C. Perform an emergency plant shutdown using AOP-05, then shutdown RC-3A.

D. Trip the reactor, then shutdown RC-3A and enter EOP-00.

Tuesday, August 07, 2012 16:16:07 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 087 Five days ago the Army Corps of engineers notified the plant that the level of the Missouri River at the plant was expected to rise to 1007 feet MSL over the next 24-36 hours.

Current plant conditions:

The plant has been placed in cold shutdown with shutdown cooling in service The river level at the plant is 1004 feet MSL and rising All 4160 volt buses are being supplied by 161 KV offsite power Which one of the following actions should be taken per FCS procedures?

A. Transfer loads on one vital bus to a diesel generator per AOP-01, "Acts of Nature."

B. Transfer loads on both vital buses to the diesel generators per AOP-01, "Acts of Nature."

C. Transfer loads on 480 V Buses 1B3A, 1B4A and 1B3A-4A to 13.8 KV supply per EOP/AOP Attachment 5, "Energizing 480 V Buses From 13.8 KV."

D. Transfer loads on 480 V Buses 1B3C, 1B4C and 1B3C-4C to 13.8 KV supply per EOP/AOP Attachment 5, "Energizing 480 V Buses From 13.8 KV."

Wednesday, August 08, 2012 16:44:31 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 87 Rev 0 KA #: 064000 2.4.20 Tier 2 Group 1: Emergency Diesel Generators Knowledge of operational implications of EOP warnings, cautions, and notes.

Importance 3.8 / 4.3 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Describe how the plant responds to the following: Flood.

Question Pedigree New question.

K/A Fit:

Question addresses procedural actions for diesel generators related to several notes in the AOP for external flooding Choice A:

Correct answer: AOP-01 directs transferring loads to one diesel generator if the river level is expected to exceed 1006 feet 6 inches. Only one diesel generator is used to conserve fuel.

Choice B:

Distractor: Plausible because vital loads are transferred to the diesel generators.

Incorrect because only one diesel generator is used.

Choice C:

Distractor: Plausible because there is an EOP/AOP Attachment for powering 480 volt buses from 13.8 KV. Incorrect because the attachment is not used with AOP-01 and does not provide power to the listed 480 volt buses.

Choice D:

Distractor: Plausible because there is an EOP/AOP Attachment for powering the listed 480 volt buses from 13.8 KV. Incorrect because the attachment is not used with AOP-01.

KA#: 064000 2.4.20 Bank Ref #: N/A LP# / Objective: 0717-01 01.02A Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

AOP-01 Handout: NONE Wednesday, August 08, 2012 16:44:31 2

AOP-01 Page 4 of 124 Section I - Flood 3.0 PRECAUTIONS The following specific cautions and notes apply prior to or throughout this procedure.

A. CAUTIONS None B. NOTES

1. The National Weather Service and the U.S. Army Corps of Engineers provide continuous forecasts regarding the weather and the Missouri River flood stage.

2. 1004 feet is the elevation of the plant site and the Security Building floor.

3. Terminals in the 161KV Switchyard Control Building are flooded at an elevation of 1005 feet 6 inches.

4. Terminals in the West 345KV Switchyard Control Building are flooded at an elevation of 1005 feet 9 inches.

5. 1007 feet 6 inches is the elevation of the Intake Structure floor and the elevation at which the Turbine Building Truck Bay overflows and where the terminals in the East 345KV Switchyard Control Building are flooded.

6. Installation of Flood Barriers in the Intake Structure and Auxiliary Building will protect those areas to at least 1014 feet. Internal building leakage will need to be controlled by temporary dewatering pumps.

7. Maintenance is responsible for the installation and removal of flood control barriers and sandbagging operations.

8. Technical Specification 2.16, River Level, addresses operating limitations associated with river level.

9. On an upstream dam failure river level at the station is automatically assumed to exceed 1004 feet. A failure of the nearest upstream dam (i.e., Gavins Point Dam) would result in the flood surge reaching the Station in approximately 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br />. The estimated rate of rise is two inches per hour once it arrives.

R31

AOP-01 Page 5 of 124 Section I - Flood 3.0 PRECAUTIONS (continued)

B. NOTES (continued)

10. AOP-01 directs flood protection of vital SSCs. Non-vital protection is directed by FCSG-64, External Flooding of the Site.

11. To convert the Blair NE river level gauge to the elevation at Fort Calhoun Station add 975.74 ft.

12. Following the loss of off-site power, utilizing a single Emergency Diesel Generator to power station loads minimizes Diesel Fuel Oil consumption.

13. River level is the water level as measured at the FCS Intake Structure by L1900 (ERF) or approved equivalent level instrumentation.

R31

AOP-01 Page 19 of 124 Section I - Flood INSTRUCTIONS CONTINGENCY ACTIONS NOTES

1. The goal of transferring 4160V loads to a diesel-generator before losing off-site power should be balanced against the goal of maximizing fuel oil inventory. The decision of when to transfer loads should consider the rate at which the flood level is rising.

2. Bus loading should be less than 1550kw for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 1381kw thereafter to ensure a 7 day supply of Diesel Fuel without an available offsite source.

11. IF river level reaches 1004 feet and is expected to exceed 1005 feet 6 inches THEN transfer 4160 volt power to the Emergency Diesel Generators by performing the following:

a. Ensure the Reactor is Tripped.

b. IF Bus 1A3 is to remain energized b.1 IF Bus 1A4 is to remain energized, THEN perform the following: THEN perform the following:

1) Deenergize Bus 1A3 by 1) Deenergize Bus 1A4 by opening BOTH of the opening BOTH of the following following breakers: breakers:

1A13 1A24 1A33 1A44 (continue) (continue)

R31

AOP-01 Page 20 of 124 Section I - Flood INSTRUCTIONS CONTINGENCY ACTIONS 11.b. (continued) b.1 (continued)

2) Ensure that Bus 1A3 is 2) Ensure that Bus 1A4 is powered from Diesel powered from Diesel Generator 1. Generator 2.

3) IF TWO CCW Pumps are running, THEN stop ONE CCW Pump, AC-3A or AC-3C.

(continue)

(continue)

R31

AOP-01 Page 21 of 124 Section I - Flood INSTRUCTIONS CONTINGENCY ACTIONS NOTE A LPSI pump is required to be started only if on Shutdown Cooling 11.b. (continued) b.1 (continued)

3) Attempt to start ALL of the 4) Attempt to start ALL of the following equipment powered following equipment powered from Bus 1A3: from Bus 1A4:

ONE CCW Pump, AC-3B, CCW Pump AC-3A/C ONE Raw Water Pump, ONE Raw Water Pump, AC-10B/D AC-10A/C AC-9B, Bearing Water AC-9A, Bearing Water Pump Pump Air Compressor CA-1B ONE Air Compressor ONE Charging Pump, CA-1A/C CH-1B/C CH-1A, Charging Pump VA-7D, Containment Vent VA-7C, Containment Fan Vent Fan SI-1B, LPSI Pump SI-1A, LPSI Pump (continue)

(continue)

R31

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 088 The plant was in hot shutdown with T-cold at 532°F and all CEAs inserted when the high containment trip bistables all tripped on high containment pressure PPLS,CPHS and SGLS actuations all occurred Three charging pumps, Two HPSI pumps, 2 LPSI pumps and Containment Spray Pump SI-3B are operating Containment Spray Pump SI-3A Failed to start and cannot be manually started Containment Cooling Fans, VA-3A and VA-3B, are running Containment Cooling Fans, VA-7C and VA-7D, are not running Containment pressure is 35 psig and slowly lowering What action should be taken in response to these conditions?

A. Enter AOP-40, "Overcooling/Excessive Steam Demand," and manually start Containment Spray Pump SI-3C.

B. Enter AOP-40, "Overcooling/Excessive Steam Demand," and manually start VA-7C and VA-7D.

C. Enter EOP-00, "Standard Post Trip Actions," and manually start Containment Spray Pump SI-3C.

D. Enter EOP-00, "Standard Post Trip Actions," and manually start VA-7C and VA-7D.

Wednesday, August 08, 2012 10:06:04 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 88 Rev 0 KA #: 013000 A2.02 Tier 2 Group 1: Engineered Safety Features Actuation System Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; Excess steam demand Importance 4.3 / 4.5 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

STATE plant conditions that require manual initiation of engineered safeguards.

Question Pedigree New question.

K/A Fit:

Question address using procedures to mitigate a steam line break.

Choice A:

Distractor: Plausible because the plant is shutdown and only one containment spray pump is running. Incorrect because SO-O-1 requires entry into EOP-00 if T-cold is 525°F or greater at the start of the event and EOP-05 contains a caution that says "Do not run SI-3B and SI-3C at the same time."

Choice B:

Distractor: Plausible because the plant is shutdown. Incorrect because SO-O-1 requires entry into EOP-00 if T-cold is above 525°F at the start of the event.

Choice C:

Distractor: Plausible because EOP-00 should be entered and only one containment spray pump is operating. Incorrect because EOP-05 contains a caution that says "Do not run SI-3B and SI-3C at the same time."

Choice D:

Correct answer: EOP-00 should be entered because T-cold was above 525°F at the start of the event. EOP-00 directs starting VA-7C and VA-7D.

KA#: 013000 A2.02 Bank Ref #: N/A LP# / Objective: 0712-14 03.01 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

EOP-00 Handout: NONE Wednesday, August 08, 2012 10:06:04 2

EOP-00 Page 28 of 44 INSTRUCTIONS CONTINGENCY ACTIONS

15. (continued)

f. Verify that BOTH of the following f.1 IF Containment pressure is Containment conditions exist: greater than or equal to 3.0 psig, OR Containment temperature is Containment pressure is less than 3.0 psig greater than or equal to 120°F, THEN initiate Containment Containment temperature is less than 120°F (ERF "CI" Cooling by performing the display) following:

1) Start ALL of the following available Containment Vent Fans:

VA-3A VA-7C VA-3B (continue) VA-7D

2) Ensure CCW is supplied to the coils of the operating Containment Vent Fans.

(continue)

R29

EOP-05 Page 4 of 58 4.0 PRECAUTIONS The following specific cautions and notes apply prior to or throughout this procedure.

A. CAUTIONS

1. UHEs inside Containment can magnify instrument inaccuracies. A single indication may be unreliable. All available indications and actual Plant response should be used to evaluate actions and Plant conditions. TSC staff should assist in evaluating these uncertainties through use of the Electrical Equipment Qualification Manual.

2. Do not allow Diesel Generator loads to exceed power and current rating limits.

3. Do not run SI-3B and SI-3C at the same time.

4. No more than three RCPs shall be in operation when RCS temperature is less than 500°F.

B. NOTES

1. Floating Step BB, Minimizing DC Loads, requires operator action within 15 minutes of loss of either battery charger.

R25

FORT CALHOUN STATION SO-O-1 STANDING ORDER INFORMATION USE PAGE 67 OF 74 5.19.3 Guidance on Terminating a Procedure Before it is Completed 5.19.3.A If the procedure (OI, OP, ST etc.) being performed must be terminated prior to completion, the steps or sections between the point of termination and the steps for restoration shall be marked N/A. The authority to N/A shall be as specified in SO-G-7 Operating Manual, or SO-G-23 Surveillance Test Program.

The system may then be returned to normal lineup in accordance with the restoration steps of the procedure being used. If the procedure modified the plant system (e.g., installed test equipment) and it is preferred not to remove the modification, then the requirements of SO-O-25 Temporary Modification Control, shall be followed by indicating the modification in the Temporary Modification Index.

5.19.4 Procedural Guidance during Off-Normal Conditions 5.19.4.A The following information is offered to enable consistent, conservative response to abnormal events or conditions whereby implementation of a procedure, EOP, AOP or ARP is deemed appropriate.

5.19.4.A.1) When the generator is on-line, the Operator shall enter the Standard Post-Trip Action procedure (EOP-00) whenever an automatic or manual RPS Actuation/Reactor Trip occurs or if an EOP-00 entry condition is met.

5.19.4.A.2) When Tc 525°F and the generator is off-line, the Operator shall enter EOP-00 and subsequent EOP optimal recovery or functional recovery procedures, and follow the procedure as written to the extent possible, whenever an RPS actuation/reactor trip occurs (automatic or manual) or EOP-00 (Standard Post Trip Actions) entry condition is met. It should be noted that verbatim compliance is not required as it will be necessary to use operational judgment in applying the instructions of the EOP since plant configuration may be somewhat different than the configuration assumed when the EOPs were designed. For example, the instruction in EOP-00 which requires the Operator to verify output breakers 3451-4 and 3451-5 are open following a reactor/turbine trip would be inappropriate if 3451-4 or 3451-5 were closed for 345KV backfeed.

The EOPs do however offer sound guidance to an Operator when events occur in plant modes not assumed in current station procedures.

Specific attention should be paid to the maintenance of safety functions.

R94

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 089 Given the following plant conditions:

A loss of load event resulted in a high pressure reactor trip Both of the Power-Operated Relief Valves (PORVs) opened PORV, PCV-102-1, failed to close when pressurizer pressure lowered below its setpoint Power was not available to PORV block valve, HCV-151, during performance of EOP-00, "Standard Post Trip Actions" Pressurizer level is 100%

Power has now been restored to HCV-151 What Action should be taken as a result of these conditions?

A. Close HCV-151 and go to EOP-01, "Reactor Trip Recovery."

B. Enter EOP-03, "Loss of Coolant Accident," and close HCV-151.

C. Enter EOP-03, "Loss of Coolant Accident," and begin a RCS cooldown.

D. Enter EOP-20, "Functional Recovery Procedure," and begin a RCS cooldown.

Wednesday, August 08, 2012 16:45:05 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 89 Rev 0 KA #: 010000 A2.03 Tier 2 Group 1: Pressurizer Pressure Control System Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: PORV failures Importance 4.1 / 4.2 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

EXPLAIN the major strategy used to mitigate the consequences of a LOCA.

Question Pedigree Modified question. The source question was used on the 2010 NRC exam.

K/A Fit:

Question addresses procedural actions to mitigate a stuck open PORV.

Choice A:

Distractor: Plausible because EOP-00 does have instructions for isolating a stuck open PORV. Incorrect because EOP-03 will be entered for a LOCA, even if isolated.

Choice B:

Distractor: Plausible because EOP-03 is the correct procedure and does contain steps to isolate the PORV. Incorrect because EOP-03 has a caution that says "DO NOT isolate a PORV if the pressurizer is water solid."

Choice C:

Correct answer: EOP-03 entry conditions are met and the caution in EOP-03 states "DO NOT isolate a PORV if the pressurizer is water solid. A cooldown must be commenced before isolation."

Choice D:

Distractor: Plausible because the cooldown is correct. Incorrect because EOP-20 entry conditions are not met.

KA#: 010000 A2.03 Bank Ref #: 2010 NRC Q 76 LP# / Objective: 0718-13 01.01 Exam Level: SRO-5 Cognitive Level: HIGH Source: MODIFIED

Reference:

EOP-03 Handout: NONE Wednesday, August 08, 2012 16:45:05 2

EOP-03 Page 16 of 82 INSTRUCTIONS CONTINGENCY ACTIONS CAUTION Do NOT isolate a PORV if the pressurizer is water solid. A cooldown must be commenced before isolation.

12. Verify PORVs and PZR Code Safety 12.1 IF RCS pressure is less than Valves are closed by performing the 2300 psia, following: AND a PORV is open, THEN isolate the PORV(s) by closing

a. Check that ALL of the following ONE PORV Block Valve for the PZR Quench Tank Alarms are associated PORV:

clear (CB-1,2,3; A4):

• PCV-102-1 HCV-151

• "QUENCH TANK TEMP HI"

• "QUENCH TANK PRESS HI"

• PCV-102-2

• "QUENCH TANK LEVEL HCV-150 HI-LO"

b. Check that ALL of the following Relief Header Temperature (continue)

Alarms are clear (CB-1,2,3; A4):

• "PRESSURIZER PWR OPERATED RELIEF VALVE DISCH TEMP HI"

• "PRESSURIZER SAFETY VALVE RC-141 DISCH TEMP HI"

• "PRESSURIZER SAFETY VALVE RC-142 DISCH TEMP HI" (continue)

Continuously Applicable or Non-Sequential Step R36

Source question for question 89 QUESTIONS REPORT for 2010 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 076 Given the following plant conditions:

The plant has tripped from full power due to a High Pressurizer Pressure Trip EOP-00, STANDARD POST TRIP ACTIONS, were completed and EOP-01, REACTOR TRIP RECOVERY, was entered Several minutes later PPLS actuated All Engineered Safety Features operated as designed Pressurizer Pressure is now 931 psia The ATCO has tripped all Reactor Coolant Pumps Pressurizer Level is 100% (solid)

Representative CET Temperature is 536°F PORV, PCV-102-2, is discovered to be open Which one of the following actions should be taken?

A. Go to Diagnostic Actions of EOP-00, STANDARD POST TRIP ACTIONS, then enter EOP-03, LOSS OF COOLANT ACCIDENT, and close HCV-151, the block valve for PCV-102-2.

B. Stay in EOP-01, REACTOR TRIP RECOVERY, and reestablish letdown using EOP/AOP Attachment 23, RESTORATION OF LETDOWN.

C. Go to Diagnostic Actions of EOP-00, STANDARD POST TRIP ACTIONS, then enter EOP-03, LOSS OF COOLANT ACCIDENT, and begin a plant cooldown.

D. Go to EOP-20, FUNCTIONAL RECOVERY PROCEDURE, implement EOP/AOP Floating Step A, HPSI STOP AND THROTTLE CRITERIA.

Wednesday, August 08, 2012 10:18:15 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 090 Given the following plant conditions:

The reactor is operating at full power Instrument Air Dryer, CA-12, is in service The following annunciators are in alarm:

AIR DRYER TROUBLE PLANT AIR PRESS LO INSTRUMENT AIR PRESS LO Instrument air pressure is steady at 78 psig What action should be taken in response to these conditions?

A. Enter AOP-17,"Loss of Instrument Air," direct a building operator to open CA-197, "Bypass Control Valve PCV-1752 Bypass Valve," then place Instrument Air Dryer, CA-31, in service per OI-CA-1.

B. Enter AOP-17,"Loss of Instrument Air," Close PCV-1849A/B, Instrument Air Containment Isolation Valves then direct a building operator to place Instrument Air Dryer, CA-31, in service per OI-CA-1.

C. Trip the reactor, complete EOP-00,"Standard Post Trip Actions." Then enter EOP-20,"Functional Recovery Procedure, MVA-IA." direct a building operator to open CA-197, "Bypass Control Valve PCV-1752 Bypass Valve," then place Instrument Air Dryer, CA-31, in service per OI-CA-1.

D. Trip the reactor, complete EOP-00,"Standard Post Trip Actions. Then enter EOP-20,"Functional Recovery Procedure, MVA-IA." Close PCV-1849A/B, Instrument Air Containment Isolation Valves, then direct a building operator to place Instrument Air Dryer, CA-31, in service per OI-CA-1.

Wednesday, August 08, 2012 16:45:43 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 90 Rev 0 KA #: 078000 A2.01 Tier 2 Group 1: Instrument Air System Ability to (a) predict the impacts of the following malfunctions or operations on the IAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Air dryer and filter malfunctions Importance 2.4 / 2.9 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Explain the principles of Abnormal operation of the Compressed Air System in terms of flow paths, major parameters, (temperature, pressure, flow, etc.), alarms and control devices.

Question Pedigree New question.

K/A Fit:

Question addresses using procedures to mitigate an air dryer malfunction.

Choice A:

Correct answer: Instrument air pressure will be holding at 78 psig due to automatic operation of PCV-1752. AOP-17 directs that CA-197 be opened and CA-31 be placed in service.

Choice B:

Distractor: Plausible because AOP-17 is entered and closing PCV-1849A/B is an action in AOP-17. Incorrect because that step in AOP17 will not be reached for a dryer failure.

Choice C:

Distractor: Plausible because AOP-17 does direct tripping the plant if instrument air pressure gets low enough. Incorrect because instrument air pressure must fall below 50 psig before the plant is tripped.

Choice D:

Distractor: Plausible because AOP-17 does direct tripping the plant if instrument air pressure gets low enough. Incorrect because instrument air pressure must fall below 50 psig before the plant is tripped.

KA#: 078000 A2.01 Bank Ref #: N/A LP# / Objective: 0711-07 01.05 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

AOP-17 Handout: NONE Wednesday, August 08, 2012 16:45:43 2

AOP-17 Page 6 of 46 INSTRUCTIONS CONTINGENCY ACTIONS

3. IF the "AIR DRYER TROUBLE" alarm (CB-10,11; A11) annunciates, THEN direct an operator to proceed to AI-79 to determine the cause of the alarm (Room 19).

4. Direct all available operators to search for the source of the air leakage.

5. IF Instrument Air pressure is less than 5.1 IF PCV-1753 is NOT closed, 80 psig, THEN close CA-121, "SERVICE AIR THEN verify PCV-1753, "SERVICE SUPPLY SYSTEM MANUAL AIR SYSTEM AUTOMATIC ISOLATION VALVE" (Room 19).

ISOLATION VALVE", is closed (Room 19).

6. IF Instrument Air pressure returns to a normal 98-108 psig after isolating Service Air, THEN GO TO Section 5.0, Exit Conditions.

7. IF Instrument Air pressure is less than 7.1 IF Instrument Air pressure is less than 78 psig, 78 psig THEN verify that PCV-1752, "AIR AND PCV-1752 is closed, DRYERS CA-31 & CA-12 BYPASS THEN open CA-197, "BYPASS VALVE", is open (Room 19). CONTROL VALVE PCV-1752 BYPASS VALVE" (Room 19).

R14

AOP-17 Page 7 of 46 INSTRUCTIONS CONTINGENCY ACTIONS

8. IF ONE of the following valves is open (Room 19):

PCV-1752, "AIR DRYERS CA-31 &

CA-12 BYPASS VALVE" CA-197, "BYPASS CONTROL VALVE PCV-1752 BYPASS VALVE" AND Instrument Air pressure returns to a normal 98-108 psig, THEN restore Instrument Air by performing the following steps:

a. Place CA-31, BACKUP HEATLESS AIR DRYER, in service PER the Startup of Instrument Air Dryer CA-31 section of OI-CA-1, Compressed Air System Normal Operation.

b. GO TO Section 5.0, Exit Conditions.

R14

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 091 Given the following sequence of events:

The plant was operating at full power A group 4 CEA dropped into the core AOP-02, "CEA and Control System Malfunctions," was entered One minute later, another group 4 CEA dropped into the core What action should be taken as a result of these events?

A. Reduce power to less than 70% using AOP-05, "Emergency Shutdown."

B. Realign the dropped CEAs to within 12 inches of all other CEAs in group 4 or place the reactor within hot shutdown using OP-4, "Load Change and Normal Power Operations."

C. Trip the reactor and perform EOP-00, "Standard Post Trip Actions" then re-enter AOP-02, "CEA and Control System Malfunctions."

D. Trip the reactor and perform EOP-00, "Standard Post Trip Actions" then enter EOP-01, "Reactor Trip Recovery."

Wednesday, August 08, 2012 16:46:24 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 91 Rev 0 KA #: 014000 2.4.08 Tier 2 Group 2 ; Rod Position Indication System Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Importance 3.8 / 4.5 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Use the CEA and Control System Malfunctions Procedure to mitigate the consequences of a malfunction of a CEA, the CEA control system or CEA position indication.

Question Pedigree New question.

K/A Fit:

Question addresses transition from an AOP to an EOP during a dropped CEA event.

Choice A:

Distractor: Plausible because this is the action in AOP-02 for one dropped CEA.

Incorrect because dropping two CEA;s requires a reactor trip.

Choice B:

Distractor: Plausible because this is the action in AOP-02 after power is lowered to 70%.

Choice C:

Distractor: Plausible because AOP-02 directs a reactor trip for two dropped CEAs.

Incorrect because AOP-02 says "GO TO" EOP-00 and all trippable rods will be fully inserted after the trip.

Choice D:

Correct answer: AOP-02 directs the operators to trip the reactor and go to EOP-00 if more than one CEA is misaligned greater than 18 inches from any other CEA in its group.

KA#: 014000 2.4.08 Bank Ref #: N/A LP# / Objective: 0717-02 01.00 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

TBD-AOP-02 Handout: NONE Wednesday, August 08, 2012 16:46:24 2

TBD-AOP-02 Page 24 of 55 Section III - Misaligned Group 4 CEA 4.0 INSTRUCTIONS/CONTINGENCY ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS

1. IF ANY of the following conditions exist:

A dropped CEA occurs during a Reactor startup More than one CEA is misaligned greater than 18 inches from any other CEA in its group THEN initiate a Reactor shutdown by performing the following steps:

a. Trip the Reactor.

b. GO TO EOP-00, Standard Post Trip Actions.

The requirement to trip the Reactor if a CEA drops during startup complies with Operating Procedures. The requirement to trip the Reactor if more than one CEA is misaligned >18" is in compliance with PED-FC-89-2482 (R1).

2. Stop all CEA movement.

All CEA movement is stopped to preclude any further power peaking and CEA movement during any power reduction.

3. Adjust Turbine load to match Reactor power.

4. Establish steady Reactor power.

5. Ensure RCS pressure is 2075 to 2150 psia.

R8

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 092 The plant was operating at 50% power due to condensate pumps FW-2A and FW-2C being out of service when the following sequence of events occurred:

The running feedwater pump tripped and the standby feedwater pumps could not be started The reactor tripped on low S/G level Auxiliary feedwater pumps, FW-6, FW-10 and FW-54 could not be started EOP-00, Standard Post Trip Actions, were completed and EOP-06, Loss of All Feedwater, was entered S/G pressure was reduced to less than 550 psia and condensate pump, FW-2B, was used to feed the S/Gs Condensate pump, FW-2B, tripped and could not be restarted.

Assuming no further equipment failures occur, what actions should be taken as a result of these events?

A. Stay in EOP-06, "Loss of All Feedwater," and initiate once-through-cooling when required.

B. Enter AOP-28, "Auxiliary Feedwater System Malfunctions," and establish flow to the S/G's from the Blair Water System.

C. Enter EOP-20, "Functional Recovery Procedure," and establish flow to the S/G's using demineralized water.

D. Enter OCAG-1, "Operational Contingency Action Guidelines," and establish flow to the S/G's from the FCS fire truck.

Wednesday, August 08, 2012 16:47:00 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 92 Rev 0 KA #: 056000 A2.04 Tier 2 Group 2: Condensate System Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of condensate pumps Importance 2.6 / 2.8 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

DEMONSTRATE the knowledge required to use EOP-06, Loss of All Feedwater (LOAF), to mitigate the consequences of a loss of all feedwater.

Question Pedigree New question.

K/A Fit:

Question addresses using procedures to mitigate the loss of a condensate pump.

Choice A:

Correct answer: EOP-06 contains instruction for initiating once-through-cooling.

Choice B:

Distractor: Plausible because all AFW has been lost and Blair Water can be used to feed the emergency feedwater storage tank. Incorrect because AOP-28 does not address a loss of all feedwater and condensate.

Choice C:

Distractor: Plausible because EOP-20 does contain directions for feeding the S/G's using demineralized water. Incorrect because this would only be performed if Once-through-cooling failed.

Choice D:

Distractor: Plausible because OCAG-1 does establish flow to the S/G's using the fire truck. Incorrect because this event is mitigated by the EOPs.

KA#: 056000 A2.04 Bank Ref #: N/A LP# / Objective: 0718-16 01.00 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

EOP-06 Handout: NONE Wednesday, August 08, 2012 16:47:00 2

EOP-06 Page 17 of 41 INSTRUCTIONS CONTINGENCY ACTIONS CAUTIONS

1. Depressurization of S/Gs to establish a low pressure Feedwater source will result in cooldown of the RCS and the potential for SGLS actuation.

2. RCS pressure and inventory control will be affected and SGIS may occur if SGLS is not blocked as the depressurization proceeds.

3. When TC is 178°F or greater, the maximum RCS cooldown rate is 100°F/hr.

When TC is less than 178°F, the maximum RCS cooldown rate is 50°F/hr.

14. IF any Condensate Pumps are operating AND a flow path to at least one S/G is available, THEN perform the following steps to maintain S/G levels 35-85% NR (73-94% WR):

a. Place all of the Feed Pumps Control Switches, FW-4A/B/C, in "PULL-TO-LOCK".

(continue)

Continuously Applicable or Non-Sequential Step R17

EOP-06 Page 18 of 41 INSTRUCTIONS CONTINGENCY ACTIONS

14. (continued)

b. Locally open ALL of the following Feed Pump Discharge Valves (North wall, Turbine Building Basement):

• "DISCH VALVE HCV-1150A"

• "DISCH VALVE HCV-1150B"

• "DISCH VALVE HCV-1150C"

c. Ensure all of the Steam Generator Feed Pump Recirc valves, FCV-1151A/B/C are closed.

d. Start all of the Feed Pump Lube Oil Pumps, FW-30A/B/C.

(continue)

Continuously Applicable or Non-Sequential Step R17

EOP-06 Page 19 of 41 INSTRUCTIONS CONTINGENCY ACTIONS

14. (continued)

e. Reduce S/G pressure to less than 550 psia using ANY of the following methods:

• Steam Dump and Bypass Valves

• HCV-1040, Atmospheric Dump Valve

• MS-291 and MS-292, Air Assisted Main Steam Safety Valves

f. Maintain PZR level 10-70% by control of Charging and Letdown.

g. Maintain RCS pressure PER Attachment 2, RCS Pressure-Temperature Limits, by control of PZR Heaters and Spray.

h. Locally ensure FCV-1172, Condensate Pump Recirc Valve, is closed (Turbine Building Mezzanine).

(continue)

Continuously Applicable or Non-Sequential Step R17

EOP-06 Page 20 of 41 INSTRUCTIONS CONTINGENCY ACTIONS

14. (continued)

i. WHEN S/G pressure is less than 550 psia, THEN ensure SGLS is blocked by performing the following steps:

1) Block SGLS-A and SGLS-B.

2) Verify at least one of the 2).1 IF neither SGLS Blocked following SGLS Blocked alarm annunciates, alarms annunciates THEN continue attempts to (CB-4; A8): block SGLS until at least one alarm annunciates.

• "SGLS "A" BLOCKED"

• "SGLS "B" BLOCKED"

j. Ensure BOTH of the Feed Header Isolation Valves are open:

• HCV-1386

• HCV-1385

k. Ensure BOTH of the Feed Reg Block Valves are closed:

• HCV-1103

• HCV-1104 (continue)

Continuously Applicable or Non-Sequential Step R17

EOP-06 Page 21 of 41 INSTRUCTIONS CONTINGENCY ACTIONS

14. (continued)

l. Reduce secondary pressure to achieve feed flow via BOTH of the Feed Reg Bypass Valves:

• HCV-1105

• HCV-1106 Time:

m. Continue efforts to restore Main or Auxiliary Feedwater.

15. Verify adequate RCS heat removal via 15.1 IF ANY of the following criteria are the S/G(s) by BOTH of the following satisfied:

conditions:

• The least affected S/G wide range level less than 27%

• At least one S/G has wide range

• An uncontrolled rise in RCS TC of level greater than or equal to 27% greater than 5°F

• RCS TC temperatures are stable or lowering THEN perform step a or b:

a. IF BOTH Vital 4160V buses are energized, THEN perform the following:

(continue) 1) Stop all RCPs

2) Deenergize all PZR Heaters.

(continue)

Continuously Applicable or Non-Sequential Step R17

EOP-06 Page 22 of 41 INSTRUCTIONS CONTINGENCY ACTIONS

15. (continued) 15.1.a (continued)

3) Initiate PPLS by placing BOTH of the following switches in TEST:

• 86A/PPLS TEST SWITCH

• 86B/PPLS TEST SWITCH

4) Ensure two HPSI pumps SI-2A/B or SI-2B/C start.

(continue) 5) Ensure all of the HPSI Loop Injection Valves are open.

6) Verify all Charging Pumps CH-1A/B/C start.

7) Ensure BOTH of the PORV Block Valves are open:

• HCV-150

• HCV-151

8) Open the PORVs.

Time:

(continue)

Continuously Applicable or Non-Sequential Step R17

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 093 Given the following plant conditions:

A LOCA occurred inside containment EOP-03, Loss of Coolant Accident, was entered after completion of EOP-00 standard post trip actions All Reactor Coolant Pumps have been tripped All charging pumps, HPSI Pumps SI-2A and SI-2B and LPSI Pumps SI-1A and SI-1B are all operating Reactor Vessel Level is 0%

SI flow is less than Attachment 3, Safety Injection Flow vs. Pressurizer Pressure The countrates on all of the source range NI's are rising WR NIs indicate 4 x 10-7% power A RCS cooldown is in progress RCS Pressure is 600 psia CETs indicate 510°F RAS has NOT occurred What action should be taken in response to these conditions?

A. Enter OI-RC-12, "Post Accident Venting of Non-Condensable Gases from the Reactor Coolant System," and vent RV head to Containment.

B. Enter OI-RC-12, "Post Accident Venting of Non-Condensable Gases from the Reactor Coolant System," and vent RV head to the Pressurizer Quench Tank C. Enter EOP-20, RC-1,"Reactivity Control," and insert the group "N" CEAs D. Enter EOP-20, IC-2, "RCS Inventory Control" and start HPSI Pump, SI-2C Wednesday, August 08, 2012 12:48:07 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 93 Rev 0 KA #: 015000 A2.05 Tier 2 Group 2: Nuclear Instrumentation System Ability to (a) predict the impacts of the following malfunctions or operations on the NIS; and (b)based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Core void formation Importance 3.3 / 3.8 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Explain the effects on the WR NIS with voiding in the core.

Question Pedigree New question.

K/A Fit:

Question address procedural action to mitigate a core uncovery event that is affecting nuclear instrumentation.

Choice A:

Distractor:Plausible because core uncovery has occured and non-condensible gases have most likely formed. Incorrect because EOPs direct starting another HPSI pump.

Choice B:

Distractor:Plausible because core uncovery has occured and non-condensible gases have most likely formed. Incorrect because EOPs direct starting another HPSI pump.

Choice C:

Distractor: Plausible because source range NI's show rising counts. Incorrect because WR NI's are less that 10-5 % power.

Choice D:

Correct answer: EOP-20, IC-2 is entered because the Inventory Control Safety Function is not met with reactor vessel level at 0%. EOP-20 directs starting all idle pumps if SI flow is not maximized. SI flow is less than required by attachment 3.

KA#: 015000 A2.05 Bank Ref #: N/A LP# / Objective: 0712-18 02.10 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

EOP-03 Handout: NONE Wednesday, August 08, 2012 12:48:08 2

EOP-20 Page 205 of 527 14.0 RCS INVENTORY CONTROL SAFETY FUNCTION: RCS Inventory Control SUCCESS PATH: Safety Injection: IC-2 RESOURCE TREE: Tree C IC-2 INSTRUCTIONS CONTINGENCY ACTIONS CAUTIONS

1. LOCAs inside Containment can greatly raise instrument inaccuracies. A single indication may be unreliable. All available indications and actual plant response should be used to evaluate actions and plant conditions. TSC staff should assist in evaluating these uncertainties through use of the Electrical Equipment Qualification Manual.

2. Do not allow Diesel Generator loads to exceed power and current rating limits.

1. IF RCS pressure is less than or equal to 1600 psia, THEN verify Engineered Safeguards are actuated by performing the following:

a. Ensure Emergency Boration is in progress.

b. Ensure SI flow is acceptable PER Attachment 3, Safety Injection Flow vs. Pressurizer Pressure.

Continuously Applicable or Non-Sequential Step R25

EOP-20 Page 209 of 527 14.0 RCS INVENTORY CONTROL IC-2 INSTRUCTIONS CONTINGENCY ACTIONS CAUTION Do not allow Diesel Generator loads to exceed power and current rating limits.

3. IF SIAS has actuated, 3.1 IF Safety Injection and Charging flow is THEN maximize Safety Injection flow NOT maximized, and Charging flow to the RCS by THEN restore Safety Injection and operating ALL of the following Charging flow by performing the available pumps: following:

Either HPSI Pumps, SI-2A/B or a. Restore electrical power to valves SI-2B/C LPSI Pumps, SI-1A/B and pumps.

Charging Pumps, CH-1A/B/C

b. Ensure ALL o. the following SI valves are open to align SI for injection:

HPSI Loop Injection Valves LPSI Loop Injection Valves HPSI Discharge Cross-Connect Valves HPSI Header Isolation Valves SI Pump Suction SIRWT (continue) Isolation Valves (continue)

Continuously Applicable or Non-Sequential Step R25

EOP-20 Page 210 of 527 14.0 RCS INVENTORY CONTROL IC-2 INSTRUCTIONS CONTINGENCY ACTIONS

3. (continued) 3.1 (continued)

c. Start ALL of the following idle pumps:

HPSI Pumps, SI-2A/B/C LPSI Pumps, SI-1A/B Charging Pumps, CH-1A/B/C

d. Verify SI flow is acceptable PER Attachment 3, Safety Injection Flow vs. Pressurizer Pressure.

4. IF high RCS pressure is preventing adequate SI flow, THEN depressurize the RCS to less than 1200 psia by performing ANY of the following:

a. Control RCS heat removal.

b. Control PZR Heaters.

c. Control PZR Main or Auxiliary Spray.

d. Operate the PORVs.

Continuously Applicable or Non-Sequential Step R25

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 094 With the plant in hot shutdown, the Shift Technical Advisor slipped on a stairway and injured his leg 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into the shift. An EMT was called and the STA was transported to the Blair Hospital. The CRS is also STA qualified.

What action should be taken due to this event?

A. No action is required.

B. Have a qualified STA on-site within two hours.

C. Have a qualified STA on-site within four hours.

D. Have the CRS fill the STA position for the remainder of the shift.

Thursday, August 09, 2012 15:28:30 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 94 Rev 0 KA #: 000000 2.1.05 Tier 3 Group 4: Generic Knowledges and Abilities Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc..

Importance 2.9 / 3.9 CFR Number: 55.43(b)(1)

Conditions and limitations in the facility license.

Fort Calhoun Objective:

Operations and STA staffing requirements (SRO only)

Question Pedigree Bank question used on the 1997 exam with slight rewording. Not counted as a modified question.

K/A Fit:

Question addresses shift staffing requirements per Technical Specifications.

Choice A:

Distractor:Plausible because reactor is shutdown. Incorrect because a STA is required in hot shutdown.

Choice B:

Correct answer: A STA is required in hot shutdown and there is a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time period.

Choice C:

Distractor: Plausible because a STA is required. Incorrect because the time period is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Choice D:

Distractor: Plausible because CRS is STA qualified. Incorrect because CRS can not fill two positions.

KA#: 000000 2.1.05 Bank Ref #: ADM-OPS 008 LP# / Objective: 0762-08 10.01 Exam Level: SRO-1 Cognitive Level: LOW Source: BANK

Reference:

TS TABLE 5.2-1 Handout: NONE Thursday, August 09, 2012 15:28:30 2

TECHNICAL SPECIFICATIONS TABLE 5.2-1 MINIMUM SHIFT CREW COMPOSITION(ii)

Operating License Core Cold Shutdown or or Hot Category Alteration Refueling Shutdown Shutdown Modes Senior Operator License 2(i) 1 2(iii)

Operator License 2 1 2(iv)

Non-Licensed (As required) 1 2 Shift Technical None None 1 Advisor (i) This includes the individual with Senior Operator License supervising Core Alterations.

(ii) Shift crew composition may be one less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 5.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewmember being late or absent.

(iii) At least one of these individuals must be in the control room at all times.

(iv) At least one of these individuals (or the second senior licensed operator, if both senior licensed operators are in the control room) must be present at the controls at all times.

5.0 - Page 3 Amendment No. 9,24,54,78, 202

Source question for question 94 QUESTIONS REPORT for ILO EXAM BANK ADM-OPS 008 With the plant in hot shutdown, the Shift Technical Advisor slipped on a stairway and injured his ankle. An off-duty Shift Manager leaving the site has taken him to the emergency room. Which of the following actions should be taken due to this event?

A. Make a four hour report to the NRC.

B. Have a System Engineer who is training to be an STA fill the position.

C. Have a qualified STA on-site within two hours.

D. Declare a Notification of Unusual Event based on transport of an injured person.

Wednesday, August 08, 2012 12:53:20 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 095 The performance of SE-PM-AE-1008, "Inspection of Station Blackout (SBO)

Restoration Structural Components," is being planned. Access to the switchyard will be required. What is the Shift Manager's responsibility for pre-planning this work?

A. Inform the Security Shift Supervisor to provide an escort for the personnel performing the inspection while they are in the switchyard per FCSG-32, "Work Week Management."

B. Determine if proposed work can be accomplished without adversely affecting plant operation per FCSG-32, "Work Week Management."

C. Inform the Security Shift Supervisor to provide an escort for the personnel performing the inspection while they are in the switchyard per NOD-QP-36, "Control of Switchyard Activities at Fort Calhoun Station"?

D. Determine if proposed work can be accomplished without adversely affecting plant operation per NOD-QP-36, "Control of Switchyard Activities at Fort Calhoun Station"?

Wednesday, August 08, 2012 13:17:32 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 95 Rev 0 KA #: 000000 2.1.13 Tier 3 Group 4 Generic Knowledges and Abilities Knowledge of facility requirements for controlling vital / controlled access.

Importance 2.5 / 3.2 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective Given a copy of the procedure NOD-QP-36, Control of SWYD Activities at FCS, the student will be able to perform the following: Describe the responsibilities associated with NOD-QP-36.

Question Pedigree New question.

K/A Fit:

Question addresses SRO responsibility for access to the switchyard.

Choice A:

Distractor: Plausible if Applicant believes security must supply escort. Incorrect because the Shift Manager does not inform the Security Shift Supervisor and switchyard access is not controlled by FCSG-32.

Choice B:

Distractor: Plausible because this the Shift Managers responsibility. Incorrect because switchyard access in not controlled by FCSG-32.

Choice C:

Distractor: Plausible because NOD-QP-36 is used to control switchyard access.

Incorrect because the Shift Manager does not inform the Security Shift Supervisor.

Choice D:

Correct answer: The Shift Manager determines if proposed work can be accomplished without adversely affecting plant operation per NOD-QP-36, "Control of Switchyard Activities at Fort Calhoun Station" KA#: 000000 2.1.13 Bank Ref #: ADM-CONTROL 017 LP# / Objective: 0713-01 01.06C Exam Level: SRO-5 Cognitive Level: LOW Source: NEW

Reference:

NOD-QP-36 Handout: NONE Wednesday, August 08, 2012 13:17:32 2

NOD-QP-36 Information Use Page 9 of 57 Grid Operations and Control of Switchyard at FCS Rev. 21 4.3.1 (continued)

F. The Shift Manager will:

1. Determine if the activity can be accomplished without adversely impacting plant operation.

2. Ensure conditions have not changed since initial draft of the activity.

3. Review and give final approval to the substation activity.

G. The Transmission Operator will notify the Shift Manager just prior to commencing the stated activity to obtain final approval.

4.3.2 Emergent activities affecting the availability of FCS substations circuits A. The T&D System Scheduler or Transmission Operator will notify the Shift Manager at least one day in advance, if possible, via OPPD normal telephone lines (preferred) or the emergency radio prior to performing any manipulations affecting the FCS Switchyard.

B. The Shift Manager will contact the Systems Analysis Department, System Engineer and the Work Week Manager for assistance in evaluating and resolving any schedule or plant risk conflicts if the grid is in a degraded condition. Reference Standing Order SO-M-100 for additional risk assessment information.

C. The Transmission Operator will notify the Shift Manager just prior to commencing the stated activity to obtain final approval.

4.3.3 Emergency Activities affecting the availability of FCS Switchyard circuits A. The Transmission Operator will notify the Shift Manager, as soon as practical, when unanticipated conditions require any manipulation of circuits affecting FCS Switchyard circuits.

B. The Shift Manager will evaluate out-of-service safety related equipment and determine the need to expedite the return of equipment that may have an impact on the plant risk assessment. The Shift Manager will continue to monitor conditions and will direct the necessary actions to maintain FCS in a safe condition. Transmission Operator will provide high priority support in the event of a failure that results in a loss of operability of FCS house service (T1A-3 and T1A-4) or auxiliary transformers (T1A-1 and T1A-2) or the loss of transmission line supplies (Circuits 1587 and 3423) to these transformers. If these are not energized within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, a plant shutdown is required.

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 096 The performance of RE-ST-RX-0003, "Moderator Temperature Coefficient Determination Using Center CEA," requires the department manager or above to attend the pre-job briefing.

Where is that requirement found?

A. SO-G-3, "Special Procedures" B. SO-G-23, Surveillance Test Program" C. SO-G-92, "Conduct of Infrequently Performed Procedures" D. SO-G-117, "Reactivity Management" Wednesday, August 08, 2012 13:28:01 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 96 Rev 0 KA #: 000000 2.2.07 Tier 3 Group 4:Generic Knowledges and Abilities Knowledge of the process for conducting special or infrequent tests.

Importance 2.9 / 3.6 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

STATE some of the activities, covered by Standing Orders, which require written procedures per Regulatory Guide 1.33.

Question Pedigree New question.

K/A Fit:

Question addresses the process for conducting an infrequent test.

Choice A:

Distractor: Plausible if the Applicant believes this is covered by SO-G-3. Incorrect because the requirement is found in SO-G-92.

Choice B:

Distractor: Plausible if the Applicant believes this is covered by SO-G-23. Incorrect because the requirement is found in SO-G-92.

Choice C:

Correct answer. This requirement is found in SO-G-92.

Choice D:

Distractor: Plausible if the Applicant believes this is covered by SO-G-117. Incorrect because the requirement is found in SO-G-92.

KA#: 000000 2.2.07 Bank Ref #: N/A LP# / Objective: 0762-01 02.00 Exam Level: SRO-5 Cognitive Level: LOW Source: NEW

Reference:

SO-G-92 Handout: NONE Wednesday, August 08, 2012 13:28:01 2

FORT CALHOUN STATION SO-G-92 STANDING ORDER INFORMATION USE PAGE 9 OF 11 Attachment 1 - Infrequently Performed Procedures NUMBER DESCRIPTION MD-RR-MX-1002 Establishing and Maintaining Freeze Seals by Vendor Personnel MD-RR-MX-1003 Establishing and Maintaining Freeze Seals on Stainless Steel Piping MD-RR-MX-1004 Establishing and Maintaining Freeze Seals on Carbon Steel Piping MD-RR-MX-1005 Establishing and Maintaining Freeze Seals Using Portable Refrigerant Units MM-RR-CEDM-1201 CEDM Coupling (Note 1)

OI-CC-1 CCW System Normal Operations; any operation or maintenance activity that directs removal of CCW system overpressure OI-RC-2A RCS Fill and Drain Operations (Lowered/Reduced Inventory Operations Only) (Note 1)

OI-SFP-7 Temporary Spent Fuel Pool Cooling via Chillers (Note 1)

OP-2A Plant Startup (Approach to Critical Operation Only) (Note 1)

OP-ST-ESF-0002 Diesel Generator No. 1 and No. 2 Auto Operation (Note 1)

OP-ST-RC-3004 Power Operated Relief Valves (PORVs) Low Temperature Low Pressure Exercise Test (PCV-102-1 and PCV-102-2) (Note 1)

PE-HT-MX-1000 Hydrostatic Testing of Safety-Related Plant System and Components PE-OT-SFP-0001 Temporary Spent Fuel Cooling System Installation/Removal (Note 1)

RE-CPT-RX-0001 Post Refueling Core Physics Testing and Power Ascension (Note 1)

RE-RR-DFS-0001 DSC/TC Prep for Fuel Loading Operations RE-RR-FE-0001 Reconstitution of Irradiated Fuel Assemblies (Note 1)

RE-RR-FE-0002 Modification of Fuel Assemblies (Note 1)

RE-ST-RX-0003 Moderator Temperature Coefficient Determination Using Center CEA (Note 1)

RP-208 Radiography RP-212 Diving Operations Within Radiologically Controlled Areas SE-ST-ILRT-0001 Containment Integrated Leak Rate Test (ILRT)

Note 1: Department Manager or above must attend activity pre-job brief [SOER 10-2].

R16

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 097 With the reactor operating at full power, an engineering evaluation determined that several pipe support snubbers in both trains of the emergency core cooling system are unable to perform their associated support function.

What action is required by Technical Specifications?

A. Place the plant in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

B. Restore the snubbers to operable status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or place the plant in hot shutdown within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

C. Restore the snubbers to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or place the plant in hot shutdown within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. Restore the snubbers to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or place the plant in hot shutdown within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Wednesday, August 08, 2012 13:39:50 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 97 Rev 0 KA #: 000000 2.2.22 Tier 3 Group 4: Generic Knowledges and Abilities Knowledge of limiting conditions for operations and safety limits.

Importance 4.0 / 4.7 CFR Number: 55.43(b)(2)

Facility operating limitations in the technical specifications and their bases.

Fort Calhoun Objective:

Given a copy of Technical Specifications, APPLY the requirements to a given condition covered by an LCO.

Question Pedigree New question.

K/A Fit:

Question address LCO that applies to numerous plant systems.

Choice A:

Distractor: Plausible if Applicant does not realize Tech Specs allow 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to restore snubbers to operable. Incorrect because we have 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Choice B:

Correct answer: Tech Spec 2.0.1 requires that we restore the snubbers to operable status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or place the plant in hot shutdown within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Choice C:

Distractor: Plausible if Applicant does not know the time limit. Incorrect because we have 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Choice D:

Distractor: Plausible because this would be correct if only one train was affected.

Incorrect because snubbers in both trains are affected.

KA#: 000000 2.2.22 Bank Ref #: N/A LP# / Objective: 0762-08 05.00 Exam Level: SRO-2 Cognitive Level: LOW Source: NEW

Reference:

TS 2.0.1 Handout: NONE Wednesday, August 08, 2012 13:39:50 2

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.0.1 General Requirements Applicability Applies to the operable status of all systems, subsystems, trains, components, or devices covered by the Limiting Conditions for Operation.

Objective To specify corrective measures to be employed for system conditions not covered by or in excess of the Limiting Conditions for Operation.

Specification (1) In the event a Limiting Condition for Operation and/or associated action requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, in at least subcritical and < 300°F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, unless corrective measures are completed that permit operation under the permissible action requirements for the specified time interval as measured from initial discovery or until the reactor is placed in an Operating Mode in which the specification is not applicable. Exceptions to these requirements shall be stated in the individual specifications.

(2) When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and:

a. the snubbers not able to perform their associated support function(s) are associated with only one train or subsystem of a multiple train or subsystem supported system or are associated with a single train or subsystem supported system and are able to perform their associated support function within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; or

b. the snubbers not able to perform their associated support function(s) are associated with more than one train or subsystem of a multiple train or subsystem supported system and are able to perform their associated support function within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

At the end of the specified period the required snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not met.

2.0 - Page 1 Amendment No. 52, 238, 264

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 098 Given the following conditions:

Spent Fuel is being moved from the Spent Fuel Pool to the Independent Spent Fuel Storage Installation (ISFSI)

A Transfer Cask/Dry Shielded Canister (TC/DSC) is being removed from the Spent Fuel Pool and transferred to the work platform.

Anticipated radiation levels are 1200 mrem/hr at 30 cm from the radiation source How will the Spent Fuel Pool Area be posted and access controlled during this operation?

A. The area will be posted as a Radiation Area; available gates and doors to the area will be locked.

B. The area will be posted as a Radiation Area; available gates and doors to the area will not be locked.

C. The area will be posted as a Restricted High Radiation Area; available gates and doors to the area will be locked.

D. The area will be posted as a Restricted High Radiation Area; available gates and doors to the area will not be locked.

Wednesday, August 08, 2012 13:58:21 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 98 Rev 0 KA #: 000000 2.3.12 Tier 3 Group 4: Generic Knowledges and Abilities Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Importance 3.2 / 3.7 CFR Number: 55.43(b)(4)

Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Fort Calhoun Objective:

DEFINE a restricted high radiation area and LIST the posting requirements.

Question Pedigree Bank question used on the 2009 NRC exam.

K/A Fit:

Question addresses access to a restricted high radiation area during spent fuel transfer.

Choice A:

Distractor: Plausible if area was posted as a Radiation Area, Incorrect because it is posted as a Restricted High Radiation Area.

Choice B:

Distractor: Plausible if area was posted as a Radiation Area, Incorrect because it is posted as a Restricted High Radiation Area.

Choice C:

Correct answer: RP-222 states that the area will be posted as a RHRA and doors and gates will be locked.

Choice D:

Distractor: Plausible because it is posted as a Restricted High Radiation Area. Incorrect because access controls to a RHRA include locking doors and gates.

KA#: 000000 2.3.12 Bank Ref #: 19-24-03 003 LP# / Objective: 1924-03 01.06 Exam Level: SRO-4 Cognitive Level: HIGH Source: BANK

Reference:

RP-222 Handout: NONE Wednesday, August 08, 2012 13:58:21 2

FORT CALHOUN STATION RP-222 RADIATION PROTECTION REFERENCE USE PAGE 7 OF 13 7.15 Area postings and controls:

7.15.1 If conditions warrant access shall be controlled as a RHRA when the TC/DSC is being removed from the spent fuel pool and transferred to the work platform or moving from work platform to transport trailer in railroad siding by using combination of one or more of the following:

7.15.1.A Lock or guard and post the gate from room 69 to the SFP.

7.15.1.B Lock or guard and post the doors in Rooms 25A, 67 and 68.

7.15.1.C Station RHRA guards in the BAST Area and at the Stairs to Room 69.

7.15.1.D Suspend all work in room 69 and the BAST area to RRS hallway.

7.15.2 After the TC/DSC is placed in room 68, RP will re-survey the area and update postings accordingly.

7.15.3 After the following steps have been completed:

The DSC is inside room 68 Up to 750 gallons Water Drained from DSC TC/DSC annulus full (within approximately 1 foot of top)

TC neutron shield full Top Shield Plug in place and included in axial shielding Inner Top Cover Plate in place and included in axial shielding Automated Welding System (AWS) with integral shield in place and included in axial shielding Obtain a gamma and neutron dose rate at the vertical centerline of DSC, 3 feet and 30 cm from AWS shield.

Ensure the sum of the dose rates obtained does not exceed 200 mrem/hr at 3 feet.

7.15.4 DSC Removable Surface Contamination Limits 7.15.4.A.1) 2,200 dpm/100 cm2 for beta-gamma sources 7.15.4.A.2) 220 dpm/100 cm2 for alpha sources R3

RP-204 Information Use Page 4 of 24 Radiological Area Controls Revision 61 2.12 Posting - A conspicuous sign bearing a magenta or black radiation symbol on a yellow background and the words "CAUTION", or "DANGER" or "GRAVE DANGER" as described in 10CFR20, which is used to post areas or rooms due to radiation or radioactive materials. Posting shall identify the presence of radiation or radioactive materials. Extra information which may aid individuals in minimizing their exposure to radiation or radioactive materials may also be provided.

2.13 Radiation Area - An Area, accessible to individuals, in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 0.005 rem (0.05 mSv) in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates.

2.14 Radiologically Controlled Area - An area posted and controlled for the purpose of protecting personnel from radiation or radioactive materials. Access to any RCA requires a Radiation Work Permit.

2.15 Restricted Area - An area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. At FCS this area is normally established by the security fence and gate house.

2.16 Restricted High Radiation Area - An area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving a deep dose equivalent in excess of 1 rem (10 mSv) in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates.

2.17 Step-Off Pad (SOP) - A delineated area on a surface outside the boundary of a Contaminated or Highly Contaminated Area, which establishes access/egress points for such areas.

2.18 Temporarily Accessible - For the purpose of identifying, demarcating and surveying, temporarily accessible is defined to be any location that is greater than eight feet off the floor and is in such a location that an individual can, access within 30 centimeters, for less than seven calendar days (via a ladder, scaffolding or climbing on pipes).

2.19 Very High Radiation Area - An area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving an absorbed deep dose in excess of 500 rads (5 grays) in one hour at 1 meter from a radiation source or from any surface that the radiation penetrates.

2.20 Whole Body - For purposes of external exposure, head, trunk (including male gonads),

arms above elbow, or legs above the knee.

3.0 RESPONSIBILITIES 3.1 The Manager-Radiation Protection is responsible for:

3.1.1 Enforcing compliance with radiological postings by all personnel at FCS.

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 099 A large steam line break inside containment occurred on steam generator RC-2A coincident with a 50 gpm tube rupture on steam generator RC-2B.

What action should be taken after EOP-00, "Standard Post Trip Actions have been completed?

A. Enter EOP-04, "Steam Generator Tube Rupture," and isolate S/G RC-2B B. Enter EOP-05, "Uncontrolled Heat Extraction," and isolate S/G RC-2A C. Enter EOP-20, "Functional Recovery Procedure," and isolate S/G RC-2A D. Enter EOP-20, "Functional Recovery Procedure," and isolate S/G RC-2B Wednesday, August 08, 2012 14:06:40 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 99 Rev 0 KA #: 000000 2.4.06 Tier 3 Group 4: Generic Knowledges and Abilities Knowledge EOP mitigation strategies.

Importance 3.7 / 4.7 CFR Number: 55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

EXPLAIN how an Excessive Heat Removal Event in conjunction with another event would be handled by the operators.

Question Pedigree New question.

K/A Fit:

Question addresses EOP mitigation strategies for multiple simultaneous events.

Choice A:

Distractor: Plausible because a tube rupture has occurred. Incorrect because EOP-20 is used to mitigate simultaneous events.

Choice B:

Distractor: Plausible because an uncontrolled heat extraction has occurred. Incorrect because EOP-20 is used to mitigate simultaneous events.

Choice C:

Correct answer: EOP-20 is entered for multiple events. The most affected steam generator is isolated.

Choice D:

Distractor: Plausible because EOP-20 is entered for multiple events. Incorrect because RC-2A, the most affected S/G, should be isolated.

KA#: 000000 2.4.06 Bank Ref #: 07-15-20 022 LP# / Objective: 0715-20 02.07 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

EOP-20 Handout: NONE Wednesday, August 08, 2012 14:06:40 2

EOP-20 Page 322 of 527 16.0 RCS AND CORE HEAT REMOVAL HR-3 INSTRUCTIONS CONTINGENCY ACTIONS

10. Determine the S/G with the Tube Rupture by performing the following:

a. Review the S/G analysis for activity.

b. Monitor Steam Line radiation levels.

c. Monitor S/G Blowdown radiation levels.

d. Monitor S/G levels.

11. WHEN RCS TH is less than or equal to 510 F, THEN isolate the most affected S/G by performing the following:

a. IF RC-2A is most affected, a.1 IF RC-2B is most affected, THEN isolate RC-2A by performing THEN isolate RC-2B by performing the following: the following:

(continue) (continue)

Continuously Applicable or Non-Sequential Step R25

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 QUESTION NUMBER: 100 A General Emergency has been declared at Fort Calhoun Station. The TSC and EOF are both fully staffed and operational. The Shift Manager is in the Command and Control position at this time.

Which of the following duties can the Shift Manager delegate to the CRS?

A. Ensuring that the event is classified properly.

B. Ensuring appropriate Protective Action Recommendations are provided to offsite officials.

C. Authorizing the issuance of Potassium Iodide to OPPD emergency workers.

D. Authorizing deviations from Technical Specifications needed to mitigate the event.

Wednesday, August 08, 2012 16:48:11 1

QUESTIONS REPORT for 2012-2 FCS NRC WRITTEN EXAM Rev 0 Question # 100 Rev 0 KA #: 000000 2.4.40 Tier 3 Group 4: Generic Knowledges and Abilities Knowledge of SRO responsibilities in emergency plan implementation.

Importance 2.7 / 4.5 CFR Number: 55.43(b)(1)

Conditions and limitations in the facility license.

Fort Calhoun Objective IDENTIFY the ERO positions that can have overall Command and Control.

Question Pedigree Bank question used on the 2005 NRC exam.

K/A Fit:

Question addresses SRO responsibilities during an emergency plan event.

Choice A:

Distractor: Plausible if Applicant does not know the Command and Control Responsibilities that can not be delegated. Incorrect because this responsibility can not be delegated.

Choice B:

Distractor: Plausible if Applicant does not know the Command and Control Responsibilities that can not be delegated. Incorrect because this responsibility can not be delegated.

Choice C:

Distractor: Plausible if Applicant does not know the Command and Control Responsibilities that can not be delegated. Incorrect because this responsibility can not be delegated.

Choice D:

Correct answer: This is not one of the responsibilities that can not be delegated.

KA#: 000000 2.4.40 Bank Ref #: ADM-EP LP# / Objective: 1070-101 01.13 Exam Level: SRO-1 Cognitive Level: LOW Source: BANK

Reference:

EPIP-OSC-2 Handout: NONE Wednesday, August 08, 2012 16:48:11 2

EPIP-OSC-2 Reference Use Page 32 of 36 Command and Control Position Revision 56 Actions/Notifications Attachment 6.8 - Command and Control Position Responsibilities The following responsibilities CAN NOT BE DELEGATED by the Command and Control position. The responsibility of their completion rests with the Command and Control position until relieved by another qualified individual or the emergency is terminated. The Command and Control position may assign other personnel to assist in conducting the actions necessary.

1. Overall COMMAND AND CONTROL of the Emergency Response Organization.

2. Ensuring the proper CLASSIFICATION AND DECLARATION of the emergency situation is made in accordance with EPIP-OSC-1 and is periodically reviewed to determine if the classification should be upgraded, downgraded or terminated.

3. Ensuring all required NOTIFICATIONS are made to appropriate state, local and federal officials.

4. Ensuring any appropriate PROTECTIVE ACTION RECOMMENDATIONS (PARs) are provided to offsite officials.

5. Authorizing OPPD emergency worker exposure extensions beyond the Federal Radiation Protection Guidance.

6. Authorizing issuance of Potassium Iodide for OPPD emergency workers.

THE COMMAND AND CONTROL POSITION ALSO HAS THE FOLLOWING RESPONSIBILITIES WHICH MAY BE DELEGATED TO OTHER PERSONNEL, AS NECESSARY.

7. Request for assistance from federal agencies.

8. Authorizing any emergency information to be released to the media or the general public.

9. Coordinating the transfer of emergency information from the Emergency Response Organization (ERO) to other OPPD and outside organizations called upon to assist.

10. Ensuring a timely and complete turnover of information to any qualified relief.

11. Providing information to authorized representatives of the states of Nebraska, and Iowa, and associated local governments.

12. Ensuring plant operations are in compliance with Technical Specifications and other license conditions. If deviations are necessary to protect the public health and safety, they must be evaluated with respect to 10CFR50.54(x) and (y) and approved, as a minimum, by a senior licensed operator, prior to taking the action.