ML111730060

2011-06 Final Written Exam

ML111730060
Person / Time
Site:	Cooper
Issue date:	06/13/2011
From:	Kelly Clayton Operations Branch IV
To:
laura hurley
References
50-298/11-301
Download: ML111730060 (440)

Text

ES-401 Site-Specific RO Written Examination Form ES-401-7 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Name:

Date: 6-13-2011 Facility/Unit: Cooper Nuclear Station Region: I II III IV Reactor Type: W CE BW GE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination begins.

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

Applicants Signature Results Examination Value __________ Points Applicants Score __________ Points Applicants Grade __________ Percent

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 263000.K1.03 Importance Rating 2.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-07-02 Section X of the USAR 10.3.5.3 Battery Rooms Exhaust Fans (Attach if not previously provided)

(including version/revision number) 29 10-25-00 Learning Objective: See Attached (As available)

Question Source: Bank # 15139 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y 1 15139 02 02/28/2011 None SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 3 Multiple Choice Topic Area Description Electrical COR0020702, DC Electrical Distribution Related Lessons COR0020702 OPS DC ELECTRICAL DISTRIBUTION Related Objectives COR0020702001110B Predict the consequences of the following events on the DC Electrical Distribution System: Loss of Battery ventilation COR002070200 I 0200 Given conditions and/or parameters associated with the DC Electrical Distribution System, determine if related Technical Specification and Technical Requirements Manual Limiting Condition for Operation are met Related References COR0020702 USAR X 10 Related Skills (K/A) 263000.K1.03 Knowledge of the physical connections and/or cause/effect relationships between D.C. ELECTRICAL DISTRIBUTION and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Battery ventilation (2.6 /

2.8)

QUESTION: 1 15139 (l point(s))

What is the main concern in the Battery Rooms as stated in the USAR during a failure of the Battery Room ventilation exhaust fans?

a. Temperatures that will result in premature battery charger failures.

b. Temperatures that will cause rapid vaporization of battery electrolyte.

c. Humidity that will result in excessive battery post corrosion and reduced current flow.

d. Temperatures and hydrogen concentration levels that will make safe operation of equipment uncertain.

ANSWER: 1 15139

d. Temperatures and hydrogen concentration levels that will make safe operation of equipment uncertain.

Explanation:

On a loss of Battery Room Ventilation removes the capability for the removal of H2 from the rooms. From the Student text - 1. The battery room exhaust fans remove the hydrogen produced by the 250 VDC and the 125 VDC batteries. 3. A loss of Battery Room Ventilation could result in excessive hydrogen levels in the room. This is a special concern when batteries are being charged, since the charging process produces hydrogen as a byproduct.

Section X of the USAR 10.3.5.3 Battery Rooms Exhaust Fans The battery rooms air is both exhausted directly to the atmosphere and recirculated to the Control Building Ventilation System. The exhaust to atmosphere is controlled by two manual exhaust fans (one operating and one standby). During periods of low battery room temperature, ventilation flow is reduced by motor-operated dampers which isolate a portion of the battery room ventilation supply air. The dampers reopen when high battery room temperature conditions are present.

The battery room exhaust system ductwork insures constant air movement through potential pocket areas, so that explosive concentrations of hydrogen cannot accumulate.

Since the exhaust system ductwork is common to battery rooms 1A and 1B, fire dampers have been installed in the exhaust ducting for each battery room. These fire dampers close on high temperature in the exhaust duct to preclude the spread of fire from one room to the other.

A standby fan is provided to automatically start upon failure of the fan in the operating mode, eliminating possible accumulation of an explosive mixture of hydrogen.

All battery room exhaust fans are automatically tripped when the essential Control Building ventilating system is in operation. Under this condition the essential Control Building ventilating system provides adequate air movement to eliminate the accumulation of an explosive mixture of hydrogen.

The battery rooms ventilation ducting that exhausts to atmosphere is welded of acid resistant material. However, where exhaust ventilation passes through the Main Control Room and Cable Spreading Rooms, the ducting consists of heavy wall, explosion-proof piping. The piping exits the Main Control Room to the Control Corridor, where the battery room exhaust fans exhaust to atmosphere, via ductwork, through the roof of the corridor.[30]

Distracters:

a. Temperatures that will result in premature battery charger failures. Although this would be a concern for the chargers the main concern according to the Student text is the accumulation of Hydrogen in the battery rooms that might get to an explosive level.

b. Temperatures that will cause rapid vaporization of battery electrolyte. The elevated temperatures will cause some electrolyte to evaporate, main concern according to the Student text is the accumulation of Hydrogen in the battery rooms that might get to an explosive level.

c. Humidity that will result in excessive battery post corrosion and reduced current flow. A concern, however the main concern according to the Student text is the accumulation of Hydrogen in the battery rooms that might get to an explosive level.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 261000.K1.12 Importance Rating 3.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-28-02 (Attach if not previously provided)

(including version/revision number) 19 Learning Objective: See Attached (As available)

Question Source: Bank # 1347 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y 2 1347 01 04/30/2007 None SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 3 Multiple Choice Topic Area Description Systems COR002280200 1050A, COR002280200 1050B Standby Gas System Related Lessons COR0022802 OPS STANDBY GAS TREATMENT Related Objectives COR002280200 1050A Describe the interrelationships between SGT and the following:

Reactor Building Ventilation System COR002280200 1050B Describe the interrelationships between SGT and the following:

Primary Containment Related References COR0022802 Related Skills (K/A) 261000.K1.12 Knowledge of the physical connections and/or cause/effect relationships between STANDBY GAS TREATMENT SYSTEM and the following: (CFR:

41.2 to 41.9 / 45.7 to 45.8) Primary containment purge system:

Plant-Specific (3.1 / 3.2)

QUESTION: 2 Where does the Standby Gas Treatment System line up to take a suction on an automatic initiation due to a refueling accident when the plant is in MODE 5?

a. Reactor building exhaust plenum and the SGTS room air.

b. HPCI gland steam condenser exhauster and the Primary Containment.

c. Reactor building exhaust plenum and the HPCI gland steam condenser exhauster.

d. Primary Containment and the SGTS room air.

ANSWER: 2

a. Reactor building exhaust plenum and the SGTS room air.

Explanation:

From Student Text Automatic Initiation - The SGT system can be automatically started on either a high drywell pressure (< 1.84 psig) or low-low reactor water level (> -42 inches) initiation signal or high radiation in the exhaust plenum initiation (< 49 mR/hr).

This signal is caused by a Group 6 containment isolation signal. Both SGTS fans will start and their respective inlet, outlet, and dilution air supply valves will open. The Group 6 isolation isolates the Reactor Building by closing the MG set ventilation valves, tripping the Reactor Building supply and exhaust fans and by isolating the normal ventilation. The SGTS suction from the Reactor Building Exhaust Plenum and SGTS room air valves draw air from the Reactor Building through the two parallel filter trains, to the fans, and then through the differential pressure control valves to the Elevated Release Point.

Distracters:

b. HPCI gland steam condenser exhauster and the Primary Containment. This is incorrect because it is only aligned here on a HPCI run with the main suction coming from the Building. Primary Containment must be manually aligned.

c. Reactor building exhaust plenum and the HPCI gland steam condenser exhauster. This is incorrect because HPCI Exhauster is aligned with the main suction coming from the Building.

d. Primary Containment and the SGTS room air. This is incorrect because its suction from Primary Containment must be manually aligned.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 262001.K2.01 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR001-01-01 (Attach if not previously provided)

(including version/revision number) 35 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 4 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: Y AC Electrical 3 00 09/21/10 SRO: Y Distribution NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 3 Multiple Choice N Topic Area Description Systems List the power supplies for Off-Site sources.

Related Lessons COR0010101 OPS AC Electrical Distribution Related Objectives LO 7.a. State the electrical power supplies to the following: Off-Site Sources of Power Related References COR0010101R35-S-OPS AC ELECTRICAL DISTRIBUTION P&ID Burns and Rowe 3001 10CFR55.41 b (4)

Related Skills (K/A) ROI SROI 262001.K2.01 Knowledge of electrical power supplies to the following: (CFR: 3.3 3.6 41.7) Off-site sources of power (3.3 / 3.6)

QUESTION: 3 What are the Offsite sources to the Cooper Stations 4160 Busses?

a. 161KV line from Auburn; and 69KV line from OPPD only.

b. 12.5KV South Underground Line; and the 69KV line from OPPD only.

c. 345KV switchyard via the Auto Transformer or the 161KV line from Auburn; and 69KV line from OPPD.

d. 345KV switchyard via the Auto Transformer and the 69KV line through the Corn Field Substation to the 12.5KV North Overhead Line.

ANSWER: 3

c. Auto Transformer via 345KV switchyard or the 161KV line from Auburn; and 69KV line from OPPD Explanation:

The Start-up Transformer is energized from the Auto Transformer via OCB-1604 or from the 161KV line via OCB-1606, The Emergency Transformer is energized by the 69KV line from Omaha. The 12.5 KV Distribution System is an underground and overhead system which provides power to electrical boilers 1C and 1D and also to the various 480V AC outdoor substations. The 12.5 KV system is not considered an off-site source, there is no way to power the 4160 Critical and non-critical busses from the 12.5 KV system.

USAR Section VIII - ELECTRICAL POWER SYSTEMS 1.0

SUMMARY

DESCRIPTION The offsite power sources at CNS are a startup station service transformer which connects to the CNS 161 kV switchyard and the 345/161 kV, 300 MVA auto-transformer connected to the 345 kV switchyard, and a separate emergency station service transformer energized by a 69 kV line. The 161 kV switchyard is connected to one 161 kV line which terminates in a switchyard near Auburn, Nebraska, and the 345/161 kV, 300 MVA auto-transformer which connects to the CNS 345 kV switchyard. The 345 kV switchyard has five (5) lines which terminate in switchyards near Booneville, Iowa; Hallam, Nebraska; St. Joseph, Missouri; Fairport, Missouri, and Nebraska City, Nebraska. The emergency station service transformer is fed by a 69 kV line which is part of a sub transmission grid of another utility (OPPD). If the normal station service transformer (powered by the main generator) is lost, the startup station service transformer, which is normally energized, will automatically energize 4160 volt buses 1A and 1B as well as their connected loads, including the critical buses. If the startup station service transformer fails to energize the critical buses, the emergency station service transformer, which is normally energized, will automatically energize both critical buses. If the emergency station service transformer were also to fail, the DGs would automatically energize their respective buses.

Distracters:

a. 161KV line from Auburn; and 69KV line from OPPD only. The normal feed into the station is from the 345KV switchyard was omitted from this option.

b. 12.5KV South Underground Line; and the 69KV line from OPPD only. Some of the 12.5 KV system is powered from the South Underground Line, but it is not considered part of the power supplies to the station.

d. Auto Transformer via 345KV switchyard; and 69KV line through the Corn Field Substation to the 12.5KV North Overhead Line. The 161 KV line was omitted and the North Overhead does not count as a power supply to the station.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 215004.K2.01 Importance Rating 2.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-30-02 (Attach if not previously provided)

(including version/revision number) 12 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: Y AC Electrical 4 00 09/21/2010 SRO: Y Distribution NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 3 Multiple Choice N Topic Area Description Systems What are the power supplies to the SRM Channels and Detectors?

Related Lessons COR0023002R12-S-Source Range Monitor Related Objectives 8.b Predict the consequences a malfunction of the following would have on the SRM system: 24/48 VDC power Related References COR0023002R12-S-Source Range Monitor 10CFR55.41 b.(7)

Related Skills (K/A) ROI SROI 215004.K2.01Knowledge of electrical power supplies to the SRM 2.6 2.8 channels/detectors

QUESTION: 4 What are the power supplies to the SRM Channels/detectors?

SRM Channels

a. A & B are powered from Division I + 24VDC System and C & D are powered from Division II + 24VDC System.

b. A & C are powered from Division I + 24VDC System and B & D are powered from Division II + 24VDC System.

c. A & B are powered from Division I 125 VDC System and C & D are powered from Division II 125 VDC System.

d. A & C are powered from Division I 125 VDC System and B & D are powered from Division II 125 VDC System.

ANSWER: 4

b. A & C are powered from Division I +24VDC System and B & D are powered from Division II +24VDC System.

Explanation:

From the Student Text for Lesson COR002-30-02 and GE Print 791E258 Sh.10, the Power Supplies for the SRM subsystem is powered from the + 24 VDC system. Channel A and C are powered from Div 1 and channel B and D are powered from Div 2.

Distracters:

a. A & B are powered from Division I + 24VDC System and C & D are powered from Division II + 24VDC System is incorrect because SRM B and D are DIV II SRMs and A and C are the DIV I SRMs.

c. A & B are powered from Division I 125 VDC System and C & D are powered from Division II 125 VDC System. This is the normal power supply form 125 VDC DIV I components and logics, however the SRMs are powered from the 24 VDC system.

d. A & C are powered from Division I 125 VDC System and B & D are powered from Division II 125 VDC System. This is the normal power supply form 125 VDC DIV II components and logics, however the SRMs are powered from the 24 VDC system.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 215005.K3.07 Importance Rating 3.2 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-01-02 (Attach if not previously provided)

(including version/revision number) 22 Learning Objective: See Attached (As available)

Question Source: Bank # 6137 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 6 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: Y 05 01 12/19/2005 Licensed Operator SRO: Y 6137 NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 4 Multiple Choice N Topic Area Description Systems COR0022402, 100% power with 26-27 selected. APRM E is 80% while all other APRMs are 100%.

Related Lessons COR0020102 OPS AVERAGE POWER RANGE MONITOR COR0022402 OPS ROD BLOCK MONITOR SKL0124224 ROD BLOCK MONITOR SYSTEM Related Objectives COR0020102001070E Given a specific APRM malfunction, determine the effect on any of the following: Rod Block Monitoring System (RBM)

COR0022402001040A Describe the RBM design features and/or interlocks that provide for the following: Prevent control rod withdrawal.

SKL012422400A030A Given plant conditions, predict changes in the following Rod Block Monitor System components/parameters: Trip reference Related References COR0020102 OPS AVERAGE POWER RANGE MONITOR COR0022402 OPS ROD BLOCK MONITOR Related Skills (K/A) ROI SROI 215005.K3.07 Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following: (CFR:

41.7 / 45.4) Rod block monitor: Plant-Specific (3.2 / 3.3)

QUESTION: 5 6137 The plant is operating at 100% power with control rod 26-27 selected. Average Power Range Monitor (APRM) channel E is reading 80% while all other APRMs are reading 100%.

What effect will this have on Rod Block Monitor (RBM) channel A?

RBM channel A will

a. initiate a Flow Reference Off-Normal rod block.

b. immediately initiate an RBM Downscale rod block.

c. enforce a non-conservative RBM Upscale rod block.

d. automatically transfer to APRM C as its reference APRM.

ANSWER: 5 6137

c. enforce a non-conservative RBM Upscale rod block.

Explanation:

The APRM's provide a reference power level for use in the Rod Block Monitor system. If the APRM used to set up the RBM trip references is indicating 30% (*) power, the RBM is zeroed, and RBM outputs are bypassed. If the reference APRM is bypassed, the reference signal is automatically provided by a second APRM. The reference APRM for RBM Channel A is APRM Channel E, with alternate APRM Channel C. The reference APRM for RBM Channel B is APRM Channel B, with alternate APRM Channel D.

(*) The actual set point number is 27.5%, but the Tech Spec Basis number for assuming RBM will mitigate the consequences of a RWE event with a peripheral control rod not selected is 30%.

When the reference APRM reads low the RBM assumes that power is lower than it actually is and will use the trip references for that power level band to assign rod blocks. The reference trip for 80% power is less than the one for 90 and above.

Distracters:

a. The Rod Block Monitor uses Recirc Flow for the Flow Reference Off-Normal rod block, not the Reference APRM.

b. The RBM looks at the reference APRM signal to determine if it is downscale from it or not.

So in this case it would not be.

d. APRM C will not function as the alternate for RBM A unless the normal reference APRM is Bypassed with the joystick.

In my opinion 10.CFR55.41 b (6) Design, components, and functions of reactivity control mechanisms and instrumentation. Is a better fit than (7) as stated in NUREG 1123.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 217000.K3.02 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-18-02 (Attach if not previously provided)

(including version/revision number) 18 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 6 00 12/15/2010 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Systems How do Reactor pressure and RCIC speed response to flow controller failure?

Related Lessons COR0021802 RCIC Related Objectives COR0020602001050A Describe the Core Spray system design features and/or interlocks that provide for the following: Prevention of over pressurization of Core Spray piping Related References COR0021802 Related Skills (K/A) 217000.K3.02 Knowledge of the effect that a loss or malfunction of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) will have on Reactor vessel pressure

QUESTION: 6 A Reactor scram occurred 30 minutes ago. The Core Isolation Cooling (RCIC) System is operating in the pressure control mode in accordance with Attachment 1 of Procedure 2.2.67.1 Pressure Control with RCIC Hard card. The following alignment exists:

• RCIC-FIC-91, RCIC flow controller is in Automatic.

• RCIC-FIC-91, RCIC flow controller Set tape is set to 400 GPM.

Shortly after the alignment was established the following occurs:

• RCIC-FT-58, discharge flow transmitter has failed high such that the flow sensed by the RCIC-FIC-91 is 500 gpm irrespective of actual RCIC flow.

How do Reactor pressure and RCIC speed response to this failure?

Reactor Pressure RCIC Speed

a. slowly rises rises and stabilizes at 5000 RPM

b. slowly rises lowers to idle speed

c. slowly lowers rises and stabilizes at 5000 RPM

d. slowly lowers lowers to idle speed ANSWER: 6

b. slowly rises lowers to idle speed Explanation:

The maximum demand (500 gpm) to the RCIC flow controller will cause FIC-91 to run speed at minimum, idle speed. The RCIC Turbine speed indication on panel 9-4 is 0 to 6000 rpm and the RCIC Turbine mechanical overspeed (125%) of rated, is 5625 rpm. So a speed of 5000 rpm is achievable which makes the distracters containing 5000 rpm plausible.

Distracters:

a. It is true that reactor pressure will rise because the RCIC Turbine is no longer providing pressure control for the vessel. The speed will lower not rise.

c. Reactor pressure will rise because the RCIC Turbine is no longer providing pressure control for the vessel, and the speed will lower not rise.

d. Reactor pressure will rise because the RCIC Turbine is no longer providing pressure control for the vessel.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 209001.K4.01 Importance Rating 3.2 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-06-02, GE Print 791E265 Sh 3 & 4 (Attach if not previously provided)

(including version/revision number) 21 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 7 00 03/23/2011 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 4 Multiple Choice Topic Area Description Systems What CS Valve interlock prevents over pressurizing the CS System?

Related Lessons COR0020602 CORE SPRAY Related Objectives COR0020602001050A Describe the Core Spray system design features and/or interlocks that provide for the following: Prevention of over pressurization of Core Spray piping Related References COR0020602 G.E. Print 791E265 Sheets 3 and 4 Related Skills (K/A) 209001.K4.01 Knowledge of LOW PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) Prevention of over pressurization of core spray piping (3.2 / 3.4)

QUESTION: 7 What interlock concerning the CS-MO-11 OUTBD INJECTION VLV and CS-MO-12 INBD INJ THROTTLE VLV, prevents the Core Spray piping from being overpressurized?

With Reactor Pressure at

a. 420 psig the CS-MO-12 must be opened first, and then the CS-MO-11 may be opened.

b. 420 psig the CS-MO-11 must be opened first, and then the CS-MO-12 may be opened.

c. 540 psig the CS-MO-12 must be opened first, and then the CS-MO-11 may be opened.

d. 540 psig the CS-MO-11 must be opened first, and then the CS-MO-12 may be opened.

ANSWER: 7

b. 420 psig the CS-MO-11 must be opened first then the 12 may be opened.

Explanation:

The Injection piping up stream of the MO-11 A/B is low pressure piping and cannot withstand the pressure of the vessel if both the 11 and the 12 were opened at the same time while the Reactor was operating at rated pressure. From the Student text for Core Spray COR002-06-02 When reactor pressure is > 291 psig and < 436 psig, MO-11 and MO-12 may be opened, using the control switches, only by opening MO-11 first.

Distracters:

a. 420 psig the CS-MO-12 must be opened first, and then the CS-MO-11 may be opened.

Print 791E265 SH 3 and 4 (see next page) for the CS-MO-11 and 12 respectively, show that the interlocks for the opening of the CS-MO-11 has to see the CS-MO-12 closed if no initiation signal is present to allow its opening.

c. 540 psig the CS-MO-12 must be opened first, and then the CS-MO-11 may be opened.

This is the wrong pressure and the valves are reversed.

d. 540 psig the CS-MO-11 must be opened first, and then the CS-MO-12 may be opened.

This is the wrong pressure. The pressure will prevent the 11 and 12 from opening and existing in an open condition at the same time.

K-13 is closed when there is an initiation LS6 closed if signal, Pwr CS-MO-11 is Avail, and low closed Rx Pressure 291-436 psig Control Switch Low Rx Pressure 291-436 psig K-13 is closed K-27 will only close if the CS-MO-12 is when there is an full closed.

initiation signal, Pwr Avail, and low Rx Pressure 291-436 psig Control Switch

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 262002.K4.01 Importance Rating 3.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR001-01-01 (Attach if not previously provided)

(including version/revision number) 35 Learning Objective: See Attached (As available)

Question Source: Bank # 1538 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: Y 8

02 09/27/2010 Licensed Operator SRO: Y 1538 NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 5 Multiple Choice N Topic Area Description Systems How does the NBPP react to a loss of normal power?

Related Lessons COR0010102 AC Electrical Distribution Related Objectives COR0010102001090G Describe the AC Electrical Distribution System design feature(s) and/or interlock(s) that provide for the following: Transfer from preferred power to alternate power supplies Related References LP COR0010102 Related Skills (K/A) ROI SROI 262002.K4.01 Knowledge of UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) design feature (s) and/or interlocks which provide for the following: (CFR: 41.7) Transfer from preferred power to alternate power supplies (3.1/3.4) (3.1 / 3.4)

QUESTION: 8 How does the No-Break Power Panel respond to a loss of its normal power supply?

The NBPP will transfer from

a. Inverter 1A to Inverter 1B.

b. Inverter 1A to MCC-R.

c. MCC-R to Inverter 1A.

d. MCC-R to selected Critical Distribution Panel.

ANSWER: 8

b. Inverter 1A to MCC-R.

Explanation:

Power to the No-Break Power Panel (NBPP) #1 is normally supplied from 250 VDC bus 1A through inverter 1A and a static switch.

An emergency (alternate) AC power source for the NBPP #1 is provided from MCC-R through a step-down transformer in the event that inverter 1A fails.

Distracters:

a. The NBPP will not automatically aligned to Inverter B

c. The NBPP is normally aligned to Inverter A therefore MCC-R is the backup power supply, not the other way around.

d. The NBPP is normally aligned to Inverter A therefore MCC-R is the backup power supply, not the other way around.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 203000.K5.02 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-23-02 (Attach if not previously provided)

(including version/revision number) 27 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: Y 09 00 03/15/2011 Licensed Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice N Topic Area Description Systems At what time can LPCI injection be lowered by throttling closed the RHR-MO-27B valve?

Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001060D Given an RHR control manipulation, predict and explain changes in the following: Reactor parameters (level, pressure, temperature)

Related References COR002-23-02 RHR Lesson Rev 27 Related Skills (K/A) ROI SROI 203000.K5.02 Knowledge of the operational implications of Core cooling 3.5 3.7 methods as they apply to RHR/LPCI: INJECTION MODE:

QUESTION: 9 The plant is operating at 100% power when a LOCA occurs and the following sequence occurs:

10:00 Rx level is -50 inches Wide Range and dropping slowly.

10:01 Drywell Pressure is 1.85 psig and rising.

10:01 Reactor Pressure is 600 psig and dropping.

10:03 Reactor Pressure is 436 psig and dropping.

10:06 Drywell Pressure is 4.5 psig and rising.

10:08 Reactor Pressure is 291 psig and dropping.

At what time can LPCI injection be lowered by throttling closed the RHR-MO-27B, OUTBD INJECTION valve?

a. 10:01

b. 10:03

c. 10:06

d. 10:08 ANSWER: 9

d. 10:08 Explanation:

During LPCI operation, with reactor pressure greater than 436 psig the RHR system starts on either of the two signals (RPV Level -113 inches or Drywell Pressure 1.84 psig). LPCI Pumps run on minimum flow, until the pressure lowers below 436 psig to provide an open signal to the RHR-MO-27B Injection Valve. Once that pressure permissive clears a five minute timer starts to prevent the closure of that valve. When five minutes has elapsed the Operator will be able to close the valve to lower core cooling.

From Lesson COR002-23-02 Outboard injection valve MO-27A(B)

Normally open valve, opens with the control switch on Panel 9-3 (27B *) positioned to OPEN (spring return to AUTO) and one of the following permissives:

a. MO-25A(B) closed, OR

b. Reactor pressure < 436 psig.

Opens if closed and a LPCI initiation signal is received and reactor pressure is < 436 psig.

It is interlocked open for 5 minutes to ensure full flow to the vessel. The 5 minute timer does not start until reactor pressure is reduced to <436 psig.

Can be throttled with control switch on Panel 9-3 unless interlocked open. This allows the Control Room operator to control system flow and reactor water level.

The interlocks for MO-27B are removed when its ASD switch is in isolate.

Distracters:

a. The initiation signal has been sent for LPCI Injection when Drywell Pressure rises above 1.84 psig. The Pumps start but the injection valve RHR-MO-25 is maintained closed to protect the low pressure piping of the RHR system. With RPV Pressure above 436 psig there is no injection so closing the 27B valve could not lower LPCI Injection.

b. The initiation signal is still present and at this pressure the RHR-MO-27B is interlocked open for 5 minutes, so it cannot be closed at this time.

c. This is the five minute mark from the initiation signal, but the five minute timer does not start for interlocking the RHR-MO-27B open until RPV pressure is below 436 psig. An additional two minutes must past until the RHR-MO-27B will be able to close.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 215003.K5.01 Importance Rating 2.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-12-02 (Attach if not previously provided)

(including version/revision number) 13 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 10 03 03/16/2011 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 2 Multiple Choice Topic Area Description Systems How will this affect the operation of A IRM versus the others when subjected to the same neutron field?

Related Lessons COR0021202 INTERMEDIATE RANGE MONITOR COR001-10-01 OPS Introduction to Neutron Monitoring Related Objectives COR0021202001060D Given a specific IRM malfunction, determine the effect on any of the following: Reactor power indication COR0011001 Obj 3 Concerning the Neutron Monitoring Systems:

a. State which detectors operate in the ionization region and which detectors operate in the proportional region of the gas conductivity curve.

b. State the reason for operating a detector in the specified region of the gas conductivity curve.

c. Explain the principle of operation of a fission chamber detector.

Related References COR0021202 Rev 12 Intermediate Range Monitor COR001-10-01 OPS Introduction to Neutron Monitoring Related Skills (K/A) 215003.K5.01 Knowledge of the operational implications of the following concepts as they apply to INTERMEDIATE RANGE MONITOR (IRM) SYSTEM: (CFR: 41.5 /

45.3) Detector operation (2.6/2.7)

QUESTION: 10 IRM A detector was installed with half the argon fill pressure than the other IRM detectors.

How will this affect the operation of A IRM versus the others when subjected to the same neutron field?

IRM A is

a. more sensitive and therefore will read higher than the others.

b. less sensitive and therefore will read higher than the others.

c. more sensitive and therefore will read lower than the others.

d. less sensitive and therefore will read lower than the others.

ANSWER: 10

d. less sensitive and therefore will read lower than the others.

Explanation:

From the Student Text COR001-10-01 Rev 12.

The IRM detector is very similar to the SRM detector, but is less sensitive due to the following major differences; the IRM detectors have: Less uranium, Lower argon gas pressure, Lower operating voltage.

Distracters:

a. The detector will lose sensitivity because there are fewer argon atoms (at lower pressure) to interact with the neutrons. Fewer atoms, fewer interactions therefore lower readings.

b. Fewer atoms, fewer interactions therefore lower readings.

c. The detector will lose sensitivity because there are fewer argon atoms (at lower pressure) to interact with the neutrons.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 205000.K6.04 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Abn. Procedure 2.4SDC (Attach if not previously provided)

(including version/revision number) 12 Learning Objective: See Attached (As available)

Question Source: Bank # 7763 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 11 7763 01 02/21/2005 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 3 Multiple Choice Topic Area Description Systems COR0022302, RESIDUAL HEAT REMOVAL Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001030K Describe RHR System design feature(s) and/or interlocks which provide for the following: Low reactor water level isolation COR0022302001050B Briefly describe the following concepts as they apply to the RHR system: Valve operation COR0022302001080K Predict the consequences a malfunction of the following will have on the RHR system: Reactor water level Related References 10CFR55.41 Written examinations: Operators 2.4SDC Shutdown Cooling Abnormal Related Skills (K/A) 205000. K6.04 Knowledge of the effect that a loss or malfunction of Reactor water level will have on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE): 3.6 3.6

QUESTION: 11 7763 The plant is shutdown with a cooldown in progress.

• Reactor pressure is 50 psig (stable)

• RPV level is +45 inches (NR stable)

• RHR loop B suction path is lined up for Shutdown Cooling RPV level then lowers to -150 inches (WR) due to a recirculation suction line break.

What is the first control room valve manipulation(s) required to establish injection with LPCI Loop B?

a. OPEN RHR-MO-27B, LPCI Injection Valve.

b. CLOSE RHR-MO-15B and 15D, S/D Cooling Suction Valves.

c. CLOSE RHR-MO17 and 18, S/D Cooling Suction Isolation Valves.

d. OPEN RHR-MO-13B and 13D, Torus Cooling Suction Valves.

ANSWER: 11 7763

b. CLOSE RHR-MO-15B and 15D, Shutdown Cooling Suction Valves.

Explanation:

With the SDC suction valves open the loss of reactor level results in a SDC isolation. In order to establish B loop LPCI injection the suction has to be realigned to the torus. MO-15B and 15D must be closed before 13B and 13D can be reopened.

Distracters:

a. is incorrect as this valve will open and will not close on the SDC isolation.

c. is incorrect because this action does not need to be performed by the operator.

d. is incorrect because the torus suction valves cannot be opened unless the SDC suctions are closed. The PCIS isolation will close these valves.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 400000.K6.07 Importance Rating 2.7 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-19-02 (Attach if not previously provided)

(including version/revision number) 21 Learning Objective: See Attached (As available)

Question Source: Bank # 18245 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: Y 12 01 03/20/2005 NRC Style Question SRO: Y 18245 NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice N Topic Area Description Systems COR0021902, REC pump response following Loss of Off-site power Related Lessons COR0021902 REACTOR EQUIPMENT COOLING Related Objectives COR0021902001070D Predict the consequences a malfunction of the following would have on the REC system: Loss of Normal AC power Related References CFR 10CFR55.41 Related Skills (K/A) ROI SROI 40000.K6.07 Knowledge of the effect that a loss or malfunction of the following will have on the CCWS: (CFR: 41.7 / 45.7) Breakers, relays, and disconnects (2.7 / 2.8)

QUESTION: 12 The plant is operating normally at full power with the following REC alignment:

• REC pumps 1B, 1C, and 1D are running

• REC pumps 1B and 1C selector switches are in STANDBY The plant experiences a loss of all offsite power.

• MCC S feeder breaker trips open when power is restored to the 4160 Buses.

What is the expected REC pump status 1 minute after the loss of all offsite power?

a. Only pump 1C will be running.

b. Only pump 1B will be running.

c. Only pumps 1B and 1C will be running.

d. Pumps 1B, 1C, and 1D will be running.

ANSWER: 12

b. Only pump 1B will be running.

Explanation:

Following the loss of all offsite power, under-voltage on 4160V bus 1F and 1G will start DG-1 and DG-2 and will result in a loss of power to MCC-K and to MCC-S and the subsequent trip of all running REC pumps. When DG-1 and DG-2 have reached rated speed and voltage and breakers EG-1 and EG-2 have closed, then REC pumps 1C and 1B will auto restart 20 seconds after MCC-K and MCC-S are re-energized via bus 1F and 1G. When the feeder breaker to MCC-S trips that removes the power to the 1C pump that should be running.

Distracters:

a. is incorrect. It is powered from MCC-S and its feeder breaker tripped.

c. is incorrect. This would have been the correct answer if MCC-S had not tripped and de-energized the MCC.

d. is incorrect. Because pump 1D will trip and will not automatically restart and MCC-S is de-energized.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 264000.A1.04 Importance Rating 2.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-08-02 (Attach if not previously provided)

(including version/revision number) 24 Learning Objective: See Attached (As available)

Question Source: Bank # 19288 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 8 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 13 19288 02 12/17/2004 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Systems COR0020802, Loss of SW to DG with Auto start present Related Lessons COR0020802 OPS DIESEL GENERATORS Related Objectives COR0020802001060H Describe the interrelationship between Diesel Generators and the following: Service Water System COR0020802001140C Given plant conditions, determine if the following should occur: Diesel Generator trip COR0020802001010G State the purpose of the following items related to the Diesel Generators: Cooling Water subsystem Related References NONE Related Skills (K/A) 264000.A1.04 Ability to predict and/or monitor changes in parameters associated with operating the EMERGENCY GENERATORS (DIESEL/JET) controls including: Crank case temperature and pressure (2.6/2.7)

QUESTION: 13 19288 The following conditions exist:

• A LOCA has occurred.

• DG1 has been supplying its loads for 5 minutes.

• SW-2797A (Service water admission valve to DG1) remains closed.

How is DG1 affected?

a. DG1 trips on low lube oil pressure.

b. DG1 trips on high jacket water temperature.

c. DG1 continues to operate using Division II service water supply.

d. DG1 continues to operate with high crank case temperature and pressure.

ANSWER: 13 19288

d. DG1 continues to operate with high lube crank case temperature and pressure.

Explanation:

A loss of service water will cause jacket water temperatures to rise and the diesel will continue to run since this trip is bypassed by the LOCA signal. As Jacket Water temperature rises the lube oil will start heating up. As the lube oil temperature rises both crank case temperature and pressure will rise.

Cooper Bessemer Print KSV47-8

SW-2797A Div I SW Div II SW From USAR Section X Subsection 8: The SW system consists of four vertical SW pumps located in the Intake Structure, and two associated strainers, piping, valving, and instrumentation (See Burns and Roe Drawings 2006, Sheets 1 through 4, and 2036, Sheet 1). The SW pumps discharge to a common header from which independent piping supplies two Seismic Class IS cooling water loops and one Turbine Building loop. In the event of a loss of header pressure below 20 psig, automatic valving is provided to shutoff all supply to the Turbine Building loop, thus assuring supply to the Seismic Class IS loops. Each Seismic Class IS loop feeds one diesel generator two RHR service water booster pumps, and one REC heat exchanger.

Valves are included in the common discharge header to permit the SW system to be operated as two independent loops. Either loop can supply normal cooling water to the REC Critical Loops and the diesel generators. Either loop can also supply the RHR service water booster pump room fan coil unit, and Control Room air conditioning units.

Distracters:

a. DG oil temperatures will rise, but only a high oil temperature alarm is provided no trip.

The low lube oil pressure trip is also bypassed.

b. The DG jacket water high temperature trip is bypassed with the diesel started by a LOCA signal.

c. The DG service water header admission SW-2797 valve is on the inlet to the cooling systems and downstream of the other division supply, so there is no way the other divisions SW cooling could get to the diesel if this valve failed closed. Students that remembered incorrectly from the description from the USAR that states that either loop of SW is capable of supplying cooling to the diesel, which is the normal alignment, however with the failure of the SW-2797 this cannot be satisfied. If a student failed to remember where the SW-2797 is located in the system (i.e. thinking that it was in the divisional SW supply line and not downstream of the cross-tie lines) would choose this answer.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 206000.A1.08 Importance Rating 4.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedure 2.2.33.1; Lesson COR002-11-02 (Attach if not previously provided)

(including version/revision number) 29 27 Learning Objective: See Attached (As available)

Question Source: Bank # 3744 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 14 3744 04 09/28/2010 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Systems SKL0124211, COR0021102 High Pressure Coolant Injection Related Lessons COR0021102 OPS High Pressure Coolant Injection System Related Objectives COR0021102001120C Given plant conditions, determine if the following HPCI actions should occur: Minimum flow valve change of position COR0021102001050E Describe the interrelationship between HPCI and the following:

ECST COR0021102001050F Describe the interrelationship between HPCI and the following:

Suppression chamber Related References 2.2.33.1 High Pressure Coolant Injection System Operations Related Skills (K/A) 206000.A1.08 Ability to predict and/or monitor changes in parameters associated with operating the HIGH PRESSURE COOLANT INJECTION SYSTEM controls including: System lineup:

(4.1* 4.0)

QUESTION: 14 3744 (1 point(s))

The High Pressure Coolant Injection (HPCI) System is started in RPV PRESSURE CONTROL MODE at a flow rate of 2000 gpm and a discharge pressure of 900 psig.

The Reactor Operator then lowers the FLOW CONTROLLER HPCI-FIC-108 tape setpoint to establish a flow rate of 400 gpm for 12 minutes.

After 20 minutes of HPCI operation at the reduced flow rate, what is the effect on Suppression Pool level AND the Emergency Condensate Storage Tanks (ECST) level?

a. Suppression Pool level will rise.

ECST level will rise.

b. Suppression Pool level will rise.

ECST level will lower.

c. Suppression Pool level will lower.

ECST level will remain the same.

d. Suppression Pool level will remain the same.

ECST level will lower.

ANSWER: 14 3744

b. Suppression Pool level will rise.

ECST level will lower.

Explanation:

In the RPV PRESSURE CONTROL MODE alignment, HPCI suction is aligned to the ECST with a return flow path to the ECST. However the operators action to lower the flow controller to 400 gpm causes the minimum flow valve to open and it returns to the torus. Water is then diverted from the ECST to the TORUS.

Distracters:

a, c, & d. are incorrect because the Full Flow Alignment has HPCI suction aligned to the ECST and at 400 gpm the min flow valve to the Torus is OPEN, diverting ECST water to the TORUS.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 212000.A2.08 Importance Rating 4.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedure 2.1.5 (Attach if not previously provided)

(including version/revision number) 64 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 2 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: Y 15 00 09/29/10 License Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice N Topic Area Description Systems Scram and Startup FCVs response on low level Related Lessons COR0022102 OPS Reactor Protection System Related Objectives COR0022102001090C Predict the consequences a malfunction of the following would have on the RPS system: Nuclear boiler instrumentation Related References COR0022102 OPS Reactor Protection System Rev20 Related Skills (K/A) ROI SROI 212000.A2.08 Ability to (a) predict the impacts of Low reactor level on the 4.1 4.2 REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(4.1*/4.2*)

QUESTION: 15 The reactor is at 100% power with the Startup Master Level Controller in Manual due to a Triconex failure when the following occurs:

• RPV water level starts lowering.

• RFC-LI-94A, B & C Rx NR LEVEL read 0 inches.

What is the status of the RPS Trip System and what must be performed on the RVLC HMIs for the Startup valves FCV-11AA or FCV-11BB to allow feeding the vessel to recover reactor water level?

RPS TRIP FCV-11AA or Logic FCV-11BB Action

a. Tripped Press GREEN UP ARROW on the Startup Master Level

b. Tripped Press AUTO button on the Startup Master Level

c. Not Tripped Press GREEN UP ARROW on the Startup Master Level

d. Not Tripped Press AUTO button on the Startup Master Level ANSWER: 15

a. Tripped Press GREEN UP ARROW on the Startup Master Level Explanation:

With all three level indicators RFC-LI-94A, B & C reading 0 inches would indicate that the NBI-LIS-101A through D would read the same, as they come off the same reference and variable instrument legs. The NBI-LIS-101A, B, C, D feed the RPS system to produce a Reactor Scram on low level at 3 inches Tech Spec and 10 inches actual setpoint. For the Startup FCVs, they are controlled by the FWLCS and if in AUTO on a reactor scram, will regulate to control RPV level when Level Set-Down is enabled. In this setup, they will not, since they are in manual. IAW Procedure 2.1.5 Attachment 3 REACTOR WATER LEVEL CONTROL; step 1.3.6 Adjust STARTUP MASTER controller using UP/DOWN arrows or RAMP FUNCTION to adjust LEVEL SETPOINT as desired.

Distracters:

b. The failure prevents selecting AUTO to control the STARTUP FCVs to prevent overfeeding.

c. RPS Trip Logic will be tripped.

d. RPS Trip Logic will be tripped, the failure prevents selecting AUTO to control the STARTUP FCVs to prevent overfeeding.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 261000.A2.04 Importance Rating 2.5 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-28-02; Ann Proc. 2.3_K-1 (Attach if not previously provided)

(including version/revision number) 19 13 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 1164 (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: Y 16 00 09/29/10 License Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice N Topic Area Description Systems What is the major impact to the system and what must be done to mitigate the consequences?

Related Lessons COR0022802 OPS STANDBY GAS TREATMENT Related Objectives COR0022802001100I Predict the consequences of the following on the Standby Gas Treatment system: High train moisture content Related References COR0022802 OPS Standby Gas Treatment System R19 AP 2.3K-1 Alarm Procedure 2.3_K-1 Rev 13 Related Skills (K/A) ROI SROI 261000.A2.04 Ability to (a) predict the impacts of High train moisture content 2.5 2.7 on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) (2.5 / 2.7)

QUESTION: 16 The plant is operating at 100% power with the A SGT train in service to support HPCI full flow surveillance when the following occurs:

• Annunciator K-1/A-2 SGT A HIGH MOISTURE alarms What is the major impact to the system due to this condition and what must be done to mitigate the consequences of the high moisture condition?

Impact Action

a. Reduced iodine Check SGT-DPIC-546 adsorption in the for proper operation.

charcoal filter.

b. Reduced iodine Check proper operation adsorption in the of the heater.

charcoal filter.

c. Potential fire hazard Check SGT-DPIC-546 in the Inlet High for proper operation.

Efficiency filter.

d. Potential fire hazard Check proper operation in the Inlet High of the heater.

Efficiency filter.

ANSWER: 16

b. Reduced iodine Check proper operation adsorption in the of the heater.

charcoal filter.

Explanation:

From the student study material: Excessive moisture or organic materials (such as lubricants) will reduce the iodine adsorption capability of the charcoal filter if these are not previously removed by the demister, heater, or filter 1.

When Alarm K-1 / A-2 SGT A HIGH MOISTURE alarms the annunciator procedure directs them to perform or check the following:

1. OPERATOR OBSERVATION AND ACTION 1.1 Send Operator to investigate.

1.1.1 Check heater operation.

1.1.2 Check for charcoal filter fire.

1.2 If during post-accident conditions, remove decay heat per Procedure 2.2.73.

1.3 If cause cannot be identified or easily corrected, start SGT B train and secure SGT A train per Procedure 2.2.73.

1.4 If fire confirmed:

1.4.1 Concurrently enter Procedure 5 .1 INCIDENT.

1.4.2 Open FP-454, SYSTEM 22 AND SYSTEM 23 SUPPLY SHUTOFF (R-976-W) for - 5 minutes.

These are the only actions stated in the procedure to perform.

Distracters:

a. Reduced iodine adsorption in the charcoal filter is the correct item of concern but checking SGT-DPIC-546 for proper operation is the wrong action to take.

c. Potential fire hazard in the Inlet High Efficiency filter is the wrong component. There is a higher potential for fire in the charcoal filter rather than the high efficiency filter. Also checking SGT-DPIC-546 for proper operation is also wrong action.

d. Potential fire hazard in the Inlet High Efficiency filter is the wrong component. There is a higher potential for fire in the charcoal filter rather than the high efficiency filter. But the action is correct.

MODIFIED QUESTION: Bank question that was modified - 1164 With regard to the Standby Gas Treatment (SGT) system, which one of the following is a consequence of an inefficient Moisture Separator?

a. Overload trip of the associated SGT fan.

b. Excessively high water level in the Z sump.

c. Reduced iodine adsorption in the charcoal Filter.

d. Potential fire hazard in the Inlet High Efficiency Filter.

ANSWER: 1 1164

c. Reduced iodine adsorption in the Charcoal Filter.

REFERENCE:

STCOR002-28-02, page 14, section II.F.2, rev. 10.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 215005.A3.02 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 2.3_9-5-1 Panel 9 Annunciator 9-5-1 (Attach if not previously provided)

(including version/revision number) 24 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 2 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 17 00 03/16/2011 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 2 Multiple Choice Topic Area Description Systems COR0020102, AVERAGE POWER RANGE MONITOR Related Lessons COR0020102 OPS AVERAGE POWER RANGE MONITOR Related Objectives COR0020102001050D Describe the interrelationships between the Average Power Range Monitor System and the following: Local Power Range Monitoring System (LPRM)

Related References 2.3_9-5-1 Panel 9 Annunciator 9-5-1 Related Skills (K/A) 215005.A3.02 Ability to monitor automatic operations of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM including: (CFR:

41.7 / 45.7) Full core display (3.5 / 3.5)

QUESTION: 17 The plant is operating at 100% power when the following full core display is observed:

What are the setpoints for these indications?

a. 3 watts/cm2, 100 watts/cm2

b. 3 watts/cm2, 75 watts/cm2

c. 5 watts/cm2, 100 watts/cm2

d. 5 watts/cm2, 75 watts/cm2 ANSWER: 17

a. 3 watts/cm2, 100 watts/cm2 Explanation:

Annunciator Procedure 2.3_9-5-1 Window B-7 LPRM UPSCALE SETPOINT 100 watts/cm2.

Window B-7 LPRM DOWNSCALE SETPOINT 3 watts/cm2.

OPERATOR OBSERVATION AND ACTION NOTE - CNS OE is that Areva LPRMs that have spiked once have exhibited a shorter frequency and larger magnitude change between events and occupy core locations:

04-21, 12-21, 12-29, 28-05, 36-13, and 44-21. (Interim Change) 1.1 Determine from Panel 9-5 full core display or Panel 9-14 LPRM alarm lights which LPRM(s) is alarming.

1.2 Select a control rod so that alarming LPRM(s) display on 4 rod display.

1.3 If power level is rising in area of alarming LPRM, enter Procedure 2.4RXPWR.

1.4 If power level is stable, enter Procedure 10.19.

USAR Section VII 5.8.2.5 Trip Functions The trip circuits for the LPRMS provide trip signals to activate lights, instrument inoperative signals, and annunciators. These trip circuits use the DC power supply and are set to trip on loss of power; they also trip when power is not available for the LPRM amplifiers. USAR Table VII-5-3 indicates the trips.

The trip levels can be adjusted to within +/- 0.5% of full scale deflection and are accurate to +/- 1% of full scale deflection in the normal operating environment.

TABLE VII-5-3 LPRM TRIPS Trip Function Trip Action LPRM Downscale White Light and Annunciator LPRM Upscale Amber Light and Annunciator LPRM Bypass White Light, Annunciator, and APRM averaging compensation Distracters:

b. 3 watts/cm2, 75 watts/cm2. This is incorrect because the setpoint for the Upscale alarm and light is 100 watts/cm2.

c. 5 watts/cm2, 100 watts/cm2. This is incorrect because the setpoint for the Downscale alarm and light is 3 watts/cm2.

d. 5 watts/cm2, 75 watts/cm2 This is incorrect because the setpoint for the Downscale alarm and light is 3 watts/cm2 and the setpoint for the Upscale alarm and light is 100 watts/cm2.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 217000.A3.04 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-18-02 (Attach if not previously provided)

(including version/revision number) 18 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: Y 18 00 10/04/10 License Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 4 Multiple Choice N Topic Area Description Systems Without operator action, how is RCIC flow affected, if at all?

Related Lessons COR0021802 OPS Reactor Core Isolation Cooling Related Objectives COR00218020011300 Briefly describe the RCIC system response to an initiation signal when in a normal standby alignment.

Related References COR0021802 OPS Reactor Core Isolation Cooling Rev 18 Related Skills (K/A) ROI SROI 217000.A3.04 Ability to monitor automatic operations of the REACTOR 3.6 3.5 CORE ISOLATION COOLING SYSTEM (RCIC) including: (CFR: 41.7 / 45.7)

System flow (3.6 / 3.5)

QUESTION: 18 The reactor is at 100% power with RCIC in a full flow test mode using the test potentiometer at 300 gpm, when a loss of feedwater occurs.

• Reactor scrams and water level lowers to - 45 inches Wide Range.

Without operator action, how is RCIC flow affected, if at all?

RCIC flow

a. increases to 400 gpm and that flow is directed to the vessel.

b. increases to 400 gpm and RCIC remains in the test mode.

c. remains at 300 gpm and that flow is directed to the vessel.

d. remains at 300 gpm and RCIC remains in the test mode.

ANSWER: 18

a. increases to 400 gpm and that flow is directed to the vessel.

Explanation:

Even though RCIC is in the full flow test alignment on the test potentiometer an initiation signal will cause RCIC to shift out of the test mode to a full automatic mode using the normal setpoint for flow (i.e. 400 gpm injecting into the vessel). RCIC starts on a level 2 setpoint of - 42 inches.

Lesson COR002-18-02 page 24 paragraph k. A Test Potentiometer on Panel 9-4 provides an additional method of controlling turbine speed. The Test Potentiometer sends a signal to open or close the governor valve as desired. To use the test potentiometer, the Test switch is placed to TEST and the Test Power switch is placed in ON. The initiation signal must be reset if present. If an initiation occurs while using the test potentiometer, flow control will be returned to the GEMAC flow controller.

RCIC Controller print GE 791E264 Sh 5 K5 relay is the Initiation signal -42 inches RPV water level When the K5 is energized it will close the K5 3/4 and 7/8 contacts to place the RCIC Flow Controller in the circuit. At the same time the K5 1/2 and 5/6 contacts open to remove the RCIC Test Potentiometer from the circuit. See drawing below.

Test Potentiometer RCIC Flow Controller Distracters:

b. The system shifts to the flow controller on an initiation signal and injects into the vessel.

c. The system shifts to the flow controller on an initiation signal and injects into the vessel.

d. The system shifts to the flow controller on an initiation signal and injects into the vessel.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 223002.A4.02 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) T.S. 3.3.6.1 Procedure 2.1.22 (Attach if not previously provided)

(including version/revision number) Amendments 178, 231, 212 55 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 19 00 12/02/2010 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Systems Given a set of plant conditions determine if a Group 3 should occur Related Lessons COR0020302 OPS CONTAINMENT Related Objectives 21b. Given plant conditions, determine if the following should have occurred: Any of the PCIS group isolations Related References 2.1.22 Recovering From A Group Isolation Related Skills (K/A) 223002.A4.02 Ability to manually operate and/or monitor in the control room: Manually initiate the system (3.9/3.8)

QUESTION: 19 The Plant is operating at 100% power when the following events occur:

• RWCU System Heat Exchanger Room temperature is 185F and going up at 1F per minute.

• RWCU System Flow is 181% of rated and steady.

When will the RWCU system isolate and what actions should be taken if it does not?

Use Tech Spec Allowable Value when considering your answer.

a. immediately; Close RWCU-MO-15, INBD ISOL VLV Only.

b. immediately; Close both RWCU-MO-15 and RWCU-MO-18, OUTBD ISOL VLV.

c. in 10 minutes; Close RWCU-MO-15 Only.

d. in 10 minutes; Close both RWCU-MO-15 and RWCU-MO-18.

ANSWER: 19

d. in 10 minutes; Close both RWCU-MO-15 and RWCU-MO-18.

Explanation:

With a temperature rise in the RWCU System Heat Exchanger Room the PCIS system should react and cause an isolation of the RWCU system when temperature reaches 195F in accordance with Procedure 2.1.22. This is the Tech Spec Allowable Value. The rate of rise is given as 1F/minute so it will take ten more minutes to take the room temperature from 185F to 195F. The other RWCU parameter that is approaching its isolation setpoint is system flow.

The isolation does not occur until flow is equal to or exceeds 191% of rated in accordance with its Tech Spec allowable value.

Distracters:

a. This time is too short, an isolation will not occur until the temperature in the room reaches 195F. If that does not occur then the operator is required to take manual action to cause the failed automatic action to happen at the setpoint. Both isolation valves should be closed because the auto action would have closed both valves.

b. This time is too short, an isolation will not occur until the temperature in the room reaches 195F. If that does not occur then the operator is required to take manual action to cause the failed automatic action to happen at the setpoint. Both isolation valves should be closed because the auto action would have closed both valves.

c. This time is correct however the operator is required to take manual action to cause the failed automatic action to happen at the setpoint. Both isolation valves should be closed because the auto action would have closed both valves.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 259002.A4.07 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-07-02, Abn. Proc. 2.4RXLVL (Attach if not previously provided)

(including version/revision number) 29 24 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 20 01 06/26/2008 05/23/2010 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 3 Multiple Choice Topic Area Description Systems Monitor RVLCS when 1 Steam Flow instrument is removed from service.

Related Lessons COR0023202 OPS REACTOR VESSEL LEVEL CONTROL Related Objectives COR0023202001010C State the purpose of the following items related to the Reactor Vessel Level Control System: Steam flow instruments COR0023202001020A Describe the interrelationship between RVLC and the following:

Main Steam COR0023202001050B Briefly describe the following concepts as they apply to the RVLC system: Steam flow/Feed Flow Mismatch Related References 2.4RXLVL RPV Water Level Control Trouble Related Skills (K/A) 259002.A4.07 Ability to manually operate and/or monitor in the control room: All individual component controllers when transferring from automatic to manual mode (3.8)

QUESTION: 20 The plant is at 75% power. The Reactor Level Control system is maintaining RPV level at +35 inches in three (3) element control. HMI shows mFT0051C_INVALID, MS Flow Ch. C INVALID.

If the operator were to then bypass MS-PT-56 Turbine 1st stage pressure, what would be the status of the RVLCS?

RVLCS would

a. remain in 3 element control with level at +35 inches.

b. transfer to single element control with level at +35 inches.

c. remain in 3 element control and develop and send a runback of the Reactor Recirc Pumps toward 45%.

d. transfer to single element control and develop and send a runback of the Reactor Recirc Pumps toward 45%.

ANSWER: 20

b. transfer to single element control with level at +35 inches.

Explanation:

Level will transfer to single element due to 1 invalid steam flow instrument when the turbine first stage pressure instrument is bypassed. Level will remain at 35. The runback would be generated on the following:

Recirculation Runback To survive loss of a feedwater pump without a scram, the system is designed to reduce reactor power by running back RRMG speed. The system will initiate a recirculation runback on a trip or loss of flow from a feedwater pump coincident with reactor power above the capacity of the remaining feedwater pump and RPV water level < 27.5 inches.

The system determines the current reactor power using validated 1st stage pressure or validated main steam flow if 1st stage pressure is unavailable. If reactor power is above 9 mlb/hr steam flow and the system has experienced a loss of a RFPT the system will generate a 45%

recirculation runback signal to reduce power. This recirculation runback will stay in effect until the system feedwater demand can be serviced by a single feed pump.

Distracters:

a. The RVLCS will transfer to single element when the first stage pressure transmitter is bypassed.

c. The RVLCS will transfer to single element when the first stage pressure transmitter is bypassed. A runback is not generated based on these particular conditions but would be if level dropped and there were a low flow condition.

d. A runback is not generated based on these particular conditions but would be if level dropped and there were a low flow condition.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 239002 G 2.4.45 Importance Rating 4.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 2.4SRV 2.3_9-3-1 Annunciator Panel Procedure (Attach if not previously provided)

(including version/revision number) 11 28 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 21 00 01/11/2011 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 3 Multiple Choice Topic Area Description Systems COR0021602, What temp corresponds to the pressure the amber light is lit?

Related Lessons COR0021602 OPS NUCLEAR PRESSURE RELIEF Related Objectives COR0021602001040A Given a Nuclear Pressure Relief system component manipulation, predict and explain the changes in the following parameters: Tail pipe temperatures COR0021602001060D Briefly describe the following concepts as they apply to NPR: Tail pipe temperature monitoring Related References 2.4SRV 2.3_9-3-1 Annunciator Panel Procedure Related Skills (K/A) 239002 SRV 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR:

41.10 / 43.5 / 45.3 / 45.12) IMPORTANCE RO 4.1 SRO 4.3.

QUESTION: 21 The plant is operating at 100% power when Annunciator Panel 9-4-1 Window D-2 RELIEF VALVE ACCUMULATOR LOW PRESSURE Alarms.

What indication is available to help the operator to determine the urgency of his/her actions and what action is required?

a. Local pressure gauges; Notify the RP Department to sample the atmosphere of the Reactor Building 903 ft south for increased Nitrogen concentration.

b. Local pressure gauges; Align Nitrogen to the pneumatic system and isolate the Instrument Air leak to maintain air pressure.

c. Panel 9-4 Amber lights; Notify the RP Department to sample the atmosphere of the Reactor Building 903 ft south for increased Nitrogen concentration.

d. Panel 9-4 Amber lights; Align Nitrogen to the pneumatic system and isolate the Instrument Air leak to maintain air pressure.

ANSWER: 21

c. Panel 9-4 Amber lights; Notify the RP Department to sample the atmosphere of the Reactor Building 903 ft south for increased Nitrogen.

Explanation:

The SRVs are supplied with pneumatic Nitrogen/Instrument Air system as a motive force to cycle the SRVs. Each SRV also has its own accumulator to allow cycling when the pneumatic system is non-functional.

As seen in the portion of Plant Drawing 2010 Sh. 2 Rx Building Air/Nitrogen, there are no pressure gauges available locally just pressure switches that send the low pressure alarm to panel 9-4-1 and to one or more amber lights associated with the accumulators which have low pressure.

Drawing 3048 Sh.1 Distracters:

a. Local pressure gauges; Notify the RP Department to sample the atmosphere of the Reactor Building 903 ft south for increased Nitrogen. This is incorrect because there are no local gauges on the individual SRV Accumulators, however there are pressure gauges located on different accumulators throughout the Reactor Building. A candidate, who thought that these accumulators had the pressure gauges, might select this answer.

Therefore making this a credible misconception.

b. Local pressure gauges; Align Nitrogen to the pneumatic system and isolate the Instrument Air leak to maintain air pressure. This is incorrect because there are no local gauges on the individual SRV Accumulators, however there are pressure gauges located on different accumulators throughout the Reactor Building. A candidate, who thought that these accumulators had the pressure gauges, might select this answer. Therefore making this a credible misconception. Also Nitrogen is already aligned to supply pneumatics when the plant is operating at 100% power.

d. Panel 9-4 Amber lights; Align Nitrogen to the pneumatic system and isolate the Instrument Air leak to maintain air pressure. This is incorrect because Nitrogen is already aligned to supply pneumatics when the plant is operating at 100% power.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 218000 G 2.1.19 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-17-02 and Proc. 2.2.1 (Attach if not previously provided)

(including version/revision number) 16 37 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 3297 (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 22 00 12/07/2010 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Systems Determine if the ADS valves should be open and how they would be displayed on SPDS.

Related Lessons COR0021602 OPS NUCLEAR PRESSURE RELIEF Related Objectives COR0021602001050B Describe the Nuclear Pressure Relief system design features and/or interlocks that provide for the following: ADS logic control COR0021602001060A Briefly describe the following concepts as they apply to NPR: ADS logic operation Related References 2.2.1 Nuclear Pressure Relief System Related Skills (K/A) 218000 ADS 2.1.19 Ability to use plant computers to evaluate system or component status. (CFR: 41.10 /

45.12) IMPORTANCE RO 3.9 SRO 3.8

QUESTION: 22 The following conditions have been present for 2 minutes:

• RPV water level indicates -148 inches on the wide range RPV level instruments.

• Reactor pressure is 300 psig.

• Drywell pressure is 22 psig.

Which one of the following describes the current status of the ADS valves, and how are the SRVs displayed on SPDS 10 PMIS Display Suppression Pool Mimic?

The ADS valves are

a. closed.

The ADS Valves are highlighted green and the LLS valves are highlighted red.

b. closed.

The ADS Valves are highlighted green and the LLS valves are highlighted green.

c. open.

The ADS Valves are highlighted red and the LLS valves are highlighted red.

d. open.

The ADS Valves are highlighted red and the LLS valves are highlighted green.

ANSWER: 22

d. open.

The ADS Valves are highlighted red and the LLS valves are highlighted green.

Explanation: With Reactor Water level below the -113 inch setpoint the ADS Timers will start and 109 seconds later the ADS valves will open if there are ECCS Pumps running as indicated by a discharge pressure. With water level below the automatic initiation setpoints for the ECCS Pumps, the pumps will be running on minimum flow.

SPDS 10 Suppression Pool Mimic displays the status of the SRVs, both the ADS and the LLS valves in either green (closed) or red (open), along with the tailpipe temperatures, the suppression pool temperature and the status of both HPCI and RCIC.

Distracters:

a. ADS valve logic is satisfied and the valves are open. SPDS 10 would indicate the ADS Valves as red, but the LLS (low low set) valves would indicate green.

b. ADS valve logic is satisfied and the valves are open. SPDS 10 would indicate the ADS Valves as red, but the LLS (low low set) valves would indicate green.

c. ADS valve logic is satisfied and the valves are open. SPDS 10 would indicate the ADS Valves as red, but the LLS (low low set) valves would indicate green.

REFERENCE:

COR0021602, Procedure 2.2.1; Procedure 2.4CSCS Modified question # 3297

MODIFIED QUESTION: 3297 The following conditions have been present for 2 minutes:

• RPV water level indicates -148 inches on the wide range RPV level instrument.

• Reactor pressure is 300 psig.

• Drywell pressure is 22 psig.

Assume ALL equipment operates as designed.

Which one of the following describes the current status of the ADS valves, and the actions necessary to close OR maintain them closed?

The ADS valves are

a. open.

The ADS A INHIBIT and the ADS B INHIBIT switches must be placed in INHIBIT.

b. closed.

The ADS A INHIBIT and the ADS B INHIBIT switches must be placed in INHIBIT.

c. closed.

The ADS LOGIC A TIMER and the ADS LOGIC B TIMER pushbuttons must be depressed at least every 90 seconds.

d. open.

The ADS A INHIBIT and the ADS B INHIBIT switches must be placed in INHIBIT and then the ADS LOGIC A TIMER and ADS LOGIC B TIMER pushbuttons must be depressed.

ANSWER:

a. open.

The ADS A INHIBIT and the ADS B INHIBIT switches must be placed in INHIBIT.

FOILS:

b. ADS valve logic is satisfied and the valves are open.

c. ADS valve logic is satisfied and the valves are open.

d. Depressing the reset push buttons is not required.

REFERENCE:

COR0021602, PR 2.2.1 Section 4.1, PR 2.4.4.1 Section 4.1.3

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 211000.A4.07 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-29-02 (Attach if not previously provided)

(including version/revision number) 18 Learning Objective: See Attached (As available)

Question Source: Bank # 3268 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 23 3268 02 12/07/2010 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 4 Multiple Choice Topic Area Description Systems COR0022902001050D, COR0022902001080A, COR0022902001080H Standby Liquid Control System Related Lessons COR0022902 OPS STANDBY LIQUID CONTROL Related Objectives COR0022902001080H Given a SLC component manipulation, predict and explain the changes in the following: Lights and alarms Related References COR0022902 OPS STANDBY LIQUID CONTROL Related Skills (K/A) 211000.A4.07 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /

45.5 to 45.8) Lights and alarms 3.6 / 3.6

QUESTION: 23 3268 The plant is operating at 100% power when an ATWS occurs; you are given the order to inject SLC.

What indication that the Squib Valves operated correctly do you have on Panel 9-5?

a. Photohelic milliamp meters read zero.

b. White indicating lights are off.

c. Alarm 9-5-2 / G-7 Loss of continuity to the squib valves is NOT alarming.

d. SLC Pump Discharge pressure greater than Reactor pressure.

ANSWER: 23 3268

b. White indicating lights are off.

Explanation: When the SLC system pumps are started, the squib valves fire, setting off a charge that disrupts the continuity of the firing circuit. The way that this is indicated in the control room on panel 9-5 are the extinguishment of the two white lights for the squib valves.

Distracters:

a. The Photohelic milliamp meter reading zero is a good indication that the squib valves have been fired, but those meters are not available to the operator on Panel 9-5.

c. The only indication for the squib valves is the continuity lights; Loss of continuity to the squib valves alarm is a good indication but it would need to be in alarm to indicate that the squib valves have fired.

d. This indication would indicate that the SLC system pumps were running but not that the squib valves have operated correctly. Since these are positive displacement pumps they will develop a discharge pressure even if the relief valve (1400 psig setpoint) were the only flow path. The pressure instrument is upstream of the squib valve.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 400000.A3.01 Importance Rating 3.0 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-19-02 (Attach if not previously provided)

(including version/revision number) 21 Learning Objective: See Attached (As available)

Question Source: Bank # 24839 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 24 24839 01 12/07/2010 SRO: Y Operator NLO: Y Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 2 Multiple Choice Topic Area Description Systems The effect a loss of REC will have on the RWCU system Related Lessons COR0021902 REACTOR EQUIPMENT COOLING Related Objectives COR0021902001060J Given a specific REC malfunction, determine the effect on any of the following: RWCU system Related References COR0021902 REACTOR EQUIPMENT COOLING Related Skills (K/A) 400000.A3.01 Ability to monitor automatic operations of the CCWS including: (CFR: 41.7 /

45.7) Setpoints on instrument signal levels for normal operations, warnings, and trips that are applicable to the CCWS 3.0 / 3.0

QUESTION: 24 24839 The Plant is operating at near rated power, when the following occurs:

• REC system supply header (DIV I, REC-PS-452A) experiences a low pressure of 59 psig for 1 minute.

What system will be affected FIRST?

a. The Service Water MO-36 and 37 will isolate.

b. Reactor Water Cleanup system will isolate.

c. Service Air Compressors will trip.

d. SW Quad temperatures will rise.

ANSWER: 24 24839

b. RWCU system will isolate.

Explanation:

With Div I header pressure below the isolation setpoint of 61.2 psig for more than 40 seconds the MO 700 goes closed. This isolates non-critical REC loop which causes a loss of cooling non-critical loads. Loss of cooling to the RWCU NRHX will result in RWCU high temperature and RWCU isolates on NRHX high temp.

Distracters:

a. This isolation is caused when SW pressure gets low, not REC, since the SW system provides cooling to the REC Hx a candidate might mistake which low pressure is received and select the MO-36 and 37 closing.

c. The Service Air Compressors will lose REC cooling however they are provided with a TEC Backup cooling line and therefore will not trip on a loss of REC Cooling.

d. REC-MO-11 remains open so there is no affect on the quad fan cooling units.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 262001.K4.03 Importance Rating 3.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR001-01-01 Proc 2.2.18 (Attach if not previously provided)

(including version/revision number) 35 135 Learning Objective: See Attached (As available)

Question Source: Bank # 1098 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 25 1098 01 09/20/2005 SRO: Y Operator NLO: Y Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Systems COR0010102, If breaker 1FA is closed before 1AF, which of the following describes the status of 4160V bus 1F AND its supply breakers Related Lessons COR0010102 AC Electrical Distribution Related Objectives COR0010102001090B Describe the AC Electrical Distribution System design feature(s) and/or interlock(s) that provide for the following: Circuit breaker automatic trips Related References 2.2.18 4160V Auxiliary Power Distribution System Related Skills (K/A) 262001.K4.03 Knowledge of A.C. ELECTRICAL DISTRIBUTION design feature (s) and/or interlocks which provide for the following: (CFR: 41.7) Interlocks between automatic bus transfer and breakers (3.1/3.4)

QUESTION: 25 1098 Given the following conditions:

• A loss of power to 4160V Bus 1A occurred.

• 4160V Bus 1F is supplied from the Emergency Transformer.

• Power has been restored to 4160V bus 1A.

• ALL breaker control switches have been flagged to match actual breaker position.

• 4160V Bus 1F is to be transferred to bus 1A.

If breaker 1FA is closed before 1AF, which of the following describes the status of 4160V bus 1F and its supply breakers?

Bus 1F will

a. deenergize THEN reenergize from the diesel.

b. remain energized BUT breaker 1FA will trip open.

c. deenergize THEN reenergize from the emergency transformer.

d. deenergize AND remain deenergized until manual operator action is taken.

ANSWER: 25 1098

a. deenergize THEN reenergize from the diesel.

Explanation:

A caution plate on Panel C states that breaker 1AF (1BG) must be closed prior to closing breaker 1FA (1GB). There is an interlock between these two breakers to prevent back feeding 4160V bus 1A (1B) from 4160V bus 1F (1G). This is accomplished by a trip circuit for breaker 1AF (1BG) which senses its control switch in Normal-After-Close and both 1AN (1BN) and 1AS (1BS) breakers open simultaneously.

Distracters:

b. The bus de-energizes 1AF trips open.

c. Breaker 1AS trips open

d. The diesel is not isolated from the bus and will re-energize the buss on a loss of power.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 239002.A2.03 Importance Rating 4.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedures 2.4SRV, 2.1.4, 2.1.4.1, & 2.1.5 (Attach if not previously provided)

(including version/revision number) 11, 128, 32, 64 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 3397 (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 26 00 12/07/2010 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Abnormal/Emergency INT0320125A0A0400 CNS Abnormal Procedures Reactor Procedures Pressure Control Related Lessons INT0320125 CNS Abnormal Procedures (RO) Reactor Pressure Control Related Objectives INT0320125N0N0100 Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Related References 2.4SRV Stuck Open Relief Valve Related Skills (K/A) 239002.A2.03 Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) Stuck open SRV 4.1/ 4.2*

QUESTION: 26 Given the following conditions:

• Reactor power is 85%.

• Temperature down stream of Relief Valve MS-RV-71F is 310°F AND rising.

• Annunciator, RELIEF VALVE OPEN, is alarming.

• Annunciator, SUPPR POOL DIV I WATER HIGH TEMP, is alarming.

• Suppression Pool temperature is 107°F AND rising rapidly.

• Subsequent actions have been unsuccessful in closing MS-RV-71F.

What actions are now required?

a. Reduce power until feedwater flow is between 5.2 to 6.5 Mlbm/hr and then place the Mode Switch in SHUTDOWN.

b. Coordinate load reduction with Load Dispatcher. Ensure all IRM Range switches have been cycled. Ensure the Rx pressure and temperature recorders are in service.

c. Inform Ops Management, Outage Director, Load Dispatcher and the GMPO of the intent to rapidly shutdown plant. Have RE provide guidance for continuous rod insertion during shutdown. Have Chemistry/RP install SJAE off-gas rad monitor bug sources.

d. Lower core flow to 40x106 Ibs/hr and transfer 4160V Buses 1A through 1D to the Startup Transformer, if time permits. Scram the Reactor and perform the Mitigating Scram Actions.

ANSWER: 26

d. Lower core flow to 40x106 Ibs/hr and transfer 4160V Buses 1A through 1D to the Startup Transformer, if time permits. Scram the Reactor and perform the Mitigating Scram Actions.

Explanation:

Procedure 2.4SRV step 4.3 If at any time, average suppression pool temperature cannot be maintained below 110°F, SCRAM and concurrently enter Procedure 2.1.5. With temperature of the suppression pool at 107F and rising rapidly, there is enough information for the operator to conclude that 110F will not be able to be maintained.

Distracters:

a. Old steps from the rapid power reduction procedure.

b. First steps out of the current Normal Shutdown Procedure, 2.1.4.

c. First steps out of the current Rapid Shutdown Procedure, 2.1.4.1.

Modified: 3397

MODIFIED QUESTION: 3397 Given the following conditions:

• Reactor power is 85%.

• Temperature down stream of Relief Valve MS-RV-71F is 310°F AND rising.

• Annunciator, RELIEF VALVE OPEN, is alarming.

• Annunciator, SUPPR POOL DIV I WATER HIGH TEMP, is alarming.

• Suppression Pool temperature is 97 °F AND rising.

• Main generator load has lowered.

• Subsequent actions have been unsuccessful in closing MS-RV-71F.

Which of the below actions is now required?

a. Enter procedure 2.1.4, Normal Shutdown.

b. Enter procedure 2.1.4.1, Rapid Shutdown.

c. Enter procedure 2.1.5, Emergency Shutdown from Power.

d. Reduce power until feedwater flow is between 5.2 to 6.5 Mlbm/hr, THEN place the Mode Switch in SHUTDOWN.

ANSWER:

c. Enter procedure 2.1.5, Emergency Shutdown from Power.

REFERENCES:

PR 2.4SRV

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 230000.K1.01 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR0022302 Residual Heat Removal (Attach if not previously provided)

(including version/revision number) 27 Learning Objective: See Attached (As available)

Question Source: Bank # 1741 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 8 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 27 1741 01 12/08/2010 05/23/2010 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 2 Multiple Choice Topic Area Description Systems What is the purpose of the Residual Heat Removal Suppression Pool Spray mode of operation?

Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001040B Describe the interrelationship between the RHR system and the following: Suppression Pool COR0022302001060L Given an RHR control manipulation, predict and explain changes in the following: Containment parameters (pressure, temperature)

Related References COR0022302 Residual Heat Removal Related Skills (K/A) 230000.K1.01 Knowledge of the physical connections and/or cause- effect relationships between RHR/LPCI: TORUS/SUPPRESSION POOL SPRAY MODE and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Suppression pool (3.6/3.7)

QUESTION: 27 1741 How is the RHR System Suppression pool spray mode of operation initiated and what is its purpose?

a. manually initiated to reduce the Suppression Pool Temperature.

b. automatically initiated to reduce the Suppression Pool Temperature.

c. manually initiated to draw the non-condensables from the Drywell into the Torus.

d. automatically initiated draw the non-condensables from the Drywell into the Torus.

ANSWER: 27 1741

c. manually initiated to draw the non-condensables from the Drywell into the Torus.

Explanation:

From the EOP Flowchart 3A Lesson INT0080613 - Torus sprays are started between a torus pressure of 1.84 psig (high drywell pressure scram set-point) and 10 psig (Suppression Chamber Spray Initiation Pressure). Below 1.84 psig, normal methods of pressure control are to be employed. The actual setpoint of the pressure switches in the RHR logic prevent placing sprays in service until drywell pressure reaches ~ 2.5 psig.

The Suppression Chamber Spray Initiation Pressure, SCSIP, is defined to be the lowest torus pressure which can occur when 95% of the non-condensables (N2) in the drywell have been transferred to the torus. This SCSIP is used to preclude chugging (the cyclic condensation of steam at the downcomer openings of the drywell vents).

Distracters:

a. it is manually initiated, however the purpose is the transfer the non-condensables to the torus. Suppression Pool Temperature will lower as a result, but it is not the purpose of Torus Sprays.

b. it is manually initiated, however the purpose is the transfer the non-condensables to the torus. Suppression Pool Temperature will lower as a result, but it is not the purpose of Torus Sprays.

d. it is manually initiated. There are no automatic initiation functions of this mode.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 202002.K2.02 Importance Rating 2.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-22-02 (Attach if not previously provided)

(including version/revision number) 26 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 28 0 12/08/2010 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 2 Multiple Choice Topic Area Description Systems What is the power supply to the A Rx Recirc MG-Set?

Related Lessons COR0022202 Reactor Recirculation Related Objectives 10A Identify the power supplies to the following: a. RRMG set drive motors Related References COR0022202 Reactor Recirculation Related Skills (K/A) 202001.K2.02 Knowledge of electrical power supplies to the following: MG sets: Plant -

Specific (3.2 / 3.3)

QUESTION: 28 What is the normal power supply to the A Reactor Recirculation Pump Motor Generator Set?

a. 4160 VAC Bus A.

b. 4160 VAC Bus B.

c. 4160 VAC Bus C.

d. 4160 VAC Bus D.

ANSWER: 28

c. 4160 VAC Bus C.

Explanation:

The Reactor Recirc Pump Motor Generator Sets can be supplied via two power supplies, the Normal Transformer and the Startup Transformer. One pump is normally aligned to the Normal Transformer during power operation, and typically it is the A RR Pump through breaker 1CN.

The B RR Pump receives its power typically from the Startup transformer through breaker 1DS.

Distracters:

a. This bus supplies equipment like the Circ Water and Condensate Pumps, but not the Reactor Recirc Pumps.

b. This bus supplies equipment like the Circ Water and Condensate Pumps, but not the Reactor Recirc Pumps.

d. This bus supplies the B Reactor Recirc Pumps.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 219000.K3.01 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-23-02 (Attach if not previously provided)

(including version/revision number) 27 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 1752 (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 29 00 12/09/2010 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 2 Multiple Choice Topic Area Description Systems COR0022302001080A, COR0022302001150B Residual Heat Removal System Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001080A Predict the consequences a malfunction of the following will have on the RHR system: A.C. electrical power (including RPS)

COR0022302001150B Given plant conditions, determine if the following should occur: RHR pump start Related References COR0022302 RESIDUAL HEAT REMOVAL Related Skills (K/A) 219000.K3.01 Knowledge of the effect that a loss or malfunction of the RHR/LPCI:

TORUS/SUPPRESSION POOL COOLING MODE will have on following:

(CFR: 41.7 / 45.4) Suppression pool temperature control (3.9 / 4.1)

QUESTION: 29 The plant is operating at 100% power with the HPCI Full Flow Surveillance in progress and B RHR pump in Suppression Pool Cooling, maintaining torus water temperature at 93F, when the following happens:

• The B RHR Pump trips on an electrical fault.

The Operator places the Control Switch for the D RHR in the Start position within 5 seconds of the B pump tripping.

What will be the D RHR pump response and what affect will this have on Suppression Pool Temperature?

D RHR Pump Torus water temperature

a. will not start starts and continues rising

b. will not start remains constant

c. starts remains constant

d. starts starts and continues lowering ANSWER: 29

c. starts remains constant Explanation:

With the B and D pumps being in the same loop but powered off different power supplies, the electrical fault would only be experienced in the B Pump. The D pump would start right away and the valves would still be aligned at their same position before the B pump trip, so the D pump would take the place of the B pump delivering the same cooling reduction and would maintain Torus temperature the same.

Distracters:

a. D pump is not affected by the electrical fault on the G Buss.

b. D pump is not affected by the electrical fault on the G Buss. Therefore torus temperature would rise if the pump could not be started.

d. The D Pump starts but is at the same flow rate and cooling rate as the B pump, so the temperature should remain constant, not lower.

Modified 1752

MODIFIED QUESTION: 1752 B Loop of RHR is operating in Suppression Pool Cooling when the operating pump trips due to loss of power on the 4160V supply bus. What will be the RHR pump response when the bus is re-energized? (Assume no operator action).

a. None of the RHR pumps start.

b. RHR pumps A and D immediately start.

c. RHR pumps B and C immediately start.

d. All RHR pumps start after a 5 second delay.

ANSWER: 1752

a. None of the RHR pumps start.

REFERENCE:

RHR Text

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 201003.K4.08 Importance Rating 2.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-05-02 (Attach if not previously provided)

(including version/revision number) 10 Learning Objective: See Attached (As available)

Question Source: Bank # 2092 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 30 2092 00 08/12/1999 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 3 Multiple Choice Topic Area Description Systems COR0020502001020L, COR0020502001050H, COR0020502001120E Control Rod Drive Mechanisms Related Lessons COR0020502 CONTROL ROD DRIVE MECHANISM Related Objectives COR0020502001020L State the purpose of the following major CRDM components:

Position Indicator Probe Thermocouple COR0020502001050H Given the CRDM design features and/or interlocks that provide for the following: monitoring CRDM temperatures COR0020502001120E Determine the interrelationships between the CRDMs and the following: CRDM Temperature Monitor Related References COR0020502 CONTROL ROD DRIVE MECHANISM Related Skills (K/A) 201003.K4.08 Knowledge of CONTROL ROD AND DRIVE MECHANISM design feature(s) and/or interlocks which provide for the following: Monitoring CRD mechanism temperature (2.6 / 2.7)

QUESTION: 30 2092 What is the function of the Position Indicator Probe thermocouple located in each Control Rod Drive Mechanism?

The Position Indicator Probe thermocouple provides

a. remote indication of CRDM operating temperatures.

b. the signal directly to the high CRDM temperature annunciator on panel 9-5.

c. local indication of CRDM operating temperatures.

d. post-LOCA (loss of coolant accident) temperature indication for accident analysis.

ANSWER: 30 2092

a. remote indication of CRDM operating temperatures.

Explanation:

In accordance with the Student Text COR0020502, the Rod Position Information System (RPIS) provides for the following:

• The position indicating probe (PIP) supplies the input to the Rod Position Information System.

• The thermocouple located in the PIP gives indication of CRDM temperature.

Distracters:

b. The actual signal comes from the PMIS Computer, not the thermocouples directly. The thermocouples send their outputs to the PMIS computer and they are processed and displayed on any PMIS screen. Alarm ranges are assigned within the computer and a signal is sent to the Annunciators from PMIS.

c. unlike some BWRs the CRD Temperatures are not provided locally, but are sent to the control room to be processed.

d. The temperature probes are not credited for any post LOCA analysis.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 233000.K5.07 Importance Rating 2.5 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) USAR X Section 5 / Procedure 2.4FPC (Attach if not previously provided)

(including version/revision number) 2/21/07 22 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 31 00 12/15/2010 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 4 Multiple Choice Topic Area Description Systems FP Cooling - What is HIGHEST temperature expected and what are the implications?

Related Lessons COR0010602 FUEL POOL COOLING Related Objectives COR0010602001100D Briefly describe the following concepts as they apply to FPC: Heat loading Related References USAR X Section 5 PROC 2.4FPC.

Related Skills (K/A) 233000.K5.07 Knowledge of the operational implications of the following concepts as they apply to FUEL POOL COOLING AND CLEAN-UP: (CFR: 41.5 / 45.3)

Maximum (abnormal) heat 102d load (2.5 / 2.8)

QUESTION: 31 What is HIGHEST fuel pool temperature expected under design heat load and what are the implications for exceeding that temperature?

a. 125 °F; RHR sub-system would have to be placed in service to support cooling.

b. 125 °F; RWCU system would have to be placed in service to support cooling.

c. 150 °F; RHR sub-system would have to be placed in service to support cooling.

d. 150 °F; RWCU system would have to be placed in service to support cooling.

ANSWER: 31

c. 150 °F; RHR sub-system would have to be placed in service to support cooling.

Explanation:

USAR - X Section 5: During normal refueling operations, the maximum expected spent fuel pool temperature of 150°F results from the decay heat of the full core load of fuel at the end of the fuel cycle plus the remaining decay heat of the spent fuel discharged at previous refuelings. Prior to the spent fuel pool reaching this temperature, the Residual Heat Removal system is manually aligned to operate in conjunction with the fuel pool cooling and demineralizer system to reduce the spent fuel pool temperature and maintain it at or below 150°F.

Distracters:

a. This is the entry condition for a loss of fuel pool cooling. But RHR is the eventual supplemental cooling system.

b. This is the entry condition for a loss of fuel pool cooling. And RWCU is not the eventual supplemental cooling system.

d. This is the correct temperature, but RWCU is not the supplemental cooling system.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 214000.K6.01 Importance Rating 2.5 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 5.3NBPP Lesson COR002-20-02 (Attach if not previously provided)

(including version/revision number) 12 20 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 2107 (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 32 00 12/15/2010 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 3 Multiple Choice Topic Area Description Systems What does RPIS indication look like on Panel 9-5 on a loss of NBPP?

Related Lessons COR0022002 OPS REACTOR MANUAL CONTROL SYSTEM Related Objectives COR0022002001100D State the electrical power supplies to the following: RPIS COR0022002001050C Predict the consequences the following would have on the RMCS and/or RPIS: RPIS failure Related References COR0022002 OPS REACTOR MANUAL CONTROL SYSTEM PROC. 5 .3NBPP Related Skills (K/A) 214000.K6.01 Knowledge of the effect that a loss or malfunction of the following will have on the ROD POSITION INFORMATION SYSTEM: (CFR: 41.7 / 45.7) A.C.

electrical power (2.5 / 2.6)

QUESTION: 32 The Plant is operating at approximately 75% with Control Rod 26-27 selected.

• There is a loss of the No Break Power Panel (NBPP).

What is the expected condition of the Control Rod Position Indication System (Full Core Display) and the indication displayed in the four-rod display on the vertical Panel 9-5?

Full Core 4 Rod Display Display

a. No Lights No position

b. Lights No Position

c. Lights Position Indicated

d. No Lights Position Indicated ANSWER: 32

a. No Lights No position Explanation:

The No Break Power System powers all indications for Control Rod Positions. Both the red and green lights on the full core display and the numeric indications located on the four rod display.

From 5.3NBPP - A loss of NBPP causes the following Distracters:

b. There are no red and green lights on the Full Core Display.

c. There are no red and green lights on the Full Core Display.

d. There are no numerals displayed in the four rod display.

Modified: 2107

MODIFIED QUESTION: 2107 As the reactor operator, you note a loss of control rod position indication on the four-rod display and suspect a failure of the Rod Position and Information System's power supply. What electrical distribution panel should you instruct your station operator to check?

a. Critical Power Panel

b. 125 VDC Distribution Panel

c. No Break Power Panel

d. Reactor Protection System Power Panel ANSWER:

c. No Break Power Panel

REFERENCE:

Reactor Manual Control System Text

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 241000.A1.06 Importance Rating 3.2 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-09-02 (Attach if not previously provided)

(including version/revision number) 16 Learning Objective: See Attached (As available)

Question Source: Bank # 2399 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 33 2399 02 12/16/2010 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 3 Multiple Choice Topic Area Description Systems What happens when a bypass opens at 70% power?

Related Lessons COr0020702001080B Digital Electro-Hydraulic Control Related Objectives COR0020902001070D Given a specific DEH Control system malfunction, determine the effect on any of the following: Main turbine steam flow Related References COR0020702 Digital Electro-Hydraulic Control Related Skills (K/A) 241000.A1.06 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR/TURBINE PRESSURE REGULATING SYSTEM controls including: (CFR: 41.5 / 45.5) Main turbine steam flow (3.2 / 3.2)

QUESTION: 33 2399 The Plant is operating at 70% power with DEH in mode 4, when the following event occurs:

• One bypass valve partially opens What is the expected response on Steam Flow and overall plant?

a. Steam flow increase, steam header pressure decrease, Group I isolation on low Rx.

Pressure.

b. Steam flow decrease, steam header pressure increase, scram on Rx. high pressure.

c. Steam flow increase, steam header pressure maintained by governor valves, generator output increase.

d. Steam flow and steam header pressure remain constant, generator output decrease.

ANSWER: 33 2399

d. Steam flow and steam header pressure remain constant, generator output decrease.

Explanation:

The LOAD CONTROL mode is entered when the first of either generator output breakers is closed (3310 or 3312). DEH is normally in this mode for only a brief period of time as generator output is automatically ramped up until the turbine BPVs close. In this case, the bypass valve opens thus transferring DEH back into MODE 3. The turbine BPVs which are controlling pressure, if they open, throttle pressure lowers and the Governor Valves close down to maintain throttle pressure. Total Steam Flow (that is going through the Governor valves and the bypass valve) remain constant, however generator load decreases as the governor valves close.

Distracters:

a. Total steam flow remains constant as the governor valve close down to compensate for the open bypass valve. Therefore maintaining Reactor Pressure constant. The group 1 would only occur if pressure continued to lower, which would be the case if the governor valves did not reposition.

b. Total steam flow remains constant as the governor valve close down to compensate for the open bypass valve. Therefore maintaining Reactor Pressure constant. If the bypass valve were open then closed and the governor valves did not respond pressure would increase and a reactor scram on high pressure is possible.

c. Total steam flow remains constant as the governor valve close down to compensate for the open bypass valve. Therefore maintaining Reactor Pressure constant. When the governor valves reposition it robs steam from the Main Turbine and generator output lowers a corresponding amount.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 215001.A2.07 Importance Rating 3.4 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-31-02 (Attach if not previously provided)

(including version/revision number) 14 Learning Objective: See Attached (As available)

Question Source: Bank # 21332 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 34 21332 00 02/08/2005 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 5 Multiple Choice Topic Area Description Systems COR0023102, TIP Loss of Power Related Lessons COR0023102 OPS TRAVERSING IN-CORE PROBE Related Objectives COR0023102001140B Predict the consequences of the following on the TIP system: A.C.

Electrical power failure Related References 4.1.4 Traversing In-Core Probe System Related Skills (K/A) 215001.A2.07 Ability to (a) predict the impacts of the following on the TRAVERSING IN-CORE PROBE; and (b) based on those predictions, use procedures to correct...: (CFR: 41.5 / 45.6) ?Failure to retract during accident conditions:

Mark-I&II(Not-BWR1) (3.4/3.7)

QUESTION: 34 21332 (1 point(s))

The plant is operating at power with TIP "A" near the core top limit when annunciator C-4/F-6, CRIT INST & CONT PNL CPP LOSS OF VOLT alarms. While investigating the cause of the loss of CPP, a reactor scram occurs due to high drywell pressure. Drywell pressure is 5 psig and rising.

The operator refers to procedure 4.1.4 TRAVERSING IN-CORE PROBE SYSTEM. What TIP action(s) is/are required in order to isolate the TIP tube? (Choose the answer that contains ONLY required TIP actions).

a. Place TIP "A" MANUAL switch to REV only.

b. Place the key lock switch for TIP "A" shear valve to FIRE.

c. Place TIP "A" MAN. VALVE CONTROL switch to CLOSED only.

d. Place TIP "A" MANUAL switch to REV and place MAN. VALVE CONTROL switch to CLOSED.

ANSWER: 34 21332

b. Place the key lock switch for TIP "A" shear valve to FIRE.

Explanation:

The loss of CPP results in the loss of TIP instrumentation and controls. Since an isolation is required and the TIP cannot be withdrawn from the core the shear valve should be fired.

Distracters:

a. If control power were still available then this would withdraw the TIP into the shield if the Ball Valve control switch is in the closed position. Since there is no control power this option will not work.

c. If the TIP is still past the ball valve the valve would close on the cable and not isolate the line, the TIP must be withdrawn first. Since there is no control power the TIP Drive and valves have lost power.

d. If control power were still available then this would withdraw the TIP into the shield and closing the Ball Valve would isolate the line. Since there is no control power this option will not work.

Psychometric Review: GPJ 2004 Biennial Exam 04LORRO01 Question # 21055

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 256000.A3.05 Importance Rating 3.0 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-02-02 (Attach if not previously provided)

(including version/revision number) 30 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 3963 (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 35 00 12/16/2010 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 4 Multiple Choice Topic Area Description Systems COR0020202, Which Cond & Booster pumps are running after power loss Related Lessons COR0020202 OPS CONDENSATE AND FEEDWATER Related Objectives COR0020202001080E Predict the consequences a malfunction of the following would have on the Condensate and Feedwater system: AC Power Related References COR0020202 OPS CONDENSATE AND FEEDWATER Related Skills (K/A) 256000.A3.05 Ability to monitor automatic operations of the REACTOR CONDENSATE SYSTEM including: (CFR: 41.7 / 45.7) Lights and alarms (3.0 / 2.9)

QUESTION: 35 The plant is at 100% power when a LOSS of both the Normal and Startup Transformers occurs.

What do the Condensate Pump indications show?

Condensate Pump Amps Meters Indicating lights

a. 80 Amps Green Light off, Red light on

b. 80 Amps Green Light on, Red light off

c. 0 Amps Green Light off, Red light on

d. 0 Amps Green Light off, Red light off ANSWER: 35

c. 0 Amps Green Light off, Red light on Explanation:

On a loss of 4160 Volt buses feeding the Condensate Pumps, the pumps will stop operating, however they do not have an under voltage trip so the red breaker closed light will remain on and the current will be 0 amps.

Distracters:

a. The 4160 Bus that feeds the Condensate pumps will be de-energized so there should be no current read on the Pump Amp Meters.

b. The 4160 Bus that feeds the Condensate pumps will be de-energized so there should be no current read on the Pump Amp Meters.

d. The 4160 Bus that feeds the Condensate pumps will be de-energized but there are no under voltage trips on the breakers feeding them, so the red light should remain on. The breaker position indication comes from 125 VDC power not from the breaker. If the candidate believed that the indication came from the breakers power they might think that the lights would be off.

Modified - 3963

MODIFIED QUESTION: 3963 The plant is at 75% power when a LOSS of the Normal Station Transformer occurs.

• EXCEPT for Bus 1B, ALL 4160 VAC buses transfer successfully.

• Breaker 1BS fails to close automatically.

• One (1) second later, breaker 1BS is closed by the BOP operator AND energizes bus 1B.

What is the state (running or stopped) of the Condensate (COND) and Condensate Booster (CB) pumps powered from Bus 1B just after Bus 1B is energized?

COND PUMP CB PUMP

a. RUNNING RUNNING

b. RUNNING STOPPED

c. STOPPED RUNNING

d. STOPPED STOPPED ANSWER:

a. RUNNING RUNNING EXPLANATION OF ANSWER: a. Correct. CP and CBP breakers do not have UV protection and will remain closed upon loss of power. Pumps will coast down during the power loss and re-start when the bus is re-energized.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 290001.A4.01 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedure 2.3_R-2 (Attach if not previously provided)

(including version/revision number) 14 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 36 00 01/11/2011 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 4 Multiple Choice Topic Area Description Systems Deviation from normal Secondary Containment pressure Related Lessons COR0020302 OPS CONTAINMENT Related Objectives Related References Procedure 2.3_R-2 Related Skills (K/A) 290001.A4.01 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /

45.5 to 45.8) Reactor building differential pressure: Plant-Specific (3.3 /

3.4)

QUESTION: 36 The Reactor is operating at 100% power when the Reactor Operator notes that Reactor Building differential pressure is -0.36 inches of water.

Is this pressure better or worse than the normal pressure and what is the trip setpoint for the reactor building fans? (Neglect the time delay)

a. better; - 0.15 inches of water.

b. better; + 0.15 inches of water.

c. worse; - 0.15 inches of water.

d. worse; + 0.15 inches of water.

ANSWER: 36

a. better; - 0.15 inches of water.

Explanation:

Normal differential pressure for the reactor building is - 0.30 to -0.33 inches of water.

- 0.36 inches is more negative and therefore a better pressure to maintain. The trip setpoint for the Reactor Building fans is - 0.15 inches of water in accordance with annunciator procedure 2.3_R-2, window R-2/A-4. SETPOINT (4703) -0.15" wg. Or -0.45 inches of water.

Distracters:

b. even though the pressure is better, the fans trip on a negative pressure of 0.15 not a positive pressure.

c. a slightly more negative pressure is better than a more positive pressure.

d. a slightly more negative pressure is better than a more positive pressure, and the trip setpoint is not +0.15.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 272000 G 2.4.50 Importance Rating 4.2 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 2.3_9.4.1 Ann. Card COR001-18-2 (Attach if not previously provided)

(including version/revision number) 39 21 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 37 New 00 5/11/2011 N/A SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 5 Multiple Choice Topic Area Description Systems During Start up and the MSL RM trip what actions are required by the Operator if they do not occur automatically?

Related Lessons COR0011802 Radiation Monitoring Related Objectives COR0011802001050N Describe the interrelationship between the RM system and the following: Primary containment isolation system Related References 2.3_9.4.1 Annunciator Card Related Skills (K/A) 2.4.50 Ability to verify system alarm setpoints and operate controls identified in l the alarm response manual. (CFR: 41.10 / 43.5 / 45.3) IMPORTANCE RO 4.2 SRO 4.0

QUESTION: 37 The plant is starting up and is operating in the IRM ranges below 1% power with the mechanical vacuum pumps drawing a vacuum on the main condenser. Chemistry is using the Recirc Sample points for vessel chemistry sampling requirements. The following occur in sequence:

• The Main Steam Line Rad Monitor starts and continues to rise due to a dropped rod and fuel failure.

• Annunciator 9-4-1/A-5 MAIN STM LINE HI RAD alarms.

• Annunciator 9-4-1/A-4 MAIN STM LINE HI HI RAD alarms.

• The Annunciator display screen shows the following conditions.

o 1785 Main Stm Line Chan A Hi Hi Rad o 1787 Main Stm Line Chan B Hi Hi Rad o 1786 Main Stm Line Chan C Hi Hi Rad o 1788 Main Stm Line Chan D Hi Hi Rad The Reactor Operator notes that none of the Automatic Actions have occurred.

What actions are required by Annunciator Procedure 9-4-1/A-5 MAIN STM LINE HI RAD?

a. Close the RR-AO-740 and RR-AO-741 valves only.

b. Close the RR-AO-740 and RR-AO-741 valves and trip the Mechanical Vacuum Pump and ensure the isolation valves close.

c. Close the Main Steam Line Isolation Valves (MSIVs) only.

d. Close the Main Steam Line Isolation Valves (MSIVs) and trip the Mechanical Vacuum Pump and ensure the isolation valves close.

ANSWER: 16 2429

b. Close the RR-AO-740 and RR-AO-741 valves and trip the Mechanical Vacuum Pump and ensure the isolation valves close.

Explanation:

From Annunciator procedure 9-4-1/A-4 MAIN STM LINE HI HI RAD, page 9

1. AUTOMATIC ACTIONS 1.1 If Logic A trips, RR-AO-741, INBD ISOL VLV, closes.

1.2 If Logic B trips, RR-AO-740, OUTBD ISOL VLV, closes.

1.3 If both Logics RR-AO-740 and RR-AO-741 close, mechanical vacuum pumps trip and associated inlet and outlet valves close.

2. OPERATOR OBSERVATION AND ACTION 2.1 Concurrently enter Procedures 2.1.22 and 5.2FUEL.

The Annunciator system indicates that all four MSL Rad Monitors have tripped, initiating a Group 7 Sample System Isolation; and the Mechanical Vacuum pumps should also have tripped and isolated. Since no Automatic actions took place, the Operator is required to take those actions.

From COR001-18-2 Operation Radiation Monitoring Lesson D. Basic System Operation Four channels monitor the gamma radiation from the main steam lines. The detectors are physically located near the main steam lines downstream of the outboard main steam line isolation valves. The detectors are geometrically arranged to allow the earliest practical detection of a gross fuel failure.

The system is divided into two divisions with two channels in each. Channels A and C are one division and are powered from REACTOR PROTECTION SYSTEM (RPS) bus A. Channels B and D are the other division and are powered from RPS bus B.

Trip signals are transmitted to the PRIMARY CONTAINMENT ISOLATION SYSTEM (PCIS)

(Group 7), and to the air removal system. Upon receipt of the high radiation trip signals, PCIS initiates closure of reactor water sample valves; and operates secondary relays in the air removal system that close the mechanical vacuum pump inlet and outlet valves and trip the mechanical vacuum pumps.

The radiation trip setting is selected so that a design basis rod drop accident results in a trip. The selected setting is high enough that background radiation level in the vicinity of the main steam lines do not cause spurious trips at rated power. Yet, the setting is low enough that the monitors respond to the fission products released during the design basis rod drop accident and maintain the radioactive materials released to the environs within the limits of 10CFR100.

Four instrumentation channels are used to provide a resistance to inadvertent isolation as a result of instrumentation malfunction. The output trip signals of each monitoring channel are combined in a one out of two, taken twice logic; consistent with RPS requirements. Thus, failure of any one monitoring channel does not result in inadvertent isolation.

Each channel has four trip circuits as follows:

1. Alarm trip (high): 1 2 times normal full power background.

2. Isolation trip (high-high): 3 times normal full power background.

3. Downscale: 1.05 mr/hr.

4. Inop (isolation trip): Keylock to INOP, or hardware failure detected.

The trip circuits for each monitoring channel energize relay when the trip is not active. When a trip setpoint is reached, the relays will de-energize and cause the associated action so that failures in which power to monitoring components is interrupted, result in a trip signal. The environmental capabilities of the components of each monitoring channel are selected in consideration of the locations in which the components are to be placed.

Distracters:

a. Since the Mechanical Vacuum Pump is in service and all four MSL Rad Monitors tripped, the MVP should have tripped and isolated.

c. The Main Steam Line Isolation Valves are no longer on the group 7 signal and remain open. The 5.2FUEL procedure will direct their closure once the reactor is scrammed and the plant is shutdown. Also the Mechanical Vacuum Pump should also be tripped.

d. The Main Steam Line Isolation Valves are no longer on the group 7 signal and remain open. The 5.2FUEL procedure will direct their closure once the reactor is scrammed and the plant is shutdown.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 216000.K3.29 Importance Rating 3.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-15-02 & COR002-22-02 (Attach if not previously provided)

(including version/revision number) 24 26 Learning Objective: See Attached (As available)

Question Source: Bank # 1063 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 38 1063 00 06/22/1999 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Systems COR0022202, REACTOR RECIRCULATION Related Lessons COR0021502 OPS NUCLEAR BOILER INSTRUMENTATION COR0022202 REACTOR RECIRCULATION SKL0124222 OPS REACTOR RECIRCULATION SYSTEM Related Objectives COR0021502001060K Given a specific NBI malfunction, determine effect on any of the following: Core flow/Jet Pump monitoring COR0022202001060B Given a specific Reactor Recirculation system or the Recirculation Flow Control system malfunction, determine the effect on any of the following: Core Flow (normal and reduced forced flow conditions)

SKL012422200A030E Given plant conditions, predict changes in the following Reactor Recirculation System components/parameters: Core flow Related References 2.2.68.1 Reactor Recirculation System Operations COR002-02-22 Nuclear Boiler Instrumentation Related Skills (K/A) 202002.K3.01 Knowledge of the effect that a loss or malfunction of the RECIRCULATION FLOW CONTROL SYSTEM will have on following: (CFR: 41.7 / 45.4)

Core flow (3.5/3.5) 216000.K3.29 Knowledge of the effect that a loss or malfunction of the NUCLEAR BOILER Instrumentation will have on following: (CFR: 41.7 / 45.4) K3.29 Jet pump flow monitoring: Plant-Specific (3.1 / 3.2)

QUESTION: 38 1063 Given the following conditions:

• The "B" Recirculation Pump has tripped.

• MO-53B, "B" Recirculation Pump discharge valve was closed and is now open.

• LOOP B JET PUMP FLOW (FI-92B) indicates 2 Mlbm/hr.

• LOOP A JET PUMP FLOW (FI-92A) indicates 35 Mlbm/hr.

• Annunciator E-7 on Panel 9-4-3, RECIRC LOOP B OUT OF SERVICE is NOT alarming.

What are the expected values for indicated Total Core Flow as indicated on Panel 9-5 Recorder NBI-FR/DPR-95 AND what is Actual Core Flow?

Indicated Total Actual Core Core Flow Flow

a. 37 Mlbm/hr 33 Mlbm/hr

b. 37 Mlbm/hr 37 Mlbm/hr

c. 33 Mlbm/hr 33 Mlbm/hr

d. 33 Mlbm/hr 37 Mlbm/hr ANSWER: 38 1063

a. 37 Mlbm/hr 33 Mlbm/hr Explanation:

With one Recirc Pump out of service, a reverse flow will exist through the idle Jet Pumps but the Jet Pump Flow instrumentation will indicate a positive flow. Annunciator E-7 not in alarm indicates Core Flow circuitry is not functioning properly for single loop (i.e. The Loop Jet Pump flows are being added vice subtracted). Total Core Flow will indicate 37 on dPD/FR-95. Since reverse flow exists in the idle loop, Actual core flow will be the difference between Loop A & B Jet Pump flows.

Distracters:

b. With the Core Flow summing circuit malfunctioning, indicated Total Core Flow will be the sum of Loop A & B Jet Pump flows (37) and Actual Core Flow will be the difference between Loop A & B Jet Pump Flows (33).

c. With the Core Flow summing circuit malfunctioning, indicated Total Core Flow will be the sum of Loop A & B Jet Pump flows (37) and Actual Core Flow will be the difference between Loop A & B Jet Pump Flows (33).

d. With the Core Flow summing circuit malfunctioning, indicated Total Core Flow will be the sum of Loop A & B Jet Pump flows (37) and Actual Core Flow will be the difference between Loop A & B Jet Pump Flows (33).

COR002-22-02, PR 2.2.68.1 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295028.EK1.01 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) PSTGs (Attach if not previously provided)

(including version/revision number)

Learning Objective: See Attached (As available)

Question Source: Bank # 24832 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 14 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 39 24832 00 07/18/2009 SRO: Y Operator NLO: Y Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 2 Multiple Choice Topic Area Description Integrated Plant EOP 3A, why monitor DW temperature Related Lessons COR0021502 OPS NUCLEAR BOILER INSTRUMENTATION INT0080618 OPS EOP AND SAG GRAPHS AND CAUTIONS Related Objectives INT00806180020400 Using the Cautions provided in the EOP and SAG Flowcharts, explain the bases behind each of the Cautions.

COR0021502001040F Briefly describe the following concepts as they apply to NBI:

Elevated containment temperature effects on level indication COR0021502001050G Predict the consequences of the following items on the NBI:

Elevated containment temperature Related References NONE Related Skills (K/A) 295028.EK1.01 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE: CFR: 41.8 to 41.10) Reactor water level measurement (3.5/3.7)

QUESTION: 39 24832 Which one of the following is a reason why Drywell Temperature is monitored and controlled by EOP-3A, PRIMARY CONTAINMENT CONTROL?

a. Ensure NPSH limits for ECCS pumps are not exceeded.

b. Verify proper operation of the Drywell Hydrogen detectors.

c. Prevent or minimize inaccurate indications of RPV pressure instruments.

d. Prevent or minimize inaccurate indications of RPV water level instruments.

ANSWER: 39 24832

d. Prevent or minimize inaccurate indications of RPV water level instruments.

Explanation:

In accordance with the EPGs and the EOP Bases, RPV water level indications may be unreliable or must be considered invalid due to the effects of increased Drywell temperatures.

Distracters:

a. NPSH limit for ECCS pumps is ensured by Torus minimum level.

b. Drywell temperature control is not a reason operation of the Hydrogen detectors is verified.

c. Torus pool temperature is monitored, not air temperature.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295024.EK1.01 Importance Rating 4.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson INT008-06-13 ; INT008-06-18 (Attach if not previously provided)

(including version/revision number) 14 18 Learning Objective: See Attached (As available)

Question Source: Bank # 19220 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 40 19220 01 07/28/2003 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 4 Multiple Choice Topic Area Description Emergency Operating INT0080618, Basis for PSP ED Procedures Related Lessons INT0080613 OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL INT0080618 OPS EOP AND SAG GRAPHS AND CAUTIONS Related Objectives INT00806130011200 Given plant conditions and EOP flowchart 3A, PRIMARY CONTAINMENT CONTROL, state the reasons for the actions contained in the steps.

INT0080613001040C State the basis for primary containment control actions as they apply to the following: Graphs reference on Flowchart 3A INT00806180010200 For each graph used in the flowcharts, identify the action(s) required if the parameters associated indicate operation in the restricted or prohibited area.

Related References INT0080613 OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL INT0080618 OPS EOP AND SAG GRAPHS AND CAUTIONS Related Skills (K/A) 295024.EK1.01 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE: (CFR: 41.8 to 41.10) Drywell integrity: Plant-Specific (4.1/ 4.2*)

QUESTION: 40 19220 Why is the RPV Emergency Depressurized if Pressure Suppression Pressure (EOP Graph 10) is exceeded?

a. Failure of primary containment may occur if drywell sprays are initiated.

b. Failure of primary containment may occur if a primary system rupture develops.

c. Failure of SRV Tailpipes may occur due to steam bypassing the suppression pool.

d. Failure of SRV Tailpipes may occur due to inadequate differential pressure across the balancing disc.

ANSWER: 40 19220

b. Failure of primary containment may occur if a primary system rupture develops.

Explanation:

INT008-06-13 If primary containment water level rises above 16.5 ft, a return flowpath for non-condensibles from the torus to the drywell through the vacuum breakers may be prevented. Drywell sprays must therefore be stopped. If adequate core cooling cannot be assured (unless combustible gas concentrations exist in the primary containment), drywell sprays must be stopped.

PC/P If torus or drywell sprays could not be started or if their operation was not effective in reducing primary containment pressure, the RPV is depressurized when torus pressure cannot be maintained below the Pressure Suppression Pressure, PSP. The Alarm Statement following step PC/P-4 is needed to alert the CRS that an override in Flowchart 1A (or Flowchart 6A/7A for failure-to-scram events) must be reviewed. RPV depressurization minimizes further release of energy from the RPV to the primary containment. See INT008-06-18, GRAP10 for discussion of the PSP.

INT008-06-18 Pressure Suppression Pressure (GRAP10)

Definition - The Pressure Suppression Pressure, PSP, is the highest torus pressure which can be maintained without steam in the torus airspace.

Use - The PSP is a function of primary containment water level. It is utilized in the EOPs to ensure that pressure suppression capability sufficient to accommodate emergency depressurization is maintained while the RPV is at pressure. It is utilized in the SAGs to ensure that pressure suppression capability sufficient to accommodate a low pressure release of core debris is maintained when RPV breach by core debris is anticipated.

Flowchart 3A step PC/P-4 requires emergency RPV depressurization when torus pressure cannot be maintained within the PSP.

If RPV water level is below the bottom of active fuel and RPV injection flow is below MDRIR, SAG 2 Strategy E is performed based on the ability to maintain parameters within the PSP; Strategy F, if unable to maintain parameter within PSP.

PSTG Appendix B page 6-34 Rev. 3 PSTG/SATG Step PC/P-3 When suppression chamber pressure cannot be maintained below the Pressure Suppression Pressure, EMERGENCY RPV DEPRESSURIZATION IS REQUIRED.

Discussion If suppression pool and/or drywell sprays cannot be initiated or are ineffective in reversing the increasing trend of primary containment pressure, as evidenced by not being able to maintain suppression chamber pressure below the Pressure Suppression Pressure, the RPV is depressurized to minimize further release of energy from the RPV to the primary containment.

This action serves to terminate, or reduce as much as possible, any continued primary containment pressure increase.

The Pressure Suppression Pressure (PSP) is the lesser of:

• The highest suppression chamber pressure which can occur without steam in the suppression chamber airspace. The CNS PSP is limited by this concern.

• The highest suppression chamber pressure at which initiation of RPV depressurization will not result in exceeding Primary Containment Pressure Limit A before RPV pressure drops to the Decay Heat Removal Pressure.

• The highest suppression chamber pressure which can be maintained without exceeding the suppression pool boundary design load if SRVs are opened.

This pressure is a function of suppression pool water level. It is used in the PSTGs to ensure that pressure suppression capability sufficient to accommodate emergency RPV depressurization is maintained while the RPV is at pressure. Refer to Section 16 of this appendix for a detailed discussion of the PSP.

No explicit direction to enter the RPV Control guideline is included in this step since drywell pressure above the scram setpoint is a defined RPV Control entry condition.

16.24 Pressure Suppression Pressure The Pressure Suppression Pressure (PSP) is the lesser of:

• The highest suppression chamber pressure which can occur without steam in the suppression chamber airspace. (This concern defines the CNS PSP).

• The highest suppression chamber pressure at which initiation of RPV depressurization will not result in exceeding Primary Containment Pressure Limit A before RPV pressure drops to the Decay Heat Removal Pressure.

• The highest suppression chamber pressure which can be maintained without exceeding the suppression pool boundary design load if SRVs are opened.

The PSP is a function of primary containment water level. It is utilized in the PSTGs to ensure that pressure suppression capability sufficient to accommodate emergency depressurization is maintained while the RPV is at pressure. It is utilized in the SATGs to ensure that pressure suppression capability sufficient to accommodate a low pressure release of core debris is maintained when RPV breach by core debris is anticipated.

The derivation of the PSP is shown graphically in Figure B-16-19.

Line 1 is the suppression pool water level corresponding to the elevation of the downcomer vent openings. If suppression pool water level is below this elevation, the RPV may not be kept in a pressurized state since steam discharged through the vents may not be condensed.

Line 2 is the maximum suppression pool water level corresponding to the Maximum Pressure Suppression Primary Containment Water Level (MPSPCWL) defined as the bottom of the ring header. Above this elevation, the pressure suppression capability of the primary containment may be insufficient to accommodate an RPV breach by core debris. The PSP is therefore vertical at this elevation.

Line 3 corresponds to the highest suppression chamber pressure which can occur without steam in the suppression chamber airspace. This pressure is determined by calculating the pressure that would exist as a function of suppression pool water level with all drywell noncondensibles purged to the suppression chamber and suppression pool temperature at the Heat Capacity Temperature Limit corresponding to the lowest SRV lift pressure. Higher suppression pool water levels result in higher pressures since the airspace volume is smaller.

Line 4 corresponds to the highest suppression chamber pressure from which an emergency depressurization will not raise suppression chamber pressure above Primary Containment Pressure Limit A before RPV pressure drops to the Decay Heat Removal Pressure. This curve is calculated by subtracting the rise in suppression chamber pressure during blowdown from Primary Containment Pressure Limit A. The calculation assumes the blowdown is initiated at the lowest SRV lift pressure and compensates for changes in suppression pool heat capacity with changes in suppression pool water level (as defined by the Heat Capacity Temperature Limit). As suppression pool water level increases, a larger heat sink is available to absorb blowdown energy. Consequently, the difference in suppression pool temperature before and after the blowdown decreases, causing the rise in suppression chamber pressure to decrease. Since Primary

Containment Pressure Limit A is constant in this range, Line 4 rises with increasing suppression pool water level.

Line 5 corresponds to the highest suppression chamber pressure at which SRVs can be opened without exceeding the suppression pool boundary design load. This curve is the suppression pool boundary design pressure less (1) the suppression pool boundary loads imposed by SRV actuation and (2) the hydrostatic head between the suppression pool water level and the level assumed in the design calculation.

The PSP is thus the envelope defined by Lines 1 and 2 and the most limiting values of Lines 3, 4 and 5. As shown in Figure 2, Line 3 is most limiting over the acceptable range of suppression pool water levels.

The PSP is determined assuming:

1. The highest suppression chamber pressure which can occur without steam in the suppression chamber air space is the lowest suppression chamber pressure which results when all the noncondensibles in the containment are in the suppression chamber and the suppression pool temperature is at the Heat Capacity Temperature Limit temperature which corresponds to the lowest SRV lift setpoint pressure.

2. The highest suppression chamber pressure at which initiation of RPV depressurization will not result in exceeding Primary Containment Pressure Limit A before RPV pressure drops to the Decay Heat Removal Pressure is the Primary Containment Pressure Limit less the maximum suppression chamber airspace pressure increase which results from depressurizing the RPV from the lowest SRV lift setpoint pressure to the Decay Heat Removal Pressure, compensating for change in suppression pool heat capacity with change in primary containment water level.

3. The highest suppression chamber pressure which can be maintained without exceeding the suppression pool boundary design load if SRVs are opened is the suppression pool boundary design pressure less (1) the maximum suppression pool boundary load which results from any actuation and (2) the hydrostatic head between the primary containment water level and the suppression pool water level which was assumed in determining the maximum suppression pool boundary load which results from SRV actuation.

4. Suppression pool cooling is unavailable.

5. The suppression pool and suppression chamber atmosphere are in thermal equilibrium.

CNS input data required to calculate the PSP are:

1. Primary Containment Pressure Limit A.

2. Heat Capacity Temperature Limit.

3. Maximum Pressure Suppression Primary Containment Water Level.

4. Suppression pool boundary design load.

5. Maximum suppression pool boundary load which results from SRV actuation and the suppression pool water level which was assumed in determining this load.

6. Elevation of the suppression pool water level instrument zero.

7. Elevation of the downcomer openings.

8. Elevation of the suppression chamber pressure instrument tap.

9. Free volume of the suppression chamber airspace as a function of suppression pool water level.

10. Free volume of the drywell and vent system.

11. Maximum normal operating drywell and suppression chamber temperatures.

12. Minimum normal operating drywell and suppression chamber pressures.

13. Minimum temperature of the suppression pool when the containment is flooded.

The PSP is referenced in the following PSTG/SATG steps:

• PSTG Step PC/P-3

• Decision box before SATG Step RC/F-1

• First, third and fourth overrides at the beginning of SATG Step RC/F-5

• Third and fourth overrides at the beginning of SATG Step RC/F-6 The CNS PSP is illustrated in Figure B-16-20.

a

Psc-hp2 Line 3 Psc-srv Suppression Chamber Pressure (psig)

Line 4 Line 5 Psc-hp1 Line 2 Line 1 Suppression Pool Water Level (ft)

Figure B-16-19: Pressure Suppression Pressure Derivation Figure B-16-20: CNS Pressure Suppression Pressure

Distracters:

a. PSP is not based on drywell spray restrictions (DWSIL is met).

c. SRV Tail Pipe Level Limit is based on the amount of stresses placed on the tailpipe and tee-quenchers when a larger than normal water slug is forced through the system and out of the holes in the tee-quencher. The right hand line on the graph bounds this. Not by a bypass of steam into the air space, this is the result of some type of breach in the suppression function boundary.

d. SRV Tail Pipe Level Limit is based on the amount of stresses placed on the tailpipe and tee-quenchers when a larger than normal water slug is forced through the system and out of the holes in the tee-quencher. The right hand line on the graph bounds this. Not by the balancing disc (Primary Containment Pressure Limit is met). This is a cause and affect relationship between primary containment and the downstream side of the pilot disc in the SRVs.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295023.AK1.03 Importance Rating 3.7 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) USAR XIV (Attach if not previously provided)

(including version/revision number) 2/5/10 Learning Objective: See Attached (As available)

Question Source: Bank # 19212 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC CNS 2002 Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 41 19212 01 03/17/2007 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 4 Multiple Choice Topic Area Description Systems State the purpose of the refueling interlocks.

Related Lessons COR001-21-01 OPS Refueling Related Objectives

6. State the purpose of the refueling interlocks.

Related References USAR VII - 6.0 REFUELING INTERLOCKS Related Skills (K/A) 295023.AK1.03 Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS: (CFR: 41.8 to 41.10) Inadvertent criticality. (3.7/4.0)

QUESTION: 41 19212 What is the basis for the refueling interlocks associated with the Reactor Mode Switch in the REFUEL position?

a. To prevent criticality during refueling by ensuring that fuel assemblies are not loaded into the core unless all control rods are fully inserted.

b. To prevent control rod/fuel assembly configurations during refueling resulting in fuel enthalpy above 280 cal/gm by preventing multiple control rod withdrawal.

c. To ensure that fuel assembly loading sequence and configurations are restricted to maintain an adequate shutdown margin by monitoring control rod position and refueling grapple load status.

d. To ensure that radioactivity releases as a result of a refueling accident are maintained below a small fraction of the 10CFR100 limits by preventing more than one hoist from being loaded at a given time.

ANSWER: 41 19212

a. To prevent criticality during refueling by ensuring that fuel assemblies are not loaded into the core unless all control rods are fully inserted.

REFERENCE:

USAR VII - 6 Explanation:

USAR VII - CONTROL AND INSTRUMENTATION 6.0 REFUELING INTERLOCKS 6.1 Safety Objective The refueling interlocks are designed to back up procedural core reactivity controls during refueling operations; specifically, the interlocks prevent an inadvertent criticality during refueling operations.

6.2 Safety Design Basis

1. During fuel movements in or over the reactor core, all control rods shall be in their fully-inserted positions.

2. No more than one control rod shall be withdrawn from its fully inserted position at any time when the reactor is in the refuel mode.

6.3 Description During a refueling operation, the reactor vessel head is removed, allowing direct access to the core. Refueling operations include the removal of reactor vessel upper internals and the movement of spent and fresh fuel assemblies between the core and the fuel storage pool. The service platform, refueling platform, and the equipment handling hoists on the platforms are used to accomplish the refueling task. The refueling interlocks reinforce operational procedures that prohibit taking the reactor critical under certain situations

encountered during refueling operations by restricting the movement of control rods and the operation of refueling equipment.

The refueling interlocks include circuitry which senses the condition of the refueling equipment and the control rods. Depending on the sensed condition, interlocks are actuated which prevent the movement of the refueling equipment or withdrawal of control rods (rod block).

Distracters:

b. 280 cal/gm is a RWM bases (N/A during refueling). The refueling interlocks place no restrictions on fuel assembly configuration.

c. The refueling interlocks do not ensure an adequate shutdown margin.

d. The refueling interlocks do not prevent loading of multiple hoists at the same time and the loading of the hoists is not related to the 10CFR100 release rates.

THIS QUESTION WAS USED ON THE 2002 CNS NRC EXAMINATION AS RECORD # 17887.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295018.AK2.01 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Emergency Procedure 5.2REC (Attach if not previously provided)

(including version/revision number) 12 Learning Objective: See Attached (As available)

Question Source: Bank # 25186 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 42 25186 02 03/29/2010 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 4 Multiple Choice Topic Area Description Systems Coolable REC loads following an REC system low pressure Related Lessons COR0021902 REACTOR EQUIPMENT COOLING Related Objectives COR0021902001050A Briefly describe the following concepts as they apply to REC:

Leak or lowering system pressure during accident and transient conditions COR0021902001060A Given a specific REC malfunction, determine the effect on any of the following: REC header pressure Related References 2.2.65 Reactor Equipment Cooling Water System 5.2REC Loss Of REC Related Skills (K/A) 295018.AK2.01 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER and the following: (CFR: 41.7 / 45.8)

System loads (3.3/3.4)

QUESTION: 42 25186 A Loss of Offsite Power (LOOP) occurred. #1 and #2 Emergency Diesel Generators automatically started and powered the critical busses. Two (2) minutes later, REC pressure has stabilized at 50 psig. Access to the Reactor Building is prohibited by radiation levels.

Which of the following loads can be provided long term cooling using the REC system?

a. Augmented Off Gas (AOG)

b. "A" Control Rod Drive pump

c. "B" Drywell Fan Coil Unit

d. HPCI Room Fan Coil Unit ANSWER: 42 25186

d. HPCI Room Fan Coil Unit Explanation:

With an isolation signal present REC-MO-702MV can be reopened, however, the REC-MO-712 and 713 will auto close on the low pressure and cannot be overridden. This will isolate REC to the non-critical loops/components.

Distracters:

a. AOG is isolated

b. This would require a manual operation.

c. The Drywell is isolated

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295005.AK2.01 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) USAR XIV - 5 (Attach if not previously provided)

(including version/revision number) 9/19/00 Learning Objective: See Attached (As available)

Question Source: Bank # 24799 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 43 24799 00 07/15/2009 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 3 Multiple Choice Topic Area Description Updated Safety Analysis M. Gen output breakers open, cause of Rx Scram Report Related Lessons INT0060119 Anticipated Operational Transients and Special Events COR0022102 REACTOR PROTECTION SYSTEM Related Objectives COR0022102001100E Describe the interrelationship between the RPS and the following:

DEH INT00601190010200 Given a Anticipated Operational Transient that is analyzed in the CNS USAR, select an action or actions that will terminate the transient.

COR0022102001040M Describe the RPS design features and/or interlocks that provide for the following: Related system inputs to RPS Related References NONE Related Skills (K/A) 295005.AK2.01 Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following: (CFR: 41.7 / 45.8) RPS. (3.8/3.9)

QUESTION: 43 24799 The Plant is operating at near rated power, with the following conditions:

• PCB-3312 is Tagged open due to maintenance in the 345 Switch yard

• PCB-3310 Auto Reclosure Switch is in OFF Then the following sequence of events occur:

• A line fault causes PCB 3310 to Open

• Automatic Reactor Scram is initiated What is the cause of the Reactor Scram?

a. High Reactor Pressure Scram

b. HI APRM Scram (Neutron monitoring Scram)

c. Reactor Scram due to Turbine Stop Valve Closure.

d. Reactor Scram due to Turbine Control valve fast closure.

ANSWER: 43 24799

d. Reactor Scram due to Turbine Control valve fast closure.

Explanation:

As shown in USAR Figure XIV-5-1, as soon as turbine control valve fast closure is sensed, a scram is initiated. This occurs in advance of the high neutron flux and high RPV pressure scram signals thereby limiting the peak neutron flux to about 180 percent of rated. The average surface heat flux reaches a peak of about 115 percent of rated. The small increase in average surface heat flux coupled with the slight increase in core flow ensures that nucleate boiling is maintained throughout the transient. The Reactor Scram due to Turbine Stop valve closure is a separate analyzed transient. The stop valve closure assumes a direct turbine trip signal not a load rejection.

Distracters:

a. High Reactor Pressure Scram, is incorrect because the Turbine Control Valve trip is faster.

b. HI APRM Scram (Neutron monitoring Scram), is incorrect because the Turbine Control Valve trip is faster.

c. Reactor Scram due Turbine Stop Valve Closure, is incorrect because the Turbine Control Valve trip is faster.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295019.AK2.01 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-04-02 ANN Proc 2.3_9.5.2 (Attach if not previously provided)

(including version/revision number) 23 28 Learning Objective: See Attached (As available)

Question Source: Bank # 3996 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 44 3996 02 10/14/2005 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Emergency / Abnormals COR0020502001110A Control Rod Drive Mechanisms Related Lessons COR0020502 CONTROL ROD DRIVE MECHANISM COR0020402 OPS CONTROL ROD DRIVE HYDRAULICS Related Objectives COR0020502001110A Given a specific CRDM malfunction, determine the effect on any of the following: Rod Movement COR0020402001110A Predict the consequences a malfunction of the following would have on the CRDH systems: Loss of plant air system COR0020402001130F Describe the interrelationships between the Control Rod Drive Hydraulic System (CRDH) and the following: Plant Air Systems Related References 2.3_9-5-2 Panel 9 Annuciator 9-5-2 LP Lesson COR002-04-02 CRD Hydraulics Related Skills (K/A) 295019.AK2.01 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: (CFR: 41.7 / 45.8) CRD hydraulics (3.8/3.9)

QUESTION: 44 3996 Given the following conditions;

• The plant is at 100% power.

• IA-PRV-PRV614, CRD Reliable Air Supply regulator fails so that output pressure slowly fails to 0 psig.

• Annunciator 9-5-2/F-5, SCRAM VALVE PILOT AIR LOW PRESSURE is received.

• NO operator action is taken.

Which of the following statements describes the status of the Control Rod Drive Hydraulics System for these conditions?

a. ALL Control Rods will scram randomly.

Normal rod insertion AND withdrawal will NOT be available

b. ALL Control Rods will scram randomly.

Normal rod insertion AND withdrawal will remain available

c. Control Rod Scram capability will be lost.

Normal rod insertion AND withdrawal will remain available

d. Control Rod Scram capability will be lost.

Normal rod insertion AND withdrawal will NOT be available ANSWER: 44 3996

a. ALL Control Rods will scram randomly.

Normal rod insertion AND withdrawal will NOT be available.

Explanation:

PRV614 supplies air to the Scram Pilot Air Header and Flow Control Valves. The Flow Control Valves fail closed on loss of air causing a loss of normal drive capability. A loss of air to the Scram Pilot Air Header will cause the Scram Valves to drift open and rods will drift in as each Scram Valve starts to open.

Distracters:

b, The Flow Control Valves fail closed on loss of air causing in a loss of normal drive capability.

c. The Flow Control Valves fail closed on loss of air causing in a loss of normal drive capability.

d. A loss of air causes Scram Valves to open and rods to insert.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295021.AK3.01 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-23-02 Procedure 2.4SDC (Attach if not previously provided)

(including version/revision number) 27 12 Learning Objective: See Attached (As available)

Question Source: Bank # 2751 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 45 2751 00 08/25/1999 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 4 Multiple Choice Topic Area Description Systems COR0022302001090D Residual Heat Removal System Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR0022302001090D Explain the significance of the following as they apply to a loss of Shutdown Cooling: Natural circulation Related References COR0022302 Residual Heat Removal 2.4SDC Loss of Shutdown Cooling Related Skills (K/A) 295021.AK3.01 Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING: (CFR: 41.5 / 45.6) Raising reactor water level (3.3 / 3.4)

QUESTION: 45 2751 With a loss of shutdown cooling, vessel water level is to be maintained > 48 inches.

What is the significance of this level?

a. To ensure adequate core cooling.

b. To ensure natural circulation will occur.

c. To maintain required NPSH for the jet pumps.

d. To maintain adequate NPSH for the Reactor Recirculation pumps.

ANSWER: 45 2751

b. To ensure natural circulation will occur.

Explanation:

From System Lesson COR002-23-02 Natural Circulation - During plant cooldown, the reactor water level is maintained 48 inches to ensure natural circulation will occur should forced circulation be lost.

From Abnormal Procedure 2.4SDC - Contingency actions for complete loss of SDC:

• Commence monitoring plant heatup rate per Procedure 6 .RCS.601.

NOTE: The Preferred level indication is NBI-LI-86, SHUTDOWN LVL. RFC-LI-94A,

• RFC-LI-94B, or RFC-LI-94C, RX NR LEVEL, may indicate up to 9" higher than actual during cold conditions.

• Control RPV level > 48" to aid in thermal convection flow.

• Monitor following temperatures and pressures frequently and log in Control Room Log every 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

From the PSTG on the EOPs:

Definitions of Key Words and Phrases The meaning of the following terms is discussed in the context of their use within the PSTGs/SATGs. This information is provided to facilitate a consistent and technically accurate understanding of the guidelines.

Adequate core cooling: Heat removal from the reactor sufficient to prevent rupturing the fuel clad. Within the PSTGs, three viable mechanisms for establishing adequate core cooling are definedcore submergence, spray cooling, and steam cooling.

Submergence is the preferred method for cooling the core. The core is adequately cooled by submergence when it can be determined that RPV water level is at or above the top of the active fuel. All fuel nodes are then assumed to be covered with water and heat is removed by boiling heat transfer.

Adequate spray cooling is provided, assuming a bounding axial power shape, when design spray flow requirements are satisfied and RPV water level is at or above the elevation of the jet pump

suctions. The covered portion of the core is then cooled by submergence while the uncovered portion is cooled by the spray flow.

Student Text on Reactor Recirc COR002-22-02:

Power to Flow Map.

h. Do Not Operate Region This region contains the Recirculation pump net positive suction head limit, the net positive suction head admin limit, the jet pump net positive suction head limit and the low feed water Recirculation runback line. This area should not be entered to insure the NPSH requirements for the jet pumps and Recirculation pumps are met.

4. Pump NPSH Requirements

a. NPSH is as a measure of the difference between the saturation pressure and the total pressure felt at the inlet of the pump. The total pressure is made up of two elements; the height of the column of water above the pump, and the amount of subcooling at the pump inlet. If the total pressure at the suction of the pump drops below the required NPSH for the pump, cavitation will occur. Cavitation causes excessive noise, pump vibration and reduction in pumping capacity. This leads to reduced pump efficiency and possible pump damage.

b. During low power operations the significant factor affecting the Recirc pump NPSH is the height of the column of water above the pump, about 57 ft (feedwater subcooling effects though present, are minimal). This provides adequate NPSH to the pumps as long as water level is maintained in the normal operating band. Lowering water level will reduce the NPSH for both the Recirc pump and the jet pumps. Carry-under (saturated steam entrained in the water of the downcomer) is also increased by low water level on the steam separator skirt, excessive carry-under reduces overall plant efficiency and can lead to a loss of NPSH to the jet pumps and Recirc pumps.

c. At high power operations the subcooling at the Recirc pump suction becomes the significant factor affecting the NPSH. Since incoming feedwater provides the subcooling, anything which reduces feedflow will reduce the subcooling and the pumps NPSH.

Feedwater provides about 20F subcooling when operating at 100% power, so the NPSH supplied to the Recirc pump is equivalent to about 437 ft. of water.

An interlock prevents increasing the Recirc pump speed above the minimum value (22%)

unless feedwater flow is at least 20% in order to supply the required NPSH.

Distracters:

a. Adequate Core Cooling is assured by submergence in this situation.

c. The Jet Pumps are not being used during a loss of Shutdown Cooling.

d. The Reactor Recirc Pumps are not being used during a loss of Shutdown Cooling.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295003.AK3.06 Importance Rating 3.7 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-03-02 (Attach if not previously provided)

(including version/revision number) 27 Learning Objective: See Attached (As available)

Question Source: Bank # 25257 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 46 25257 00 04/14/2010 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 3 Multiple Choice Topic Area Description Systems Using panel indications discern RPS power condition.

Related Lessons COR0020302 OPS CONTAINMENT COR0022102 REACTOR PROTECTION SYSTEM Related Objectives COR0020302001060M Describe the interrelationship between PCIS and the following: AC Distribution COR0020302001200A Briefly explain the reason for the following: Containment isolation on partial or complete loss of AC power COR0020302001210B Given plant conditions, determine if the following should have occurred: Any of the PCIS group isolations.

COR0022102001040L Describe the RPS design features and/or interlocks that provide for the following: Under/over voltage and frequency protection COR0022102001050C Briefly describe the following concepts as they apply to RPS: EPA operation Related References LP COR002-03-02 Related Skills (K/A) 295003.AK3.06 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: (CFR: 41.5 / 45.6)

Containment isolation. (3.7/3.7)

QUESTION: 46 25257 Using the photograph below, of the 9-5 PCIS Group Isolation Indications:

What condition or malfunction caused the indications shown above?

Power has been lost or interrupted to

a. only "A" RPS power panel.

b. only "B" RPS power panel.

c. both "A" and "B" RPS power panels.

d. neither "A" nor "B" RPS power panels.

ANSWER: 46 25257

a. only "A" RPS power panel.

Explanation:

On a loss of "A" RPS panel, a full Group 3 isolation will result due to the loss of the RWCU NRHX temperature switch relay. This relay is in both PCIS divisions. The full PCIS Group 6 is due to losing power to a Group 2 relay that is seen by the Gp 6 isolation circuit.

Distracters:

b. The B RPS Power loss would have caused the DIV II side of the lights for Groups 1, 2, 3, 6, & 7 to extinguish, not the DIV I side.

c. Since there are some logics that feed this panel that are 125 VDC, there might be lights on because of them, but since Groups 4 and 5, and part of Group 1 lights are fed from DC some of the lights could be illuminated. As stated above in the explanation section, all four lights will be extinguished for both the Group 3 and 6. This adds some confusion when relying just on the indications for the group isolations on Panel 9-5. Thus, making this a valid distracter.

d. RPS feeds the indication to the Group 2, 3, 6, & 7 lights so if some of those lights are illuminated then that RPS bus has power it should indicate that the associated division 120 VAC RPS Bus is energized.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295037.EK3.03 Importance Rating 4.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) PSTG Appendix B and Lesson INT008-06-10 (Attach if not previously provided)

(including version/revision number) 3 22 Learning Objective: See Attached (As available)

Question Source: Bank # 14478 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 47 14478 02 12/12/2008 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 3 Multiple Choice Topic Area Description Emergency Operating INT0080610, How far to lower water level and why Procedures Related Lessons INT0080610 OPS EOP FLOWCHART 7A - RPV LEVEL (FAILURE-TO-SCRAM)

Related Objectives INT00806100010900 Given an EOP flowchart 7A, RPV LEVEL (FAILURE TO SCRAM) step, state the reason for the actions contained in the step.

INT00806100010800 Given plant conditions and EOP flowchart 7A, RPV LEVEL (FAILURE TO SCRAM), determine required actions.

Related References PSTG INT008-06-10 Related Skills (K/A) 295037.EK3.03 Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: (CFR: 41.5 / 45.6) Lowering reactor water level (4.1*/4.5*)

QUESTION: 47 14478 The plant has experienced ATWS conditions with the following indications

• The feedwater system in operation

• Reactor power 4% (stable)

• Reactor Pressure 935 psig (stable)

• Reactor water level +13 inches (NR) (stable)

• Average torus water temperature 87ºF (stable)

How long must RPV injection be stopped and prevented and why?

Stop and prevent injection until

a. the reactor is subcritical below the heating range to ensure reactor power is below the threshold for thermal hydraulic instabilities.

b. reactor water level is less than - 60 inches (corrected FZ) to prevent the uncontrolled injection of large amounts of cold unborated water at low core flows.

c. reactor water level is less than - 60 inches (corrected FZ) to ensure incoming feedwater is heated.

d. the reactor is subcritical below the heating range to ensure required feedwater injection rates reduce the chance of uncontrolled injection of large amounts of cold unborated water.

ANSWER: 47 14478

c. reactor water level is less than - 60 inches (corrected FZ) to ensure incoming feedwater is heated.

Explanation:

Reactor water level at - 60 inches (FZ) is below the level of the feedwater spargers which allows the incoming feedwater to be heated by the steam in downcomer.

PSTG/SATG Step C5-3 If reactor power is above 3% (APRM downscale trip) or cannot be determined and RPV water level is above -60 in. 100.94 in. (FZ) (24 in. below the feedwater sparger nozzles),

lower RPV water level to below -60 in. 100.94 in. (FZ) (24 in. below the feedwater sparger nozzles) by terminating and preventing all injection into the RPV except from boron injection systems, RCIC, and CRD, defeating interlocks if necessary.

Discussion To prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities, RPV water level is lowered sufficiently below the elevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet

subcooling. For conditions that are susceptible to oscillations, the initiation and growth of oscillations is principally dependent upon the subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude.

If reactor power is at or below the APRM downscale trip setpoint, it is highly unlikely that core bulk boiling boundary would be below that which provides suitable stability margin for operation at high powers and low flows. (A minimum boiling boundary of 4 ft above the bottom of active fuel has been shown to be effective as a stability control because a relatively long two-phase column is required to develop a coupled neutronic/ thermal-hydraulic instability.) Furthermore, flow/density variations would be limited with reactor power this low since the core has a relatively low average void content. Therefore, there is significant stability margin with power at or below the APRM downscale trip setpoint.

Twenty-four inches below the lowest nozzle in the feedwater sparger has been selected as the upper bound of the RPV water level control band. This water level is sufficiently low that steam heating of the injected water will be at least 65% to 75% effective (i.e., the temperature of the injected water will be increased to 65% to 75% of its equilibrium value in the steam environment.

This water level is sufficiently high that the capability to bypass the low RPV water level MSIV isolation should be able to control RPV water level with feedwater pumps to preclude the isolation.

Distracters:

a. is incorrect. Because water level need only be lowered to - 60 inches (FZ) even if reactor power remains elevated because 3 of the 4 level power conditions do not exist.

b. is incorrect. Because the reason that level is lowered in this case is to ensure heating of the incoming feedwater. Stopping and preventing injection prior to emergency depress ensures no uncontrolled injection of relatively cold water.

d. is incorrect. Because water level need only be lowered to - 60 inches (FZ) even if reactor power remains elevated because 3 of the 4 level power conditions do not exist.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 700000.AA1.01 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 5.3GRID Lesson COR001-13-01 (Attach if not previously provided)

(including version/revision number) 31 22 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 48 00 01/12/2011 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 5 Multiple Choice Topic Area Description None INT0320131, CNS Abnormal Procedures (RO) Electrical Related Lessons INT0320131 CNS Abnormal Procedures (RO) Electrical Related Objectives INT0320131S0S0100 Given plant condition(s), determine from memory the appropriate Abnormal/Emergency Procedure(s) to be utilized to mitigate the event(s).

COR001-13-01 Objective

11. © Concerning the Main Generator capability curves:

a. Describe the basis for each part of the curve.

b. Given generator conditions and the Main Generator capability curve, determine if operation is within the acceptable region of the curve.

Related References 5.3GRID System Lesson COR001-13-01 Main Generator and Auxiliaries Related Skills (K/A) 700000.AA1.01 Ability to operate and/or monitor the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: (CFR: 41.5 and 41.10

/ 45.5, 45.7, and 45.8 ) Grid frequency and voltage (3.6 / 3.7)

QUESTION: 48 A Main System Grid disturbance is causing voltage and frequency changes on Coopers Main Generator and 5.3GRID, DEGRADED GRID VOLTAGE is entered. The current plant conditions are as follows:

• Reactor power 95%.

• Main Generator frequency is 59.8 MHz.

• Main Generator voltage is 21.60 KVolts.

• Main Generator MVARs are -160 MVARs.

What actions should be taken on Panel C to correct the electrical transient?

a. Raise voltage

b. Lower voltage

c. Raise frequency

d. Lower frequency ANSWER: 48

a. Raise voltage Explanation:

From the Main Generator Lesson:

Reactive Load - Once the generator output breakers are closed, the generators reactive load is controlled by the Main Generator field current. When the Gen Voltage Adjust control switch is in the OFF or TEST position, field current is adjusted with the Base Adjuster control switch. With the Gen Voltage Adjust switch in the ON position, field current is adjusted by the voltage adjust control switch.

The lowermost line shows the limit of reactive load (MVARs) when the generator is being operated underexcited, that is to say that the generator has become a reactive load to the grid, due to the additional current draw from the grid needed to supplement the rotor field in this condition where the regulator is not supplying enough field current on its own. (This is also known as the vars in or leading power factor region of the curve.) In this condition, part of the rotor field strength is induced by current in the stator. The lines of flux on the stator in this condition are highly concentrated on the stator core ends. High current flows result in the stator (core) ends, which cause localized overheating to occur. This heating in the stator core ends (retainer sections) is the basis for the limits of the curve in this region.

From 5.3GRID Grid oscillations can be induced by CNS generator regulator malfunction or other grid generators malfunctioning. Grid generated oscillations could affect 345 kV, 161 kV, and/or 69 kV voltages.

When determining if oscillating voltage has exceeded the lower limit, average oscillating voltage is used.

NPPD System frequency < 59.5 indicates NPPD system is unstable and outside of its secure operating state. If frequency is degraded, it will likely be a rapidly occurring event that will cause automatic system load shedding to stabilize system.

Maintain generator MVAR per DCC System Operator direction and Main Generator MEGAWATTS vs. MEGAVARS per Attachment 2 (Page 8) with GEN BASE ADJUST by performing following: To pick up positive (OUT) MAIN GENERATOR MVAR. Maintain MEGAWATTS vs. MEGAVARS in area bound by: 0.85 LAGGING PF line (positive or over-excited). 0.95 LEADING PF line (negative or under-excited).

Distracters:

b. Lower voltage - would make the situation worse by decreasing MVARs even further and cause more instability in the generator.

c. Raise frequency - with the Frequency at 59.8 MHz, this is still within the normal operating range for frequency and no adjustments need to be made.

d. Lower frequency - with the Frequency at 59.8 MHz, this is still within the normal operating range for frequency and no adjustments need to be made.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295016.AA1.02 Importance Rating 2.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 5.1ASD (Attach if not previously provided)

(including version/revision number) 13 Learning Objective: See Attached (As available)

Question Source: Bank # 19280 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 49 19280 01 12/17/2004 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 4 Multiple Choice Topic Area Description Abnormal/Emergency INT0320134, OPS CNS Abnormal Procedures - Fire Procedures Related Lessons INT0320134 OPS CNS Abnormal Procedures (RO) Fire Related Objectives INT0320134H0H0100 Given plant condition(s), and the applicable Abnormal/Emergency Procedure, determine the correct subsequent actions required to mitigate the event(s).

Related References 5.1ASD Related Skills (K/A) 295016.AA1.02 Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT: (CFR: 41.7 / 45.6) Reactor/turbine pressure regulating system (2.9* / 3.1*)

QUESTION: 49 19280 The plant is at power when the following events occur:

• A fire in the Control Room.

• Control Room is evacuated.

• Alternate Shutdown Room is manned.

• A 90F/hr cooldown is commenced from normal operating pressure.

What indication must the ASD operator use to initially monitor RPV cooldown rate during a shutdown outside the Control Room?

a. RPV bottom drain thermocouple taken locally by I&C.

b. Reactor Recirc Suction Temperature on Rack 25-5 and 25-6.

c. HPCI Turbine Steam pressure on the ASD HPCI Control Panel.

d. HPCI Pump Discharge pressure on the ASD HPCI Control Panel.

ANSWER: 49 19280

c. HPCI Turbine Steam pressure on the ASD HPCI Control Panel.

Explanation:

HPCI Turbine Steam Pressure is used to calculate cooldown during these conditions.

Reference:

5.4FIRE-SD Attachment 1 Caution prior to step 2.1.3.

Distracters:

a. is incorrect. RPV bottom head temperature indication is feed to the control room and is not an approved method of determining changes in vessel coolant temperature.

b. is incorrect. Reactor pressure on Rack 25-5 is used after HPCI is isolated on low pressure, not recirc suction temperatures.

d. is incorrect. HPCI pump discharge pressure will be dependent upon HPCI pump speed and not representative of reactor pressure.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295030.EA1.05 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) EOP 3A EOP SAG Graphs (Attach if not previously provided)

(including version/revision number) 13 14 Learning Objective: See Attached (As available)

Question Source: Bank # 9724 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y 50 9724 00 04/24/2000 None SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 3 Multiple Choice Topic Area Description None COR0021102001100N Related Lessons COR0021102 OPS High Pressure Coolant Injection System Related Objectives COR0021102001100N Predict the consequences of the following on the HPCI system: Low suppression pool level Related References EOP SAG Graphs Related Skills (K/A) 295030.EA1.05 Ability to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: (CFR: 41.7 / 45.6) HPCI (3.5 / 3.5)

QUESTION: 50 9724 The plant is recovering from a scram and the following conditions exist:

• Control rods fully inserted

• RPV water level is -180 inches on Fuel Zone Instrument on Panel 9-3 (slowly lowering)

• RPV pressure is 800 psig

• CRD is injecting to RPV at maximum flow

• HPCI is injecting to RPV at rated flow

• No other sources of RPV injection available

• Primary containment water level is 10.9 feet (lowering)

What action is required?

a. Remove HPCI from service.

b. Emergency depressurize the RPV.

c. Raise injection from HPCI, disregarding Vortex Limits.

d. Enter the Steam Cooling Procedure when RPV water level drops to -125 inches Corrected FZ.

ANSWER: 50 9724

a. Remove HPCI from service.

Explanation:

Step SP/L-10 of the Primary Containment Control Procedure because the HPCI turbine exhaust is not submerged when torus level drops below 11 feet.

Distracters:

b. is incorrect. Because ED is not required until the RPV water level drops below -158 inches corrected FZ.

c. is incorrect. Because the HPCI turbine is exhausting directly into the torus air space and there is no Vortex Limit for the HPCI pumps.

d. is incorrect. Because the operating crew will go to Steam Cooling ONLY when there is no vessel injection available.

PROVIDE THE STUDENTS WITH EOP Graph 14.

EOP/SAG Graphs ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295026.EA2.02 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) EOP/SAG Graphs (Attach if not previously provided)

(including version/revision number) 14 Learning Objective: See Attached (As available)

Question Source: Bank # 24497 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 51 24497 02 03/22/2010 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Emergency Operating INT0080613 obj 11 & INT0080618 obj 3, EOP 3A, pressure for Procedures HCTL (2008 BIENNIAL EXAM)

Related Lessons INT0080613 OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL INT0080618 OPS EOP AND SAG GRAPHS AND CAUTIONS Related Objectives INT00806130011100 Given plant conditions and EOP Flowchart 3A, PRIMARY CONTAINMENT CONTROL, determine required actions.

INT00806180010300 Given plant conditions and the EOP and SAG Graphs Flowchart, determine if operation is within the allowed region of a graph.

Related References INT0080613 Flowchart 3A - Primary Containment Control INT0080618 EOP and SAG Graphs and Cautions 5.8 Emergency Operating Procedures (EOPs)

Related Skills (K/A) 295026.EA2.02 Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: (CFR: 41.10 /

43.5 / 45.13) Suppression pool level (3.8 / 3.9)

QUESTION: 51 24497 The plant has just experienced a small break LOCA resulting in the following conditions:

• Reactor Pressure 800 psig (steady)

• Torus pressure is 5 psig (rising slowly)

• Primary Containment water level is 11 feet (steady)

• Torus water temperature is 180° F (rising slowly)

In accordance with EOP 3A, PRIMARY CONTAINMENT CONTROL; what is the HIGHEST Torus Water Temperature allowed before the plant is required to be Emergency Depressurized?

a. 185F

b. 195F

c. 210F

d. 263F ANSWER: 51 24497

b. 195F Explanation:

In accordance with EOP 3A Torus Temperature leg, Step SP/T-5 When average torus water temperature and RPV pressure cannot be maintained within HCTL (Graph 7) EMERGENCY DEPRESSURIZE. At 800 psig and a primary containment level of 11 feet the lines of the HCTL graph intersect at 200F.

PROVIDE THE STUDENTS WITH HCTL (Graph 7)

Distracters:

a. 185F is the temperature for 1080 psig the next line before the correct one.

c. 210F is the temperature for 600 psig the next line after the correct one.

d. 263F is the temperature for 50 psig the last line before it turns red.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 600000.AA2.05 Importance Rating 2.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Proc 2.2.84 and COR0010801 HVAC Lesson (Attach if not previously provided)

(including version/revision number) 49 22 Learning Objective: See Attached (As available)

Question Source: Bank # 3724 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 52 3724 06 03/15/2011 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 4 Multiple Choice Topic Area Description Systems SKL0124108A0A030E HEATING, VENTILATION, AIR CONDITIONING Related Lessons SKL0124108 HEATING, VENTILATION, AIR CONDITIONING COR0010802 OPS HEATING, VENTILATION AND AIR CONDITIONING Related Objectives SKL012410800A030E Given plant conditions, predict changes in the following:

Starting/stopping of fans COR0010802001160D Predict the consequences a malfunction of the following would have on the Control Room HVAC system: Fire protection Related References 2.2.84 HVAC Main Control Room and Cable Spreading Room Related Skills (K/A) 600000.AA2.05 Ability to determine and interpret Ventilation alignment necessary to secure affected area as they apply to PLANT FIRE ON SITE (2.9 / 3.0)

QUESTION: 52 3724 A fire has occurred in the Cable Spreading Room with the plant initially operating at 100% power.

The fire did not spread beyond the Cable Spreading Room.

What impact does this fire have on the Control Room Ventilation System?

a. Fire Dampers DO NOT isolate the Control Room. CREF starts and supplies the Control Room with outside air.

b. Fire Dampers isolate the Control Room. CREF starts and supplies the Control Room with outside air.

c. Fire Dampers DO NOT isolate the Control Room. CREF does NOT start.

d. Fire Dampers isolate the Control Room. CREF does NOT start.

ANSWER: 52 3724

d. Fire Dampers isolate the Control Room. CREF does NOT start.

Explanation:

Procedure 2.2.84 ATTACHMENT 1 INFORMATION SHEET FUNCTION - The system provides HVAC to the Control Room and Cable Spreading Room for personnel comfort and optimum equipment performance.

OPERATING CHARACTERISTICS Fire/Smoke Dampers HV-AD-AD1544, HV-AD-AD1545, HV-AD-AD1546, HV-AD-AD 1547, HV-AD-AD 1581, and HV-AD-AD 1582 automatically close when fire or smoke is detected locally at the damper or when smoke is detected in the Cable Spreading Room to prevent smoke from spreading to the Control Room when there is a fire in the Cable Spreading Room.

When smoke is detected by SD-1001 (Cable Spreading Room return duct), Supply Fans SF-C-1A and SF-C-1B receive trip signals and fire/smoke Dampers HV-AD-AD1544, HV-AD-AD1545, HV-AD-AD1546, HV-AD-AD1547, HV-AD-AD1581, and HV-AD-AD1582 close.

From COR0010801 HVAC Lesson A bypass filter system, Control Room Emergency Filtration System (CREFS), is also installed to insure safe operation for personnel in the Main Control Room. This system can be initiated by PCIS Group 6 isolation or by manual initiation.

Control Room Emergency Filtration System (CREFS)

The emergency bypass supply fan is intended to prevent the introduction of unfiltered air into the Control Room envelope (which includes the cable spreading room) by maintaining the Control

Room envelope at a positive pressure as compared to the adjoining buildings and the outside atmosphere under calm wind conditions.

The PCIS group 6 isolation signal is an indicator of a Loss of Coolant Accident or Fuel Handling accident. Combining the group 6 isolation signal with the rapid stroke times of the isolation valves, ensure dose for Control Room personnel to remain below limits.

Upon PCIS group 6 isolation, an alarm is annunciated in the Control Room, and the emergency supply fan starts.

Once the fan starts, the intake valves automatically shift to ensure that all outside air passes through the filter train. Another valve which also repositions is the kitchen/toilet exhaust valve (closes). If the emergency supply fan fails to start, then the dampers will not shift. Also the pantry/toilet exhaust fan is interlocked to trip whenever the emergency supply fan starts.

Distracters:

a. The fire dampers do isolate the control room. The Emergency Bypass Train (CREF) does NOT start.

b. The Emergency Bypass Train (CREF) does NOT start.

c. The fire dampers do isolate the control room.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295001.AA2.05 Importance Rating 3.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-21-02, Procedure 2.4RxPWR (Attach if not previously provided)

(including version/revision number) 20 4 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: Y 53 00 09/29/10 License Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice N Topic Area Description Abnormal What action is required for a Jet Pump Failure?

Related Lessons COR0022102 OPS Reactor Protection System Related Objectives COR0022202001060G Given a specific Reactor Recirculation system or the Recirculation Flow Control system malfunction, determine the effect on any of the following: Reactor Vessel Internals (jet pumps, stratification, bottom head drain temperature, pump starts)

Related References COR0022102 OPS Reactor Protection System Rev20 2.4RxPWR Reactor Power Anomalies Rev 4 Related Skills (K/A) ROI SROI 295001.AA2.05 Ability to determine and/or interpret the following as they apply 3.1 3.4 to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: (CFR: 41.10 / 43.5 / 45.13) Jet pump operability: Not-BWR-1&2 (3.1/3.4)

QUESTION: 53 The reactor is at 100% power when Reactor Power lowers unexpectedly. The following conditions are noted:

• Generator output has decreased.

• Indicated Total Core flow has increased.

• Core plate D/P has decreased.

• RR pump speeds have not changed.

• RR Loop A indicated flow is 34x106 lbm/hr and stable.

• RR Loop B indicated flow is 36x106 lbm/hr and stable.

What action is required?

a. Lock the A Reactor Recirculation Pump Scoop Tube.

b. Perform Jet Pump operability.

c. Lower the speed of the B RR Pump to within 5% of the A RR Pump.

d. Raise the speed of A RR Pump to within 5% of the speed of the B RR Pump.

ANSWER: 53

b. Perform Jet Pump operability Explanation:

Core flow has risen with no change of recirc pump speeds this does NOT comply with the reactor power and core plate dIp lowering; which indicates a lowering of core flow. Recirc pump speed has not changed therefore it appears recirc pump B flow has risen while its speed remained the same this indicates a failed jet pump. For these indications 2.4RxPWR Attachment 2 Step 1.3 requires performing jet pump operability.

Distracters:

a. is incorrect because there has been no change in recirc pump speeds. In accordance with 2.4RR, the scoop tube is locked when recirc pump speed is rising.

c. is incorrect because the requirement is to balance speeds is within 10% when core flow is greater than or equal to 70% rated core flow. There is not a 10% imbalance.

d. is incorrect because the speed should be lowered on the faster pump additionally the requirement is to balance speeds is within 10% when core flow is greater than or equal to 70% rated core flow. There is not a 10% imbalance.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295006 G 2.4.45 Importance Rating 4.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 2.4TURB (Attach if not previously provided)

(including version/revision number) 24 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 12134 (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y 54 00 12/12/2010 LOR SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 3 Multiple Choice Topic Area Description None INT0320127O0O0100 CNS Abnormal Procedures (RO)

Turbine/Generator Related Lessons INT0320127 CNS Abnormal Procedures (RO) Turbine/Generator Related Objectives INT0320127O0O0100 Given plant condition(s), determine from memory if a manual reactor scram or an emergency shutdown from power is required due to the event(s).

Related References 2.4TURB Main Turbine Abnormal 10CFR55.41 Written examinations: Operators Related Skills (K/A) 295006. SCRAM Generic 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12) (4.1 / 4.3)

QUESTION: 54 Power ascension is in progress with reactor power at 25%.

• 9-5-2/C-4, TSV & TCV CLOSURE TRIP BYP is in alarm.

• The Main Turbine is being placed in service and is currently rotating at 1800 RPM.

• Rising thrust bearing metal temperatures are noted.

• Thrust bearing metal temperature on computer point T079 is 220°F and T080 is 230°F.

• The crew also noted that lube oil cooler outlet temperature is 145°F.

Which action is required next?

a. Trip the Main Turbine only.

b. Commence a normal shutdown.

c. Scram the reactor and trip the Main Turbine.

d. Send a Station Operator to reduce lube oil cooler outlet temperature.

ANSWER: 54

a. Trip the Main Turbine only.

Explanation:

Since power is below the set point for 9-5-2/C-4, TSV & TCV CLOSURE TRIP BYP annunciator a reactor scram is not required however the Main Turbine must be tripped in accordance with 2.4TURB.

Distracters:

b. A normal shutdown will be commenced if the temperature is not corrected; however the turbine trip must be performed as the set points have been exceeded.

c. A turbine trip is required; however the Reactor need not be scrammed.

d. A turbine trip is required; sending a station operator to investigate the problem is a little late.

Modified: 12134

MODIFIED QUESTION: 12134 At 2230 a power ascension was in progress with reactor power at 90%. Rising thrust bearing metal temperatures were noted. Thrust bearing metal temperature on computer point T079 was 220°F and T080 is 230°F. The crew also noted that lube oil cooler outlet temperature was 145°F.

Which action is required next?

a. Reduce turbine load by 10%.

b. Commence a normal shutdown.

c. Scram the reactor and trip the turbine.

d. Reduce lube oil cooler outlet temperature.

ANSWER:

c. is correct. Since power remains above the setpoint for 9-5-2/C-4, TSV & TCV CLOSURE TRIP BYP annunciator a reactor scram and then a turbine trip are required.

a. is incorrect. Because a turbine trip is required.

b. is incorrect. Because a turbine trip is required.

d. is incorrect. Because a turbine trip is required.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295038 G 2.2.44 Importance Rating 4.2 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedure 5.7.17 (Attach if not previously provided)

(including version/revision number) 35 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 13 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 55 00 01/02/2011 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 3 Multiple Choice Topic Area Description Emergency Procedures Related Lessons GEN0030401 Emergency Plan for Licensed Operators Related Objectives GEN0030401E0E0200 State the primary method used to quantify the source term.

GEN0030401E0E0300 List the monitored release path at CNS.

Related References 5.7.17 Dose Projection Related Skills (K/A) 295038 High Off-site Release Rate Generic 2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5 / 45.12) IMPORTANCE RO 4.2 SRO 4.4

QUESTION: 55 The reactor was scrammed due to a steam leak into Secondary Containment, with the following indications:

Question 55 (continued)

The following conditions are present:

• MAX Safe Temperatures and Radiation levels are exceeded in one (1) area in Secondary Containment.

• Reactor Water level is -20 inches slowly lowering on the Wide Range Instruments.

The Operator places the Mode switch for RMP-RM-452A, RMP-RM-452B, RMP-RM-452C, and RMP-RM-452D, to TRIP TEST then back to operate in accordance with Procedure 2.1.22 Recovering From A Group Isolation.

What is the status of Standby Gas Treatment (SGT), and how do these Operator actions affect The release from the Reactor Building?

The Standby Gas Treatment Trains

a. are running and the release from the Reactor Building is now being filtered.

b. are running and the release is unaffected until there are additional operator actions to align the SGTs.

c. are not running and there is no change to the release from the Reactor Building.

d. are not running and the release from the Reactor Building is unaffected until there are some additional Operator actions to start and align the SGTs.

ANSWER: 55

a. are running and the release from the Reactor Building is now being filtered.

Explanation:

With a Primary System discharging into the Reactor Building and all four Reactor Building Exhaust Ventilation Rad Monitors reading below their set points for a Group 6 isolation, the Operator must take the mode selector switches for each of the four rad monitors to Trip/Test and back to operate to initiate a group 6 and start the Standby Gas Treatment Trains. Once that is done, no other action by the Operator is required to align the system to start filtering the Reactor Building atmosphere and discharging it out of the Elevated Release Point. Both SGTs should be running taking a suction on the Reactor Building.

With Reactor Water level at -20 inches on the wide range level instruments, level is not low enough to have caused a Group 6 isolation and a Standby Gas Treatment train initiation.

Distracters:

b. are running and the release is unaffected until there are additional operator actions to align the SGTs, is incorrect because there are no actions necessary to align the SGT to the Reactor Building to start the filtering process.

c. are not running and there is no change to the release from the Reactor Building, is incorrect because the Operator action to place the Rx Build Rad Monitor switches to trip test and back to operate will initiate a Group 6 and Initiate the SGT.

d. are not running and the release from the Reactor Building is unaffected until there are some additional Operator actions to start and align the SGTs, is incorrect because the Operator action to place the Rx Build Rad Monitor switches to trip test and back to operate will initiate a Group 6 and Initiate the SGT. Also no additional actions are necessary to align SGT for treatment.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295004 G 2.4.4 Importance Rating 4.5 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 5.3DC125 (Attach if not previously provided)

(including version/revision number) 20 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 56 00 02/14/2011 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 4 Multiple Choice Topic Area Description Abnormal/Emergency INT0320131, Loss of DC reason the reactor scrams.

Procedures Related Lessons INT0320131 CNS Abnormal Procedures (RO) Electrical Related Objectives INT0320131T0T0100 Given plant condition(s), determine from memory any automatic actions listed in the applicable Abnormal/Emergency Procedure(s) which will occur due to the event(s).

Related References 5.3DC125 Loss of 125 VDC Related Skills (K/A) 295004 Partial of Complete Loss of Forced Core Flow Circulation. Plus a Generic 2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (CFR:

41.10 / 43.2 / 45.6) 4.5 SRO 4.7

QUESTION: 56 The plant is operating at 100% power when the following annunciators are displayed and alarm:

• Panel 9-5-2 / E-7 ARI/ATWS RPT Logic B Power Failure

• Panel 9-4-3 / C-6 RRMG B Scoop tube Lockout

• Panel 9-4-3 / A-5 RRMG B BKR 1DS Trip

• Panel 9-5-1 / C-7 LPRM Downscale The Operator enters 5.3DC125 and starts executing the procedure. Which Panel has been lost and what additional Abnormal Procedures should be entered along with 5.3DC125?

a. BB2, 2.4RX-PWR,REACTOR POWER ANOMALIES.

b. BB2, 2.4RR,REACTOR RECIRCULATION ABNORMAL.

c. BB3, 2.4RX-PWR,REACTOR POWER ANOMALIES.

d. BB3, 2.4RR, REACTOR RECIRCULATION ABNORMAL.

ANSWER: 56

d. BB3, 2.4RR, REACTOR RECIRCULATION ABNORMAL.

Explanation:

Emergency Procedure 5.3DC125, The Entry Conditions are Multiple alarms on remaining Control Room annunciator panels, indicating loss of D C control power.

When 125VDC panel BB3 is lost, the associated annunciators are for the Reactor Recirc MG Set and ARI/ATWS RPT Trip Logic. The loss of the Reactor Recirc Pump causes a reduction of power to approximately 63% which has caused the LPRM Downscale alarm.

Knowing the specific power panel which feed the logic to ARI/ATWS RPT will help deduce that BB3 is the panel that has lost power.

The additional Abnormal Procedure to be entered, is 2.4RR, Reactor Recirculation Abnormal, because one of the entry conditions is Trip of Reactor Recirculation pump.

Distracters:

a. BB2, 2.4RX-PWR,REACTOR POWER ANOMALIES. This selection is incorrect because the Power Panel that is lost is incorrect. However a candidate might choose this answer if they believed the logic power for ARI/AWTS RPT was powered by panel BB2. BB2 supplies logic power for the DIV 2 ECCS Logic, but not the ARI/ATWS RPT Logic. Also the additional Abnormal Procedure to enter is incorrect. However a candidate might

choose this answer if they believed the LPRM Downscale alarm was an unexplained drop in reactor power as indicated by: APRMs, Reactor steam flow, Reactor feed flow, Main generator electrical output, and Reactor total core flow. These are the entry conditions for 2.4RX-PWR.

b. BB2, 2.4RR,REACTOR RECIRCULATION ABNORMAL. This selection is incorrect because the Power Panel that is lost is incorrect. However a candidate might choose this answer if they believed the logic power for ARI/AWTS RPT was powered by panel BB2.

BB2 supplies logic power for the DIV 2 ECCS Logic, but not the ARI/ATWS RPT Logic.

c. BB3, 2.4RX-PWR,REACTOR POWER ANOMALIES. This selection is incorrect because the additional Abnormal Procedure to enter is incorrect. However a candidate might choose this answer if they believed the LPRM Downscale alarm was an unexplained drop in reactor power as indicated by: APRMs, Reactor steam flow, Reactor feed flow, Main generator electrical output, and Reactor total core flow.

These are the entry conditions for 2.4RX-PWR.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295025.EA2.02 Importance Rating 4.2 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 2.3_9-5-2 (Attach if not previously provided)

(including version/revision number) 28 Learning Objective: See Attached (As available)

Question Source: Bank # 19077 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y NRC Style 57 19077 02 12/31/2008 SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Systems COR0022102, RPS Main Turbine trip with MTBVs initially open Related Lessons COR0022102 REACTOR PROTECTION SYSTEM Related Objectives COR0022102001040J Describe the RPS design features and/or interlocks that provide for the following: Bypassing of selected scram signal (manually and automatically)

SKL012422100A030I Given plant conditions, predict changes in RPS components/parameters: Reactor power.

Related References 2.3_9-5-2 Panel 9 Annuciator 9-5-2 Related Skills (K/A) 295025.EA2.02 Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: (CFR: 41.10 / 43.5 / 45.13) Reactor power (4.2* /

4.2)

QUESTION: 57 19077 With the plant initially at 35% power and all conditions normal, a voltage surge in the DEH control system results in the following plant conditions:

• Main Generator load is 192 MWe.

• Main Turbine Bypass Valves are partially open.

• Total Main Steam flow is 35%.

• Annunciator 9-5-2/C-4, TSV & TCV CLOSURE TRIP BYP CHAN A/B, is in alarm.

1. (2704) TSV & TCV CLOSURE TRIP BYPASSED CHAN A1

2. (2705) TSV & TCV CLOSURE TRIP BYPASSED CHAN A2

3. (2706) TSV & TCV CLOSURE TRIP BYPASSED CHAN B1

4. (2707) TSV & TCV CLOSURE TRIP BYPASSED CHAN B2 What is the expected plant response if a Main Turbine trip should subsequently occur?

a. Reactor scrams on high reactor pressure.

b. Reactor power lowers to 25% due to Rx Recirc runback.

c. Reactor continues to operate at 35% power.

d. Reactor scrams on TSV closure.

ANSWER: 57 19077

a. Reactor scrams on high reactor pressure.

Explanation:

Since the TSV/TCV closures are bypassed due to the bypass valves being open a turbine trip will not initiate a scram but will increase reactor pressure causing on a scram on high reactor pressure.

Distracters:

b. is incorrect. Reactor power will not automatically lower to bypass valve capacity; operator action must be taken to reduce reactor power.

c. is incorrect. When the turbine trips reactor power is greater than bypass valve capacity, so reactor power cannot continue steady at 35% power.

d. is incorrect. The TSV closure scram signal is bypassed as indicated by annunciator 9-5-2, C-4.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295031.EK1.01 Importance Rating 4.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) INT008-06-18 (Attach if not previously provided)

(including version/revision number) 18 Learning Objective: See Attached (As available)

Question Source: Bank # 12325 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 8 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 58 12325 02 03/28/2006 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 4 H 1 6 Multiple Choice Topic Area Description Emergency Operating INT0080607, FLOWCHART 2A - EMERGENCY RPV Procedures DEPRESSURIZATION/ STEAM COOLING Related Lessons INT0080607 OPS EOP Flowchart 2A - Emergency RPV Depressurization & Steam Cooling Related Objectives INT00806070010800 Given plant conditions and EOP flowchart 2A, EMERGENCY RPV DEPRESSURIZATION/STEAM COOLING, state the reasons for the actions contained in the steps.

INT00806070010700 Given plant conditions and EOP flowchart 2A, EMERGENCY RPV DEPRESSURIZATION/STEAM COOLING, determine required actions.

Related References 10CFR55.41 Written examinations: Operators INT0080607 Flowchart 2A Emergency RPV Depressurization INT008-06-18 OPS EOP and SAG Graphs and Cautions Related Skills (K/A) 295031.EK3.04 Knowledge of the reasons for the following responses as they apply to REACTOR LOW WATER LEVEL: (CFR: 41.5 / 45.6) Steam cooling.(4.0/4.3*)

295031.EK1.01 Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL: (CFR: 41.8 to 41.10)

Adequate core cooling. (4.6*/4.7*)

QUESTION: 58 12325 The plant was operating at power when a station blackout occurred. HPCI and RCIC will not inject into the RPV. The crew is able to operate RCIC in a pressure control mode. Reactor water level lowered to -158 inches (corrected FZ) and the crew entered steam cooling. RPV pressure stabilized using the RCIC system at a reactor pressure of 800 psig.

Five minutes later, the following plant conditions were present:

• Reactor water level -183 inches (corrected FZ) and slowly lowering

• Reactor pressure 810 psig

• Average drywell temp 210ºF

• RCIC turbine speed 4500 RPM

• RCIC Flow 500 gpm What action is required and why?

a. Emergency depressurize because the core may not be adequately cooled.

b. Alternate emergency depressurize to preserve post depressurization inventory.

c. Continue to operate RCIC because adequate core cooling is assured until reactor level reaches -202 inches (corrected FZ).

d. Allow pressure to rise and be controlled by SRV pressure relief setpoint(s) to ensure adequate steam flow to cool the core.

ANSWER: 58

c. Continue to operate RCIC because adequate core cooling is assured until reactor level reaches -202 inches (corrected FZ).

Explanation:

Emergency Depressurization is required at -202 inches FZ (corrected). Operation of RCIC is allowed to slow the pressure rise allowing continued steam cooling. Boil-off is occurring while level is slowly lowering maximizing the time available for steam cooling to occur.

INT008-06-18 Both RO and SRO ENABLING OBJECTIVES

1. Using the graphs provided in the EOP and SAG Graphs Flowchart, explain how the shape of each curve or family of curves was determined.

2. For each graph used in the flowcharts, identify the action(s) required if the parameters associated indicate operation in the restricted or prohibited area.

3. Given plant conditions and the EOP and SAG Graphs Flowchart, determine if operation is within the allowed region of a graph.

4. Using the Cautions provided in the EOP and SAG Flowcharts, explain the bases behind each of the Cautions.

Distracters:

a. -183 inches is the level you emergency depressurize at if you have an injection source available. In this case you do not so you are directed to wait until level drops to -202 inches.

b. Alternate emergency depressurize is not allowed if you should wait for -202 inches to perform an emergency depressurization.

d. If RCIC is still incapable of injection, there is no reason to remove it from pressure control and allow pressure to rise to the SRV setpoints.

Provide to Candidate: EOP flowchart 2A with cautions removed.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295008.AK1.03 Importance Rating 3.4 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Tech Specs Bases 3.3.2.2 (Attach if not previously provided)

(including version/revision number) 0, 1, 2/20/07 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 59 00 01/12/2011 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 4 Multiple Choice Topic Area Description Abnormal/Emergency What is the reason for the high level trip of the main turbine?

Related Lessons COR0023202 OPS REACTOR VESSEL LEVEL CONTROL INT0320135 CNS Abnormal Procedures (RO) - Condensate/Feedwater Related Objectives COR0023202001090D Given a specific RVLC system malfunction determine the effect on any of the following: RPV water level INT0320135H0H0100 Given plant condition(s), determine from memory any automatic actions listed in the applicable Abnormal/Emergency Procedure(s) which will occur due to the event(s).

Related References Tech Specs Bases 3.3.2.2 Related Skills (K/A) 295008.AK1.03 Knowledge of the reasons for the following responses as they apply to HIGH REACTOR WATER LEVEL: (CFR: 41.5 / 45.6) Main turbine trip (3.4 / 3.5)

QUESTION: 59 Why does the Main Turbine trip on a high reactor water level?

a. Provide anticipatory Scram to prevent exceeding MCPR safety limit.

b. Provide anticipatory Main Turbine trip when both Reactor Feed Pumps trip.

c. To prevent moisture carryover into the Main Turbine that could cause turbine blade damage.

d. To prevent a rapid cooldown of the Main Steam Lines and water hammer that could potentially cause system damage.

ANSWER: 59

c. To prevent moisture carryover into the Main Turbine that could cause turbine blade damage.

Explanation:

Tech Specs 3.3.2.2 Bases - BACKGROUND - The feedwater and main turbine high water level trip instrumentation is designed to detect a potential failure of the Feedwater Level Control System that causes excessive feedwater flow. With excessive feedwater flow, the water level in the reactor vessel rises toward the high water level, Level 8 reference point, causing the trip of the two feedwater pump turbines and the main turbine.

Reactor Vessel Water Level-High, Level B signals are provided by level sensors that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level in the reactor vessel (variable leg). Three channels of Reactor Vessel Water Level-High, Level B instrumentation are provided as input to a-two-out-of-three initiation logic that trips the two feedwater pump turbines and the main turbine.

Each channel consists of a level transmitter loop and a trip relay that compares measured input signals with pre-established-setpoints. When the setpoint is exceeded, the channel outputs a main feedwater and main turbine trip signal to the trip logic. A trip of the feedwater pump turbines limits further increase in reactor vessel water level by limiting further addition of feedwater to the reactor vessel. A trip of the main turbine and closure of the stop and control valves protects the turbine from damage due to water entering the turbine.

Distracters:

a. Although tripping the main turbine provides an anticipatory Scram on high water level it is not there to prevent exceeding MCPR safety limit. The Reactor Feed Pump trip to provide this function.

b. There is no anticipatory Main Turbine trip on a trip of both Reactor Feed Pumps. This answer is a combination of two of the other answers that are incorrect.

d. This is the reason the MSIVs are closed at +90 inches if an overfill condition were to occur.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295002.AK2.08 Importance Rating 3.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR001-02-01 (Attach if not previously provided)

(including version/revision number) 24 Learning Objective: See Attached (As available)

Question Source: Bank # 2590 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 60 2590 00 01/03/2011 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 4 Multiple Choice Topic Area Description Systems COR0010202001040A Circulating Water Related Lessons COR0010202 OPS Circulating Water Related Objectives COR0010202001040A Given a specific Circulating Water malfunction, determine the effect on any of the following: Main Condensate System Related References COR0010202 OPS Circulating Water Related Skills (K/A) 295002.AK2.08 Knowledge of the interrelations between LOSS OF MAIN CONDENSER VACUUM and the following: (CFR: 41.7 / 45.8) Condenser circulating water system 3.1 / 3.2

QUESTION: 60 2590 The plant is operating at 60% power when one of the operating circulating water pumps trip.

What is the effect on the plant when this pump trips?

a. Main Generator output rises.

b. Condenser vacuum will degrade.

c. Overall plant efficiency improves.

d. Condensate Pump NPSH increase.

ANSWER: 60 2590

b. Condenser vacuum will degrade.

Explanation:

At 60% power the loss of one Circulating Water Pump will cause cooling to decrease in the Main Condenser effectively lowering condenser vacuum.

Distracters:

a. The turbine efficiency decreases therefore causing an output reduction

c. The turbine efficiency decreases therefore overall plant efficiency lowers

d. The turbine efficiency decreases therefore causing an increase in the hotwell temperatures and a reduction of NPSH to the Condensate Pumps

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295015.AK3.01 Importance Rating 3.4 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-26-02 , 5.8.3 (Attach if not previously provided)

(including version/revision number) 19 13 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 5 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 61 New 00 01/07/2011 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 3 Multiple Choice Topic Area Description Emergency Operating Rod insertion during an ATWS why bypass the insertion blocks?

Procedures Related Lessons INT0080606 FLOWCHART 6A - RPV PRESSURE/POWER (FAILURE-TO-SCRAM)

Related Objectives INT00806060010600 List the methods of alternate rod insertion.

Related References PRO 5.8.3 LP COR002-26-02 Related Skills (K/A) 295015.AK3.01 Knowledge of the reasons for the following responses as they apply to INCOMPLETE SCRAM: (CFR: 41.5 / 45.6) Bypassing rod insertion blocks (3.4 / 3.7)

QUESTION: 61 The Plant was operating at near rated power when a steam line rupture in the HPCI room occurred. A manual scram was attempted. An ATWS occurred, with the following conditions:

• 25 control rods are at position 24 or GREATER.

• All scram valves are open.

The CRS has directed you to enter 5.8.3, ALTERNATE ROD INSERTION METHODS and insert control rods with RMCS.

Why is the Rod Worth Minimizer bypassed for this evolution?

a. To prevent all Rod Block Monitor rod blocks.

b. To prevent inadvertent rod withdrawals during the ATWS.

c. To allow the Control Rods to be driven in with the Rod Movement Control Switch.

d. To allow the Control Rods to be selected on the select matrix and driven in with the Emergency Notch Override Control Switch.

ANSWER: 61

d. To allow the Control Rods to be selected on the select matrix and driven in with the Emergency Notch Override Control Switch.

Explanation:

In accordance with Emergency Procedure 5.8.3 the RWM is bypassed in step ARI-20 to allow the Rod Insertion Blocks to be cleared so the emergency notch override switch can be used to insert the control rods. The step is in the flowchart that is scanned into the procedure so a word search will not locate it.

Distracters:

a. Bypassing the Rod Worth Minimizer does not prevent the rod blocks from the RBM.

b. Bypassing the Rod Worth Minimizer will not prevent the operator from driving the control rod in the wrong direction (i.e. OUT)

c. The control rods are not driven with the normal Rod Movement Control Switch in the event there is an ATWS, Emergency Notch Override Switch is positioned in the INSERT direction.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295032.EA1.05 Importance Rating 3.7 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedure 2.1.22 (Attach if not previously provided)

(including version/revision number) 55 Learning Objective: See Attached (As available)

Question Source: Bank # 2622 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 62 2622 01 08/24/1999 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 3 Multiple Choice Topic Area Description Systems COR0020302001060D Containment Related Lessons COR0020302 OPS CONTAINMENT Related Objectives COR0020302001060D Describe the interrelationship between PCIS and the following: HPCI Related References PR 2.1.22 Related Skills (K/A) 295032.EA1.05 Ability to operate and/or monitor the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: (CFR: 41.7 / 45.6)

Affected systems so as to isolate damaged portions (3.7 /3.9)

QUESTION: 62 2622 The HPCI System auto initiated but shortly after the following conditions were indicated:

• HPCI System Flow Normal

• Reactor Level +20 inches

• HPCI Area Temperature 205F

• D/W Pressure 1.2 psig

• HPCI Steam Supply Pressure 925 psig What automatic actions will occur and why?

a. A Group 4 isolation from steam line low pressure.

b. A Group 4 isolation from high area temperature.

c. A Group 5 isolation from high area temperature.

d. A Group 5 isolation from high drywell pressure.

ANSWER: 62 2622

b. A Group 4 isolation from high area temperature.

Explanation:

With the HPCI Area Temperature 205F this has exceeded the trip setpoint for the isolation.

Group 4 is HPCI valves and Group 5 is RCIC.

Distracters:

a. HPCI Steam Supply Low pressure isolation is approximately 100 psig we are well above that.

c. Group 5 are the RCIC Valves not the HPCI ones, RCIC does not isolate on a HPCI Steam Line high area temperature.

d. Group 5 are the RCIC Valves not the HPCI ones, RCIC does not isolate on high drywell pressure.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295022.AA2.01 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Tech Specs 3.1.5 (Attach if not previously provided)

(including version/revision number) Amendment 178 Learning Objective: See Attached (As available)

Question Source: Bank # Pilgrim Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Pilgrim 2009 Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 63 00 01/07/11 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 4 Multiple Choice Topic Area Description Technical Specifications, INT0070502, CNS Tech Spec 3.1, Reactivity Control Systems ODAM, TRM Related Lessons INT0070502 CNS Tech. Spec. 3.1, Reactivity Control Systems Related Objectives INT00705020010300 Given a set of plant conditions that constitutes non-compliance with a Section 3.1 LCO, determine the ACTIONS that are required.

Related References 3.1.5 Control rod scram accumulators Related Skills (K/A) 295022.AA2.01 Ability to determine and/or interpret the following as they apply to LOSS OF CRD PUMPS: (CFR: 41.10 / 43.5 / 45.13) Accumulator pressure (3.5 / 3.6)

QUESTION: 63 A Plant startup is in progress with the following conditions:

• The Reactor MODE Switch is in the Startup position.

• Procedure 2.1.1, STARTUP PROCEDURE is at the point of placing the first Reactor Feed Pump in service.

• CRD Pump A is out of service for bearing replacement.

When the following occurs:

• CRD Pump B trips on over-current

• CRD ACCUM LOW PRESS OR HIGH LEVEL Alarms

• Full Core display indicates that there are two (2) accumulator alarms on control rods that are at position 48.

• The Station Operator verifies that they are caused by low pressure.

In accordance with Technical Specifications 3.1.5 Control Rod Scram Accumulators what action is required?

a. Declare the associated control rods inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

b. Declare the associated control rod scram times slow within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

c. Place the Reactor MODE Switch in the shutdown position immediately.

d. Restore CRD Charging Header pressure to > 940 psig within 20 minutes.

ANSWER: 63

c. Place the Reactor MODE Switch in the shutdown position immediately.

Explanation:

With the plant at the point of placing the first feed pump in service, Reactor Pressure should be around 400 psig in accordance with the Startup Procedure. When the running CRD Pump trips and charging header pressure drops below 940 psig, Tech Specs 3.1.5 Condition C states to verify that the associated control rods are fully inserted. In this case they were full out. So condition C could not be met and the operator is required to enter condition D, that states to place the reactor mode switch in the shutdown position immediately.

Distracters:

a. Only allowed by Tech Specs if Reactor Steam pressure is above 940 psig.

b. Only allowed by Tech Specs if Reactor Steam pressure is above 940 psig.

d. Only allowed by Tech Specs if Reactor Steam pressure is above 940 psig.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295033 G 2.4.1 Importance Rating 4.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 5.8.4, EOP- 1A (Attach if not previously provided)

(including version/revision number) 16, 15 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 64 00 02/03/2011 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Systems COR00223020011600, COR0022302 Residual Heat Removal Related Lessons COR0022302 RESIDUAL HEAT REMOVAL Related Objectives COR00223020011600 Given plant conditions including a Shutdown Cooling isolation, determine actions required to place RHR in the LPCI mode.

Related References COR002-23-02 PR 5.8.4 ALTERNATE INJECTION SUBSYSTEMS Related Skills (K/A) 295033 High Secondary Containment Area Radiation Levels Generic 2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. (CFR: 41.10 /

43.5 / 45.13) IMPORTANCE RO 3.8 SRO 4.2

QUESTION: 64 The Plant is operating at 40% power when a fuel failure causes the plant to be scrammed and cooled down.

A LOCA then occurs with the following plant conditions:

• RPV water level is -165 inches lowering.

• RVP Pressure is 100 psig.

• All normal injection systems are unavailable.

• DRYWELL RAD MONITOR RMA-RM-40A, B are both reading 5 x 104 rem/hour.

• The TSC is NOT operational yet.

Which Alternate Injection Subsystem(s) (Table 4) can be used at this time?

Alternate Injection using

a. RHRSW Cross-Tie.

b. SLC System from Demin Water.

c. ECCS Pressure Maintenance.

d. CRD System.

ANSWER: 64

a. RHRSW Cross-Tie.

Explanation:

If DRYWELL RAD MONITOR RMA-RM-40A (PNL 9-02) or DRYWELL RAD MONITOR RMA-RM-40B (PNL 9-10) is reading > 104 rem/hour, entry into Secondary Containment is prohibited until TSC is operational and personnel can be dispatched per Procedure 5.7.15.

In accordance with EOP-1A RPV Control, with reactor water level below -158 inches and there are no normal injection Subsystems (Table 5) lined up for injection with a pump running, step RC/L-13 states Make available for injection with pumps running Alternate Injection Subsystems (Table 4). From EOP-5.8.4 Alternate Injection Subsystems (Table 4), the systems are, RHR SW Cross-tie, SLC from the Boron Tank or Demin Water, ECCS Pressure Maintenance, or injection with CRD System.

The only system that can be totally operated from outside Secondary Containment is RHRSW.

The other systems require an operator to access the Reactor Building to perform some actions to align the systems for injection.

Distracters:

b. Requires the operator to operate valves that are located in the Rx Building on the 976 foot elevation. This is prohibited until the TSC is operational in accordance with procedure 5.8.4 section 3, Dispatching Personnel for EOP Actions.

c. Requires RPV Pressure to be less than 50 psig and to fully implement the procedure the operator has to operate valves that are located in the Rx Building on the 881 foot elevation in the NW Quad. This is prohibited until the TSC is operational in accordance with procedure 5.8.4 section 3, Dispatching Personnel for EOP Actions.

d. Requires the operator to operate valves that are located in the Rx Building on the 903 foot elevation SE. This is prohibited until the TSC is operational in accordance with procedure 5.8.4 section 3, Dispatching Personnel for EOP Actions.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295029 G 2.4.9 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 5.8.4, Lesson COR002-23-02 (Attach if not previously provided)

(including version/revision number) 16 27 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 65 New 0 01/07/2011 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 4 Multiple Choice Topic Area Description Emergency Operating INT0080613 OPS EOP FLOWCHART 3A - PRIMARY Procedures CONTAINMENT CONTROL Related Lessons NT0080613 OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL Related Objectives NT0080613 OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL Related References EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL Related Skills (K/A) 295029 High Suppression Pool Water Level Generic 2.4.1: Knowledge of EOP entry conditions and immediate action steps. (CFR: 41.10 /

43.5 / 45.13) IMPORTANCE RO 4.6 SRO 4.8

QUESTION: 65 The plant is operating at 100% power when valve CS-66 Core Spray Pump A Condensate Supply starts leaking by. There are no Operator actions taken.

At what level in the suppression pool will entry into EOP-3A, PRIMARY CONTAINMENT CONTROL be required?

a. + 1.5 inches

b. + 2.0 inches

c. - 1.5 inches

d. - 2.0 inches ANSWER: 65

b. + 2.0 inches Explanation:

With the Core Spray Suction Valve leaking, suppression pool level will be rising, because the ECST is at a higher elevation than the water in the torus. As with water level begins to rise +1.5 inches will cause the Torus High Water Level Alarm to sound, however, it is not the entry level for the EOPs. Level has to rise above + 2.0 inches to actually cause entry into EOP 3A Primary Containment Control.

Distracters:

a. This is the high alarm setpoint, not the entry condition.

c. This is the low alarm setpoint, and should not come in because the leaking of the suction valve should cause level in the torus to rise not fall.

d. This is the low entry point for EOP 3A, the leaking of the suction valve should cause level in the torus to rise not fall.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G 2.1.5 Importance Rating 2.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedure 2.0.3 (Attach if not previously provided)

(including version/revision number) 73 Learning Objective: See Attached (As available)

Question Source: Bank # 16465 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 66 16465 01 01/01/11 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 4 Multiple Choice Topic Area Description None Crew manning Related Lessons INT0320103 CNS Administrative Procedures Conduct of Operations and General Alarm Procedures (Formal Classroom/Pre-OJT Training)

INT0070513 CNS Technical Specifications 5.0, Administrative Controls Related Objectives INT032010300C010H Discuss the following as described in conduct of Operations Procedure 2.0.3, Conduct of Operations: Control Room and Station Shift Staffing Requirements INT032010300C0400 Discuss the following as described in conduct of Operations Procedure 2.0.3, Conduct of Operations: Given a Control Room staffing level, determine if the proper staffing requirements are met.

INT00705130010100 Given a set of plant conditions, recognize non-compliance with a Chapter 5.0 Requirement.

INT00705130010200 Given a set of conditions that constitutes non-compliance with a Chapter 5.0 Requirement, determine the actions that are required.

Related References Procedure 2.0.3, Conduct of Operations Related Skills (K/A) 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (CFR: 41.10 / 43.5 / 45.12) IMPORTANCE RO 2.9* SRO 3.9

QUESTION: 66 16465 Given the following conditions:

• The plant is in MODE 4

• Shutdown cooling in service

• No work is scheduled Which one of the following crew compliments meets the MINIMUM requirement for ACTIVE LICENSED OPERATOR personnel required to be physically present in the Control Room?

a. Only 1 licensed operator; must be a SRO.

b. Only 1 licensed operator; either an RO or SRO.

c. 2 licensed operators; must be 2 SROs.

d. 2 licensed operators; must be an RO and SRO.

ANSWER: 66 16465

b. Only 1 licensed operator; either an RO or SRO.

Explanation:

Two active licensed operators (one of which must be an SRO) are required in MODE 4 (Cold Shutdown). Only one licensed operator must be in the Control Room at the controls. This will normally be an RO; however, per section 10, note 2, and higher grade licensed operators may take the place of lower grade licensed operators, therefore, an SRO meets the requirement.

REFERENCE:

2.0.3 Distracters

a. Can be an RO or SRO. An SRO is not required.

c. Only one licensed operator must be in the Control Room at the controls. The operator may believe two licensed operators are required (which is likely because 2 people but not necessarily licensed operators are required in the control room for security purposes). If this was the case, then this could be 2 SROs because step 10.2.4.1, higher grade licensed operators may take the place of lower grade licensed operators, therefore, an SRO meets the requirement.

d. Only one licensed operator must be in the Control Room at the controls. The candidate may believe two licensed operators are required (which is likely because 2 people are required in the control room for security purposes).

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.17 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedure 2.0.3 (Attach if not previously provided)

(including version/revision number) 73 Learning Objective: See Attached (As available)

Question Source: Bank # 19155 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 67 19155 01 12/04/2008 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 2 Multiple Choice Topic Area Description Administrative INT0320103, "Round of Containment Parameters" requirements Related Lessons OTH0151003 Focus Area Standards INT0320103 CNS Administrative Procedures Conduct of Operations and General Alarm Procedures (Formal Classroom/Pre-OJT Training)

Related Objectives OTH0151003001010D From memory define the following terms in accordance with in procedure 2.0.3, Conduct of Operations, and Operations Instruction

1. 7: Briefs and updates INT032010300C010J Procedure 2.0.3, Conduct of Operations: Discuss the following as described in Conduct of Operations Procedure 2.0.3, Conduct of Operations: Announcing Parameters and Trends Related References 2.0.3 Conduct of Operations Related Skills (K/A) 2.1.17 Ability to make accurate, clear, and concise verbal reports. (CFR: 41.10 / 45.12 / 45.13)

IMPORTANCE RO 3.9 SRO 4.0

QUESTION: 67 19155 In accordance with Procedure 2.0.3, CONDUCT OF OPERATIONS; what is required to be provided during a report of critical parameter that are out of the assigned band during abnormal or emergency situation?

Parameter value,

a. trend shall be given and rate should be provided.

b. trend, rate and reason why the parameter is out of band.

c. rate and the system being used to control the parameter only.

d. rate, trends and the system being used to control the parameter only.

ANSWER: 67 19155

b. trend, rate and reason why the parameter is out of band.

Explanation:

From Conduct of Operations Procedure 2.0.3, Section 7 All disciplines communicating with the Control Room shall use three (3)-way communication for all orders/directions that involve operation of plant equipment or exchange of critical information related to a given evolution.

When reporting a parameter, the VALUE and TREND shall be given. If the parameter is also outside of the assigned band, the reason should also be stated.

As plant conditions permits and when it can be established, a RATE should be reported with the given parameter.

Distracters:

a. trend shall be given and rate should be provided would be correct if the parameter was not out of the specified band. This will also require the reason why the parameter is out of the band.

c. rate and the system being used to control the parameter only is not correct because the report of the current system being used is not required. This information might be thought to be in the required report, but just the reason is required. Also the trend is required.

d. rate, trends and the system being used to control the parameter only, is incorrect because the reason why the parameter is out of band is required, not the system being used to control it.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # G 2.2.14 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedure 0.9 (Attach if not previously provided)

(including version/revision number) 74 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date 68 00 3-28-11 NRC Style Question RO: Y SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 6 Multiple Choice N Topic Area Description Admin What is one approved method for including this control switch to the Section Tags List?

Related Lessons SKL0080306R01-L-NOMS Tagout Preparer Related Objectives

4. Use the NOMS Equipment module to update equipment information in a tagout or section.

Related References 0.9 Related Skills (K/A) ROI SROI 2.2.14 Knowledge of the process for controlling equipment configuration or status. (CFR: 41.10 / 43.3 / 45.13) IMPORTANCE RO 3.9 SRO 4.3

QUESTION: 68 When tagging a control switch in a position that does not reflect the actual position of the remote component, the position of the remote component shall be included in the Section Tags list.

What is the approved method from Procedure 0.9 TAGOUT, for including this control switch to the Section Tags List?

a. No-Tag, with required hang component position identified.

b. No-Tag, with a note in the Workers Notes section that states a total system lineup is to be performed.

c. Step Tag, with required positions for hang and/or release for the remote component.

d. Step Tag, with required post maintenance procedure steps identified to be performed during the hanging of the Tagout Order.

ANSWER: 68

c. Step Tag, with required positions for hang and/or release for the remote component.

Explanation:

Procedure 0.9 4.19 CONTROL SWITCH REQUIREMENTS 4.19.1 A control switch should not be used as the sole energy isolation device for personnel or equipment protection unless no other isolation methods are available.

4.19.1.1 If it is necessary that a control switch provide the only barrier for personnel or equipment protection, then this hazard shall be listed in the Worker Notes and Hazards Section of the associated Tagout and the circuit treated as live during subsequent maintenance activities.

4.19.2 When tagging a control switch in a position that does not reflect the actual position of the remote component, the position of the remote component shall be included in the Section Tags list by one of following methods:

4.19.2.1 Step Tag with required positions for hang and/or release for the remote component.

4.19.2.2 No-Tag with required release position if a placement position is not required.

Distracters:

a. No-Tag, with required component position identified. This is incorrect, because there is no requirement to list the required component position.

b. No-Tag, with a note in the Workers Notes section that states a total system lineup is to be performed.

d. Step Tag, with required post maintenance procedure steps identified to be performed during the hanging of the Tagout Order. This is incorrect because a Tagout order should not be used to direct post maintenance testing during the hanging of the order.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # G 2.2.14 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedure 0.31 (Attach if not previously provided)

(including version/revision number) 61 Learning Objective: See Attached (As available)

Question Source: Bank # 12203 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 69 12203 03 12/04/2008 Licensed RO: Y Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 3 Multiple Choice Topic Area Description Administrative INT0320101, System Component Checklist is being performed; a valve is found out of position. What actions are required for this condition and why?

Related Lessons INT0320101 CNS Administrative Procedures Volume 0, Administrative Procedures (Formal Classroom/Pre-OJT Training)

Related Objectives INT032010100H010N Discuss the following as described in Administrative Procedure 0.31, Equipment Status Control: System line-up deviations Related References 10CFR55.41 Written examinations: Operators 0.31 Equipment Status Control Related Skills (K/A) 2.2.14 Knowledge of the process for controlling equipment configuration or status. (CFR:

41.10 / 43.3 / 45.13) IMPORTANCE RO 3.9 SRO 4.3

QUESTION: 69 12203 The plant is recovering from a refueling outage. While the HPCI System Component Checklist is being performed, the independent verification performer finds that HPCI-V-23, MAIN PUMP VENT is open, which is the incorrect position in accordance with the line-up checklist.

What actions are required for this condition?

Immediately notify the Shift Manager and

a. wait for the SM to direct its closing, then contact the Work Control Supervisor and write a CR.

b. write a CR, notify the Work Control Supervisor and have the QA group perform an investigation into the out of position valve before closing it.

c. close the valve when directed by the SM, initial the checklist, write a CR, and record the discrepancy on the discrepancy sheet and attach it to the Checklist.

d. initial the checklist, write a CR, notify the QA group, and record the discrepancy on the discrepancy sheet and attach it to the Checklist.

ANSWER: 69 12203

c. close the valve when directed by the SM, initial the checklist, write a CR, and record the discrepancy on the discrepancy sheet and attach it to the Checklist.

Explanation:

Procedure 0.31 Page 18: System Line-Up Deviations When there are multiple positions listed as a Normal position for a component, the AS FOUND position of the component shall be documented. If the AS FOUND position matches the applicable comment, no discrepancy exists. If the AS FOUND position does not match the applicable comment, the actions specified in Step 17.2 shall be taken.

If while performing a Performer Verification or an Independent Verification of a System Component Checklist a component is found in other than the Normal position, perform following:

• Immediately notify the SM.

• Position the component as directed by the SM and initial the System Component Checklist for Performed By.

• The SM shall ensure an Independent/Concurrent Verification is performed.

• Ensure CR is generated addressing all applicable information available.

• Record component description, AS FOUND position and AS LEFT position on a Discrepancy Sheet and ensure it is attached to the System Component Checklist.

Distracters:

a. is incomplete, you must also initial the checklist, write a CR, and record the discrepancy on the discrepancy sheet and attach it to the Checklist. The WCC Supervisor is not required to be contacted.

b. writing a notification and have the QA group look into the out of position valve may be appropriate but not before closing it.

d. ADMINISTRATIVE PROCEDURE O-QA-11 OVERSIGHT STANDARDS AND EXPECTATIONS states in step 3.8.1.5 Notify the Supervisor/Leader if problems occur during procedure performance due to procedure deficiencies. This is not a procedure deficiency so there is no need to contact QA Supervisor.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.11 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson INT008-06-17 (Attach if not previously provided)

(including version/revision number) 16 Learning Objective: See Attached (As available)

Question Source: Bank # 23483 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 12 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: Y 70 00 02/10/2011 NRC Style Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 4 Multiple Choice N Topic Area Description Emergency Operating INT0080617, Why do you restart TB Vent while in 5A? (ILT 2006 Procedures NRC EXAM)

Related Lessons INT0080617 OPS FLOWCHART 5A - SECONDARY CONTAINMENT AND RADIOACTIVITY RELEASE CONTROL Related Objectives INT00806170010700 Given plant conditions and EOP flowchart 5A, SECONDARY CONTAINMENT CONTROL and RADIOACTIVITY RELEASE CONTROL, state the reasons for the actions contained in the steps.

Related References 10CFR55.41 (B)(12)

Related Skills (K/A) ROI SROI 2.3.11 Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10)

(3.8/4.3) (2.7 / 3.2)

QUESTION: 70 How are radiation releases controlled when there is a large steam leak in the Turbine Building?

a. Restart Turbine Building Ventilation if it is not running.

b. Start additional Elevated Release Point Dilution Fans.

c. Startup and place the Augmented Radwaste system in service.

d. Manually initiate a Group 6 Isolation and Standby Gas Treatment initiation.

ANSWER: 70

a. Restart Turbine Building Ventilation if it is not running.

Explanation:

Continued personnel access to the turbine building, radwaste and augmented radwaste may be essential for responding to emergencies. These structures are not air tight and radioactivity release inside them would not only limit personnel access, but would eventually lead to an unmonitored ground level release. Operation of ventilation in these structures preserves accessibility, and assures that radioactivity is discharged through an elevated, monitored release point.

Distracters:

b. 2.4OG Off-gas Abnormal has a section to start the dilution fans, but that is used during off-gas problems, not steam leaks.

c. Augmented Radwaste will only help for elevated radiation levels within the off-gas stream not leaks within the building.

d. These are the actions if the leak were in the Reactor Building.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.4 Importance Rating 3.2 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Rad Procedure 9.ALARA.1 (Attach if not previously provided)

(including version/revision number) 39 Learning Objective: See Attached (As available)

Question Source: Bank # 23484 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC CNS 2006 Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 12 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date 71 00 06/24/2006 NRC Style Question RO: Y 23484 SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 3 Matching N Topic Area Description Administrative INT0320115, Knowledge of radiation exposure limits (ILT 2006 NRC EXAM)

Related Lessons INT0320115 OPS CNS Administrative Procedures Radiation Protection and Chemistry Procedures (Formal Classroom/Pre-OJT Training)

Related Objectives INT0320115D0D010I Discuss the following as described in Rad Protection Procedure 9.ALARA.1, Personnel Dosimetry and Occupational Radiation Exposure Program: Lifetime TEDE Guideline Related References 10CFR55.41 (B)(12)

Related Skills (K/A) ROI SROI 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4 / 45.10) (3.2/3.7)

(2.5 / 3.1)

QUESTION: 71 If you have exceeded your Lifetime TEDE Guideline how much exposure are you allowed during the year at CNS? What authority, if any, may grant extension to this allowed exposure?

a. 1000 mrem Radiological Manager and Site Vice President may authorize an extension.

b. 0 mrem No extensions are allowed.

c. 1000 mrem No extensions are allowed.

d. 0 mrem Radiological Manager and Site Vice President may authorize an extension.

ANSWER: 71

c. 1000 mrem No extensions are allowed.

Explanation:

The Lifetime TEDE Guideline states that NPPD shall normally limit an individual's lifetime TEDE in rem to the individual's age in years. In addition an individual exceeding the lifetime TEDE Guideline will be limited to a TEDE of 1000 mrem and will not be granted an extension.

Distracters:

a. is incorrect even though 1000 mrem are allowed no extension to this dose is allowed.

b. is incorrect because 1000 mrem TEDE is allowed.

d. is incorrect because 1000 mrem TEDE is allowed and now extensions are allowed.

Source: Direct Cognitive Level 1 Difficulty 3 10CFR55.41(b)12

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.43 Importance Rating 3.2 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR001-03-02 (Attach if not previously provided)

(including version/revision number) 20 Learning Objective: See Attached (As available)

Question Source: Bank # 1927 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date 72 00 08/10/1999 Licensed Operator RO: Y 1927 SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 2 Multiple Choice N Topic Area Description Systems COR0010302001020H Communication System Related Lessons COR0010302 Communication Related Objectives COR0010302001020H State the purpose of the following major components in the Communications system: Emergency Notification System (ENS)

Related References COR0010302 Communication Related Skills (K/A) ROI SROI 2.4.43 Knowledge of emergency communications systems and techniques. (CFR: 41.10 / 45.13) (3.2/3.8) (2.8 / 3.5)

QUESTION: 72 What is the primary communications system that links the site and the NRC during a radiation release condition at the plant?

a. Emergency Notification System (ENS)

b. Health Physics Network (HPN)

c. Reactor Safety Counterpart Link (RSCL)

d. Protective Measures Counterpart Link (PMCL)

ANSWER: 72

a. Emergency Notification System (ENS)

Explanation:

Lesson COR001-03-02: Communication The Federal Telecommunication System provides emergency communication between the station and Federal authorities. The system is made up of seven networks.

a. Emergency Notification System (ENS)

b. Health Physics Network (HPN)

c. Reactor Safety Counterpart Link (RSCL)

d. Protective Measures Counterpart Link (PMCL)

e. Emergency Response Data System Channel (ERDS)

f. Management Counterpart Link (MCL)

g. Local Area Network Access (LAN)

In accordance with Communications Text (COR001-03-02); The ENS is intended as the primary means of reporting emergencies and other significant events at the station to the NRC.

When a station emergency occurs, the ENS becomes the dedicated and continuous line to the NRC for the transmission of operational data.

ENS designated telephones are located in the Control Room, NRC Resident Inspectors Office Technical Support Center, and Emergency Operations Facility.

Distracters:

b. The Health Physics Network (HPN) is not part of the ENS; however is one method of communicating with a branch of the NRC. A candidate might choose this answer if they incorrectly identify the section of the NRC that should be notified in the event of an emergency.

c. Established initially with the base team, and then with the NRC site team representatives once they arrive at the site, to conduct internal NRC discussions on plant and equipment

conditions separate from the licensee, and without interfering with the exchange of information between the licensee and NRC.

d. Established initially with the base team, and then with the NRC site team representatives once they arrive at the site, to conduct internal NRC discussions on radiological releases and meteorological conditions, and the need for protective actions separate from the licensee and without interfering with the exchange of information between the licensee and NRC.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.22 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) EOP 1A (Attach if not previously provided)

(including version/revision number) 15 Learning Objective: See Attached (As available)

Question Source: Bank # 8933 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 73 8933 02 02/01/2005 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Emergency Operating INT0080609, Following a LOCA with HPCI as only injection Procedures source, what is required?

Related Lessons INT0080609 OPS EOP FLOWCHART 1A - RPV CONTROL, RPV LEVEL Related Objectives INT00806090011100 Given plant conditions and EOP flowchart 1A, RPV CONTROL, determine required actions.

Related References INT0080609 Flowchart 1A RPV Level Related Skills (K/A) 2.4.22 Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations. (CFR: 41.7 / 41.10 / 43.5 / 45.12) (3.6/4.4)

QUESTION: 73 8933 Following a Loss of Coolant Accident, the following conditions exist:

• All control rods inserted to or beyond position 02.

• Reactor Pressure 445 psig (slowly lowering)

• Reactor water level -185 inches corrected FZ (slowly lowering)

• SRVs All closed

• HPCI is the ONLY high pressure injection source currently available (injecting at rated flow).

What action is now required and why?

a. Steam Cooling; HPCI is inadequate.

b. Steam Cooling; adequate core cooling is not being maintained.

c. Emergency Depressurization; Injection with Alternate Injection Subsystems.

d. Emergency Depressurization; Injection with low pressure ECCS is required.

ANSWER: 73 8933

d. Emergency Depressurization; Injection with low pressure ECCS is required.

Explanation:

With Reactor Pressure greater than the shutoff head of the low pressure injection systems and level is -185 and lowering, the reactor must be depressurized to inject with low pressure injection systems to restore reactor water level.

Distracters:

a. Low pressure injection is available, but pressure must be lowered to allow it to inject and restore level.

b. Steam Cooling is not appropriate. RPV injection is available.

c. Alternate Injection Subsystems are not required, the normal Low Pressure Coolant Injection Systems are available.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.32 Importance Rating 3.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 2.4ANN (Attach if not previously provided)

(including version/revision number) 9 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 10 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date 74 00 01/08/2011 NRC Style RO: Y Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 4 Multiple Choice Topic Area Description Abnormal/Emergency Loss of all Annunciators Related Lessons INT0320136 OPS-CNS Abnormal Procedures (RO) Miscellaneous Related Objectives INT0320136M0M0100 Given plant condition(s), determine from memory any automatic actions listed in the applicable Abnormal/Emergency Procedure(s) which will occur due to the event(s).

Related References 2.4ANN Related Skills (K/A) 2.4.32 Knowledge of operator response to loss of all annunciators. (CFR: 41.10 / 43.5 /

45.13) IMPORTANCE RO 3.6 SRO 4.0

QUESTION: 74 The plant is operating at 100%. The following indications are observed:

• A Transient occurs that should have caused multiple alarms on all of the Annunciator panels.

• No alarms windows light or sound.

• No CRTs display any alarms.

• No printers print any alarms.

How should the Panel Operator (BOP) respond to this loss?

Enter Abnormal Procedure 2.4ANN, stop all testing in progress and

a. send Station Operators into the field to hard boot the annunciator servers.

b. ensure the computer automatically fails-over to the backup PMIS computer.

c. test all of the annunciator lamps and horns to determine the extent of the failure.

d. pull individual annunciator output cards until the bad card is found and write a CR to have it replaced.

ANSWER: 74

c. test all of the annunciator lamps and horns to determine the extent of the failure.

Explanation:

In accordance with 2.4ANN;

• Stop all tests, evolutions, and power changes, until extent of failure known.

• If alarms continuously sound and Panel Annunciator Buttons are ineffective, then use of Master Silence feature can be used to control Control Room noise levels.

• If Annunciator A-1/F-1, ANNUNCIATOR SYSTEM FAILURE, is alarming with both lights lit, then consider both divisions lost and go to Step 4.9.

• Determine which lamps are lit for Annunciator A-1/G-1, ANUNCIATOR SYSTEM TROUBLE:

• Test all Annunciators lamps and horns.

• From all Video Indicating Displays, determine color of COMM A and B letters on display (left side of display, center).

• If both COMM A and B lights are red on all Video Indicating Displays and the Master Alarm Log, perform following:

• Test one or more individual alarms, such as the Radwaste Trouble or CST'A' Level alarm.

Do not use a Mux 18/Remote Input Chassis (RIC) 10 alarm points.

• If no Annunciator tile(s) and horn is received, then consider Annunciation lost and perform Step 4.9.

Distracters:

a. Rebooting of the annunciator servers is not an operations function and IT and I&C would have to investigate and repair this type of problem..

b. The PMIS computer does not fail over to the backup during a loss of Annunciators, it will on a loss of computers however.

d. Pulling of the annunciator output cards is not an operations function and a work order should be generated to have I&C and IT perform these actions.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.27 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Lesson COR002-19-02 (Attach if not previously provided)

(including version/revision number) 21 Learning Objective: See Attached (As available)

Question Source: Bank # 2527 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC NA Exam Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 7 55.43 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: Y Licensed 75 2527 01 08/20/1999 SRO: Y Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 2 Multiple Choice Topic Area Description Systems COR0021902001010A Reactor Equipment Cooling Related Lessons COR0021902 REACTOR EQUIPMENT COOLING Related Objectives COR0021902001010A State the purpose of the following items related to REC: REC System Related References COR0021902 REACTOR EQUIPMENT COOLING Related Skills (K/A) 2.1.27 Knowledge of system purpose and/or function. (CFR: 41.7) IMPORTANCE RO 3.9 SRO 4.0

QUESTION: 75 2527 What is the purpose of the Reactor Equipment Cooling (REC) system?

The REC system provides critical and non-critical cooling water to contaminated or potentially contaminated components in the Reactor Building

a. Radwaste and Control buildings only.

b. Augmented Radwaste and Control buildings only.

c. Radwaste and Augmented Radwaste buildings only.

d. Radwaste, Augmented Radwaste, and Control buildings only.

ANSWER: 75 2527

d. Radwaste, Augmented Radwaste, and Control buildings only.

Explanation:

From Lesson COR002-19-02 SYSTEM BRIEF DESCRIPTION A. System Purpose The REACTOR EQUIPMENT COOLING (REC) system provides cooling water to both the critical and non-critical, contaminated or potentially contaminated components located in the Reactor, Radwaste, Augmented Radwaste, and Control buildings.

Distracters:

a. Augmented Radwaste is also a load that is cooled by REC.

b. Radwaste is also a load that is cooled by REC.

c. Control Building is also a load that is cooled by REC.

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

Date: 6-13-2011 Facility/Unit: Cooper Nuclear Station Region: I II III IV Reactor Type: W CE BW GE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.

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

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

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295005 G 2.2.25 Importance Rating 4.2 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Tech Spec 3.8.4 and Bases (Attach if not previously provided)

(including version/revision number) 178 & 236 0, 04/22/10, and 04/11/06 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 55.43 2 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: N S1 NRC Style 00 05/13/2011 SRO: Y 76 Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 4 Multiple Choice Topic Area Description Technical Specifications, INT0070509, CNS Tech. Spec. 3.8, Electrical Power System ODAM, TRM Related Lessons INT0070509 OPS Tech. Spec. 3.8, Electrical Power Systems Related Objectives INT00705090010700 From memory, in MODES 1, 2, and 3, state the actions required in less than or equal to one hour if one 250 V DC electrical power subsystem inoperable (LCO 3.8.4).

Related References 3.8.4 DC sources - Operating B 3.8.4 Bases DC sources - Operating Related Skills (K/A) 295004 Partial or Complete Loss of D.C. Power 2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41.7 / 43.2) IMPORTANCE RO 3.2 SRO 4.2

QUESTION: S1 76 The plant is in Mode 1 when the DC system engineer calls the control room and reports that the Division 2 250 VDC Electrical Power Subsystem cannot perform its intended function.

What action is required by Technical Specification 3.8.4 DC Sources - Operating?

AND According to Tech Spec Bases, what system(s) is/are required to be declared inoperable for this failure?

a. Declare associated supported features inoperable with a Completion Time of 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B Loop LPCI only

b. Declare associated supported features inoperable with a Completion Time of Immediately.

B Loop LPCI only

c. Declare associated supported features inoperable with a Completion Time of 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

HPCI and B Loop LPCI

d. Declare associated supported features inoperable with a Completion Time of Immediately.

HPCI and B Loop LPCI ANSWER: S1 76

d. Declare associated supported features inoperable with a Completion Time of Immediately.

HPCI and B Loop LPCI Explanation:

From Tech Spec 3.8.4 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources -Operating LCO 3.8.4 The Division 1 and Division 2 125 V and 250 V DC power subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.1 Restore 125 V DC A. One 125 V DC electrical electrical power power subsystem 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> subsystems to inoperable.

OPERABLE status.

B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B. Required Action and associated Completion AND Time of Condition A not met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. One 250 V DC electrical C.1 Declare associated power subsystem supported features Immediately inoperable. inoperable.

From Tech Spec Bases B 3.8.4 C.1 With the Division 1 250 V DC electrical power subsystem inoperable, one LPCI subsystem is rendered inoperable. Loss of the Division 2 250 V DC electrical power subsystem renders HPCl and the other LPCI subsystem inoperable. Required Action C.1 therefore requires with one 250 V DC electrical power subsystem inoperable that the associated supported features be declared inoperable immediately. This declaration also requires entry into applicable Conditions and Required Actions for the associated supported features.

Distracters:

a. Declare associated supported features inoperable with a Completion Time of 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B Loop LPCI only. This answer is incorrect because the Completion Time listed is the Completion Time for 125 VDC electrical power subsystem being inoperable. Also HPCI must be declared inoperable with B Loop LPCI for a Div 2 Failure of the 250 VDC subsystem.

b. Declare associated supported features inoperable with a Completion Time of Immediately.

B Loop LPCI only. This answer is incorrect because HPCI must be declared inoperable with B Loop LPCI for a Div 2 Failure of the 250 VDC subsystem according to TS Bases 3.8.4 Condition C.

c. Declare associated supported features inoperable with a Completion Time of 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

HPCI and B Loop LPCI. This answer is incorrect because the Completion Time listed is the Completion Time for 125 VDC electrical power subsystem being inoperable not the 250 VDC Subsystem as described in the question.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295023.AA2.05 Importance Rating 4.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 5.7.1 Emergency Classification (Attach if not previously provided)

(including version/revision number) 43 Learning Objective: See Attached (As available)

Question Source: Bank # 19335 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: N S2 NRC Style 19335 04 06/24/2010 SRO: Y 77 Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 4 Multiple Choice Topic Area Description Emergency Plan GEN0030402, GEN0030401, SRO classify a dropped bundle per 5.7.1 Related Lessons GEN0030401 Emergency Plan for Licensed Operators MCR0010101 On-Shift Emergency Director GEN0030402 OPS EAL Training Part 1, Category A Related Objectives GEN0030401C0C050E Concerning event classification: Given a copy of EPIP 5.7.1 and hypothetical abnormal plant symptoms, indications, or events, determine any and all EALs which have been exceeded and specify the appropriate emergency classification.

GEN0030402001050E Concerning event classification: Given a copy of EPIP 5.7.1 and hypothetical abnormal plant symptoms, indications, or events, determine any and all EALs which have been exceeded and specify the appropriate emergency classification.

Related References 5.7.1 Emergency Classification Related Skills (K/A) 295023.AA2.05 Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS: (CFR: 41.10 / 43.5 / 45.13) ?Entry conditions of emergency plan (3.2/4.6*)

QUESTION: S2 77 19335 The plant is in MODE 5 with the following conditions:

• Refueling activities are in progress.

2

• RA-1 Refueling Floor ARM is reading 3 x 10 mRem/hr.

While transferring a fuel bundle from the Spent Fuel Pool to the Vessel the following occur:

• An irradiated fuel bundle is dropped in the cattle chute.

• REFUEL AREA HIGH RAD, 9-3-1/A-10 is in alarm (both Ronan 1448 and 1449).

4

• RA-1 Refueling Floor ARM is reading 5.5 x 10 mRem/hr.

What is the required Emergency Classification for this event?

a. Unusual Event

b. Alert

c. Site Area Emergency

d. General Emergency ANSWER: S2 77 19335

b. Alert Explanation:

An ALERT should be declared per EAL AA2.1.

Distracters:

a. is incorrect. Conditions are met for an ALERT.

c. and d. are incorrect. Only a single bundle has been dropped. Elevation to a higher level above an ALERT would require major fuel damage defined as more than ten irradiated fuel bundles.

PROVIDE THE CANDIDATES PROCEDURE 5.7.1 EAL HARDCARDS (Attachment 4)

Reference:

5.7.1 EAL AA2.1 Page 44 ATTACHMENT 2 EMERGENCY ACTION LEVEL S Category: A - Abnormal Rad Release/Rad Effluent

Subcategory: 2 - On-Site Rad Conditions And Spent Fuel Pool Event s Initiating Condition: Damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel outside the RPV EAL: AA2 .1 Alert Damage to irradiated fuel OR loss of water level (uncovering irradiated fuel outside the RPV) that causes EITHER of the following:

Valid RMA-RA-1 Fuel Pool Area Rad reading > 50 R/hr OR Valid RMP-RM-452 A-D Rx Bldg Vent Exhaust Plenum Hi-Hi alarm Mode Applicability: All NEI 99-01 Basis:

This EAL addresses increases in radiation dose rates within plant buildings and may be a precursor to a radioactivity release to the environment. These events represent a loss of control over radioactive material and represent an actual or substantial potential degradation in the level of safety of the plant.

This EAL addresses radiation monitor indications of fuel uncovery and/or fuel damage.

Increased ventilation monitor readings may be indication of a radioactivity release from the fuel, confirming that damage has occurred. Increased background at the ventilation monitor due to water level decrease may mask increased ventilation exhaust airborne activity and needs to be considered.

While a radiation monitor could detect an increase in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered.

Escalation of this emergency classification level, if appropriate, would be based on EALs in Subcategory Al.

CNS Basis:

When considering classification, information may come from:

• Radiation monitor readings.

• Sampling and surveys.

• Dose projections/calculations.

• Reports from the scene regarding the extent of damage (e.g., Refueling Crew, Radiation Protection Technicians).

This EAL is defined by the specific areas where irradiated fuel is located, such as the refueling cavity or Spent Fuel Pool (SFP).

The bases for the ventilation radiation Hi-Hi alarm is a spent fuel handling accident (Reference 2). Fuel Pool area radiation > 50 R/hr represents 100 times the high alarm setpoint (HR) and is unambiguously indicative of spent fuel damage or uncovery (Reference 1).

CNS Basis Reference(s):

1. Alarm Procedure 2 .3-9-3-1, Panel 9 Annunciator 9-3-1, Alarm A-10 .

2. Alarm Procedure 2 .3_9-4-1, Panel 9 Annunciator 9-4-1, Alarm E-4 .

SRO Justification: 10CFR55.43.b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295018.AA2.03 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: _See Attached_____

Explanation: See Attached Technical Reference(s) 5.2REC, 5.2SW (Attach if not previously provided)

(including version/revision number) 12, 22 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: N S3 NRC Style 00 02/11/11 SRO: Y 78 Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 4 Multiple Choice Topic Area Description ABN / Emergency Loss of SW flow to an REC heat exchanger, which Procedures procedure and why?

Related Lessons INT032-01-26 Cooling Water Abnormal Related Objectives L. Given plant condition(s), determine from memory the appropriate Abnormal/Emergency Procedure(s) to be utilized to mitigate the event(s).

Related References 5.2REC Loss of Reactor Equipment Cooling 5.2SW Service Water Casualties Related Skills (K/A) 295018.AA203 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: (CFR: 41.10 / 43.5 / 45.13) Cause for partial or complete loss (3.2 / 3.5)

QUESTION: S 3 78 The plant is operating at 100% power when the following conditions occur:

• SW Pressure on both Divisions have risen but are still in the green band

• REC system pressure is steady and is in the green band

• REC Surge Tank Level High alarms

• RR MG Set Oil Temperatures are rising

• Drywell temperature and pressure are rising

• RWCU F/D Inlet Temp High alarms What is the cause and which procedure should be entered to correct the problem?

a. REC Heat Exchanger SW Outlet valve failed closed; enter 5.2REC to correct the problem.

b. Service water non-critical loop isolated; enter 5.2SW to correct the problem.

c. REC system has developed a leak; enter 5.2REC to correct the problem.

d. Service water pump tripped; enter 5.2SW to correct the problem.

ANSWER: S3 78

a. REC Heat Exchanger SW Outlet valve failed closed; enter 5.2REC to correct the problem.

Explanation:

A loss of SW flow to the REC Heat Exchanger recovery is covered in both 5.2REC and 5.2SW.

If the pressure of the service water system lowers to < 38 psig the system will isolate non-critical loads. The subsequent steps will have the Operators place the other loops REC Heat Exchanger in service. For this scenario, SW Pressure rose indicating that there was some restriction in the flow path. With REC temperatures rising and systems cooled by REC alarming, the flow restriction is affecting the REC Heat Exchanger. The entry conditions for 5.2REC are met; however the entry conditions for 5.2SW are not met. Therefore, only 5.2REC should be entered.

Distracters:

b. Service water non-critical loop isolated; enter 5.2SW to correct the problem, is incorrect because SW pressures never dropped, in fact they rose.

c. REC system has developed a leak; enter 5.2REC to correct the problem, is incorrect because REC Surge Tank level rose. If there were a leak on the REC system, tank level would have lowered.

d. Service water pump tripped; enter 5.2SW to correct the problem, SW pressure rose. A tripping of a SW Pump would have made pressure lower on the SW system.

SRO Justification: 10CFR55.43. b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295038 G 2.2.42 Importance Rating 4.6 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) D3.3.2 Gaseous Effluent Monitoring (Attach if not previously provided)

(including version/revision number) 10/10/01 Learning Objective: See Attached (As available)

Question Source: Bank # 5448 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 2 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: N S4 NRC Style 5448 02 06/11/2004 SRO: Y 79 Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Off-Site Dose INT0070702, Offgas monitors inop with high release rates Assessment Manual Related Lessons INT0070702 ODAM Specifications INT0070508 OPS Tech. Specs. 3.7, Plant Systems Related Objectives INT00707020010300 Given the ODAM Appendix D, and conditions of non-compliance with any ODAM Specification Section 3.1 thru 3.5, determine the required actions.

INT00705080010100 Given a set of plant conditions, recognize non-compliance with a Chapter 3.7 LCO.

INT00705080010300 Given a set of plant conditions that constitutes non-compliance with a Chapter 3.7 LCO, determine the ACTIONS that are required.

Related References D3.3.2 Gaseous Effluent Monitoring Related Skills (K/A) 2.2.38 Knowledge of conditions and limitations in the facility license. (CFR:

41.7 / 41.10 / 43.1 / 45.13) (3.6/4.5) 2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR:

41.5 / 43.2 / 45.2) (4.0/4.7) 2.2.40 Ability to apply Technical Specifications for a system. (CFR: 41.10 /

43.2 / 43.5 / 45.3) (3.4/4.7) 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3)

(3.9/4.6)

QUESTION: S4 79 5448 The plant is at 100% power with the following conditions:

• Both channels of Offgas Radiation Monitoring are incapable of providing a trip signal.

• Rising radiation levels occur on many ARMs, Main Steam Line Radiation Monitors and Offgas radiation monitors.

• The release rate at the Steam Jet Air Ejector discharge is 2.2 Ci/sec.

• Radiation levels have been stable for the last 23 minutes.

If conditions do not improve, what are the MOST restrictive actions required per Technical Specifications and the Offsite Dose Assessment Manual?

(NOTE: The choices are listed from LEAST restrictive to MOST restrictive.)

a. Confirm that the Elevated Release Point radiation monitor is operable and perform DLCO 3.3.2 REQUIRED ACTION F.1 and F.2.

b. Perform LCO 3.7.5 REQUIRED ACTION A.1 and if not restored within the limit, perform REQUIRED ACTION B.1.

c. Perform LCO 3.4.6 REQUIRED ACTION A.1 and A.2 and if not restored within the limit, perform REQUIRED ACTION B1 and B.2.

d. Perform DLCO 3.3.2 REQUIRED ACTION K.1 and K.2 and K.3.

ANSWER: S4 79 5448

d. Perform DLCO 3.3.2 REQUIRED ACTION K1 and K.2 and K.3.

Explanation:

With rising radiation levels throughout the plant and an increased release rate in the SJAEs.

The SRO must assess the cause of the radiation level rise and its affects on Technical Specifications and the ODAM. There are three possible TS and ODAM specs that need to be reviewed.

T.S. 3.4.6 Specific Activity of the coolant. The only indication of this is the SJAE Rad Monitors and those are reading 2.2 Ci/sec. This would warrant a call to Chemistry group to sample the specific activity of the coolant.

The next TS or ODAM to assess is T.S. 3.7.5 Air Ejector Offgas. With a release rate reading of 2.2 Ci/sec. this is above the LCO limit of <1.0 Ci/sec, so Condition A applies with Required Action A.1 Restored the release rate to within limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or comply with Required Action B.1 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The next to be evaluated is the ODAM because the Off-gas Radiation Monitors will not trip.

ODAM DLCO 3.3.2 table D3.3.2-1 instrument Elevated Release Point Monitoring System, Condition referenced from B.1 is F, Take grab samples once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and Analyze for gross activity every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Also function 1.a SJAE Noble gas activity monitor will not perform

its intended function, so condition C must be entered requiring actions C.1 C.2 and C.3 to be performed, the first two immediately and the last within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. DLCO 3.3.2, condition "K" must be addressed as well, because the trip capability is not being maintained. So the Offgas isolation valves must be closed immediately, and the reactor shutdown immediately and be in Mode 4 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

This question has been proven as a good discriminating question that performs well on exams.

Distracters:

a. Taking grab samples is required once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> but this is not the MOST Limiting.

b. Tech Spec 3.7.5 allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore gross gamma activity rate to within limits.

This is required but it is still not the MOST Limiting.

c. Specific may or may not be elevated and the only way is to sample the coolant. If that reading comes back above the LCO limit, then the main steam line isolation valves must be closed in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the unit be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Still not the MOST Limiting.

SRO Justification: (2) Facility operating limitations in the technical specifications and their bases.

SRO persons assess plant conditions and determine compliance with Technical Specification and action required for non-compliance.

PROVIDE THE CANDIDATE: DLCO 3.3.2 PROVIDE THE CANDIDATE: TECH SPEC 3.4.6 and 3.7.5 DO NOT PROVIDE ODAM OR TECH SPEC BASES

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295024 G 2.4.6 Importance Rating 4.7 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) PSTG (Attach if not previously provided)

(including version/revision number) 3 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33 Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: N S5 NRC Style 00 01/10/2011 SRO: Y 80 Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 5 Multiple Choice Topic Area Description Emergency Operating INT008-06-18, OPS EOP and SAG Graphs and Cautions Procedures Related Lessons INT0080618 OPS EOP and SAG Graphs and Cautions Related Objectives INT00806180011200 For each graph used in the flowcharts, identify the action(s) required if the parameters associated indicate operation in the restricted or prohibited area.

Related References 5.8 Emergency Operating Procedures (EOPs)

EOP SAG Graphs Related Skills (K/A) 295024 High Drywell Pressure 2.4.6 Knowledge of EOP mitigation strategies. (CFR: 41.10 / 43.5 / 45.13)

IMPORTANCE RO 3.7 SRO 4.7

QUESTION: S5 80 During a LOCA the following conditions exist:

• Average drywell temperature is 210F and steady.

• Average suppression pool temperature is 205F and rising.

• Drywell maximum run temperature is 200F and steady.

• Torus water level is 14 feet and rising.

• Reactor pressure is 1000 psig and steady.

• Reactor water level +25 inches (narrow range).

IF the mitigating strategy for these conditions were Emergency RPV Depressurization, what potential consequence could result?

a. Reactor water level indication may be lost.

b. The Safety Relief Valve (SRV) tail pipe supports may fail.

c. The Primary Containment Pressure Limit (PCPL) may be exceeded.

d. The Torus-to-Drywell Vacuum Breaker capacity may be exceeded.

ANSWER: S5 80

c. The Primary Containment Pressure Limit (PCPL) may be exceeded.

PROVIDE THE STUDENTS WITH EOP/SAG GRAPH #7 Explanation:

Current conditions place the plant on the unsafe side of the HCTL Curve. An ADS initiation now may result in exceeding the PCPL due to insufficient energy absorption capacity to handle a blowdown.

From the PSTG Page B-5-29 PSTG/SATG Step (First override before RC/P-2)

If while executing the following steps:

• Suppression pool temperature cannot be maintained below the Heat Capacity Temperature Limit, maintain RPV pressure below the Limit, exceeding 100°F/hr (RPV cooldown rate LCO) cooldown rate if necessary.

• Steam Cooling is required, enter Contingency #3.

Discussion The CNS Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature at which initiation of RPV depressurization will not result in exceeding either:

• The maximum temperature capability of the suppression chamber, or

• Primary Containment Pressure Limit A before the rate of energy transfer from the RPV to the primary containment is within the capacity of the containment vent. Refer to Section 16 of this appendix for a detailed discussion of the HCTL.

Control of suppression pool temperature is addressed in subsection SP/T of the Primary Containment Control guideline. If the actions being taken to limit suppression pool temperature increase are inadequate or not effective, RPV pressure must be reduced in order to remain below the HCTL. Therefore, actions in the RPV pressure control subsection of the RPV Control guideline must accommodate these requirements. Failure to do so may lead to failure of containment or loss of equipment necessary for the safe shutdown of the plant.

The normal cooldown rate limit may be exceeded to the extent necessary to maintain RPV pressure below the HCTL. If RPV pressure cannot be maintained below the HCTL, emergency RPV depressurization will be required, possibly resulting in an even more rapid cooldown.

Distracters:

a. is incorrect 200F is below the saturation limit curve.

b. is incorrect The maximum level where a blowdown would not cause damage to the SRV tail pipe supports is 16 feet.

d. is incorrect Torus to Drywell Vacuum Breakers are designed for LOCA energy release.

SRO Justification: 10CFR55.43. b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295032 G 2.4.9 Importance Rating 4.2 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) EOP 1A and EOP 5A (Attach if not previously provided)

(including version/revision number) 15 13 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: N S6 NRC Style 00 1/10/2011 SRO: Y 81 Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 3 Multiple Choice Topic Area Description Abnormal and Shutdown cooling with a RWCU leak, what EOPs are Emergency required to be entered and what is the mitigating strategy is required?

Related Lessons INT0080617 OPS EOP Flowchart 5A Secondary Containment and Radioactivity Release Control Related Objectives

1. List the entry conditions to Flowchart 5A (including the radioactivity release path) and briefly explain each.

6. Given plant conditions and EOP flowchart 5A, SECONDARY CONTAINMENT CONTROL and RADIOACTIVITY RELEASE CONTROL, determine required actions.

Related References EOP 1A RPV Control EOP 5A Secondary Containment Control Related Skills (K/A) 295032 High Secondary Containment Area Temperature / 5 2.4.9 - Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. (CFR: 41.10 /

43.5 / 45.13) IMPORTANCE RO 3.8 SRO 4.2

QUESTION: S6 81 The plant is shutdown and in shutdown cooling with A Loop RHR, with the following:

• Reactor Water Level is 54 inches

• Reactor Pressure is 15 psig

• RWCU is in service A leak develops on the RWCU Heat Exchanger and the RWCU Heat Exchanger Room 958 ft.

South temperatures start rising. The following indications are present after a short time:

• Reactor Water Level is 10 inches lowering slowly.

• Reactor Pressure is 5 psig and lowering slowly.

• RWCU Hx Room temperature is 145F and rising slowly.

• Annunciator 9-3-1 / E-10 Area High Temp is alarming.

Which EOP(s) is(are) required to be entered and what is the mitigating strategy that must be pursued?

EOP

a. 5A then concurrently enter 1A; scram and isolate RWCU to ensure the Reactor is shutdown and to stop the LOCA in progress.

b. 5A only, isolate RWCU to maintain Secondary Containment Temperatures below Max Safe Operating values and stop the LOCA in progress.

c. 1A only, ensure Groups 2 and 3 isolation to maintain Secondary Containment Temperatures below Max Safe Operating values.

d. 5A then concurrently enter 1A; ensure Groups 2 and 3 isolation to stop the LOCA in progress.

ANSWER: S6 81

b. 5A only, isolate RWCU to maintain Secondary Containment Temperatures below Max Safe Operating values and stop the LOCA in progress.

Explanation:

EOP 5A entry condition (Any area temperature above max normal operating values) exists when RWCU temperature rises above the Alarm setpoint. The mitigating strategy for 5A is to isolate any system that is discharging into the Reactor Building not needed to suppress a fire or support EOPs. A leak on the RWCU system is classified as a primary system discharging into the reactor building. The Max Safe Operating value for this parameter is 195F and if this value was being approached, entry into EOP 1A would be required. Since the value is 145F and rising slowly and there is no indication that the system would not isolate when the valves are closed, temperature should start dropping soon after the isolation is completed.

Distracters:

a. This is incorrect, because EOP 1A is not required to be entered until the temperature approaches 195F.

c. This is incorrect, because EOP 5A is required to be entered. There is no EOP 1A entry conditions present in the stem. Also neither Group 2 nor 3 isolation setpoints have been reached.

d. This is incorrect, because EOP 1A is not required to be entered until the temperature approaches 195F. Also neither Group 2 nor 3 isolation setpoints have been reached.

SRO Justification: 10CFR55.43. b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295031 G 2.4.3 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) T.S. 3.3.3.1 Post accident monitoring (PAM) instrumentation (Attach if not previously provided)

(including version/revision number) Amendments 233 and 178 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 19967 (Note changes or attach parent)

New Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 2 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: N S7 NRC Style 00 03/15/2011 SRO: Y 82 Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 3 Multiple Choice Topic Area Description Technical Specifications, INT0070504, How long to MODE 3 with PAM instruments ODAM, TRM inop Related Lessons INT0070504 CNS Tech. Spec. 3.3, Instrumentation Related Objectives INT00705040010300 Given a set of plant conditions that constitutes non-compliance with a Section 3.3 LCO, determine the ACTIONS that are required.

INT00705040010100 Given a set of plant conditions, recognize non-compliance with a Section 3.3 Requirement.

Related References 3.3.3.1 Post accident monitoring (PAM) instrumentation Related Skills (K/A) 295031 Reactor Water Level Low 2.4.3 Ability to identify post-accident instrumentation. (CFR: 41.6 / 45.4)

IMPORTANCE RO 3.7 SRO 3.9

QUESTION: S7 82 The plant is operating at rated power with the following conditions:

• NBI-PR-2A (Wide Range Reactor Pressure) becomes inoperable on 3/28 at 1100.

• NBI-LR-1A (Fuel Zone and Wide Range RPV water level recorder) becomes inoperable on 3/29 at 0900.

• NBI-LR-1B (Fuel Zone and Wide Range RPV water level recorder) becomes inoperable on 4/1 at 1500.

IF conditions do not change, when are you required to enter MODE 3 by Technical Specifications?

a. 2100 on 4/5

b. 1500 on 4/8

c. 0300 on 4/9

d. 0300 on 5/2 ANSWER: S7 82

c. 0300 on 4/9 Provide to Candidate: T.S. LCO 3.3.3.1 Explanation:

At 1100 on 3/28 Enter 3.3.3.1.A for NBI-PR-2A with a required action to restore the required channel to operable status within 30 days. This LCO would expire at 11:00 on April 27th. At that time a 14 day report is required to be submitted.

At 0900 on 3/29 Enter 3.3.3.1.A for NBI-LR-1A with a required action to restore the required channel to operable status within 30 days. This LCO would expire at 0900 on April 28th. At that time a 14 day report is required to be submitted.

At 1500 on 4/1 Enter 3.3.3.1.A again for NBI-LR-1B and 3.3.3.1.C because two required channels are inoperable. The required action is to restore one required channel to operable status within 7 days. This LCO expires at 1500 on April 8th. At that time Condition D would be entered with a required action to enter the condition referenced in Table 3.3.3.1-1 for the channel immediately. The condition referenced in the table is 3.3.3.1.E. The required action for E is to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In other words 0300 on 4/9.

Distracters:

a. Enters Condition C on first WR Level Recorder inoperable then adding the 7 days and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. Not adding 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> once Condition E is met.

d. not entering condition C and misreading A.

SRO Justification: 10CFR55.43. b (2) Facility operating limitations in the technical specifications and their bases.

MODIFIED QUESTION: 19967 (1 point(s))

The plant is operating at rated power with the following conditions:

• NBI-PR-2A (Wide Range Reactor Pressure) becomes inoperable on 3/28 at 1100.

• NBI-LI-85A (Wide Range RPV water level) becomes inoperable on 3/29 at 0900.

• NBI-PR-2B (Wide Range Reactor Pressure) becomes inoperable on 4/1 at 1500.

IF conditions do not change, when are you required to enter MODE 3 by Technical Specifications?

a. 1500 on 4/8

b. 0300 on 4/9

c. 1500 on 5/1

d. 0300 on 5/2 ANSWER: 19967

b. 0300 on 4/9 Enter 3.3.3.1.A and 3.3.1.C at 1500 on 4/1. Enter 3.3.3.1.D at 1500 on 4/8. Enter 3.3.3.1.E at 1500 on 4/8. Be in MODE 3 by 0300 on 4/9.

Answer source: LCO 3.3.3.1 Distracters:

a. Not adding the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. Not adding 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and not entering condition C and misreading A.

d. not entering condition C and misreading A.

Provide to Candidate: T.S. LCO 3.3.3.1 and bases.

2004 Biennial Exam SRO Justification: 10CFR55.43. b (2) Facility operating limitations in the technical specifications and their bases.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295033.EA2.01 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Emergency Procedure 5.7.1 (Attach if not previously provided)

(including version/revision number) 43 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge H Comprehension or Analysis 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: N S8 Licensed 00 01/10/2011 SRO: Y 83 Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Abnormal/Emergency GEN0030403, Determine the highest EAL classification Procedures required for C S/D Related Lessons GEN0030401 Emergency Plan for Licensed Operators GEN0030403 OPS EAL Training Part 2, Category S and Category C Related Objectives GEN0030401C0C050A Concerning event classification: Given a copy of EPIP 5.7.1 and an EAL identification code, determine the EAL and its associated emergency classification.

GEN0030403001050A Concerning event classification: Given a copy of EPIP 5.7.1 and an EAL identification code, determine the EAL and its associated emergency classification.

Related References 5.7.1 Emergency Classification Related Skills (K/A) 295033.EA2.01 Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS:

(CFR: 41.10 / 43.5 / 45.13) Area radiation levels (3.8 / 3.9)

QUESTION: S8 83 The Reactor is in Cold SHUTDOWN when the following events occur:

• Normal Area Radiation Monitor readings for the NE Quad is 5 mRem/hr., SE Quad is 4 mRem/hr., NW Quad is 3 mRem/hr., SW Quad is 6 mRem/hr.

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> later, the following conditions exist:

• All vital 125 VDC buses indicate 100 Volts DC and have for the last 20 minutes.

• Two area Reactor Building Sumps have started pumping much more frequently.

• The NW Quad radiation levels are steady at 4250 mRem/hr.

• The SE Quad radiation levels are steady at 3750 mRem/hr.

Determine all applicable EAL classifications required for these conditions?

a. SAE per SS7.1 only

b. SAE per SS7.1 and NOUE per AU2.2

c. NOUE per CU6.1 only

d. NOUE per CU6.1 and NOUE per AU2.2 ANSWER: S8 83

d. NOUE per CU6.1 and NOUE per AU2.2 Explanation:

With the Reactor in Cold Shutdown the EAL should be based on Cold S/D or A all modes. In this case, the radiation levels have risen >1000 times normal in the NW Quad requiring the SRO to assess EALs A2 Onsite Rad Conditions and in the UE column under AU2.2 Unplanned vital area radiation monitor reading or survey results rise by a factor of 1,000 over normal levels is required to be declared. Also with 125 VDC voltage at 100 Volts for greater than 15 minutes, a NOUE should be declared under CU6.1 cold shutdown and loss of DC power.

Distracters:

a. If the plant had been in either modes 1,2, or 3 a SAE in accordance with SS7.1 would be required to be declared, also CU6.1 applies and should be listed.

b. If the plant had been in either modes 1,2, or 3 a SAE in accordance with SS7.1 would be required to be declared.

c. The NOUE in accordance with AU2.2 still applies and should be listed.

REQUIRED

REFERENCES:

Procedure 5.7.1 Hard cards (Attachment 4)

SRO Justification: 10CFR55.43.b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295020 G 2.2.22 Importance Rating 4.7 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) T.S. 3.3.6.1 (Attach if not previously provided)

(including version/revision number) Amendment 178 / 231 / 212 Learning Objective: See Attached (As available)

Question Source: Bank # 110 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 2 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: N S9 Licensed 110 0 04/01/1998 SRO: Y 84 Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 1 Multiple Choice Topic Area Description Technical Specifications, INT0070504, CNS Tech Spec 3.3, Instrumentation ODAM, TRM Related Lessons INT0070504 CNS Tech. Spec. 3.3, Instrumentation Related Objectives INT00705040010100 Given a set of plant conditions, recognize non-compliance with a Section 3.3 Requirement.

INT00705040010300 Given a set of plant conditions that constitutes non-compliance with a Section 3.3 LCO, determine the ACTIONS that are required.

Related References 3.3.6.1 Primary Containment Isolation Instrumentation Related Skills (K/A) 295020 Inadvertent Containment Isolation 2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 /

43.2 / 45.2) IMPORTANCE RO 4.0 SRO 4.7

QUESTION: S9 84 110 At 1335 on 1/7 it is discovered that HPCI Temperature Elements HPCI-TS 101 A through D, HPCI-TS 102 A through D, HPCI-TS 103 A through D, and HPCI-TS 104 A through D (located in RHR Injection Valve room and Torus Area West) are set to trip at 205°F.

What action(s) is/are required by Technical Specifications?

a. Place the channels associated with the HPCI Temperature Elements in trip by 2335 on 1/7.

b. Restore HPCI isolation capability by 1435 on 1/7 or isolate HPCI by 1535 on 1/7.

c. Place the channels associated with the HPCI Temperature Elements in trip by 1335 on 1/8.

d. Place only the channels associated with HPCI-TS 101 A through D and HPCI-TS 104 A through D in trip by 1335 on 1/8.

ANSWER: S9 84 110

b. Restore HPCI isolation capability by 1435 on 1/7 or isolate HPCI by 1535 on 1/7.

PROVIDE STUDENT WITH TECH SPEC 3.3.6.1 Explanation:

In accordance with Tech Specs 3.3.6.1 B. One or more functions with isolation capability not maintained. The required action is to restore isolation capability within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (14:35 on 1/7). If that is not possible then Condition C applies and required action C.1 states, Enter the Condition referenced in Table 3.3.6.1-1 for the channel. That Condition from the Table is F which states isolate the affected penetration flow path(s), within one hour (15:35 on 1/7)

Distracters:

a. This is action for functions 2.a, 2.b, 5.d, and 6.b. HPCI is function 3.d. A candidate might choose this answer if they misread the functions in the completion time column.

c. This is the required actions if trip capability is maintained, which it is not.

d. This is a combination of required actions for multiple inoperable instruments, however not allowed by Tech Specs for this case.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295029.EA2.01 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) T.S. 3.6.2.2 & Bases 3.6.2.2 (Attach if not previously provided)

(including version/revision number) 178 0 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 2 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date RO: N S 10 Licensed 00 06/02/2010 SRO: Y 85 Operator NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice Topic Area Description Technical Specifications, Suppression Pool Level High out of spec high, what is ODAM, TRM required by TS?

Related Lessons INT0070507 CNS Tech. Spec. 3.6, Containment Systems Related Objectives INT00705070010100 Given a set of plant conditions, recognize non-compliance with a Chapter 3.6 LCO.

Related References 3.6.2.2 Suppression Pool Level Related Skills (K/A) 295029.EA2.01 Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: (CFR: 41.10 / 43.5 /

45.13) Suppression pool water level 3.9* / 3.9*

QUESTION: S10 85 The Plant is in MODE 1 with the following conditions at 09:00 on September 8th:

• Suppression Pool Level as read on PC-LI-13 is +2.5 inches

• There are no methods available to reduce this level.

When is the plant required by Tech Specs to be in MODE 4?

a. 21:00 on Sept. 8th

b. 23:00 on Sept. 8th

c. 21:00 on Sept. 9th

d. 23:00 on Sept. 9th ANSWER: S10 85

d. 23:00 on Sept. 9th Explanation:

The corresponding levels indicated on the narrow range suppression pool level instrument PC-LI-13 are -2 inches and +2 inches for the Tech Spec Numbers given in LCO 3.6.2.2 of > 12 ft 7 inches and < 12 ft 11 inches. With level at 2.5 inches this would be above the TS Limit of 12 ft 11 inches therefore requiring entry into LCO 3.6.2.2. Since water level cannot be lowered the plant must be placed in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and MODE 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. There are also two hours to try to get level lowered and this time is added to the 36 total hours to be in MODE 4.

Provide the Students with Tech Spec 3.6.2.2 Distracters:

a. 21:00 on Sept 8th is incorrect; this is the time to be in mode 3 minus the extra 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

b. 23:00 on Sept 8th is incorrect; this is the time to be in mode 3.

c. 21:00 on Sept 9th is incorrect; this is the time to be in mode 4 minus the extra 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 209001.A2.09 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s)

(Attach if not previously provided)

(including version/revision number)

Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: N S 11 00 1/08/2011 Licensed Operator SRO: Y 86 NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice N Topic Area Description Emergency Procedures NPSH for CS Pump exceeded, what is required?

Related Lessons INT0080618 OPS EOP AND SAG GRAPHS AND CAUTIONS INT0080609 OPS EOP FLOWCHART 1A - RPV CONTROL, RPV LEVEL Related Objectives INT00806090011100 Given plant conditions and EOP flowchart 1A, RPV CONTROL, determine required actions.

INT00806180010300 Given plant conditions and the EOP and SAG Graphs Flowchart, determine if operation is within the allowed region of a graph.

Related References EOP 1A EOP/SAG Graphs Related Skills (K/A) ROI SROI 209001.A2.09 Ability to (a) predict the impacts of the following on 3.1 3.3 the LOW PRESSURE CORE SPRAY SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6) Low suppression pool level (3.1 / 3.3)

QUESTION: S 11 86 The plant is operating at 100% power when a LOCA occurs and torus level lowers and heats up. Core Spray A is the only non-alternate injection source available. The following parameters are noted:

• Primary Containment Water Level is 5.5 feet

• Suppression Pool Temperature is 150F

• Suppression Pool Pressure is 4 psig

• Reactor Water Level is -150 inches Wide Range

• Reactor Pressure is 54 psig Core Spray Pump 1A is started and is injecting into the vessel at a rate of 5600 gpm.

The Station Operator in the area of the pump reports hearing what sounds like cavitation. Reactor water level is recovering and is currently -100 inches.

What EOP Action should be performed next?

In accordance with EOP

a. 1A and EOP/SAG Graphs; Reduce Core Spray Pump A flow to less than 4900 gpm if Reactor water level will continue to rise.

b. 1A and 3A only; Shutdown the running Core Spray Pump and start up and inject with Alternate Injection Systems per Table 4.

c. 3A only; Shutdown the running Core Spray Pump and start up and inject with Alternate Injection Systems per Table 4.

d. 1A, 3A and EOP/SAG Graphs; Raise injection with Core Spray to 6000 gpm and restore level into the +3 to 54 inch level band as soon as possible.

ANSWER S11 86

a. 1A and EOP/SAG Graphs; Reduce Core Spray Pump A flow to less than 4900 gpm if Reactor water level will continue to rise.

PROVIDE THE CANDIDATES A COPY OF EOP/SAG GRAPHS # 3, 4, 5 and Note 3 block concerning NPSH Limits Explanation:

EOP 1A directs the operator to Restore and maintain RPV water level above -150 inches using Injection Systems (TABLE 3) and, if necessary, Alternate Injection Subsystems (TABLE 4). Core Spray is an injection system from Table 3.

Caution 3 from the same EOP states Operation of HPCI, RCIC, CS, or RHR with suction from suppression pool and pump flow above NPSH or Vortex Limit (GRAPHs 3, 4, 5, 16, 18) may result in equipment damage. It does not prohibit the running of the pump below the NPSH and Vortex limit however.

The calculation for Torus over pressure is Note 3 of the EOP/SAG Graphs:

Torus pressure (psig) 4 psig Hydrostatic head (psig)

PC water level (ft.) 5.5 Strainer level (ft.) -4 0.43 x 1.5 = +0.645 Torus overpressure (psig) 4.645 Using this calculation the instruction is to use the next lower curve on the NPSH graph, which is the 0 psig line. With Core Spray Pump A operating at 5500 gpm the NPSH for the pump is being exceeded and flow should be reduced to less than 4900 gpm with torus temperature at 150F if level is still recovering. This action will preserve the pump for future use.

Distracters:

b. 1A and 3A only; Shutdown the running Core Spray Pump and start up and inject with Alternate Injection Systems per Table 4. The Core Spray Pump does not need to be shutdown, just throttled back to maintain it in the NPSH Limit for the current torus level and temperature. Alternate Injection systems are not needed, CS is adequate to restore level.

c. 3A only; Shutdown the running Core Spray Pump and start up and inject with Alternate Injection Systems per Table 4. The Core Spray Pump does not need to be shutdown, just throttled back to maintain it in the NPSH Limit for the current torus level and temperature. Alternate Injection systems are not needed, CS is adequate to restore level.

d. 1A, 3A and EOP/SAG Graphs; Raise injection with Core Spray to 6000 gpm and restore level into the +3 to 54 inch level band as soon as possible. The Core Spray Pump is above its NPSH Limit for the current torus level and temperature.

Flow should be lowered not raised.

SRO Justification: 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 212000.A2.19 Importance Rating 3.9 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) GE Prints 791E256 Sheets 8 and 13 (Attach if not previously provided)

(including version/revision number)

Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: N S 12 NRC Style 00 1/8/2011 SRO: Y 87 Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 H 1 4 Multiple Choice N Topic Area Description Systems COR0022102, quarter scram on RPS loss Related Lessons COR0022102 REACTOR PROTECTION SYSTEM Related Objectives COR0022102001080G Given a specific RPS malfunction, determine the effect on any of the following: Scram air header solenoid operated valves Related References PR 4.5 55.43 b (5)

Related Skills (K/A) ROI SROI 295015.AK2.04 Knowledge of the interrelations between INCOMPLETE SCRAM and the following: (CFR: 41.7 /

45.8) RPS (4.0/4.1) (4.0 / 4.1) 212000.A2.19 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the...: (CFR: 41.5 / 45.6) Partial system activation (half-SCRAM) (3.8 / 3.9)

QUESTION: S 12 87 The plant is operating at 100% power with no equipment out of service. The Reactor Operator noticed that one of the four RPS Scram Group lights is NOT lit on RPS trip system B. The following occurs:

• A half-scram RPS trip system A is received.

What are the consequences of this event and what action is required?

a. Approximately one quarter of the control rods insert into the core; enter 2.4CRD CRD Trouble and manually Scram the Reactor.

b. Approximately one half of the control rods insert into the core; enter 2.4CRD CRD Trouble and manually Scram the Reactor

c. Approximately one quarter of the control rods insert into the core; enter 5.3AC120 Loss of 120 VAC and manually Scram the Reactor.

d. Approximately one half of the control rods insert into the core; enter 5.3AC120 Loss of 120 VAC and manually Scram the Reactor.

ANSWER: S 12 87

a. One quarter of the control rods insert into the core; Enter 2.4CRD CRD Trouble and manually Scram the Reactor.

Explanation:

With one (1) white SCRAM INDICTIONS GROUP B light extinguished, approximately 1/4 of the Scram Pilot Valves for RPS "B" are de-energized. When the power supply for RPS A is lost (1/2 Scram), all A channel Scram Pilot Valves will be de-energized. This will cause all Gr. 4 rods, approximately 1/4 of all rods, to scram.

2.4CRD contains the actions to Scram the reactor in this condition (more then one control rod drifting).

Distracters:

b. One half of the control rods insert into the core; Enter 2.4CRD CRD Trouble and manually Scram the Reactor, is incorrect, because that would require two Group B lights to be extinguished, meaning that 1/2 of the Scram Pilot Valves for RPS B are de-energized. Therefore half of the control rods would scram.

c. One quarter of the control rods insert into the core; Enter 5.3AC120 Loss of 120 VAC and manually Scram the Reactor. This is the wrong procedure for this failure. However, the RPS system is a 120V AC system.

d. One half of the control rods insert into the core; Enter 5.3AC120 Loss of 120 VAC and manually Scram the Reactor, is incorrect, because that would require two Group B lights

to be extinguished, meaning that 1/2 of the Scram Pilot Valves for RPS B are de-energized. Therefore half of the control rods would scram. And this is the wrong procedure.

SRO Justification: 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

The attached copies are of GE Print 791E256 Sheet 13 and 8, to show that approximately 1/4 of the rods are indicated by one of the four Rod Group Lights on Panel 9-5

Panel 9-5; 1 of 4 Rod Group white lights

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 264000 G 2.2.36 Importance Rating 4.2 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) T.S. 3.8.1 (Attach if not previously provided)

(including version/revision number) 233 Learning Objective: See Attached (As available)

Question Source: Bank # 21404 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 2 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date S 13 01 10/15/2004 Open Reference RO: N 21404 SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 6 Multiple Choice N Topic Area Description Technical Specifications, INT0070509, SRO ONLY, TS 3.8.1 and TS 3.5.1 ODAM, TRM Related Lessons INT0070509 OPS Tech. Spec. 3.8, Electrical Power Systems INT0239960 OPS SCR Events 2003 Related Objectives INT00705090010300 Given a set of plant conditions that constitutes non-compliance with a Section 3.8 LCO, determine the ACTIONS that are required.

INT02399600010600 Given a set of plant conditions that constitutes non-compliance with a Section 3.8 LCO, determine the ACTIONS that are required. (INT0070509 EO 3)

Related References 10CFR55.43 (B)(2)

TS 3.8.1 Related Skills (K/A) ROI SROI 2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.2 / 45.2) (4.0/4.7) ( / )

2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2) (3.2/4.2) ( / )

2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (CFR: 41.10 / 43.2 /

45.13) (3.1/4.2) ( / )

QUESTION: S 13 88 The plant is at 100% power with the following:

November 1 at 08:00: DG#2 is declared INOPERABLE for maintenance.

November 7 at 08:00: The ESST is declared INOP due to degraded voltage.

November 7 at 09:00: DG#2 is declared OPERABLE.

November 13 at 07:00: DG#1 is declared INOPERABLE due to low jacket water level.

November 13 at 08:00: The ESST is declared OPERABLE.

What date and time is the plant required to be in MODE 3?

a. November 13 at 21:00.

b. November 14 at 20:00.

c. November 15 at 20:00.

d. November 20 at 19:00.

ANSWER: S 13 88

c. November 15 at 20:00.

PROVIDE THE CANDIDATE A COPY OF T.S. 3.8.1 Explanation:

At 08:00 on Nov.1 DG #2 is inop requiring entry into LCO 3.8.1 Condition B. There are 4 Required Actions B.1 and B.2 and B.3.1 or B3.2 and B.4 which has two completion times listed. One 7 days and one 14 days form the discovery of failure to meet the LCO.

Six days into the 7 day LCO for DG #2 inop, the ESST (an off-site power source) is declared inoperable. This requires the entry into both Condition A and Condition D. For Condition A, 3 Required Actions A.1 and A.2 and A.3, which has the same two completion times that the inop DG had, a 7 day and a 14 day from the discovery of the failure to meet the LCO. Condition D, One Off-site circuit inop and one DG inop. There are 2 Required Actions D.1 or D.2 each with a completion time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

One hour into the time where Condition D applied, the DG is returned to operable status.

Now Condition D is exited but Condition A still applies and the LCO was never exited therefore the time clock did not reset. It is still counting down from the 08:00 on Nov. 1 start time.

Almost five days later the # 1 DG is declared inoperable, and the LCO is still in effect.

That LCO Condition B again would be entered along with D.

Two hours into the Time for Condition D the off-site source returns to service. The LCO is still not exited.

In accordance with Tech Spec 1.3, Completion Times. There is a separate completion time with a separate 14-day clock measured from the time the LCO statement was declared not met at 08:00 on November 1. Fourteen (14) days later is 08:00 on November 15. Then there is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed to be in Mode 3. The time the plant is required to be in Mode 3 is November 15th at 20:00 Distracters:

a. This would be correct until the offsite circuit is declared operable at which point Condition D is exited and the clock for DG#1 inoperable for Condition B is continued (7-day clock) but there is a 14 day completion time for the LCO statement not met.

b. There is a separate completion time with a separate 14-day clock measured from the time the LCO statement was declared not met at 08:00 on November 1.

Fourteen (14) days later is 08:00 on November 15 not 14.

d. This is correct until the offsite circuit is declared operable at which point Condition D is exited and the clock for DG#1 inoperable for Condition B is continued (7-day clock) but there is a 14 day completion time for the LCO statement not met.

2004 Biennial Exam 04LORSRO03, Question # 21208

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 262001 G 2.2.40 Importance Rating 4.7 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Tech Spec 3.8.7 (Attach if not previously provided)

(including version/revision number) Amendment 178 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # 100 (Note changes or attach parent)

New Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 2 Comments:

ES-401, Page 28 of 33

Question Question Rev Revision Last Used Exam Bank Applicability Number ID # Date Date S 14 0 03/15/2011 NRC Style RO: N 89 Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 4 H 1 5 Multiple Choice Topic Area Description Technical Specifications, INT0070501, Determine Rx MODE ODAM, TRM Related Lessons INT0070501 OPS Introduction to Technical Specifications INT0070509 Technical Specifications 3.8 Electrical Power System Related Objectives INT00705010010300 From memory, given a set of plant conditions, determine the plant MODE.

INT0070509 Obj. 3. Given a set of plant conditions that constitutes non-compliance with a Section 3.8 LCO, determine the ACTIONS that are required.

Related References T.S. 1.1 Definitions T.S. 3.8.7 Distribution Systems - Operating T.S. 3.8.2 A/C Sources - Shutdown T.S. 3.8.8 Distribution Systems - Shutdown Related Skills (K/A) 262001 AC Electrical Distribution 2.2.40 Ability to apply Technical Specifications for a system. (CFR: 41.10 / 43.2 / 43.5

/ 45.3) IMPORTANCE RO 3.4 SRO 4.7

QUESTION: S14 89 The following conditions exist:

• Reactor Mode switch REFUEL

• Average Rx Coolant Temp. 213 F

• RPV Head Closure Bolts Tensioned (all)

• Surveillances in progress None

• Control rod position ALL are fully inserted A switching mistake causes 4160 VAC Bus 1F and 480 VAC Bus 1F to de-energize and remain de-energized.

What is required by Technical Specifications?

a. Enter LCO 3.8.7 Distribution Systems - Operating and restore the 1F Bus to Operable status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> AND 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> from the failure to meet the LCO.

b. Enter LCO 3.3.8 Distribution Systems - Shutdown and Declare associated supported features inoperable immediately.

c. Enter LCO 3.8.2 A/C Sources - Shutdown and Declare associated supported features inoperable immediately.

d. Enter LCO 3.0.3 immediately.

ANSWER: S 14 89

a. Enter LCO 3.8.7 Distribution Systems - Operating and restore the 1F Bus to Operable status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> AND 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> from the failure to meet the LCO.

PROVIDE THE CANDIDATE TS 3.8.7, 3.8.8 and 3.8.2 Explanation:

From the definition of Reactor Modes the Plant is in Startup - Mode 2. LCOs 3.8.7 and LCO 3.0.3 are applicable in Mode 2. LCO 3.8.7 should be entered and with one AC electrical power distribution subsystem inoperable, restore AC electrical power distribution subsystem to Operable status within 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> from discovery of failure to meet LCO.

In accordance with Table 3.8.7-1

• Each division of the AC, the 125 V DC, and the 250 V DC electrical power distribution systems is a subsystem. Both 4160 and 480 V AC Bus 1Fs are considered a subsystem.

Distracters:

b. This TS is for modes 4 and 5, not 2

c. This TS is for modes 4 and 5, not 2

d. This would be the required action if there were two or more electrical power distribution subsystems inoperable that result in a loss of function. In this case, there is only 1 subsystem inop.

MODIFIED QUESTION: 100 The following conditions exist:

Reactor Mode switch REFUEL Average Rx Coolant Temperature 213 F RPV Head Closure Bolts Tensioned (all)

Surveillances in progress None Control rod position ALL are fully inserted What is the reactor operating mode?

a. Startup - MODE 2

b. Hot Shutdown - MODE 3

c. Cold Shutdown - MODE 4

d. Refueling - MODE 5 ANSWER: 1 100

a. Startup - Mode 2 (All head bolts are tensioned.)

Distracters:

b. Average coolant temperature is > 212F but the Reactor Mode Switch is not in SHUTDOWN.

c. Cold Shutdown is < 212F with the Mode switch in SHUTDOWN.

d. Since all RPV head bolts are tensioned, the Reactor cannot be in Mode 5.

MODE TITLE REACTOR MODE SWITCH AVERAGE REACTOR POSITION COOLANT TEMP (F) 1 Power Operation Run NA 2 Startup Refuel(a) or NA Startup/Hot Standby 3 Hot Shutdown(a) Shutdown > 212 4 Cold Shutdown(a) Shutdown < 212 5 Refueling(b) Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bolts less than fully tensioned.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 1 K/A # 215005 G 2.4.20 Importance Rating 4.3 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) EOP 7A (Attach if not previously provided)

(including version/revision number) 14 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: N S 15 00 1/08/2011 Licensed Operator SRO: Y 90 NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 3 Multiple Choice N Topic Area Description Emergency Operating INT0080610, Actions after HSBW with power rise Procedures Related Lessons INT0080610 OPS EOP FLOWCHART 7A - RPV LEVEL (FAILURE-TO-SCRAM)

Related Objectives INT00806100010800 Given plant conditions and EOP flowchart 7A, RPV LEVEL (FAILURE TO SCRAM), determine required actions.

Related References CFR 10CFR55.43 Related Skills (K/A) ROI SROI 215005 Average Power Range Monitor/Local Power Range Monitor 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10 / 43.5 /

45.13) (3.8/4.3)

QUESTION: S15 90 The plant was operating at 100% when the following occurred:

• All Circ water pumps were lost.

• Very few control rods inserted on the scram.

• When the MSIVs closed, a small leak on a Recirc suction pipe occurred.

• Reactor pressure control was established using the SRVs.

• The crew initiated SLC.

• RPV water level was intentionally lowered.

• A level band of -183"(corrected FZ) to -60"(corrected FZ) was established.

• When Hot Shutdown Boron Weight had been injected, the crew began raising water level.

When the level reached -50"(corrected FZ) the following conditions were present:

• Reactor pressure 950 psig (steady)

• All APRMs are approximately 15% (rising)

• Drywell pressure 3.0 psig (slowly lowering)

• Torus pressure 3.5 psig (slowly lowering)

• Average Torus water temperature 160ºF (rising slowly)

• Average Drywell temperature 255ºF (slowly lowering)

• Primary Containment water level 12.8' (steady)

What action is required?

a. Emergency Depressurize the RPV IAW EOP 3A, PRIMARY CONTAINMENT CONTROL.

b. Reduce reactor pressure to less than 800 psig IAW EOP 6A, RPV PRESSURE (FAILURE TO SCRAM).

c. Stop and prevent injection except for RCIC, CRD and SLC IAW EOP 7A, RPV LEVEL CONTROL (FAILURE TO SCRAM).

d. Restore and maintain RPV water level between +3" and 54" IAW EOP 7A, RPV LEVEL CONTROL (FAILURE TO SCRAM).

ANSWER: S15 90

c. Stop and prevent injection except for RCIC, CRD and SLC IAW EOP 7A, RPV LEVEL CONTROL (FAILURE TO SCRAM).

Explanation:

Even though hot shutdown boron weight has been injected, reactor water level is greater than -60"(FZ) and reactor power is greater than 3%. EOP 7A has a note that if reactor power rises and continues to rise (as given) then they must reenter EOP 7A at H. With reactor power above 3% they must transition to L and reevaluate. Therefore the crew is

required to stop and prevent injection and lower water until level is less than -60"(FZ)

AND one of the level power condition no longer exists.

Distracters:

a. no ED condition present.

b. Operation is in the safe region of the HCTL and CSBW has not been injected so cooldown is not allowed. In accordance with EOP 6A at override FS/P-4 If average torus water temperature cannot be maintained within HTLC (Graph 7)

Then maintain RPV pressure within HTLC (Graph 7) (exceed cooldown if necessary) 800 psig is the first curve that the operator would have to address and lower pressure to obtain from 950 psig, as given in the stem.

d. Power is continuing to rise therefore the note would require the crew to stop and prevent to lower level.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 223001.A2.07 Importance Rating 4.3 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) EOP 3A (Attach if not previously provided)

(including version/revision number) 13 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: N S 16 NRC Style 00 2/14/2011 SRO: Y 91 Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 4 Multiple Choice N Topic Area Description Systems COR0020302, Effect of High Drywell Pressure on Drywell Ventilation (2008 Requal EXAM 2006 NRC EXAM)

Related Lessons COR0020302 OPS CONTAINMENT Related Objectives COR0020302001130D Describe the PCIS design features and/or interlocks that provide for the following: Bypassing of selected isolations COR0020302001130E Describe the PCIS design features and/or interlocks that provide for the following: Operator action to defeat/reset isolations COR0020302001170A Predict the consequences of the following items on Primary containment: LOCA COR0020302001210C Given plant conditions, determine if the following should have occurred: Drywell cooling fan trip.

Related References 10CFR55.43 (B)(5)

PR 5.8.10 Related Skills (K/A) ROI SROI 295024.EK2.18 Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following: (CFR: 41.7 /

45.8) Ventilation. (3.3/3.4) (3.3 / 3.4) 223001.A2.07 Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES; and (b) based on those predictions, use procedures to correct, control, or mitigate the conseq...:

(CFR: 41.5 / 45 (High drywell pressure (4.2*/4.3* (4.2 /

4.3)

QUESTION: S 16 91 A small break LOCA has occurred with the following conditions:

• Reactor water level is +45 inches (NR).

• Reactor pressure is 560 psig.

• Drywell pressure is 3.1 psig.

• Drywell temperature is 195F.

• Drywell FCU control switches are in RUN.

What is the status of DW cooling and what action is required?

a. All DW FCUs are running; vent primary containment IAW 2.4PC, PRIMARY CONTAINMENT CONTROL.

b. NO DW FCUs are running; Place all DW FCUs to Override IAW 2.4PC, PRIMARY CONTAINMENT CONTROL.

c. All DW FCUs are running; vent primary containment IAW EOP 3A, PRIMARY CONTAINMENT CONTROL.

d. NO DW FCUs are running; Place all DW FCUs to Override IAW EOP 3A, PRIMARY CONTAINMENT CONTROL.

ANSWER: S 16 91

d. NO DW FCUs are running; Place all DW FCUs to Override IAW EOP 3A, PRIMARY CONTAINMENT CONTROL.

Explanation:

With a 1.84 psig signal in to trip the DW FCUs the control switches must be taken to override to clear that signal to get them to run. EOP 3A give the guidance to place DW FCU to override.

Distracters:

a. FCUs have tripped.

b. the drywell hi pressure signal can be overridden with the override switch when in the EOPs, but not in 2.4PC

c. FCUs have tripped.

SRO Justification: 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 268000 G 2.1.23 Importance Rating 4.4 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 8.8.11 Liquid Radioactive Waste Discharge Authorization (Attach if not previously provided)

(including version/revision number) 29 Learning Objective: See Attached (As available)

Question Source: Bank # 23510 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam 2006 CNS Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date S 17 00 04/02/2004 NRC Style RO: N 23510 Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 3 Multiple Choice N Topic Area Description Administrative Liquid Release Authorization/Approval and Actions for a Lost CW Pump During Release (ILT 2006 NRC EXAM)

Related Lessons INT0320115 OPS CNS Administrative Procedures Radiation Protection and Chemistry Procedures (Formal Classroom/Pre-OJT Training)

Related Objectives INT0320115B0B0100 State who, by title, authorizes releases of radioactive liquid effluents from CNS.

INT0320115B0B0300 State the number of Circulating Water Pumps required to be in service during liquid radioactive discharges.

Related References 10CFR55.43 (B)(4)

PR 8.8.11 Related Skills (K/A) ROI SROI 268000 Radwaste 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR:

41.10 / 43.5 / 45.2 / 45.6) (4.3/4.4)

QUESTION: S 17 92 The plant is operating at low power with 2 Circulating Water pumps running. De-icing is in progress. The Radwaste Operator indicates that the Floor Drain Sample Tank requires discharging.

1) Who is required to approve the release?

2) If one of the two operating circulating water pumps trip during the discharge, what action, if any, is required and why?

a. Chemistry department.

Continue the discharge sufficient dilution flow exists.

b. Duty Shift Manager.

Continue the discharge sufficient dilution flow exists.

c. Chemistry department.

Terminate the discharge insufficient dilution flow exists.

d. Duty Shift Manager.

Terminate the discharge insufficient dilution flow exists.

ANSWER: S 17 92

d. Duty Shift Manager.

Terminate the discharge insufficient dilution flow exists.

Explanation:

Procedure 8.8.11 requires that chemistry authorizes the release and the duty Shift Manager approves the release. The loss of one CW pump would reduce flow to less than the minimum required and the discharge should be terminated.

Distracters:

a. is incorrect because the discharge should be terminated.

b. is incorrect because the discharge should be terminated.

c. is incorrect because chemistry authorizes the release they do not approve the release, only the duty Shift Manager can perform this function.

SRO Justification: This is an SRO only item because in accordance with procedure 8.8.11 only the duty Shift Manager can approve liquid radioactive releases.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 2 Group # 2 K/A # 219000.A2.16 Importance Rating 3.2 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) EOP 3A and EOP 6 A (Attach if not previously provided)

(including version/revision number) 13 14 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date S 18 00 NRC Style RO: N 93 Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 3 Multiple Choice N Topic Area Description Emergency Operating INT0080613, Actions High PC Water Level.

Procedures Related Lessons INT0080613 OPS EOP FLOWCHART 3A - PRIMARY CONTAINMENT CONTROL Related Objectives INT00806130011100 Given plant conditions and EOP Flowchart 3A, PRIMARY CONTAINMENT CONTROL, determine required actions.

INT00806130011200 Given plant conditions and EOP flowchart 3A, PRIMARY CONTAINMENT CONTROL, state the reasons for the actions contained in the steps.

Related References PR 5.8 Related Skills (K/A) ROI SROI 219000.A2.16 Ability to (a) predict the impacts of the following on the RHR/LPCI: TORUS/SUPPRESSION POOL COOLING MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal condition or operations: (CFR: 41.5/45.6) High suppression pool level (2.9/3.2)

QUESTION: S 18 93 The plant was operating when a LOCA with an ATWS has occurred.

The following plant conditions are present:

• RPV water level is -60 inches wide range and rising.

• PCIS Group 1, 2, 3, 6, 7 have occurred.

• Reactor Pressure is 300 psig and slowly lowering.

• Drywell pressure is 20 psig and slowly lowering.

• Drywell Temperature is 220º F and slowly lowering.

• Suppression pool temperature is 160º F and slowly rising.

• Primary containment water level is 16.5 feet and slowly rising.

• HPCI has been stopped and prevented.

• Condensate and Feed has been stopped and prevented.

What action is required next?

a. Open 6 SRVs in accordance with EOP 6B, EMERGENCY RPV DEPRESSURIZATION (FAILURE-TO-SCRAM).

b. Place all RHR and CS pumps in PTL in accordance with EOP 7A, RPV LEVEL (FAILURE TO SCRAM).

c. Maximize Suppression Pool cooling using all RHR pumps in accordance with EOP 3A, PRIMARY CONTAINMENT CONTROL.

d. Rapidly depressurize RPV with main turbine BPVs (disregard cooldown rate) in accordance with EOP 6A, RPV PRESSURE (FAILURE TO SCRAM).

ANSWER: S 18 93

b. Place all RHR and CS pumps in PTL in accordance with EOP 7A, RPV LEVEL (FAILURE TO SCRAM).

PROVIDE THE CANDIDATE WITH EOP / SAG GRAPH 7 HCTL Explanation:

In accordance with EOP 3A Primary Containment Control when torus water level reaches 16.5 feet regardless if it could be restored, Emergency Depressurization is required. At that Torus level both the Heat Capacity Temperature Limit (HCTL) curve and the Pressure Suppression Pressure (PSP) curve have been violated.

During ATWS EOPs 6A, 6B, 7A, and 7B when an Emergency Depressurization is require, the action to Stop and Prevent EOP 7A FS/L-14 has been performed, must be completed prior to opening 6 SRVs to prevent cold unborated water from being injected into the core which would cause a power excursion.

Distracters:

a. Open 6 SRVs in accordance with EOP 6B, is incorrect because Stop and Prevent has not been completed. If six SRVs were to be opened at this point, Core Spray systems would spray cold unborated water directly onto the core and power would rise uncontrollably. Stop and Prevent must be completed prior to depressurizing the vessel.

c. Maximize Suppression Pool cooling using all RHR pumps in accordance with EOP 3A, is incorrect because when torus water level reaches 16.5 feet Emergency Depressurization is required. Additional cooling of the water in the Torus is required eventually, but the RPV must be depressurized next.

d. Rapidly depressurize RPV with main turbine BPVs (disregard cooldown rate) in accordance with EOP 6A. Anticipating Emergency Depressurization is not allowed during an ATWS, therefore EOP 6A just allows depressurizations at rates less than 100F per hour.

SRO Justification: 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # 2.1.4 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) T.S Section 5.2 (Attach if not previously provided)

(including version/revision number) 200 Learning Objective: See Attached (As available)

Question Source: Bank # 24480 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 55.43 2 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date S 19 00 12/12/2008 12/20/2010 NRC Style RO: N 24480 Question SRO: Y 94 NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 2 Multiple Choice N Topic Area Description Technical Specifications, T.S. 5.2.2 Unit Staff ODAM, TRM Related Lessons INT0070513 CNS Technical Specifications 5.0, Administrative Controls Related Objectives INT00705130010100 Given a set of plant conditions, recognize non-compliance with a Chapter 5.0 Requirement.

Related References TS 5.2.2 Related Skills (K/A) ROI SROI 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, ¿no-solo¿ operation, maintenance of active license status, 10CFR55, etc. (CFR: 41.10 / 43.2)

(3.3/3.8) (2.3 / 3.4)

QUESTION: S 19 94 The plant is preparing to return to power operations after a 30 day refueling outage.

Minimum staffing requirements for Mode 5 in accordance with Technical Specifications has been set during the outage.

According to Technical Specifications, what additional personnel are required to operate in Mode 3?

a. One NLO, One SRO

b. One NLO, One SRO, One RP

c. Two NLOs, One STE

d. Two NLOs, One RO, One STE ANSWER: S 19 94

c. Two NLOs, One STE Explanation:

From T.S. 5.2.2 Amendment 200 and 10CFR50.54 MODE 3 required positions: MODE 5 required positions: Difference 3 NLOs 1 NLO 2NLO 1 SRO in CR 1 SRO on site 1RO in CR 1 RO in CR 1 RP Tech 1 RP Tech 1 STE 1 STE Distracters:

a. is incorrect - requires two additional NLOs and an STE, not an SRO, he/she is required to be on site during all modes of operation.

b. is incorrect - requires two additional NLOs and an STE not an SRO, he/she is required to be on site during all modes of operation. An RP was already assigned.

d. is incorrect - An additional RO is not needed.

SRO Justification: 10CFR55.43 b (2) Facility operating limitations in the technical specifications and their bases.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.1 Importance Rating 4.4 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedure 2.1.1 and 10.13 and 0-CNS-61 (Attach if not previously provided)

(including version/revision number) 159 64 22 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 55.43 6 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: N S 20 00 1/8/2011 Licensed Operator SRO: Y 95 NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 2 Multiple Choice N Topic Area Description Integrated Plant Period precautions and limits approach to crit.

Related Lessons SKL0124301 OPS COLD STARTUP (All Rods In to 100% Power)

Related Objectives SKL01243010010200 Given a simulated Reactor plant with a startup in progress, the crew will withdraw Control Rods in accordance with procedure 10.13 to bring the Reactor critical and calculate the Reactor period in accordance with procedure 2.1.1.

Related References 2.1.1 STARTUP PROCEDURE 10.13 CONTROL ROD SEQUENCE AND MOVEMENT CONTROL 0-CNS-61 Reactivity Management Related Skills (K/A) ROI SROI 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13)

(4.4/4.7) 2.2.1 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity. (CFR: 41.5 / 41.10

/ 43.5 / 43.6 / 45.1) (4.5/4.4)

QUESTION: S20 95 Given the following set of conditions:

• A Reactor startup in progress.

• The RO withdraws control rod 22-19 from notch 12 to notch 14.

• Reactor period changes from 100 seconds to a stable 39 seconds.

Which of the following identifies the NEXT required action to be taken?

a. Re-insert control rods as necessary to achieve sub-criticality per GOP 2.1.1 Startup Procedure.

b. Re-insert control rods as necessary to achieve sub-criticality per NPP 10.13 Control Rod Sequence.

c. Re-insert control rod 22-19 to obtain a stable period indication of greater than 50 seconds per GOP 2.1.1 Startup Procedure.

d. Re-insert control rod 22-19 to obtain a stable period indication of greater than 50 seconds per NPP 10.13 Control Rod Sequence.

ANSWER: S20 95

c. Re-insert control rod 22-19 to obtain a stable period indication of greater than 50 seconds per GOP 2.1.1 Startup Procedure.

Explanation:

In accordance with GOP 2.1.1 "A period faster than 50 seconds shall not be maintained." The operator would reinsert 22-19 to notch 12 to establish stable period that is longer than 50 seconds. In this case there is no requirement to continue inserting control rods to achieve a sub-critical condition.

In accordance with 0-CNS-61 CNS REACTIVITY MANAGEMENT PROGRAM Estimated Critical Position (ECP) - Procedures shall incorporate the following expectations:

• Criteria with acceptable band of delta k/k for criticality shall be specified, when practical. Contingency guidance to make or maintain core subcritical if failure to meet criteria until a determination made to explain deviation.

From Procedure 2.1.1 Startup Procedure Precautions and Limitations 2.22 Conservative action is required whenever an unexpected situation arises with respect to reactivity, criticality, power level, or any other anomalous behavior of reactor core. This conservative action should include rod insertion to reduce power or a reactor scram without hesitation whenever such unanticipated or anomalous behavior is encountered.

Pg 16 CAUTION - If moderator temperature is dropping when approaching criticality, caution should be exercised. Due to greater moderation, period may become shorter, requiring rod insertion until reactor power reaches heating range and moderator temperature starts rising.

Distracters:

a. Re-insert control rods as necessary to achieve sub-criticality per GOP 2.1.1 Startup Procedure. This short period does not require the operator to place the reactor in a sub-critical condition, just to reinsert the rod and establish an approximate 100 second period (>50).

b. Re-insert control rods as necessary to achieve sub-criticality per NPP 10.13 Control Rod Sequence. The guidance for delta k/k for criticality not met is to make or maintain core subcritical until a determination is made to explain the deviation. NPP-10.13 Section 8; Emergency Power Reduction is not necessary.

Even with the guidance in 0-CNS-61 listed above.

d. Re-insert control rod 22-19 to obtain a stable period indication of greater than 50 seconds per NPP 10.13 Control Rod Sequence. Section 3; Approach to Critical 3.2.3 Confirm that rod configuration is within the +/- 1% ECP band recorded on Attachment 1. If it is not within the band, cease rod movement and contact a Reactor Engineer for evaluation. Do not proceed with rod movement until directed by Reactor Engineer.

SRO Justification: 10CFR55.43 b (6) Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 3 K/A # 2.3.13 Importance Rating 3.8 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 9.EN-RP-108 (Attach if not previously provided)

(including version/revision number) 5 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 55.43 4 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: N S 21 00 01/08/2011 Licensed Operator SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 3 Multiple Choice N Topic Area Description Administrative INT0320115, (DDE) of 1250 mrem/hour. How should the entrance to this room be posted?

Related Lessons INT0320115 OPS CNS Administrative Procedures Radiation Protection and Chemistry Procedures (Formal Classroom/Pre-OJT Training)

Related Objectives INT0320115H0H010D Discuss the following as described in Radiological Protection Procedure 9.EN-RP-108, Area Posting and Access Control: RCA/Satellite Area controls and postings Related References PR 9.EN-RP-108 TS 5.7 Related Skills (K/A) ROI SROI 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4 / 45.10)

(3.2/3.7) (2.5 / 3.1) 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, (CFR: 41.12

/ 43.4 / 45.9 / 45.10) (3.4/3.8)

QUESTION: S21 96 One of the rooms in the Reactor Building contains an area in which a person could receive a deep dose equivalent (DDE) of 1250 mRem/hour.

How should the entrance to this room be posted?

a. CAUTION - LOCKED HIGH RADIATION AREA IAW 9.EN-RP-108 Radiation Protection Posting

b. CAUTION - HIGH RADIATION AREA IAW 9.EN-RP-108 Radiation Protection Posting

c. CAUTION - HIGH RADIATION AREA IAW CNS Technical Specification 5.7.1 High Radiation Area.

d. CAUTION - LOCKED HIGH RADIATION AREA IAW Technical Specification 5.7.1 High Radiation Area.

ANSWER: S21 96

a. CAUTION - LOCKED HIGH RADIATION AREA IAW 9.EN-RP-108 Radiation Protection Posting Explanation:

9.EN-RP-108 Radiation Protection Posting, Locked High Radiation Area (LHRA) - An area, accessible to individuals, in which radiation levels from sources external to the body could result in an individual receiving a deep dose equivalent > 1 Rem (10 mSv) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm (- 12")

from the radiation source or from any surface that the radiation penetrates.

This is not covered in 5.7.1 which is each high radiation area in which the deep dose equivalent in excess of 100 mrem but less than 1000 mrem in one hour (measurement made at 12 inches from source of radiation) shall be barricaded (barricade will impede physical movement across the entrance or access to the high radiation area; i.e., doors, yellow and magenta rope, turnstile) and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Special Work Permit (SWP).

Per Technical Specifications, section 5.7.2, an area in which an individual could pick up a DDE in excess of 1000 mem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shall be classified as a locked high radiation area.

Reference:

Procedure 9.EN-RP-108, Technical Specifications SRO Justification: 10CFR55.43 b (2) Facility operating limitations in the technical specifications and their bases.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.28 Importance Rating 4.1 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedure 5.7.1 Attachment 1 EAL Matrix (Attach if not previously provided)

(including version/revision number) 43 Learning Objective: See Attached (As available)

Question Source: Bank # 20682 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis H 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date S 22 00 04/21/2004 05/23/2010 NRC Style RO: N 20682 Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 H 1 5 Multiple Choice N Topic Area Description Emergency Plan INT0320136, SECURITY EAL Related Lessons COR0090001 OPS ACP/EP REVIEW INT0320136 OPS-CNS Abnormal Procedures (RO) Miscellaneous Related Objectives COR00900010010400 Given plant conditions and appropriate Abnormal/Emergency Procedure, determine required Subsequent Operator Actions(s).

Related References PR 5.7.1 Related Skills (K/A) ROI SROI 2.4.28 Knowledge of procedures relating to a security event (non-safeguards information). (CFR: 41.10 / 43.5 /

45.13) (3.2/4.1) (2.3 / 3.3)

QUESTION: S 22 97 The plant was operating at near rated power when the Security Shift Supervisor informed the Shift Manager that heavily armed intruders were attempting to gain entry into the protected area. The reactor was scrammed and the Station Operators directed to obtain fire fighting gear and report to the Control Room. The following events occurred following initial report by the Security Shift Supervisor:

0900 Armed intruders have gained access to the Protected Area.

0905 An explosion occurs at the B CST and intruders heading to the power block.

0910 Intruders are entering the Reactor Building 903 southeast via the steam tunnel roof and the ASD room door.

0915 Intruders occupy CAS, SAS and have taken the entire operating crew hostage.

When is a General Emergency first required to be declared?

a. 0900

b. 0905

c. 0910

d. 0915 ANSWER: S 22 97

c. 0910 Explanation:

The intruders have gained access to the reactor building via ASD room which they must now control. This requires the declaration of a General Emergency. The fact that the intruders occupy ASD would indicated that station control is lost and a General Emergency is required.

Distracters:

a. is incorrect as this would only require the declaration of an alert.

b. is incorrect as this would only require the declaration of a SAE.

d. is incorrect although this requires a GE this is not the first condition requiring a GE as asked in the test item stem.

PROVIDE THE STUDENTS Copy of Procedure 5.7.1 EAL Hard cards

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 2 K/A # 2.2.43 Importance Rating 3.3 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Procedure 2.0.3 and 2.3.1 (Attach if not previously provided)

(including version/revision number) 73 59 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: N NRC Style S 23 00 NEW SRO: Y Question NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 4 Multiple Choice N Topic Area Description Administrative INT0320103, Annunciator Procedure Related Lessons INT0320103 CNS Administrative Procedures Conduct of Operations and General Alarm Procedures (Formal Classroom/Pre-OJT Training)

Related Objectives INT032010300E010D Discuss the following as described in Alarm Procedure 2.3.1, General Alarm Procedure: Annunciator disabling Related References PR 2.3.1 Related Skills (K/A) ROI SROI 2.2.43 Knowledge of the process used to track inoperable 3.0 3.3 alarms. (CFR: 41.10 / 43.5 / 45.13) (3.0/3.3)

QUESTION: S 23 98 The plant is operating at near rated power when the following annunciator alarms:

A-2/C-2 RFP TURBINE B LOW VACUUM PRE-TRIP

• The crew reviews plant conditions and determines the alarm is invalid, but will not reset.

• Further investigation by maintenance reveals the parts to fix the alarm will not be on site for 5 days.

• The decision is made to disable the alarm.

What is the acceptable method for tracking this disabled alarm?

a. Tracked on the Crew turnover sheet IAW COP 2.0.3 Conduct of Operations.

b. Tracked in the NOMs LCO tracking module IAW AP 2.3.1 General Alarm Procedure.

c. Tracked in the NOMs Narrative logs as a night order IAW Operations Instruction

1. 4 Standing and Night Orders.

d. Tracked in the NOMs Narrative logs as a narrative log entry IAW COP 2.0.2 Operations Logs and Reports.

ANSWER: S 23 98

b. Tracked in the NOMs LCO tracking module IAW AP 2.3.1 General Alarm Procedure.

Explanation:

In accordance with Procedure 2.3.1 General Alarm Procedure step 7.4 Disabled alarm evaluation shall be documented and tracked by NOMS LCO Tracking module per Operations Desk Guide 3, NOMS; or Attachment 1, Alarm Problem Evaluation (APE), if NOMS is not available. There is no mention of NOMs being unavailable in the stem and that assumption should not be made.

Alarms disabled for short durations (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) maybe tracked with a narrative log entry.

The alarm evaluation and compensatory actions should be specified in the log entry.

Distracters:

a. Tracked on the Crew turnover sheet IAW COP 2.0.3 Conduct of Operations, is incorrect, the NOMs system should be used for long term disabled annunciators.

Also the procedure referenced is incorrect. The procedure that governs shift turnover and contains the turnover sheet is COP 2.0.4 RELIEF PERSONNEL AND SHIFT TURNOVER.

c. Tracked in the NOMs Narrative logs as a night order IAW Operations Instruction

1. 4 Standing and Night Orders, is incorrect, the NOMs system should be used for long term disabled annunciators.

d. Tracked in the NOMs Narrative logs as a narrative log entry IAW COP 2.0.2 Operations Logs and Reports. If this were a short duration disabled annunciator

(<24 hours), this could be used to track the disabled annunciator. However this annunciator is a long duration (>24 hour) item and the only correct way is to track it via the NOMs LCO Tracker.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 1 K/A # G 2.1.34 Importance Rating 3.5 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) 2.4CHEM (Attach if not previously provided)

(including version/revision number) 11 Learning Objective: See Attached (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date S 24 00 03/29/2011 NRC Style RO: N 99 Question SRO: Y NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 3 L 1 3 Multiple Choice N Topic Area Description Administrative Hotwell Conductivity above limit, what action and what procedure is required?

Related Lessons INT0320115 CNS Administrative Procedures Radiation Protection and Chemistry Procedures Related Objectives INT0320115 Obj.2. Discuss the following as described in Administrative Procedure 8.3, Control Parameters and Limits: a. Precautions and Limitations 1) Technical Specifications /

TRM requirements 2) Action Levels 3) Chemistry Warning Limits b. Response to Out-of-Limit Conditions Related References 8.3 CONTROL PARAMETERS AND LIMITS 2.4CHEM, CHEMISTRY PARAMETER OUT OF LIMIT Related Skills (K/A) ROI SROI 2.1.34 Knowledge of primary and secondary plant chemistry limits. 3.5 (CFR: 41.10 / 43.5 / 45.12) IMPORTANCE RO 2.7 SRO 3.5

QUESTION: S 24 99 The Plant is operating at 20% power when a Main Condenser tube leak develops resulting in rising hotwell conductivity of 0.8 mho on MC-CR-1, Condensate Conductivity (Channels 1-4). Conductivity continues to rise slowly.

What procedure is required to be entered and what actions are required?

a. 8.3, CONTROL PARAMETERS AND LIMITS, and reduce Reactor Power to less than 10% rated.

b. 2.4CHEM, CHEMISTRY PARAMETER OUT OF LIMIT, and reduce Reactor Power to less than 10% rated.

c. 8.3, CONTROL PARAMETERS AND LIMITS, Isolate the affected waterbox.

d. 2.4CHEM, CHEMISTRY PARAMETER OUT OF LIMIT, Isolate the affected waterbox.

ANSWER: S 24 99

d. 2.4CHEM, CHEMISTRY PARAMETER OUT OF LIMIT, Isolate the affected waterbox.

Explanation:

Abnormal Procedure 2.4CHEM page 2 step 4.3 Reactor Power > 10% RTP.

4 .3 .1 If hotwell conductivity cannot be maintained below 0 .5 mho on MC-CR-1 and tube/tube sheet leakage suspected, isolate affected water box per Attachment 1 (Page 3).

Distracters:

a. 8.3, CONTROL PARAMETERS AND LIMITS, is the procedure that Chemistry will be performing during this evolution. However there is no need to reduce power the limits are the same whether you are above or below 10% power.

b. 2.4CHEM, CHEMISTRY PARAMETER OUT OF LIMIT, this is the correct procedure, However there is no need to reduce power the limits are the same whether you are above or below 10% power.

c. 8.3, CONTROL PARAMETERS AND LIMITS, is the procedure that Chemistry will be performing during this evolution. However there are no actions for Operations to isolate the waterbox in this procedure. They would have to transition to Operating Procedure 2.2.3.

SRO Justification: 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-

Reference:

Level RO SRO Tier # 3 Group # 4 K/A # 2.4.5 Importance Rating 4.3 Proposed Question: See Attached Proposed Answer: _See Attached______

Explanation: See Attached Technical Reference(s) Admin Procedure 0.1 Procedure Use And Adherence (Attach if not previously provided)

(including version/revision number) 36 Learning Objective: See Attached (As available)

Question Source: Bank # 23314 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam NA Question Cognitive Level: Memory or Fundamental Knowledge L Comprehension or Analysis 10 CFR 55 Content 55.41 55.43 5 Comments:

ES-401, Page 28 of 33

Question Revision Revision Last Used Exam Bank Applicability Number Number Date Date RO: N S 25 01 08/06/2009 01/07/2011 Licensed Operator SRO: Y 23314 NLO: N Difficulty Cognitive Point Response Question Type Inactive?

Level Level Value Time 2 L 1 3 Multiple Choice N Topic Area Description Administrative INT0320101, Explain how to perform actions when an abnormal conflicts with an EOP. (2006 ILT AUDIT EXAM)

Related Lessons INT0320101 CNS Administrative Procedures Volume 0, Administrative Procedures (Formal Classroom/Pre-OJT Training)

Related Objectives INT032010100E010A Discuss the following as described in Administrative Procedure 0.1, Introduction to CNS Operations Manual: Procedural adherence INT032010100E010E Discuss the following as described in Administrative Procedure 0.1, Introduction to CNS Operations Manual: Procedure hierarchy Related References PR 0.1 Related Skills (K/A) ROI SROI 2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions. (CFR: 41.10 / 43.5 / 45.13)

(3.7/4.3) (2.9 / 3.6)

QUESTION: S 25 100 One of the Immediate Operator Actions of 2.4VAC, LOSS OF CONDENSER VACUUM as condenser vacuum degrades is to close the MSIVs if vacuum cannot be maintained >

12" Hg. Due to a loss of alternate pressure control systems the Bypass Valves are the only system available to perform EOP 1A Pressure Control actions.

How is the Senior Reactor Operator to respond to this conflict between the procedures?

a. Invoke 10CFR50.54X and perform the actions to reopen the MSIVs even though the Abnormal directs that they be closed.

b. Get authorization from two SROs to deviate from the Abnormal and reopen the MSIVs to perform the actions in EOP 1A.

c. Follow the actions of EOP 1A and open the MSIVs, disregard the actions in AP 2.4VAC that require them to be closed as vacuum lowers.

d. Contact the work control center to expedite repairs of the Alternate Pressure Control systems and wait until they are available to continue EOP 1A actions.

ANSWER: S 25 100

c. Follow the actions of EOP 1A and open the MSIVs, disregard the actions in AP 2.4VAC that require them to be closed as vacuum lowers.

Explanation:

Ops Policy Procedure 2.0.1.2 Alarm/Abnormal/Emergency/System Operating Procedures/Instrument Operating Procedures may be carried out concurrently with an EOP. In the event that conflicting actions are directed by procedures, the EOP actions shall take precedence.

EOPs/SAMGs are Operations highest tier procedure. If an explicit operation is directed by EOPs per a 5.8 EOP Support Procedure, then transition shall be made from the Alarm/Abnormal/Emergency/System Operating/Instrument Operating Procedures (including hard cards) to the 5.8 Procedure to perform or continue performing that operation.

Distracters:

a. Invoke 10CFR50.54X and perform the actions to reopen the MSIVs even though the Abnormal directs that they be closed. This answer is incorrect, because there is guidance within the Operating Procedure and an Admin process that tells you how to handle this situation.

b. Get authorization from two SROs to deviate from the Abnormal and reopen the MSIVs to perform the actions in EOP 1A. This answer is incorrect because

Procedure 2.0.1.2 covers this situation and allows the operator to perform the higher tier procedure.

d. Contact the work control center to expedite repairs of the Alternate Pressure Control systems and wait until they are available to continue EOP 1A actions.

This answer is incorrect because there is guidance in the EOP to reopen the MSIVs.

SRO Justification 10CFR55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

The pen and ink changes that were done on the white board in the exam admin room during administration are documented in the post exam comments/analysis file and not directly on the exam due to file size issues. Rescanning a document with a few minor edits makes the file a lot larger than just noting for the record that these changes were done on the white board and then documented as mentioned above.

Final Written Exam references

COOPER NUCLEAR STATION OPERATIONS MANUAL 2A Emergency Operating Procedure 5.8 Attachment 1 EMERGENCY RPV Revision: 14 DEPRESSURIZATION STEAM COOLING 4

C 5 RC/P-9 RC/P-16 IF PC water level rise above 44 ft is anticipated THEN open inboard main steam line drain valve before PC water level reaches 44 ft IF Emergency RPV Depressurization is required THEN C (defeat isolation interlocks if necessary), EOP 5.8.20 IF RPV water level is rising THEN 1A, 2 IF RPV water level cannot be determined THEN 2B, 6 IF any means of RPV injection is available THEN Emergency RPV Depressurization AND RPV water level cannot be restored and maintained above -183 in. is required C RC/P-10 Does IF any SRV is being used to stabilize RPV pressure THEN Emergency RPV Depressurization high drywell AND pressure ECCS YES SRV continuous IA/N2 supply is or becomes unavailable initiation signal exist is required C (1.84 psig)

RC/P-11 IF RPV water level cannot be determined THEN 2B, 6 Prevent injection from CS and LPCI not required for adequate core cooling RC/P-17 NO D

Stabilize RPV pressure using following as necessary:

• SRVs only with PC water level above 6 ft, EOP 5.8.1 RC/P-12 (Restore IA/N2 supply if necessary) 2 Rapidly depressurize RPV (disregard cooldown rate): ALTERNATE EMERGENCY

• 3 4 HPCI (defeat interlocks DEPRESSURIZATION SYSTEMS, EOP 5.8.2 if necessary, EOP 5.8.20)

• IF PC water level is above 6 ft

• 2 3 4 RCIC (defeat interlocks THEN Open 6 SRVs

• Main steam lines and turbine bypass valves if necessary, EOP 5.8.20)

(restore IA/N2 supply if necessary)

• Main steam line drains

• RWCU with filters bypassed, EOP 5.8.1

• IF less than 4 SRVs are open (defeat isolation interlocks if necessary)

AND

• 2 3 4 RCIC RPV pressure is 50 psig or more above torus pressure

• Gland sealing steam RC/P-18 THEN 1. Rapidly depressurize RPV to less

• RFPT A than 50 psig above torus pressure WHEN using Alternate Emergency

• RFPT B Depressurization Systems (TABLE 2) as RPV water level necessary

• Steam jet air ejectors drops below (exceed offsite radioactivity release rate -202 in.

limits if necessary)

• RWCU in blowdown mode (defeat interlocks if necessary)

• AOG steam line drains

2. Maintain RPV pressure less than 50 psig above torus pressure

• AOG 3rd stage air ejector and preheater

• 3 4 HPCI (only with PC water level EMERGENCY RPV above 11 ft) DEPRESSURIZATION IS REQUIRED RC/P-13

• RWCU with filters bypassed IF less than 4 SRVs are open THEN D

• RPV head vent AND RPV pressure is 50 psig or more above torus pressure RC/P-14 WHEN shutdown cooling RPV pressure interlock clears (70 psig)

RC/P-15 Cool down to cold shutdown conditions with shutdown cooling (use only RHR pumps not required to maintain RPV water level above +3 in.)

IF shutdown cooling cannot be established and 2 Elevated suppression chamber pressure may trip RCIC turbine on high further cooldown is required exhaust pressure THEN continue to cool down to less than 212°F using 3 Operation of HPCI, RCIC, Core Spray, or RHR with suction from SRVs or one or more Alternate Emergency suppression pool and pump flow above NPSH or Vortex Limit Depressurization Systems (TABLE 2) (GRAPHs 3, 4, 5, 16, 18) may result in equipment damage (defeat interlocks if necessary) 4 Operation of HPCI or RCIC turbines with suction temperatures above 140°F may result in equipment damage

COOPER NUCLEAR STATION OPERATIONS MANUAL 2A Emergency Operating Procedure 5.8 Attachment 1 EMERGENCY RPV Revision: 14 DEPRESSURIZATION STEAM COOLING 4

C 5 RC/P-9 RC/P-16 IF PC water level rise above 44 ft is anticipated THEN open inboard main steam line drain valve before PC water level reaches 44 ft IF Emergency RPV Depressurization is required THEN C (defeat isolation interlocks if necessary), EOP 5.8.20 IF RPV water level is rising THEN 1A, 2 IF RPV water level cannot be determined THEN 2B, 6 IF any means of RPV injection is available THEN Emergency RPV Depressurization AND RPV water level cannot be restored and maintained above -183 in. is required C RC/P-10 Does IF any SRV is being used to stabilize RPV pressure THEN Emergency RPV Depressurization high drywell AND pressure ECCS YES SRV continuous IA/N2 supply is or becomes unavailable initiation signal exist is required C (1.84 psig)

RC/P-11 IF RPV water level cannot be determined THEN 2B, 6 Prevent injection from CS and LPCI not required for adequate core cooling RC/P-17 NO D

Stabilize RPV pressure using following as necessary:

• SRVs only with PC water level above 6 ft, EOP 5.8.1 RC/P-12 (Restore IA/N2 supply if necessary) 2 Rapidly depressurize RPV (disregard cooldown rate): ALTERNATE EMERGENCY

• 3 4 HPCI (defeat interlocks DEPRESSURIZATION SYSTEMS, EOP 5.8.2 if necessary, EOP 5.8.20)

• IF PC water level is above 6 ft

• 2 3 4 RCIC (defeat interlocks THEN Open 6 SRVs

• Main steam lines and turbine bypass valves if necessary, EOP 5.8.20)

(restore IA/N2 supply if necessary)

• Main steam line drains

• RWCU with filters bypassed, EOP 5.8.1

• IF less than 4 SRVs are open (defeat isolation interlocks if necessary)

AND

• 2 3 4 RCIC RPV pressure is 50 psig or more above torus pressure

• Gland sealing steam RC/P-18 THEN 1. Rapidly depressurize RPV to less

• RFPT A than 50 psig above torus pressure WHEN using Alternate Emergency

• RFPT B Depressurization Systems (TABLE 2) as RPV water level necessary

• Steam jet air ejectors drops below (exceed offsite radioactivity release rate -202 in.

limits if necessary)

• RWCU in blowdown mode (defeat interlocks if necessary)

• AOG steam line drains

2. Maintain RPV pressure less than 50 psig above torus pressure

• AOG 3rd stage air ejector and preheater

• 3 4 HPCI (only with PC water level EMERGENCY RPV above 11 ft) DEPRESSURIZATION IS REQUIRED RC/P-13

• RWCU with filters bypassed IF less than 4 SRVs are open THEN D

• RPV head vent AND RPV pressure is 50 psig or more above torus pressure RC/P-14 WHEN shutdown cooling RPV pressure interlock clears (70 psig)

RC/P-15 Cool down to cold shutdown conditions with shutdown cooling (use only RHR pumps not required to maintain RPV water level above +3 in.)

IF shutdown cooling cannot be established and 2 Elevated suppression chamber pressure may trip RCIC turbine on high further cooldown is required exhaust pressure THEN continue to cool down to less than 212°F using 3 Operation of HPCI, RCIC, Core Spray, or RHR with suction from SRVs or one or more Alternate Emergency suppression pool and pump flow above NPSH or Vortex Limit Depressurization Systems (TABLE 2) (GRAPHs 3, 4, 5, 16, 18) may result in equipment damage (defeat interlocks if necessary) 4 Operation of HPCI or RCIC turbines with suction temperatures above 140°F may result in equipment damage

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT AG1.1 1 2 3 4 5 DEF AS1.1 1 2 3 4 5 DEF AA1.1 1 2 3 4 5 DEF AU1.1 1 2 3 4 5 DEF CA1.1 4 5 DEF CU1.1 4 5 Any valid gaseous monitor reading > Table A-1 column GE for 15 min. (Note 1)

Any valid gaseous monitor reading > Table A-1 column SAE for 15 min. (Note 1)

Any valid gaseous monitor reading > Table A-1 column Alert for 15 min. (Note 2)

Any valid gaseous monitor reading > Table A-1 column UE for 60 min. (Note 2) 1 None None Loss of all offsite and all onsite AC power (Table C-4) to critical 4160V buses 1F and 1G for 15 min. (Note 3)

AC power capability to critical 4160V buses 1F and 1G reduced to a single power source (Table C-4) for 15 min.

Loss of such that any additional single failure would result in loss of AC Power all AC power to critical buses (Note 3)

AG1.2 1 2 3 4 5 DEF AS1.2 1 2 3 4 5 DEF AA1.2 1 2 3 4 5 DEF AU1.2 1 2 3 4 5 DEF Dose assessment using actual meteorology indicates doses Dose assessment using actual meteorology indicates doses Any valid liquid effluent monitor reading > Table A-1 column Any valid liquid effluent monitor reading > Table A-1 column 1 > 1 Rem TEDE or > 5 Rem thyroid CDE at or beyond the site boundary

> 0.1 Rem TEDE or > 0.5 Rem thyroid CDE at or beyond the site boundary Alert for 15 min. (Note 2) UE for 60 min. (Note 2) CG2.1 4 5 CS2.1 4 5 CA2.1 4 5 CU2.1 4 Offsite Rad RPV level < -158 in. for 30 min. (Note 3) With Containment Closure not established, RPV level < -42 in. RPV level cannot be restored and maintained > +3 in.

Conditions AND RPV level < -48 in. (Note 4) OR for 15 min. (Note 3) due to RCS leakage AG1.3 1 2 3 4 5 DEF AS1.3 1 2 3 4 5 DEF AA1.3 1 2 3 4 5 DEF AU1.3 1 2 3 4 5 DEF Any Containment Challenge indication, Table C-5 RPV level cannot be monitored for 15 min. (Note 3) with any unexplained RPV leakage indication, Table C-1 Field survey results indicate closed window dose rates Field survey indicates closed window dose rate > 0.1 Rem/hr Confirmed sample analyses for gaseous or liquid releases Confirmed sample analyses for gaseous or liquid releases

> 1 Rem/hr expected to continue for 60 min. at or beyond that is expected to continue for 60 min. at or beyond the site indicate concentrations or release rates > 200 x ODAM limits indicate concentrations or release rates > 2 x ODAM CG2.2 4 5 CS2.2 4 5 CU2.2 5 the site boundary (Note 1) boundary (Note 1) for 15 min. (Note 2) limits for 60 min. (Note 2)

A 2 OR OR Analyses of field survey samples indicate thyroid CDE Field survey sample analysis indicates thyroid CDE > 0.5 Rem RPV level cannot be monitored for 30 min. (Note 3) with With Containment Closure established, Unplanned RPV level drop for 15 min (Note 3) below

> 5 Rem for 1 hr of inhalation at or beyond the site boundary for 1 hr of inhalation at or beyond the site boundary core uncovery indicated by RPV level < -158 in. (Note 4) EITHER:

EITHER: RPV flange (LI-86: 206 in. normal calibration, 113.75 in.

RPV Unexplained RPV leakage indication, Table C-1 elevated calibration)

Abnorm. AA2.1 1 2 3 4 5 DEF AU2.1 1 2 3 4 5 DEF Level OR OR Rad Table A-1 Effluent Monitor Classification Thresholds CS2.3 4 5 RPV level band when the RPV level band is established Erratic Source Range Monitor indication Release Damage to irradiated fuel OR loss of water level (uncovering Unplanned water level drop in the reactor cavity or spent fuel below the RPV flange AND

/ Rad GE SAE ALERT UE irradiated fuel outside the RPV) that causes EITHER of the pool as indicated by any of the following: RPV level cannot be monitored for 30 min. (Note 3) with a Monitor Any Containment Challenge indication, Table C-5 Effluent for 15 min. for 15 min. for 15 min. for 60 min. following:

• LI-86 (calibrated to 1001' elev.) loss of inventory as indicated by EITHER:

2 Valid RMA-RA-1 Fuel Pool Area Rad reading > 50 R/hr

• Spent fuel pool low level alarm Unexplained RPV leakage indication, Table C-1

• Visual observation OR CU2.3 5 C

ERP 3.50E+08 µCi/sec 3.50E+07 µCi/sec 2.80E+06 µCi/sec 2.24E+05 µCi/sec OR AND Erratic Source Range Monitor indication Valid RMP-RM-452 A-D Rx Bldg Vent Exhaust Plenum Valid area radiation monitor reading rise on RMA-RA-1 or RPV level cannot be monitored with any unexplained Onsite Rad Hi-Hi alarm RMA-RA-2 RPV leakage indication, Table C-1 Conditions Rx Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.45E+05 µCi/sec 8.48E+04 µCi/sec GASEOUS Spent Fuel Cold SD/

Pool Turb Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.62E+05 µCi/sec 9.02E+04 µCi/sec AA2.2 1 2 3 4 5 DEF AU2.2 1 2 3 4 5 DEF Events Refuel CA3.1 4 5 CU3.1 4 5 A water level drop in the reactor refueling cavity or spent fuel Unplanned valid area radiation monitor reading or survey System RW / ARW Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.64E+05 µCi/sec 9.08E+04 µCi/sec pool that will result in irradiated fuel becoming uncovered results rise by a factor of 1,000 over normal levels*

• Normal levels can be considered as the highest reading in the past 24 Malfunct.

3 Any unplanned event results in EITHER:

RCS temperature > 212°F for > Table C-3 duration (Note 4)

OR Any unplanned event results in RCS temperature > 212°F due to loss of decay heat removal capability The lesser of: The lesser of: None None hours excluding the current peak value RCS RPV pressure increase > 10 psig due to loss of RCS 200 x calculated 2 x calculated LIQUID Temp. cooling CU3.2 4 5 Rad Waste Effluent ----- ----- alarm values alarm values AA3.1 1 2 3 4 5 DEF 3 OR monitor upscale*

OR monitor upscale* Dose rates > 15 mRem/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety Loss of all RCS temperature and RPV level indication for 15 min. (Note 3)

MCR/CAS Service Water Effluent ----- ----- 4.80E-04 µCi/cc 4.80E-06 µCi/cc functions:

Rad Main Control Room (RM-RA-20) CU4.1 4 5 5 DEF OR

• with effluent discharge not isolated CAS 4

Loss of all Table C-2 onsite (internal) communication methods None None affecting the ability to perform routine operations HA1.1 1 2 3 4 5 DEF HU1.1 1 2 3 4 5 DEF None OR Table H-1 Safe Shutdown Areas Comm. Loss of all Table C-2 offsite (external) communication Seismic event > 0.1g as indicated by SMA-3 Strong Motion Seismic event identified by any two of the following: methods affecting the ability to perform offsite notifications Accelograph or Alarm B-3/A-1 EMERGENCY SEISMIC HIGH

• SMA-3 Strong Motion Accelograph actuated or Alarm

• Reactor Building LEVEL B-3/B-1 SEISMIC EVENT AND CU5.1 4 5 5

• Control Building

• Earthquake felt in plant Earthquake confirmed by any of the following:

• National Earthquake Information Center

• Service Water Pump Room

• Earthquake felt in plant An unplanned sustained positive period observed on

• National Earthquake Information Center None None None

• Diesel Generator Building nuclear instrumentation

• Control Room indication of degraded performance of Inadvertent

• Cable Expansion Room systems required for the safe shutdown of the plant Criticality 6

HA1.2 1 2 3 4 5 DEF HU1.2 1 2 3 4 5 DEF CU6.1 4 5 Tornado striking or high winds > 100 mph resulting in EITHER: Tornado striking within Protected Area boundary < 105 VDC bus voltage indications on all Technical None None None Visible damage to any Table H-1 area structure containing OR Specification required 125 VDC buses for 15 min. (Note 3) safety systems or components Loss of Sustained high winds > 100 mph OR DC Power Control Room indication of degraded performance of safety systems HA1.3 1 2 3 4 5 DEF HU1.3 1 2 3 4 5 DEF Main turbine failure-generated projectiles result in EITHER: Main turbine failure resulting in casing penetration or 1 Visible damage to or penetration of any Table H-1 area structure containing safety systems or components OR damage to turbine or generator seals Table C-1 RPV Leakage Indications Table C-2 Communications Systems Table C-3 RCS Reheat Duration Thresholds Table C-4 AC Power Sources Onsite Offsite

• If an RCS heat removal system is in operation within this time Offsite Natural or Control Room indication of degraded performance of

• Drywell equipment drain sump level rise System frame and RCS temperature is being reduced, the EAL is not Destructive safety systems (internal) (external)

• Drywell floor drain sump level rise applicable

• Startup Station Service Transformer Phenomena 1 2 3 4 5 DEF 1 2 3 4 5 DEF Station Intercom System Gaitronics X

• Emergency Station Service HA1.4 HU1.4

• Reactor Building equipment drain sump level rise

1. RCS intact (Containment Closure N/A) 60 min.* Transformer Flooding in any Table H-1 area resulting in EITHER: Flooding in any Table H-1 area that has the potential to

• Reactor Building floor drain sump level rise Site UHF Radio Consoles X

• Backfeed 345 kv line through Main An electrical shock hazard that precludes access to operate affect safety-related equipment required by Technical Radio Paging System X Power Transformer to the Normal or monitor safety equipment Specifications for the current operating mode

• Suppression Pool water level rise Alternate Intercom X 2. Containment Closure established Station Service Transformer (Note 8)

OR

• RPV make-up rate rise 20 min.*

AND Control Room indication of degraded performance of safety Sound Power System X Onsite systems

• Observation of unisolable RCS leakage RCS not intact Notes CNS On-Site Cell Phone System X X

• DG-1

1. The Emergency Director should not wait until the applicable time has elapsed, but should declare Telephone system (PBX) X X

• DG-2 the event as soon as it is determined that the condition will likely exceed the applicable time. HA1.5 1 2 3 4 5 DEF HU1.5 1 2 3 4 5 DEF

3. Containment Closure not established Federal Telecommunications System (FTS 2001) X If dose assessment results are available, declaration should be based on dose assessment instead High river/forebay water level > 902' MSL High river/forebay water level > 899' MSL AND 0 min.

of radiation monitor values. OR OR Table C-5 Containment Challenge Indications Local Telephones (C.O. Lines) X RCS not intact (See EAL AS1.2/AG1.2.) Do not delay declaration awaiting dose assessment results. Low river/forebay level < 865' MSL Low river level/forebay < 870' MSL CNS State Notification Telephones X

2. The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed,

• Containment Closure not established (Note 4) the applicable time. In the absence of data to the contrary, assume that the release duration has HA1.6 1 2 3 4 5 DEF

• Deflagration concentrations exist inside PC exceeded the applicable time if an ongoing release is detected and the release start time is unknown. Vehicle crash resulting in EITHER: 6% H2 in drywell or torus

3. The Emergency Director should not wait until the applicable time has elapsed, but should declare Visible damage to any Table H-1 area structure containing the event as soon as it is determined that the condition will likely exceed the applicable time. safety systems or components AND

4. Containment Closure is the action taken to secure primary or secondary containment and its OR 5% O2 in drywell or torus associated structures, systems, and components as a functional barrier to fission product release Control Room indication of degraded performance of safety

• Unplanned rise in PC pressure H

under existing plant conditions. Containment Closure requirements are specified in Administrative systems Procedure 0.50.5, Outage Shutdown Safety.

• Secondary Containment area radiation

> 1000 mR/hr (EOP-5A Table 10)

5. Manual scram methods for EAL SA2.1 and EAL SS2.1 are the following: HA2.1 1 2 3 4 5 DEF HU2.1 1 2 3 4 5 DEF

• Reactor Scram push buttons

• Reactor Mode switch in SHUTDOWN 2

Fire or explosion resulting in EITHER: Fire in any Table H-1 area not extinguished within 15 min. of Hazards

• Manual or auto actuation of ARI Visible damage to any Table H-1 area containing safety Control Room notification or receipt of a valid Control Room

& 6. See Table F-1, Fission Product Barrier Matrix, for possible escalation above the Unusual Event due systems or components alarm due to fire (Note 3)

Other Fire or to RCS Leakage. OR HU2.2 1 2 3 4 5 DEF Condi- Explosion 7. If the equipment in the stated area was already inoperable, or out of service, before the event Control Room indication of degraded performance of safety tions occurred, then this EAL should not be declared as it will have no adverse impact on the ability of the systems Explosion within the Protected Area Affect- plant to safely operate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event.

ing

8. The time required to effect the backfeed is likely longer than the fifteen-minute interval. If off-normal HA3.1 1 2 3 4 5 DEF HU3.1 1 2 3 4 5 DEF Plant plant conditions have already established the backfeed, its power to the safety-related buses may be Safety considered an offsite power source. Toxic, corrosive, asphyxiant or flammable gases in amounts 3

Access to any Table H-1 area is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize that have or could affect normal plant operations operation of systems required to maintain safe operations or safely shut down the reactor (Note 7)

Hazardous 1 2 3 4 5 DEF HU3.2 Gas Recommendation by local, county or state officials to evacuate or shelter site personnel based on an offsite event HG4.1 1 2 3 4 5 DEF HS4.1 1 2 3 4 5 DEF HA4.1 1 2 3 4 5 DEF HU4.1 1 2 3 4 5 DEF A hostile action has occurred such that plant personnel are A hostile action is occurring or has occurred within the A hostile action is occurring or has occurred within the A security condition that does not involve a hostile action as 4 unable to operate equipment required to maintain safety functions Protected Area as reported by the Security Shift Supervisor Owner Controlled Area as reported by the Security Shift Supervisor reported by the Security Shift Supervisor OR OR OR A credible site-specific security threat notification Security A hostile action has caused failure of Spent Fuel Cooling A validated notification from NRC of an airliner attack threat OR Systems and imminent fuel damage is likely for a freshly within 30 min. of the site A validated notification from NRC providing information of an off-loaded reactor core in pool aircraft threat HS5.1 1 2 3 4 5 DEF HA5.1 1 2 3 4 5 DEF 5 None Control Room evacuation has been initiated AND Procedure 5.1ASD, Alternate Shutdown, or Procedure 5.4FIRE-S/D, Fire Induced Shutdown From Outside the None Control Control of the plant cannot be established within 15 min. Control Room, requires Control Room evacuation Room Evacuation HG6.1 1 2 3 4 5 DEF HS6.1 1 2 3 4 5 DEF HA6.1 1 2 3 4 5 DEF HU6.1 1 2 3 4 5 DEF Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that EITHER:

have occurred which involve EITHER: have occurred which involve EITHER: have occurred which involve EITHER: Events are in progress or have occurred which indicate a Actual or imminent substantial core degradation or melting An actual or likely major failures of plant functions An actual or potential substantial degradation of the potential degradation of the level of safety of the plant with potential for loss of containment integrity needed for protection of the public level of safety of the plant OR 6 OR Hostile action that results in an actual loss of physical control of the facility OR Hostile action that results in intentional damage or malicious acts; 1) toward site personnel or equipment OR A security event that involves probable life threatening risk to site personnel or damage to site equipment A security threat to facility protection has been initiated 1.

Notes The Emergency Director should not wait until the applicable time has elapsed, but should declare No releases of radioactive material requiring offsite response Judgment that could lead to the likely failure of or; 2) that prevent because of hostile action the event as soon as it is determined that the condition will likely exceed the applicable time.

or monitoring are expected unless further degradation of Releases can be reasonably expected to exceed EPA effective access to equipment needed for the protection safety systems occurs If dose assessment results are available, declaration should be based on dose assessment instead Protective Action Guideline exposure levels (1 Rem TEDE of the public Any releases are expected to be limited to small fractions of of radiation monitor values.

and 5 Rem thyroid CDE) beyond the site boundary the EPA Protective Action Guideline exposure levels beyond (See EAL AS1.2/AG1.2.) Do not delay declaration awaiting dose assessment results.

Any releases are not expected to result in exposure levels the site boundary 2. The Emergency Director should not wait until the applicable time has elapsed, but should declare which exceed EPA Protective Action Guideline exposure the event as soon as it is determined that the release duration has exceeded, or will likely exceed, levels (1 Rem TEDE and 5 Rem thyroid CDE) beyond the the applicable time. In the absence of data to the contrary, assume that the release duration has site boundary exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

3. The Emergency Director should not wait until the applicable time has elapsed, but should declare EU1.1 N/A the event as soon as it is determined that the condition will likely exceed the applicable time.

4. Containment Closure is the action taken to secure primary or secondary containment and its Damage to a loaded cask confinement boundary associated structures, systems, and components as a functional barrier to fission product release E

ISFSI None None None 5.

under existing plant conditions. Containment Closure requirements are specified in Administrative Procedure 0.50.5, Outage Shutdown Safety.

Manual scram methods for EAL SA2.1 and EAL SS2.1 are the following:

• Reactor Scram push buttons Prepared for NPPD by: Operations Support Services, Inc. - www.ossi-net.com (Rev. 8 4/2/10)

• Reactor Mode switch in SHUTDOWN

• Manual or auto actuation of ARI

6. See Table F-1, Fission Product Barrier Matrix, for possible escalation above the Unusual Event due to RCS Leakage.

7. If the equipment in the stated area was already inoperable, or out of service, before the event EAL Identifier occurred, then this EAL should not be declared as it will have no adverse impact on the ability of the plant to safely operate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event.

XXX.X 8. The time required to effect the backfeed is likely longer than the fifteen-minute interval. If off-normal Category (A, H, S, F, C, E) Sequential number within subcategory/classification plant conditions have already established the backfeed, its power to the safety-related buses may be considered an offsite power source.

Emergency classification (G, S, A, U) Subcategory number (1 if no subcategory)

Cooper Nuclear Station Modes: 1 Power Operation 2

Startup 3

Hot Shutdown Cold Shutdown 4 5 Refueling DEF Defueled Emergency Action Level Matrix EPIP 5.7.1 Attachment 4, Rev. 9 MODE 4, 5 or DEF

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT AG1.1 1 2 3 4 5 DEF AS1.1 1 2 3 4 5 DEF AA1.1 1 2 3 4 5 DEF AU1.1 1 2 3 4 5 DEF SG1.1 1 2 3 SS1.1 1 2 3 SA1.1 1 2 3 SU1.1 1 2 3 Loss of all offsite and all onsite AC power (Table S-3) to Loss of all offsite and all onsite AC power (Table S-3) to AC power capability to critical 4160V buses 1F and 1G Loss of all offsite AC power (Table S-3) to critical 4160V Any valid gaseous monitor reading > Table A-1 column GE Any valid gaseous monitor reading > Table A-1 column SAE Any valid gaseous monitor reading > Table A-1 column Any valid gaseous monitor reading > Table A-1 column UE for 15 min. (Note 1) for 15 min. (Note 1) Alert for 15 min. (Note 2) for 60 min. (Note 2) 1 critical 4160V buses 1F and 1G AND EITHER:

Restoration of at least one emergency bus critical 4160V buses 1F and 1G for 15 min. (Note 3) reduced to a single power source (Table S-3) for 15 min.

such that any additional single failure would result in loss of all AC power to critical buses (Note 3) buses 1F and 1G for 15 min. (Note 3)

AG1.2 AS1.2 AA1.2 1 2 3 4 5 DEF AU1.2 1 2 3 4 5 DEF Loss of in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely 1 2 3 4 5 DEF 1 2 3 4 5 DEF OR Power RPV level cannot be restored and maintained 1 Dose assessment using actual meteorology indicates doses

> 1 Rem TEDE or > 5 Rem thyroid CDE at or beyond the site boundary Dose assessment using actual meteorology indicates doses

> 0.1 Rem TEDE or > 0.5 Rem thyroid CDE at or beyond the site boundary Any valid liquid effluent monitor reading > Table A-1 column Alert for 15 min. (Note 2)

Any valid liquid effluent monitor reading > Table A-1 column UE for 60 min. (Note 2)

SG2.1

> -158 in. or cannot be determined 1 2 SS2.1 1 2 SA2.1 1 2 SU2.1 3 Offsite Rad Conditions Automatic and all manual scrams were not successful An automatic scram failed to shut down the reactor An automatic scram failed to shut down the reactor An unplanned sustained positive period observed on AND AND AND nuclear instrumentation AG1.3 1 2 3 4 5 DEF AS1.3 1 2 3 4 5 DEF AA1.3 1 2 3 4 5 DEF AU1.3 1 2 3 4 5 DEF Reactor power 3%

2 Manual actions taken at the reactor control console Manual actions taken at the reactor control console AND EITHER of the following exist or have occurred (Note 5) do not shut down the reactor as indicated (Note 5) successfully shut down the reactor as indicated Field survey results indicate closed window dose rates Field survey indicates closed window dose rate > 0.1 Rem/hr Confirmed sample analyses for gaseous or liquid releases Confirmed sample analyses for gaseous or liquid releases due to continued power generation: by reactor power 3% by reactor power < 3%

> 1 Rem/hr expected to continue for 60 min. at or beyond that is expected to continue for 60 min. at or beyond the site indicate concentrations or release rates > 200 x ODAM limits indicate concentrations or release rates > 2 x ODAM RPV level cannot be restored and maintained ATWS A

the site boundary (Note 1) boundary (Note 1) for 15 min. (Note 2) limits for 60 min. (Note 2) > -183 in. or cannot be determined OR OR Criticality OR Analyses of field survey samples indicate thyroid CDE Field survey sample analysis indicates thyroid CDE > 0.5 Rem Average torus water temperature and RPV pressure

> 5 Rem for 1 hr of inhalation at or beyond the site boundary for 1 hr of inhalation at or beyond the site boundary cannot be maintained within the Heat Capacity Temperature Limit (EOP/SAG Graph 7)

Abnorm. AA2.1 1 2 3 4 5 DEF AU2.1 1 2 3 4 5 DEF Rad Table A-1 Effluent Monitor Classification Thresholds Release

/ Rad Monitor GE SAE ALERT UE Damage to irradiated fuel OR loss of water level (uncovering irradiated fuel outside the RPV) that causes EITHER of the Unplanned water level drop in the reactor cavity or spent fuel pool as indicated by any of the following:

3 Inability to SU3.1 1 2 3 Effluent for 15 min. for 15 min. for 15 min. for 60 min. following:

• LI-86 (calibrated to 1001' elev.) None None None Plant is not brought to required operating mode within 2 ERP 3.50E+08 µCi/sec 3.50E+07 µCi/sec 2.80E+06 µCi/sec 2.24E+05 µCi/sec Valid RMA-RA-1 Fuel Pool Area Rad reading > 50 R/hr OR

• Spent fuel pool low level alarm

• Visual observation AND Reach Shutdown Conditions Technical Specifications LCO action statement time Onsite Rad Valid RMP-RM-452 A-D Rx Bldg Vent Exhaust Plenum Valid area radiation monitor reading rise on RMA-RA-1 or S

Conditions Hi-Hi alarm SS4.1 1 2 3 SA4.1 1 2 3 SU4.1 1 2 3 Rx Bldg Vent RMA-RA-2

& 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.45E+05 µCi/sec 8.48E+04 µCi/sec Loss of > approximately 75% of annunciators or indicators Unplanned loss of > approximately 75% of annunciators or Unplanned loss of > approximately 75% of annunciators or GASEOUS Spent Fuel Pool Events Turb Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.62E+05 µCi/sec 9.02E+04 µCi/sec AA2.2 1 2 3 4 5 DEF AU2.2 1 2 3 4 5 DEF System 4 None associated with safety systems on Control Room Panels 9-3, 9-4, 9-5, and C for 15 min. (Note 3) indicators associated with safety systems on Control Room Panels 9-3, 9-4, 9-5, and C for 15 min. (Note 3) indicators associated with safety systems on Control Room Panels 9-3, 9-4, 9-5, and C for 15 min. (Note 3)

A water level drop in the reactor refueling cavity or spent fuel Unplanned valid area radiation monitor reading or survey Malfunct. Inst.

AND AND EITHER:

pool that will result in irradiated fuel becoming uncovered results rise by a factor of 1,000 over normal levels* Any significant transient is in progress, Table S-1 Any significant transient is in progress, Table S-1 RW / ARW Bldg Vent 3.50E+07 µCi/sec 3.50E+06 µCi/sec 5.64E+05 µCi/sec 9.08E+04 µCi/sec AND OR

• Normal levels can be considered as the highest reading in the past 24 Compensatory indications are unavailable Compensatory indications are unavailable The lesser of: The lesser of: hours excluding the current peak value 200 x calculated 2 x calculated LIQUID SU5.1 1 2 3 5

Rad Waste Effluent ----- ----- alarm values alarm values AA3.1 1 2 3 4 5 DEF 3 OR monitor upscale*

OR monitor upscale* Dose rates > 15 mRem/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety Fuel Clad None None None SJAE monitor > 1.58E+3 mR/hr SU5.2 functions: 1 2 3 MCR/CAS Service Water Effluent ----- ----- 4.80E-04 µCi/cc 4.80E-06 µCi/cc Degradation Rad Main Control Room (RM-RA-20)

OR Coolant activity 4.0 µCi/gm dose equivalent I-131

• with effluent discharge not isolated CAS 6

SU6.1 1 2 3 HA1.1 1 2 3 4 5 DEF HU1.1 1 2 3 4 5 DEF None None None Unidentified or pressure boundary leakage > 10 gpm RCS OR Seismic event > 0.1g as indicated by SMA-3 Strong Motion Seismic event identified by any two of the following:

Table H-1 Safe Shutdown Areas Identified leakage > 30 gpm Accelograph or Alarm B-3/A-1 EMERGENCY SEISMIC HIGH

• SMA-3 Strong Motion Accelograph actuated or Alarm Leakage (Note 6)

LEVEL B-3/B-1 SEISMIC EVENT

• Reactor Building AND

• Earthquake felt in plant 7

SS7.1 1 2 3

• Control Building Earthquake confirmed by any of the following:

• National Earthquake Information Center

• Earthquake felt in plant None < 105 VDC bus voltage indications on all vital 125 VDC None None

• Service Water Pump Room

• National Earthquake Information Center Loss of buses (1A and 1B) for 15 min. (Note 3)

• Diesel Generator Building

• Control Room indication of degraded performance of DC Power systems required for the safe shutdown of the plant

• Cable Expansion Room SU8.1 1 2 3 8

HA1.2 1 2 3 4 5 DEF HU1.2 1 2 3 4 5 DEF Loss of all Table S-2 onsite (internal) communication Tornado striking or high winds > 100 mph resulting in EITHER: Tornado striking within Protected Area boundary None None None capability affecting the ability to perform routine operations Visible damage to any Table H-1 area structure containing OR Comm. OR safety systems or components Sustained high winds > 100 mph Loss of all Table S-2 offsite (external) communication OR methods affecting the ability to perform offsite notifications Control Room indication of degraded performance of safety systems HA1.3 1 2 3 4 5 DEF Main turbine failure-generated projectiles result in EITHER:

HU1.3 1 2 3 4 5 DEF Main turbine failure resulting in casing penetration or F

Fission FG1.1 1 2 Loss of any two barriers 3 FS1.1 1 2 3 Loss or potential loss of any two barriers (Table F-1)

FA1.1 1 2 3 Any loss or any potential loss of either Fuel Clad or RCS FU1.1 1 2 3 Any loss or any potential loss of Primary Containment 1 Visible damage to or penetration of any Table H-1 area structure containing safety systems or components OR damage to turbine or generator seals Product Barriers AND Loss or potential loss of third barrier (Table F-1)

(Table F-1) (Table F-1)

Natural or Control Room indication of degraded performance of Destructive safety systems Phenomena Table F-1 Fission Product Barrier Matrix HA1.4 1 2 3 4 5 DEF HU1.4 1 2 3 4 5 DEF Flooding in any Table H-1 area resulting in EITHER: Flooding in any Table H-1 area that has the potential to Fuel Clad Barrier Reactor Coolant System Barrier Primary Containment Barrier An electrical shock hazard that precludes access to operate affect safety-related equipment required by Technical or monitor safety equipment Specifications for the current operating mode OR Loss Potential Loss Loss Potential Loss Loss Potential Loss Control Room indication of degraded performance of safety A. RPV Level 1. PC flooding is required due to 5. RPV level cannot be 7. RPV level cannot be 22. PC Flooding required Notes systems None any of the following: restored and maintained restored and maintained None

1. The Emergency Director should not wait until the applicable time has elapsed, but should declare

• RPV water level cannot > -158 in. or cannot be > -158 in. or cannot be the event as soon as it is determined that the condition will likely exceed the applicable time. HA1.5 1 2 3 4 5 DEF HU1.5 1 2 3 4 5 DEF be restored and determined determined If dose assessment results are available, declaration should be based on dose assessment instead maintained of radiation monitor values. High river/forebay water level > 902' MSL High river/forebay water level > 899' MSL > -183 in.

(See EAL AS1.2/AG1.2.) Do not delay declaration awaiting dose assessment results. OR OR

• RPV water level cannot Low river/forebay level < 865' MSL Low river level/forebay < 870' MSL be restored and

2. The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, maintained the applicable time. In the absence of data to the contrary, assume that the release duration has HA1.6 1 2 3 4 5 DEF -209 in. and no core exceeded the applicable time if an ongoing release is detected and the release start time is unknown. spray subsystem flow

3. The Emergency Director should not wait until the applicable time has elapsed, but should declare Vehicle crash resulting in EITHER: can be restored and the event as soon as it is determined that the condition will likely exceed the applicable time. Visible damage to any Table H-1 area structure containing maintained 4,750 gpm

4. Containment Closure is the action taken to secure primary or secondary containment and its safety systems or components

• RPV water level cannot associated structures, systems, and components as a functional barrier to fission product release OR be determined and core under existing plant conditions. Containment Closure requirements are specified in Administrative Control Room indication of degraded performance of safety damage is occurring H Procedure 0.50.5, Outage Shutdown Safety. systems B. PC Pressure 8. PC pressure > 1.84 psig 16. PC pressure rise followed by a rapid unexplained 23. PC pressure > 56 psig and rising

5. Manual scram methods for EAL SA2.1 and EAL SS2.1 are the following: / Temperature due to RCS leakage drop in PC pressure

• Reactor Scram push buttons HA2.1 1 2 3 4 5 DEF HU2.1 1 2 3 4 5 DEF 24. Deflagration concentrations exist inside PC

• Reactor Mode switch in SHUTDOWN 17. PC pressure response not consistent with LOCA 6% H2 in drywell or torus 2

• Manual or auto actuation of ARI Fire or explosion resulting in EITHER: Fire in any Table H-1 area not extinguished within 15 min. of conditions Hazards Visible damage to any Table H-1 area containing safety Control Room notification or receipt of a valid Control Room (or cannot be determined)

6. See Table F-1, Fission Product Barrier Matrix, for possible escalation above the Unusual Event due

& to RCS Leakage. systems or components alarm due to fire (Note 3) None None None AND Other 5% O2 in drywell or torus Fire or 7. If the equipment in the stated area was already inoperable, or out of service, before the event OR HU2.2 1 2 3 4 5 DEF Condi- Control Room indication of degraded performance of safety (or cannot be determined)

Explosion occurred, then this EAL should not be declared as it will have no adverse impact on the ability of the tions plant to safely operate or safely shutdown beyond that already allowed by Technical Specifications at systems Explosion within the Protected Area

25. Average torus water temperature and RPV pressure Affect- the time of the event.

cannot be maintained within the Heat Capacity ing 8. The time required to effect the backfeed is likely longer than the fifteen-minute interval. If off-normal Temperature Limit (EOP/SAG Graph 7) plant conditions have already established the backfeed, its power to the safety-related buses may be HA3.1 1 2 3 4 5 DEF HU3.1 1 2 3 4 5 DEF Plant Safety considered an offsite power source.

Toxic, corrosive, asphyxiant or flammable gases in amounts 3

Access to any Table H-1 area is prohibited due to toxic, C. Isolation 9. Release pathway exists 13. RCS leakage > 50 gpm inside the 18. Failure of all valves in any one line to close corrosive, asphyxiant or flammable gases which jeopardize that have or could affect normal plant operations outside primary containment drywell AND operation of systems required to maintain safe operations or resulting from isolation failure Direct downstream pathway to the environment safely shut down the reactor (Note 7) in any of the following 14. Unisolable primary system exists after PC isolation signal Hazardous 1 2 3 4 5 DEF (excluding normal process discharge outside primary HU3.2 Gas system flowpaths from an containment as indicated by 19. Intentional PC venting per EOPs Recommendation by local, county or state officials to None None unisolable system): exceeding any secondary None evacuate or shelter site personnel based on an offsite event

• Main steam line containment Maximum Normal 20. Unisolable primary system discharge outside PC

• HPCI steam line Operating temperature or radiation as indicated by exceeding any secondary HG4.1 1 2 3 4 5 DEF HS4.1 1 2 3 4 5 DEF HA4.1 1 2 3 4 5 DEF HU4.1 1 2 3 4 5 DEF

• RCIC steam line value (EOP-5A Tables 9 and 10) containment Maximum Safe Operating

• RWCU temperature or radiation value (EOP-5A Tables 9 A hostile action has occurred such that plant personnel are A hostile action is occurring or has occurred within the A hostile action is occurring or has occurred within the A security condition that does not involve a hostile action as 4

• Feedwater and 10) unable to operate equipment required to maintain safety Protected Area as reported by the Security Shift Supervisor Owner Controlled Area as reported by the Security Shift reported by the Security Shift Supervisor functions Supervisor OR OR OR A credible site-specific security threat notification D. ERD 10. Emergency RPV Security None None None None None A hostile action has caused failure of Spent Fuel Cooling A validated notification from NRC of an airliner attack threat OR depressurization is required Systems and imminent fuel damage is likely for a freshly within 30 min. of the site A validated notification from NRC providing information of an off-loaded reactor core in pool aircraft threat E. Rad 2. Drywell radiation monitor 11. Drywell radiation monitor 26. Drywell radiation monitor (RMA-RM-40A/B)

HS5.1 1 2 3 4 5 DEF HA5.1 1 2 3 4 5 DEF (RMA-RM-40A/B) (RMA-RM-40A/B) > 5.00E+04 Rem/hr

> 2.50E+03 Rem/hr > 2.40E+02 Rem/hr 5 None Control Room evacuation has been initiated AND Procedure 5.1ASD, Alternate Shutdown, or Procedure 5.4FIRE-S/D, Fire Induced Shutdown From Outside the None

3. Primary coolant activity

> 300 µCi/gm dose None None None Control Control of the plant cannot be established within 15 min. Control Room, requires Control Room evacuation equivalent I-131 Room Evacuation F. Judgment 4. Any condition in the 6. Any condition in the 12. Any condition in the 15. Any condition in the opinion of 21. Any condition in the opinion of the Emergency 27. Any condition in the opinion of the Emergency HG6.1 1 2 3 4 5 DEF HS6.1 1 2 3 4 5 DEF HA6.1 1 2 3 4 5 DEF HU6.1 1 2 3 4 5 DEF opinion of the Emergency opinion of the opinion of the Emergency the Emergency Director that Director that indicates loss of the PC barrier Director that indicates potential loss of the PC Director that indicates loss Emergency Director Director that indicates loss indicates potential loss of the barrier Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the of the Fuel Clad barrier. that indicates of the RCS barrier RCS barrier Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director indicate that EITHER: Changes in the following potential loss of the have occurred which involve EITHER: have occurred which involve EITHER: have occurred which involve EITHER: Events are in progress or have occurred which indicate a parameters should also be Fuel Clad barrier Actual or imminent substantial core degradation or melting An actual or likely major failures of plant functions An actual or potential substantial degradation of the potential degradation of the level of safety of the plant considered here:

with potential for loss of containment integrity needed for protection of the public level of safety of the plant

• SJAE Radiation monitor.

OR

• DW Rad Monitor reading with 6 OR Hostile action that results in an actual loss of physical control of the facility OR Hostile action that results in intentional damage or malicious acts; 1) toward site personnel or equipment OR A security event that involves probable life threatening risk to site personnel or damage to site equipment A security threat to facility protection has been initiated normal RCS Leakage.

• Main Steam Line Radiation Monitor Readings.

No releases of radioactive material requiring offsite response Notes Judgment that could lead to the likely failure of or; 2) that prevent because of hostile action or monitoring are expected unless further degradation of effective access to equipment needed for the protection 1. The Emergency Director should not wait until the applicable time has elapsed, but should declare Releases can be reasonably expected to exceed EPA safety systems occurs of the public the event as soon as it is determined that the condition will likely exceed the applicable time.

Protective Action Guideline exposure levels (1 Rem TEDE Any releases are expected to be limited to small fractions of and 5 Rem thyroid CDE) beyond the site boundary the EPA Protective Action Guideline exposure levels beyond Table S-2 Communications Systems Table S-3 AC Power Sources If dose assessment results are available, declaration should be based on dose assessment instead Any releases are not expected to result in exposure levels the site boundary of radiation monitor values.

which exceed EPA Protective Action Guideline exposure Onsite Offsite Offsite (See EAL AS1.2/AG1.2.) Do not delay declaration awaiting dose assessment results.

System levels (1 Rem TEDE and 5 Rem thyroid CDE) beyond the Table S-1 Significant Transients (internal) (external)

• Startup Station Service Transformer 2. The Emergency Director should not wait until the applicable time has elapsed, but should declare site boundary

• Emergency Station Service the event as soon as it is determined that the release duration has exceeded, or will likely exceed, Station Intercom System Gaitronics X the applicable time. In the absence of data to the contrary, assume that the release duration has Reactor scram Transformer EU1.1 N/A Site UHF Radio Consoles X exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

Runback > 25% thermal power

• Backfeed 345 kv line through Main Power Transformer to the Normal 3. The Emergency Director should not wait until the applicable time has elapsed, but should declare Radio Paging System X the event as soon as it is determined that the condition will likely exceed the applicable time.

Damage to a loaded cask confinement boundary Electrical load rejection > 25% full electrical load Station Service Transformer (Note 8)

E Alternate Intercom X 4. Containment Closure is the action taken to secure primary or secondary containment and its ECCS injection None None None Sound Power System X Onsite associated structures, systems, and components as a functional barrier to fission product release Thermal power oscillations > 10% under existing plant conditions. Containment Closure requirements are specified in Administrative ISFSI CNS On-Site Cell Phone System X X

• DG-1 Procedure 0.50.5, Outage Shutdown Safety.

Prepared for NPPD by: Operations Support Services, Inc. - www.ossi-net.com (Rev. 8 4/2/10)

Telephone system (PBX) X X

• DG-2 5. Manual scram methods for EAL SA2.1 and EAL SS2.1 are the following:

Federal Telecommunications System (FTS 2001) X

• Main Generator

• Reactor Scram push buttons

• Reactor Mode switch in SHUTDOWN Local Telephones (C.O. Lines) X

• Manual or auto actuation of ARI CNS State Notification Telephones X 6. See Table F-1, Fission Product Barrier Matrix, for possible escalation above the Unusual Event due to RCS Leakage.

EAL Identifier 7. If the equipment in the stated area was already inoperable, or out of service, before the event occurred, then this EAL should not be declared as it will have no adverse impact on the ability of the XXX.X plant to safely operate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event.

Category (A, H, S, F, C, E) Sequential number within subcategory/classification 8. The time required to effect the backfeed is likely longer than the fifteen-minute interval. If off-normal plant conditions have already established the backfeed, its power to the safety-related buses may be Emergency classification (G, S, A, U) Subcategory number (1 if no subcategory) considered an offsite power source.

Cooper Nuclear Station Modes: Power Operation 1 2 Startup 3

Hot Shutdown 4

Cold Shutdown 5

Refueling DEF Defueled Emergency Action Level Matrix EPIP 5.7.1 Attachment 4, Rev. 9 MODE 1, 2 or 3

Liquid Effluent Monitoring D 3.3.1 D 3.3 INSTRUMENTATION D 3.3.1 Liquid Effluent Monitoring DLCO 3.3.1 The liquid effluent radiation monitoring instrumentation channels shown on Table D3.3.1-1 shall be OPERABLE with:

a. The minimum OPERABLE channel(s) in service.

b. The alarm and trip setpoints set to ensure that the limits of DLCO 3.1.1 are not exceeded.

APPLICABILITY: According to Table D3.3.1-1.

ACTIONS

---

NOTE---------------------------------------------------------------

Separate condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. Liquid effluent radiation monitoring A.1 Suspend liquid effluent Immediately instrumentation channel radiation release monitored alarm and trip setpoint by the inoperable channel.

less conservative than required. OR A.2 Declare channel inoperable. Immediately CNS ODAM D 3.3-1 10/10/01

Liquid Effluent Monitoring D 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One or more channels B.1 Enter the Condition Immediately inoperable. referenced in Table D3.3.1-1 for the channel.

AND B.2.1 Restore inoperable 31 days channel(s) to OPERABLE status.

OR B.2.2 In lieu of any other In accordance with report, explain in the the Radioactive Radioactive Effluent Effluent Release Release Report why the Report frequency.

instrument was not repaired in a timely manner.

CNS ODAM D 3.3-2 10/10/01

Liquid Effluent Monitoring D 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. As required by Required C.1 Analyze a minimum of 2 Prior to initiating a Action B.1 and referenced independent samples in release in Table D3.3.1-1. accordance with Table D3.1.1-1.

AND C.2 -------------NOTE----------

Determination Action and Verification Action will be performed by two separate technically qualified members of the Facility Staff.

Determine and Prior to initiating a independently verify the release release rate calculations and discharge valving.

D. As required by Required D.1 Collect and analyze a grab 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Action B.1 and referenced sample for gross beta in Table D3.3.1-1. radioactivity or gross AND gamma radioactivity (as applicable) at a lower limit of Once per detection 10-6 Ci/ml. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter E. As required by Required E.1 Estimate flow rate during 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Action B.1 and referenced actual release.

in Table D3.3.1-1. AND Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter CNS ODAM D 3.3-3 10/10/01

Liquid Effluent Monitoring D 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and F.1 Suspend liquid effluent Immediately associated Completion releases monitored by the Time for Condition C or E inoperable channel(s).

not met.

CNS ODAM D 3.3-4 10/10/01

Liquid Effluent Monitoring D 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.3.1.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DSR 3.3.1.2 Perform CHANNEL CHECK for each channel to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on any demonstrate OPERABILITY by verifying indication day on which of flow during periods of release. continuous, periodic, or batch releases are made DSR 3.3.1.3 Perform SOURCE CHECK. Completed prior to each release DSR 3.3.1.4 Perform SOURCE CHECK. 31 days DSR 3.3.1.5 Perform CHANNEL CALIBRATION 18 months DSR 3.3.1.6 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate automatic isolation of the pathway for instrument indication levels measured above the alarm/trip setpoint and circuit failure; and control room alarm annunciation for instrument indication levels measured above the alarm/trip setpoint, circuit failure and instrument indicating downscale failure.

CNS ODAM D 3.3-5 10/10/01

Liquid Effluent Monitoring D 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.3.1.7 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate control room alarm annunciation for instrument indication levels measured above the alarm/trip setpoint, circuit failure, instrument indicating downscale failure, and instrument controls not set in operate mode.

DSR 3.3.1.8 Perform CHANNEL FUNCTIONAL TEST. 184 days DSR 3.3.1.9 Perform LOGIC SYSTEM FUNCTIONAL TEST 184 days CNS ODAM D 3.3-6 10/10/01

Liquid Effluent Monitoring D 3.3.1 Table D3.3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation APPLICABILITY CONDITION OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE INSTRUMENT CONDITIONS OPERABLE ACTION B.1 REQUIREMENTS

1. Gross Beta or Gamma Radioactivity Monitors Providing Automatic Isolation

a. Liquid Radwaste (a) 1(b) C DSR 3.3.1.1 Effluent Line DSR 3.3.1.3 DSR 3.3.1.5 DSR 3.3.1.6 DSR 3.3.1.9

2. Gross Beta or Gamma Radioactivity Monitors Providing Alarm but not Providing Automatic Isolation

a. Service Water (a) D DSR 3.3.1.1 1

System Effluent Line DSR 3.3.1.4 DSR 3.3.1.5 DSR 3.3.1.7

3. Flow Rate Measurement Devices (a) E DSR 3.3.1.2 1

a. Liquid Radwaste DSR 3.3.1.5 Effluent Line DSR 3.3.1.8 (a) During releases via this pathway.

(b) Set to alarm and automatically close the waste discharge valve prior to exceeding the limits of DLCO 3.1.1.

CNS ODAM D 3.3-7 10/10/01

Gaseous Effluent Monitoring D 3.3.2 D 3.3 INSTRUMENTATION D 3.3.2 Gaseous Effluent Monitoring DLCO 3.3.2 The gaseous effluent radiation monitoring instrumentation channel(s) shown in Table D3.3.2-1 shall be OPERABLE with:

a. The minimum OPERABLE channel(s) in service.

b. The alarm and trip setpoints set to ensure that the limits of DLCO 3.2.1 are not exceeded.

APPLICABILITY: According to Table D3.3.2-1.

ACTIONS

---

NOTE-----------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. Gaseous effluent A.1 Suspend gaseous effluent Immediately radiation monitoring radiation release monitored instrumentation by inoperable channel.

channel alarm and trip setpoint less OR conservative than required. A.2 Declare channel inoperable. Immediately CNS ODAM D 3.3-8 03/28/06

Gaseous Effluent Monitoring D 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One or more channels B.1 Enter the Condition Immediately inoperable. referenced in Table D3.3.2-1 for the channel.

AND B.2.1 Restore inoperable 31 days channel(s) to OPERABLE status.

OR B.2.2 In lieu of any other report, In accordance with the explain in the Radioactive Radioactive Effluent Effluent Release Report Release Report why the instrument was not frequency repaired in a timely manner.

C. As required by C.1 Ensure the offgas delay Immediately Required Action B.1 system is not bypassed.

and referenced in Table D3.3.2-1. AND C.2 Ensure the Elevated Immediately Release Point Monitoring noble gas activity monitor is OPERABLE.

AND C.3 Restore inoperable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> channels to OPERABLE status.

D. Required Action and D.1 Be in MODE 2. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time for Condition C not met.

CNS ODAM D 3.3-9 03/28/06

Gaseous Effluent Monitoring D 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. As required by E.1 Estimate flowrate. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Required Action B.1 and referenced in AND Table D3.3.2-1.

Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter F. As required by F.1 Take grab samples. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Required Action B.1 and referenced in AND Table D3.3.2-1.

Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter AND F.2 Analyze for gross activity. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of sampling completion CNS ODAM D 3.3-10 03/28/06

Gaseous Effluent Monitoring D 3.3.2 ACTIONS G. As required by G.1.1 Verify one Function 2.a Immediately Required Action B.1 monitor OPERABLE.

and referenced in Table D3.3.2-1. AND G.1.2 Verify recombiner exhaust Immediately temperature change less than 10F over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND period.

Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

OR G.2.1 Verify one Function 2.a Immediately monitor OPERABLE.

AND G.2.2 Collect gas sample. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter AND G.2.3 Analyze gas sample and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from time of ensure within the limit of sampling completion DLCO 3.2.6.

OR G.3.1 Verify recombiner exhaust Immediately temperature change less than 10F over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND period.

Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

AND G.3.2 Collect gas sample. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

CNS ODAM D 3.3-11 03/28/06

Gaseous Effluent Monitoring D 3.3.2 ACTIONS AND G.3.3 Analyze gas sample and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from time of ensure within the limit of sampling completion.

DLCO 3.2.6.

H. Required Action and H.1 Discontinue operation of Immediately associated the augmented offgas Completion Time for treatment system.

Condition G not met.

I. As required by I.1 - - - - - - -NOTE- - - - - - -

Required Action B.1 When the primary and referenced in monitoring system is Table D3.3.2-1. inoperable and the backup system is in service, sampling may be discontinued for up to 30 minutes only for changing particulate filters and iodine cartridges.

Continuously collect 4 Hours samples with auxiliary sampling equipment as required in Table D3.2.3-1.

OR Immediately I.2.1 If auxiliary sampling equipment cannot be established within the specified completion time, enter the problem into the Corrective Action Program to evaluate particulate and iodine effluent releases.

AND In accordance with the Radioactive Effluent I.2.2 Report this event in the Release Report Radioactive Effluent Frequency Release Report.

CNS ODAM D 3.3-12 03/28/06

Gaseous Effluent Monitoring D 3.3.2 ACTIONS J. Required Action and J.1 Discontinue effluent Immediately associated releases via this pathway.

Completion Time for Condition E or F not met.

K. Function 1.a trip K.1 Close the offgas isolation Immediately capability not valve.

maintained.

AND AND K.2 Initiate reactor shutdown. Immediately Radiation level exceeds 1.0 ci/sec AND (prior to 30 min. delay line) for > 15 K.3 Be in MODE 4. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> consecutive minutes.

CNS ODAM D 3.3-13 03/28/06

Gaseous Effluent Monitoring D 3.3.2 SURVEILLANCE REQUIREMENTS NOTES

1. Refer to Table D3.3.2-1 to determine which DSRs apply for each instrument.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function is maintained.

SURVEILLANCE FREQUENCY DSR 3.3.2.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DSR 3.3.2.2 Perform CHANNEL CHECK. 7 days DSR 3.3.2.3 Perform SOURCE CHECK. 31 days DSR 3.3.2.4 Perform CHANNEL FUNCTIONAL TEST. 31 days DSR 3.3.2.5 Perform SOURCE CHECK. 92 days DSR 3.3.2.6 Perform CHANNEL CALIBRATION. The CHANNEL 92 days CALIBRATION shall include the use of a standard gas sample containing a percentage of hydrogen to verify accuracy of the monitoring channel in its operating range.

DSR 3.3.2.7 Perform CHANNEL FUNCTIONAL TEST. 92 days DSR 3.3.2.8 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if; the instrument indicates measured levels above the alarm/trip setpoint, circuit failure, instrument indicates a downscale failure, or instrument controls not set in operate mode.

DSR 3.3.2.9 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if; the instrument indicates measured levels above the alarm/trip setpoint or circuit failure.

CNS ODAM D 3.3-14 03/20/03

Gaseous Effluent Monitoring D 3.3.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.3.2.10 Perform CHANNEL CALIBRATION. For 18 months Function 1.a, the time delay setting for closure of the steam jet air ejector isolation valves shall 15 minutes and trip settings shall correspond to Technical Specification 3.7.5.

DSR 3.3.2.11 Perform CHANNEL FUNCTIONAL TEST. The 18 months CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if; the instrument indicates measured levels above the alarm/trip setpoint, circuit failure, instrument indicates a downscale failure, or instrument controls not set in operate mode.

DSR 3.3.2.12 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months CNS ODAM D 3.3-15 03/20/03

Gaseous Effluent Monitoring D 3.3.2 Table D3.3.2-1 (Page 1 of 3)

Radioactive Gaseous Effluent Monitoring Instrumentation APPLICABILITY CONDITION OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE INSTRUMENT CONDITIONS OPERABLE ACTION B.1 REQUIREMENTS

1. Steam Jet Air Ejector

a. Noble Gas Activity (a) 1(e) C DSR 3.3.2.1 Monitor DSR 3.3.2.3 DSR 3.3.2.8 DSR 3.3.2.10 DSR 3.3.2.11 DSR 3.3.2.12

b. Effluent System Flow (b) 1 E DSR 3.3.2.1 Rate Measuring Device DSR 3.3.2.7 DSR 3.3.2.10

2. Augmented Offgas Treatment System Explosive Gas Monitoring System

a. Hydrogen Monitor (c) 2 G DSR 3.3.2.1

( 2% monitor) DSR 3.3.2.4 DSR 3.3.2.6

3. Reactor Building Ventilation Monitoring System

a. Noble Gas Activity (b) (f) 1 F DSR 3.3.2.1 Monitor DSR 3.3.2.3 DSR 3.3.2.9 DSR 3.3.2.10

b. Iodine Sampler (b) (f) 1 I DSR 3.3.2.2 Cartridge

c. Particulate Sampler (b) (f) 1 I DSR 3.3.2.2 Filter

d. Effluent System Flow (b) 1 E DSR 3.3.2.1 Rate Measuring Device DSR 3.3.2.7 DSR 3.3.2.10

e. Sampler Flow Rate (b) 1 E DSR 3.3.2.1 Measuring Device DSR 3.3.2.7 DSR 3.3.2.10

f. Isolation Monitor (d) (d) (d) DSR 3.3.2.5 DSR 3.3.2.11 (continued)

(a) During operation of the steam jet air ejector (b) During releases via this pathway (c) During augmented offgas treatment system operation (d) See Technical Specification 3.3.6.2 (e) Second channel must either be OPERABLE or be in the tripped condition.

(f) A channel may be removed from service for up to 30 minutes for changing particulate filters or iodine cartridges or for low flow alarm check without entering Conditions or Required Actions.

CNS ODAM D 3.3-16 03/20/03

Gaseous Effluent Monitoring D 3.3.2 Table D3.3.2-1 (Page 1 of 3)

Radioactive Gaseous Effluent Monitoring Instrumentation APPLICABILITY CONDITION OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE INSTRUMENT CONDITION OPERABLE ACTION B.1 REQUIREMENTS

4. Elevated Release Point Monitoring System

a. Noble Gas Activity (b) (f) 1 F DSR 3.3.2.1 Monitor DSR 3.3.2.3 DSR 3.3.2.9 DSR 3.3.2.10

b. Iodine Sampler (b) (f) 1 I DSR 3.3.2.2 Cartridge

c. Particulate Sampler (b) (f) 1 I DSR 3.3.2.2 Filter

d. Effluent System Flow (b) 1 E DSR 3.3.2.1 Rate Measuring Device DSR 3.3.2.7 DSR 3.3.2.10

e. Sampler Flow Rate (b) 1 E DSR 3.3.2.1 Measuring Device DSR 3.3.2.7 DSR 3.3.2.10

5. Radwaste Building Ventilation Monitoring System

a. Noble Gas Activity (b) (f) 1 F DSR 3.3.2.1 Monitor DSR 3.3.2.3 DSR 3.3.2.9 DSR 3.3.2.10

b. Iodine Sampler (b) (f) 1 I DSR 3.3.2.2 Cartridge

c. Particulate Sampler (b) (f) 1 I DSR 3.3.2.2 Filter

d. Effluent System Flow (b) 1 E DSR 3.3.2.1 Rate Measuring Device DSR 3.3.2.7 DSR 3.3.2.10

e. Sampler Flow Rate (b) 1 E DSR 3.3.2.1 Measuring Device DSR 3.3.2.7 DSR 3.3.2.10

6. Turbine Building Ventilation Monitoring System

a. Noble Gas Activity (b) (f) 1 F DSR 3.3.2.1 Monitor DSR 3.3.2.3 DSR 3.3.2.9 DSR 3.3.2.10

b. Iodine Sampler (b) (f) 1 I DSR 3.3.2.2 Cartridge (continued)

(b) During releases via this pathway (f) A channel may be removed from service for up to 30 minutes for changing particulate filters or iodine cartridges or for low flow alarm check without entering Conditions or Required Actions.

CNS ODAM D 3.3-17 03/20/03

Gaseous Effluent Monitoring D 3.3.2 Table D3.3.2-1 (Page 1 of 3)

Radioactive Gaseous Effluent Monitoring Instrumentation APPLICABILITY CONDITION OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE INSTRUMENT CONDITION OPERABLE ACTION B.1 REQUIREMENTS

6. (continued)

c. Particulate Sampler (b) (f) 1 I DSR 3.3.2.2 Filter

d. Effluent System Flow (b) 1 E DSR 3.3.2.1 Rate Measuring Device DSR 3.3.2.7 DSR 3.3.2.10

e. Sampler Flow Rate (b) 1 E DSR 3.3.2.1 Measuring Device DSR 3.3.2.7 DSR 3.3.2.10

7. Multi Purpose Facility (MPF)

Building Ventilation Monitoring System

a. Iodine Sampler (b) (f) 1 I DSR 3.3.2.2 Cartridge

b. Particulate Sampler (b) (f) 1 I DSR 3.3.2.2 Filter

c. Effluent System Flow (b) 1 E DSR 3.3.2.1 Rate Measuring Device DSR 3.3.2.7 DSR 3.3.2.10

d. Sampler Flow Rate (b) 1 E DSR 3.3.2.1 Measuring Device DSR 3.3.2.7 DSR 3.3.2.10 (b) During releases via this pathway (f) A channel may be removed from service for up to 30 minutes for changing particulate filters or iodine cartridges or for low flow alarm check without entering Conditions or Required Actions.

CNS ODAM D 3.3-18 03/20/03