Text

Draft Written Exam and Outline
	ML23059A335
Person / Time
Site:	Millstone
Issue date:	02/22/2023
From:	O'Connor M; Dominion Energy Nuclear Connecticut
To:	Ray Lorson; NRC/RGN-I/DORS/OB
Shared Package
ML22300A088	List: ML23059A226; ML23059A335; ML23059A351;
References
	EPID L-2023-OLL-0036, 22-369B
	Download: ML23059A335 (1)
	v • d • e

Form 4.1-PWR Pressurized-Water Reactor Examination Outline Facility:

Millstone 2 K/A Catalog Rev. 3 Rev.

DRAFT Date of Exam:

2/22/2023 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total

1.

Emergency and Abnormal Plant Evolutions 1

3 3

18 2

2 1

1 2

1 1

8 Tier Totals 5

4 4

5 4

4 26

2.

Plant Systems 1

2 4

2 2

3 3

2 3

3 28 2

1 0

1 2

1 1

1 0

0 9

Tier Totals 3

4 3

3 4

4 3

3 3

37

3.

Generic Knowledge and Abilities Categories CO EC RC EM 6

CO EC RC EM 2

2 1

1

4. Theory Reactor Theory Thermodynamics 6

3 3

Notes: CO

=

EM =

Conduct of Operations; EC = Equipment Control; RC = Radiation Control; Emergency Procedures/Plan These systems/evolutions may be eliminated from the sample when Revision 2 of the K/A catalog is used to develop the sample plan.

These systems/evolutions are only included as part of the sample (as applicable to the facility) when Revision 2 of the K/A catalog is used to develop the sample plan.

ES-4.1-PWR PWR Examination Outline (Millstone 2)

Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO)

Item E/APE # / Name /

Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR Q#

1 (000007) (EPE 7; BW E02 & E10; CE E02) Reactor Trip, Stabilization, Recovery X

(CE02EA1.04) Ability to operate and/or monitor the following as they apply to (CE E02) STANDARD POST-TRIP ACTIONS AND REACTOR TRIP RECOVERY (CFR: 41.5 /

41.7 / 45.5 to 45.8): Rod control system 4.1 15 2

(000008) (APE 8)

Pressurizer Vapor Space Accident X

(000008AK1.05) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to (APE 8) PRESSURIZER VAPOR Space Accident (CFR: 41.5 / 41.7 / 45.7 / 45.8): Probable PZR steam space leakage paths other than PORV or code safety 3.6 9

3 (000009) (EPE 9)

Small Break LOCA X

(000009EK3.24) Knowledge of the reasons for the following responses and/or actions as they apply to (EPE 9) SMALL-Break LOCA (CFR: 41.5 / 41.10 / 45.6 / 45.13): ECCS throttling or termination criteria 3.9 34 4

(000011) (EPE 11)

Large Break LOCA X

(000011EK3.14) Knowledge of the reasons for the following responses and/or actions as they apply to (EPE 11) LARGE-Break LOCA (CFR: 41.5 / 41.10 / 45.6 / 45.13): RCP tripping requirement 4.0 22 5

(000015) (APE 15)

Reactor Coolant Pump Malfunctions X

(000015) (APE 15) Reactor Coolant Pump Malfunctions (G2.2.35) EQUIPMENT CONTROL: Ability to determine TS for mode of operation (CFR: 41.7 / 41.10 / 43.2 / 45.13) 3.6 74 6

(000022) (APE 22)

Loss of Reactor Coolant Makeup X

(000022AK3.04) Knowledge of the reasons for the following responses and/or actions as they apply to (APE 22) LOSS OF REACTOR Coolant Makeup (CFR: 41.5 / 41.10 / 45.6 /

45.13): Isolating letdown 3.7 14 7

(000025) (APE 25)

Loss of Residual Heat Removal System X

(000025AA1.20) Ability to operate and/or monitor the following as they apply to (APE 25) LOSS OF RESIDUAL Heat Removal System (CFR: 41.5 / 41.7 / 45.5 to 45.8):

ECCS 3.7 23 8

(000026) (APE 26)

Loss of Component Cooling Water X

(000026AK2.04) Knowledge of the relationship between (APE 26) LOSS OF Component Cooling Water and the following systems or components (CFR: 41.8 / 41.10 / 45.3):

SFPCS 3.6 7

9 (000027) (APE 27)

Pressurizer Pressure Control System Malfunction X

(000027AA2.14) Ability to determine and/or interpret the following as they apply to (APE 27) PRESSURIZER PRESSURE Control System Malfunction (CFR: 41.10 / 43.5

/ 45.13): RCP seal injection flow 3.0 40 10 (000029) (EPE 29)

Anticipated Transient Without Scram X

(000029EK1.03) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to (EPE 29) ANTICIPATED TRANSIENT WITHOUT SCRAM (ATWS) (CFR: 41.5 / 41.7 /

45.7 / 45.8): Addition of negative reactivity 4.2 66 11 (000038) (EPE 38)

Steam Generator Tube Rupture X

(000038EK2.20) Knowledge of the relationship between (EPE 38) STEAM GENERATOR Tube Rupture and the following systems or components (CFR: 41.8 / 41.10 / 45.3):

SGS 3.8 37 12 (000040) (APE 40; BW E05; CE E05; W E12) Steam Line Rupture - Excessive Heat Transfer X

(000040AK2.18) Knowledge of the relationship between (APE 40) STEAM LINE RUPTURE and the following systems or components (CFR: 41.8 / 41.10 / 45.3): PZR LCS 3.4 49 13 (000054) (APE 54; CE E06) Loss of Main Feedwater X

(000054) (APE 54; CE E06) Loss of Main Feedwater (G2.2.44) EQUIPMENT CONTROL: 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) 4.2 24

14 (000055) (EPE 55)

Station Blackout X

(000055) (EPE 55) Station Blackout (G2.4.20)

EMERGENCY PROCEDURES/PLAN: Knowledge of the operational implications of emergency and abnormal operating procedures warnings, cautions, and notes (CFR:

41.10 / 43.5 / 45.13) 3.8 30 15 (000056) (APE 56)

Loss of Offsite Power X

(000056AA2.88) Ability to determine and/or interpret the following as they apply to (APE 56) Loss of Offsite Power (CFR: 41.10 / 43.5 / 45.13): Conditions necessary for natural circulation 4.3 42 16 (000062) (APE 62)

Loss of Nuclear Service Water X

(000062AA1.05) Ability to operate and/or monitor the following as they apply to (APE 62) LOSS OF SERVICE WATER (CFR: 41.5 / 41.7 / 45.5 to 45.8): CCWS 3.4 45 17 (000065) (APE 65)

Loss of Instrument Air X

(000065AK1.02) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to (APE 65) LOSS OF Instrument Air (CFR: 41.5 / 41.7 / 45.7 / 45.8): Effects of water and/or particulate matter in instrument air lines (operating experience) 3.1 63 18 (000077) (APE 77)

Generator Voltage and Electric Grid Disturbances X

(000077AA2.01) Ability to determine and/or interpret the following as they apply to (APE 77) GENERATOR VOLTAGE AND ELECTRIC Grid Disturbances (CFR: 41.10 /

43.5 / 45.13): Operating point on the generator capability curve 3.7 44 K/A Category Totals:

3 3

Group Point Total:

18

ES-4.1-PWR PWR Examination Outline (Millstone 2)

Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO)

Item E/APE # / Name /

Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR Q#

19 (000001) (APE 1)

Continuous Rod Withdrawal X

(000001AA2.09) Ability to determine and/or interpret the following as they apply to (APE 1) CONTINUOUS Rod Withdrawal (CFR: 41.10 / 43.5 / 45.13): RPI 3.8 31 20 (000028) (APE 28)

Pressurizer (PZR)

Level Control Malfunction X

(000028AA1.04) Ability to operate and/or monitor the following as they apply to (APE 28) PRESSURIZER (PZR)

Level Control Malfunction (CFR: 41.5 / 41.7 / 45.5 to 45.8):

Regenerative heat exchanger and temperature limits 3.0 52 21 (000032) (APE 32)

Loss of Source Range Nuclear Instrumentation X

(000032) (APE 32) Loss of Source Range Nuclear Instrumentation (G2.2.42) EQUIPMENT CONTROL: Ability to recognize system parameters that are entry-level conditions for TS (CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3) 3.9 61 22 (000033) (APE 33)

Loss of Intermediate Range Nuclear Instrumentation X

(000033AK3.01) Knowledge of the reasons for the following responses and/or actions as they apply to (APE 33) LOSS OF INTERMEDIATE RANGE Nuclear Instrumentation (CFR:

41.5 / 41.10 / 45.6 / 45.13): Termination of startup following loss of intermediate range instrumentation 3.6 17 23 (000059) (APE 59)

Accidental Liquid Radwaste Release X

(000059AK1.04) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to (APE 59) ACCIDENTAL LIQUID Radwaste Release (CFR: 41.5 / 41.7 / 45.7 / 45.8): The relationship between background radiation intensity and the alarm setpoints on a rad 3.0 20 24 (000060) (APE 60)

Accidental Gaseous Radwaste Release X

(000060AA1.04) Ability to operate and/or monitor the following as they apply to (APE 60) ACCIDENTAL GASEOUS Radwaste Release (CFR: 41.5 / 41.7 / 45.5 to 45.8): Gaseous radwaste release isolation valve 3.7 8

25 (000061) (APE 61)

Area Radiation Monitoring System Alarms X

(000061AK1.02) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to (APE 61) AREA RADIATION MONITORING (ARM) System Alarms (CFR: 41.5 / 41.7 /

45.7 / 45.8): Adverse containment conditions 3.6 50 26 (000067) (APE 67)

Plant Fire On Site X

(000067AK2.11) Knowledge of the relationship between (APE 67) PLANT Fire On Site and the following systems or components (CFR: 41.8 / 41.10 / 45.3): Auxiliary building gas treatment system 2.9 47 K/A Category Totals:

2 1

1 2

1 1

Group Point Total:

8

ES-4.1-PWR PWR Examination Outline (Millstone 2)

Plant SystemsTier 2/Group 1 (RO)

Item System / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR Q#

27 (003) (SF4P RCP)

REACTOR COOLANT PUMP SYSTEM X

(003A2.06) Ability to (a) predict the impacts of the following on the (SF4P RCP) REACTOR COOLANT PUMP SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations (CFR:

41.5 / 45.6): CCWS malfunction 3.5 70 28 (003) (SF4P RCP)

REACTOR COOLANT PUMP SYSTEM X

(003) (SF4P RCP) REACTOR COOLANT PUMP SYSTEM (G2.4.32) EMERGENCY PROCEDURES/PLAN:

Knowledge of operator response to loss of annunciators (CFR: 41.10 /

43.5 / 45.13) 3.6 28 29 (004) (SF1; SF2 CVCS) CHEMICAL AND VOLUME CONTROL SYSTEM X

(004K6.01) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the (SF1; SF2 CVCS) CHEMICAL AND VOLUME CONTROL SYSTEM (CFR: 41.7 / 45.7):

Spray/heater combination in PZR to ensure uniform boron concentration 3.1 56 30 (005) (SF4P RHR)

RESIDUAL HEAT REMOVAL SYSTEM X

(005K1.15) Knowledge of the physical connections and/or cause and effect relationships between the (SF4P RHR)

RESIDUAL HEAT REMOVAL SYSTEM and the following systems (CFR: 41.2 to 41.9 /

45.7 to 45.8): IAS 3.3 16 31 (006) (SF2; SF3 ECCS)

EMERGENCY CORE COOLING SYSTEM X

(006A3.06) Ability to monitor automatic features of the (SF2; SF3 ECCS) EMERGENCY CORE COOLING SYSTEM, including (CFR: 41.7 / 45.7):

Valve lineups 3.7 67 32 (007) (SF5 PRTS)

PRESSURIZER RELIEF/QUENCH TANK SYSTEM X

(007) (SF5 PRTS)

PRESSURIZER RELIEF/QUENCH TANK SYSTEM (191002K1.17)

SENSORS AND DETECTORS (CFR: 41.7): (NULCEAR INSTRUMENTATION) Effects of core voiding on neutron detection 3.5 25 33 (008) (SF8 CCW)

COMPONENT COOLING WATER SYSTEM X

(008A4.12) Ability to manually operate and/or monitor the (SF8 CCW) COMPONENT COOLING WATER SYSTEM in the control room (CFR: 41.7 /

45.5 to 45.8): CRDM temperatures 2.9 5

34 (010) (SF3 PZR PCS)

PRESSURIZER PRESSURE CONTROL SYSTEM X

(010) (SF3 PZR PCS)

PRESSURIZER PRESSURE CONTROL SYSTEM (G2.2.7)

EQUIPMENT CONTROL:

Knowledge of the process for conducting infrequently performed tests or evolutions (CFR: 41.10 / 43.3 / 45.13) 2.9 58 35 (010) (SF3 PZR PCS)

PRESSURIZER PRESSURE CONTROL SYSTEM X

(010K2.05) Knowledge of electrical power supplies to the following (CFR: 41.7): (SF3 PZR PCS) PRESSURIZER PRESSURE CONTROL SYSTEM Pressure channels 3.3 33 36 (012) (SF7 RPS)

REACTOR PROTECTION SYSTEM X

(012K3.03) Knowledge of the effect that a loss or malfunction of the (SF7 RPS) REACTOR PROTECTION SYSTEM will have on the following systems or system parameters (CFR:

41.7 / 45.4): SDS 3.3 71 37 (013) (SF2 ESFAS)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM X

(013K5.12) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the (SF2 ESFAS)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (CFR: 41.5 / 45.3):

Reactor trip actuation 4.1 19 38 (022) (SF5 CCS)

CONTAINMENT COOLING SYSTEM X

(022K5.03) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the (SF5 CCS)

CONTAINMENT COOLING SYSTEM (CFR: 41.5 / 45.3):

Containment equipment subject to damage by high or low temperature, humidity, and pressure 3.1 43 39 (026) (SF5 CSS)

CONTAINMENT SPRAY SYSTEM X

(026A1.05) Ability to predict and/or monitor changes in parameters associated with operation of the (SF5 CSS)

CONTAINMENT SPRAY SYSTEM, including (CFR: 41.5

/ 45.5): Chemical additive tank level and concentration 3.0 12 40 (039) (SF4S MSS)

MAIN AND REHEAT STEAM SYSTEM X

(039A3.01) Ability to monitor automatic features of the (SF4S MSS) MAIN AND REHEAT STEAM SYSTEM, including (CFR: 41.7 / 45.7): Moisture separator reheater steam supply 2.8 10 41 (039) (SF4S MSS)

MAIN AND REHEAT STEAM SYSTEM X

(039K4.09) Knowledge of (SF4S MSS) MAIN AND REHEAT STEAM SYSTEM design features and/or interlocks that provide for the following (CFR: 41.7): Main steamline drains 2.6 62

42 (059) (SF4S MFW)

MAIN FEEDWATER SYSTEM X

(059K4.20) Knowledge of (SF4S MFW) MAIN FEEDWATER SYSTEM design features and/or interlocks that provide for the following (CFR:

41.7): Automatic feed pump recirculation flow 3.0 60 43 (061) (SF4S AFW)

AUXILIARY /

EMERGENCY FEEDWATER SYSTEM X

(061A2.07) Ability to (a) predict the impacts of the following on the (SF4S AFW)

AUXILIARY/EMERGENCY FEEDWATER SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations (CFR: 41.5 / 45.6): Air-operated valve, solenoid-operated valve, or motor-operated valve failure 4.0 6

44 (062) (SF6 ED AC)

AC ELECTRICAL DISTRIBUTION SYSTEM X

(062K1.05) Knowledge of the physical connections and/or cause and effect relationships between the (SF6 ED AC) AC ELECTRICAL DISTRIBUTION SYSTEM and the following systems (CFR: 41.2 to 41.9 /

45.7 to 45.8): Vital AC electrical instrument buses 4.2 75 45 (062) (SF6 ED AC)

AC ELECTRICAL DISTRIBUTION SYSTEM X

(062K2.02) Knowledge of electrical power supplies to the following (CFR: 41.7): (SF6 ED AC) AC ELECTRICAL DISTRIBUTION SYSTEM Breaker control power 3.5 27 46 (063) (SF6 ED DC)

DC ELECTRICAL DISTRIBUTION SYSTEM X

(063A2.05) Ability to (a) predict the impacts of the following on the (SF6 ED DC) DC ELECTRICAL DISTRIBUTION SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations (CFR: 41.5 / 45.6): Loss of all AC 4.2 26 47 (063) (SF6 ED DC)

DC ELECTRICAL DISTRIBUTION SYSTEM X

(063K2.02) Knowledge of electrical power supplies to the following (CFR: 41.7): (SF6 ED DC) DC ELECTRICAL DISTRIBUTION SYSTEM Battery room ventilation 2.5 64 48 (064) (SF6 EDG)

EMERGENCY DIESEL GENERATOR SYSTEM X

(064A4.03) Ability to manually operate and/or monitor the (SF6 EDG) EMERGENCY DIESEL GENERATOR SYSTEM in the control room (CFR: 41.7 / 45.5 to 45.8):

Synchroscope 3.8 69 49 (073) (SF7 PRM)

PROCESS RADIATION MONITORING SYSTEM X

(073A1.02) Ability to predict and/or monitor changes in parameters associated with operation of the (SF7 PRM)

PROCESS RADIATION MONITORING SYSTEM, including (CFR: 41.5 / 45.5):

Lights and alarms 3.2 39

50 (073) (SF7 PRM)

PROCESS RADIATION MONITORING SYSTEM X

(073K3.05) Knowledge of the effect that a loss or malfunction of the (SF7 PRM) PROCESS RADIATION MONITORING SYSTEM will have on the following systems or system parameters (CFR: 41.7 / 45.4):

S/GB 3.4 53 51 (076) (SF4S SW)

SERVICE WATER SYSTEM X

(076K2.09) Knowledge of electrical power supplies to the following (CFR: 41.7): (SF4S SW) SERVICE WATER SYSTEM Intake screens 2.6 59 52 (078) (SF8 IAS)

INSTRUMENT AIR SYSTEM X

(078A1.05) Ability to predict and/or monitor changes in parameters associated with operation of the (SF8 IAS)

INSTRUMENT AIR SYSTEM, including (CFR: 41.5 / 45.5):

Service air pressure 2.9 46 53 (078) (SF8 IAS)

INSTRUMENT AIR SYSTEM X

(078A4.02) Ability to manually operate and/or monitor the (SF8 IAS) INSTRUMENT AIR SYSTEM in the control room (CFR: 41.7 / 45.5 to 45.8):

Instrument air compressors 3.2 72 54 (103) (SF5 CNT)

CONTAINMENT SYSTEM X

(103K6.11) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the (SF5 CNT)

CONTAINMENT SYSTEM (CFR: 41.7 / 45.7): RCS 3.8 41 K/A Category Totals:

2 4

2 2

3 3

2 3

3 Group Point Total:

28

ES-4.1-PWR PWR Examination Outline (Millstone 2)

Plant SystemsTier 2/Group 2 (RO)

Item System / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR Q#

55 (002) (SF2; SF4P RCS) REACTOR COOLANT SYSTEM X

(002K1.02) Knowledge of the physical connections and/or cause and effect relationships between the (SF2; SF4P RCS)

REACTOR COOLANT SYSTEM and the following systems (CFR: 41.2 to 41.9 /

45.7 to 45.8): CRDS 3.5 3

56 (011) (SF2 PZR LCS)

PRESSURIZER LEVEL CONTROL SYSTEM X

(011K6.18) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the (SF2 PZR LCS) PRESSURIZER LEVEL CONTROL SYSTEM (CFR:

41.7 / 45.7): Reactor regulating system 3.3 57 57 (014) (SF1 RPI) ROD POSITION INDICATION SYSTEM X

(014A2.03) Ability to (a) predict the impacts of the following on the (SF1 RPI) ROD POSITION INDICATION SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations (CFR: 41.5 / 45.6): Dropped rod 4.0 55 58 (017) (SF7 ITM) IN CORE TEMPERATURE MONITOR SYSTEM X

(017K4.01) Knowledge of (SF7 ITM) IN CORE TEMPERATURE MONITOR SYSTEM design features and/or interlocks that provide for the following (CFR: 41.7):

Input to subcooling monitors 3.9 13 59 (035) (SF4P SG)

STEAM GENERATOR SYSTEM X

(035A1.01) Ability to predict and/or monitor changes in parameters associated with operation of the (SF4P SG)

STEAM GENERATOR SYSTEM, including (CFR: 41.5

/ 45.5): S/G level 4.1 1

60 (045) (SF4S MTG)

MAIN TURBINE GENERATOR SYSTEM X

(045K5.12) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the (SF4S MTG)

MAIN TURBINE GENERATOR SYSTEM (CFR: 41.5 / 45.3):

Role of field excitation in generator 2.8 54 61 (050) (SF9 CRV*)

CONTROL ROOM VENTILATION X

(050A3.02) Ability to monitor automatic features of the (SF9 CRV) CONTROL ROOM VENTILATION, including (CFR:

41.7 / 45.7): Initiation/failure of FPS 3.0 18

62 (068) (SF9 LRS)

LIQUID RADWASTE SYSTEM X

(068K4.03) Knowledge of (SF9 LRS) LIQUID RADWASTE SYSTEM design features and/or interlocks that provide for the following (CFR: 41.7):

Automatic system realignments 3.0 32 63 (075) (SF8 CW)

CIRCULATING WATER SYSTEM X

(075K3.06) Knowledge of the effect that a loss or malfunction of the (SF8 CW)

CIRCULATING WATER SYSTEM will have on the following systems or system parameters (CFR: 41.7 / 45.4):

Plant efficiency 3.2 21 K/A Category Totals:

1 0

1 2

1 1

1 0

0 Group Point Total:

9

Form 4.1-COMMON Common Examination Outline Generic Knowledge and Abilities Outline (Tier 3) (RO)

Category K/A #

Topic RO SRO-Only Item #

IR Q#

1.

Conduct of Operations G2.1.18 (G2.1.18) CONDUCT OF OPERATIONS: Ability to make accurate, clear, and concise logs, records, status boards, and reports (CFR: 41.10 / 45.12 / 45.13) 64 3.6 2

G2.1.20 (G2.1.20) CONDUCT OF OPERATIONS: Ability to interpret and execute procedure steps (CFR: 41.10 / 43.5 / 45.12) 65 4.6 4

Subtotal 2

2.

Equipment Control G2.2.6 (G2.2.6) EQUIPMENT CONTROL: Knowledge of the process for making changes to procedures (CFR: 41.10 / 43.3 / 45.13) 66 3

35 G2.2.40 (G2.2.40) EQUIPMENT CONTROL: Ability to apply TS with action statements of less than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (CFR: 41.10 / 43.2 /

43.5 / 45.3) 67 3.4 65 Subtotal 2

3.

Radiation Control G2.3.5 (G2.3.5) RADIATION CONTROL: Ability to use RMSs, such as fixed radiation monitors and alarms or personnel monitoring equipment (CFR: 41.11 / 41.12 / 43.4 / 45.9) 68 2.9 51 Subtotal 1

4.

Emergency Procedures /

Plan G2.4.46 (G2.4.46) EMERGENCY PROCEDURES/PLAN: Ability to verify that the alarms are consistent with the plant conditions (CFR:

41.10 / 43.5 / 45.3 / 45.12) 69 4.2 36 Subtotal 1

Tier 3 Point Total 6

Form 4.1-COMMON Common Examination Outline Theory (Tier 4) (RO)

Category K/A #

Topic RO Item #

IR Q#

Reactor Theory 192006 (192006K1.14) FISSION PRODUCT POISONS (CFR: 41.1):

Explain the methods and reasons for the reactor operator to compensate for the time-dependent behavior of xenon-135 concentration in the reactor 70 3.3 29 192007 (192007K1.05) FUEL DEPLETION AND BURNABLE POISONS (CFR: 41.1): Describe the effects of boration/dilution on reactivity during forced-flow and natural circulation conditions 71 3.2 11 192008 (192008K1.07) REACTOR OPERATIONAL PHYSICS (CFR:

41.1): (STARTUP AND APPROACH TO CRITICALITY) Calculate ECP using procedures and given plant procedures 72 3.6 38 Subtotal 3

Thermodynamics 193003 (193003K1.16) STEAM (CFR: 41.14): Define the following term: --

subcooled and compressed liquids 73 2.7 48 193004 (193004K1.15) THERMODYNAMIC PROCESS (CFR: 41.14):

(THROTTLING AND THE THROTTLING PROCESS) Determine the exit conditions for a throttling process based on the use of steam and/or water 74 2.8 68 193009 (193009K1.07) CORE THERMAL LIMITS (CFR: 41.14): Describe factors that affect peaking and hot channel factors 75 3.3 73 Subtotal 3

Tier 4 Point Total 6

Question #1 The plant is operating at 80% power.

The #2 ADV fails full open.

What is the immediate effect on #2 S/G level?

A.

Decreases B.

Increases C.

Remains the same D.

Oscillates

Question #1 RO SRO Tier # 2 Group # 2 K/A # 035 A1.01 Importance Rating: 4.1 K/A Statement: Ability to predict and/or monitor changes in parameters associated with operation of the STEAM GENERATOR SYSTEM, including: S/G level Proposed Answer: B Justification: On an up power transient (when the control valves are opened further) the flow rate increases, causing the DP across the separators to go up, increasing the pressure backup in the riser section. This forces the water in the down comer region to back up to produce sufficient driving head to push the higher flow into the tube bundle. SG water level is measured in the downcomer area, so the result of an up power transient is a swell in indicated level. (Another way of looking at the swell on an rising power transient is that the bubbles in the riser area get bigger, displacing water into the downcomer.)

Plausibility:

A. Plausible: a sudden up power also causes steam flow to exceed feed flow until the feed system catches up, resulting in lowering mass in the generators.

B. Correct.

C. Plausible: examinee may not recall shrink and swell, and may believe the feed system is able to keep up with the change in steam flow.

D. Plausible: examinee may not recall shrink and swell, and may believe the feed system needs to hunt (proportional controller) to maintain level after the step change in steam demand.

Technical Reference(s):

MILLSTONE 2 Main Steam System MSS-01-C PowerPoint, slide 88 Provided reference(s): None Learning Objective: 281742 ILT Explain the functional dependency between the Main Steam System, Steam Generator and Feedwater Control System as related to S/G Shrink and Swell.

Source: Bank #413037 Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics Comments:

Question #2 The unit experienced a reactor trip 3 days ago.

The trip was NOT logged in eSOMs.

How is the cause of the trip added to the narrative log?

A.

Edit the original log entry to add the new information.

B.

Delete the original log entry and make a new entry today with the new information.

C.

Make a late entry today with the additional information.

D.

Make a pen and ink change to the paper log that is retained by Nuclear Records.

Question #2 RO SRO Tier # 3 Group #

K/A # G2.1.18 Importance Rating: 3.6 K/A Statement: CONDUCT OF OPERATIONS: Ability to make accurate, clear, and concise logs, records, status boards, and reports Proposed Answer: C Justification: OP-AA-100, Conduct of Operations, states that during a plant transient, control of the plant takes precedence over log keeping, and to make a late entry if a required entry is discovered to be missed.

Plausibility:

A. Plausible, this would work if done on the same shift before the SM approves the logs for the shift.

B. Plausible, this would work in the same manner as A.

C. Correct.

D. Plausible, procedural guidance exists to do this to correct existing information in the logs at a later date.

Technical Reference(s):

OP-AA-100, Conduct of Operations Provided reference(s): None Learning Objective: Discuss the general requirements that apply to all formal logs (251424)

Source: New Kewaunee 2005 Audit Q67 Cognitive Level: Memory 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #3 The plant is operating at 100% power when all offsite lines are lost (LOOP).

What is the condition of:

1. the Reactor Coolant System (RCS)?

AND

2. the Control Element Drive Mechanisms (Coil Stacks)?

A.

1. Forced Circulation

2. De-energized B.

1. Forced Circulation

2. Energized C.

1. Natural Circulation

2. De-energized D.

1. Natural Circulation

2. Energized

Question #3 RO SRO Tier # 2 Group # 2 K/A # 002 K1.02 Importance Rating: 3.5 K/A Statement: Knowledge of the physical connections and/or cause and effect relationships between the REACTOR COOLANT SYSTEM (RCS) and the following systems:

CRDS Proposed Answer: C Justification: The loss of offsite power causes a turbine and reactor trip, as well as the loss of 6.9 KV busses, resulting in no forced flow. The reactor trip causes the Trip Circuit Breakers (TCBs) to open, de-energizing the Control Element Drive Mechanisms (CEDMs)

Plausibility:

A. Student may not recall that 6.9 KV busses are lost when offsite lines go away B. Student may not recall that 6.9 KV busses are lost when offsite lines go away, and may not recall that the CEDMs de-energize on a LOOP.

C. Correct D. Student may not recall that the CEDMs de-energize on a LOOP.

Technical Reference(s):

ARP 2590C-013 RC LO FLOW TRIP CH A

EOP 2528 Loss of Offsite Power/Loss of Forced Circulation, Entry Conditions Provided reference(s): None Learning Objective: Given a loss of normal power and a list of major 6900 VAC loads, assuming no operator action, determine the condition of the major components (deenergized/energized) as indicated on electrical drawing 30001, 30003 and 30042 (281517).

Source: New Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(3): Mechanical components and design features of the reactor primary system.

Comments: Changed part two of question to read CEDMs (Coil Stacks) to make question more specific. CEDS is too vague. DF 12/22

Question #4 The crew is performing a "Continuous Use" surveillance procedure.

The BOP is leading the evolution and has the master copy of the procedure in-hand.

The BOP directs a PEO to perform steps in the field.

The PEO has a field copy of the procedure in-hand.

In accordance with AD-AA-102, Procedure Use and Adherence, how will the field actions be documented in the master copy?

1. The PEO 1. required to have a copy of the procedure in hand.

AND

2. The 2. place-keep the master copy of the procedure.

A.

1. IS NOT

2. BOP can B.

1. IS NOT

2. PEO is required to C.

1. IS

2. BOP can D.

1. IS

2. PEO is required to

Question #4 RO SRO Tier # 3 Group #

K/A # G2.1.20 Importance Rating: 4.6 K/A Statement: CONDUCT OF OPERATIONS: Ability to interpret and execute procedure steps Proposed Answer: C Justification: AD-AA-102, Procedure Use and Adherence, requires step-by-step placekeeping when performing a Continuous Use procedure. Direct communications between individuals is permitted to maintain placekeeping. All persons performing actions from the procedure who are unable to view the master copy will have a copy of the procedure in hand.

Plausibility:

A. The PEO is required to have a copy of the procedure in hand. The candidate may think the requirements are the same as a Reference Level of Use procedure (i.e. The person leading the evolution is permitted to direct the procedure remotely).

B. The PEO is required to have a copy of the procedure in hand. The candidate may think the requirements are the same as a Reference Level of Use procedure (i.e. The person leading the evolution is permitted to direct the procedure remotely).

C. Correct D. The student may think the person performing the steps (field operator) is required to placekeep the master copy of the procedure as the steps are performed.

Technical Reference(s):

AD-AA-102, Procedure Use and Adherence Provided reference(s): None Learning Objective: 251853 STATE how steps in a Continuous Level of Use procedure are performed Source: Bank #369895 Cognitive Level: Memory 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #5 The plant is operating at 100% power.

Cooling water flow is lost to the A CEDM cooler.

Annunciator CEDM COOLER A FLOW LO (C-04, AA-11) alarms.

T8102, CEA COIL A AIR TEMP (PPC) is 105°F and rising.

What action is directed by ARP 2590C-079, CEDM COOLER A FLOW LO?

A.

Start the standby CEDM fan.

B.

Start the standby RBCCW pump.

C.

Close a Facility 1 CAR fan EMERG outlet valve.

D.

Trip the reactor and go to EOP 2525.

Question #5 RO SRO Tier # 2 Group # 1 K/A # 008 A4.12 Importance Rating: 2.9 K/A Statement: Ability to manually operate and/or monitor the COMPONENT COOLING WATER SYSTEM (CCW) in the control room: CRDM temperatures Proposed Answer: A Justification: CEDM cooling fans provide cooling to the CEDM coil stacks. Inadequate cooling causes the stacks to overheat, eventually leading to dropped CEAs. The alarm response (ARP 2590C-079) for low CEDM cooler cooling water flow directs starting the standby CEDM fan.

Plausibility:

A. Correct.

B. Alarms for low cooling water flow to all three fan coolers is an alternate indication of loss of Facility 1 RBCCW.

C. Removing RBCCW flow from one component causes flow to go up through remaining components on the same train.

D. ARP 2590C-079 requires a plant shutdown if all three CEDM cooling units fail.

Technical Reference(s):

ARP 2590C-079 CEDM COOLER A FLOW LO Provided reference(s): None Learning Objective: 281029Given a copy of OP 2313D, describe the operational requirements for the running of CEDM Cooling Fans, and give the reasons for those requirements, pertaining to the following issues:

A. Operation of CEDM Cooling Units following a reactor shutdown.

B. Maximum number of CEDM Cooling Units that should be operated at any one time.

Source: New Cognitive Level: Memory 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments: Added 1st bullet (flow is lost) based on validator feedback DF 12/7

Question #6 The plant is operating at 100% power.

The B AFW pump is out of service.

The plant experiences an Excess Steam Demand (ESD) on the #1 S/G.

Automatic Auxiliary Feed Water (AAFW) has actuated.

Subsequently, a loss of VA-10 occurs.

What actions are directed by EOP 2525 to isolate feed to the #1 Steam Generator?

A. Close 2-FW-44, AFW Header X-Tie, and stop the A MDAFP.

B. Swap the turbine driven AFW pump control power supply to Facility 2.

C. Place #1 OVERRIDE/MAN/START/RESET switch in 'Pull-To-Lock'.

D. Place the A AFW FRV controller (C-05) in MANUAL and closed.

Question #6 RO SRO Tier # 2 Group # 1 K/A # 061 A2.07 Importance Rating: 4.0 K/A Statement: Ability to (a) predict the impacts of the following on the AUXILIARY/EMERGENCY FEEDWATER SYSTEM (AFW) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Air-operated valve, solenoid-operated valve, or motor-operated valve failure Proposed Answer: A Justification: The loss of VA-10 will cause the #1 aux feed regulating valve to fail open. In order to maintain feedwater flow to #2 SG and prevent feeding #1 SG, EOP 2525 directs the operator to consider by shutting 2-FW-44 and feeding #2 SG with the TDAFW pump.

Plausibility:

A. Correct.

B. The control power supply to the turbine driven auxiliary feedwater pump is NOT affected; i.e., swapping power supplies will have NO impact on the turbine driven auxiliary feedwater pump or which S/G has feedwater flow. Plausible: Student may not recall that TDAFW power supply is DC (i.e. DV-20/DV-10) vice AC i.e. VA-20/VA-10).

C. Placing #1 OVERRIDE/MAN/START/RESET switch in 'Pull-To-Lock' will NOT prevent feeding #1 SG because the #1 aux feed regulating valve fails open on loss of VA-10.

Plausible: Student may not recall that FW-43A fails open on a loss of VA-10.

D. 2-FW-43A fails open on loss of power (VA-10) and CANNOT be closed from C-05.

Plausible: Student may not recall that FW-43A fails open on a loss of VA-10.

Technical Reference(s):

EOP 2525, Standard Post Trip Actions, step 7 RNO c2 Provided reference(s): None Learning Objective: Predict the impact of the following Auxiliary Feedwater malfunctions on the Steam Generator heat removal: (280922)

A) Loss of DC control power B) Loss of an Auxiliary Feedwater Pumps C) Failure of AFW flow control valves D) Loss of Vital 120 VAC Source: Modified Vision #451586 Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments: Clarification added to stem and one distractor based on validation feedback. SM 12/7

Question #7 The plant is operating at 100% power.

RBCCW cooling is lost to the Spent Fuel Pool Cooling (SFPC) heat exchangers.

Per AOP 2582, Loss of Spent Fuel Pool Cooling, how will the Spent Fuel Pool be cooled?

A.

Using the SDC heat exchanger and a Containment Spray pump.

B.

Using the SDC heat exchanger and a LPSI pump.

C.

Feed and Bleed using the RWST.

D.

Feed and Bleed using the CST.

Question #7 RO SRO Tier # 1 Group # 1 K/A # 026 AK2.04 Importance Rating: 3.6 K/A Statement: Knowledge of the relationship between Loss of Component Cooling Water and the following systems or components: SFPCS Proposed Answer: C Justification: AOP 2582, Loss of Spent Fuel Pool Cooling, has the operator feed and bleed from the RWST if cooling is lost to the SFP when SDC is NOT in service.

Plausibility:

A. The SDC Hx and Containment Spray pump can be used per AOP 2582 if the core is off-loaded. The student may forget the Spray option is lower modes only.

B. The SDC Hx and LPSI pump can be used per AOP 2582 in lower modes. The student may forget the LPSI option is lower modes only.

C. Correct D. The CST is used per AOP 2582 if the RWST is not available for feed and bleed. The student may think the CST has a longer Action statement so that may be preferable Technical Reference(s):

AOP 2582, Loss of Spent Fuel Pool Cooling Provided reference(s): None Learning Objective: 282249 Describe the effects on the SFP & Purification System for a loss of:

f) RBCCW Source: Bank 413558, 2011 NRC exam Q60 Cognitive Level: Memory 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #8 What indications are available in the Control Room that an accidental radioactive gaseous release (from the Gaseous Radwaste Systems normal discharge flowpath) is in progress?

1. Waste Gas Discharge Isolation Valves (GR-37.1 & GR-37.2) indicate closed on 1..

AND

2. Radiation Monitor 2. is rising.

A.

1. the PPC

2. RM-8168, Unit 2 Stack Rad Monitor (Kaman),

B.

1. the PPC

2. RM-8169, Millstone Stack Wide Range Rad Monitor, C.

1. RC-14

2. RM-8168, Unit 2 Stack Rad Monitor (Kaman),

D.

1. RC-14

2. RM-8169, Millstone Stack Wide Range Rad Monitor,

Question #8 RO SRO Tier # 1 Group # 2 K/A # 060 AA1.04 Importance Rating: 3.7 K/A Statement: Ability to operate and/or monitor the following as they apply to ACCIDENTAL GASEOUS Radwaste Release: Gaseous radwaste release isolation valve Proposed Answer: B Justification: The Gaseous Radwaste discharge isolation valve positions are only available in the Control Room from the PPC. Gaseous Radwaste discharges are released through the Site Stack due to its elevation providing better atmospheric dispersion. RM-8169 monitors the Site Stack. An accidental release such as relief valve GR-59 lifting would cause elevated readings on RM-8169 with the isolation valves closed.

Plausibility:

A. RM-8168 monitors the Unit 2 stack which is not connected to the Gaseous Radwaste system. The student may think the Unit 2 stack is the normal release point for Gaseous Radwaste discharges.

B. Correct C. The Gaseous Radwaste discharge isolation valves dont have indication on RC-14 (Radiation Monitors Panel). The student may think a HIGH alarm on RC-14 counts as indication of valve position. RM-8168 monitors the Unit 2 stack which is not connected to the Gaseous Radwaste system. The student may think the Unit 2 stack is the normal release point for Gaseous Radwaste discharges.

D. The Gaseous Radwaste discharge isolation valves dont have indication on RC-14 (Radiation Monitors Panel). The student may think a HIGH alarm on RC-14 counts as indication of valve position.

Technical Reference(s):

GRW-04-C R6, Gaseous Radwaste System Lesson Plan (pg 18)

PID_26021_02, PPC Gaseous Radwaste schematic (based on P&ID 25203-26021 sh.2)

Learning Objective: 287619 NLIT Describe the relationships between the Gaseous Radwaste system and the following: A) Site Stack Source: New Cognitive Level: Memory 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments: Added words to specify that the release is through the normal discharge flowpath DF12/7

Question #9 The plant is operating at 100% power.

The following conditions are noted:

Selected PZR Level L-110Y = 65% and stable.

PZR Level L-110X = 67% and rising.

Reactor Power = 100% and stable.

RCS Pressure = 2245 psia and lowering.

RCS TCOLD = 545 F and stable.

What would cause these indications?

A.

1 Main Control valve drifting open B.

1 Main Control valve drifting closed C.

L-110X reference leg leak D.

L-110X variable leg leak

Question #9 RO SRO Tier # 1 Group # 1 K/A # 008 AK1.05 Importance Rating: 3.6 K/A Statement: Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to PRESSURIZER VAPOR Space Accident:

Probable PZR steam space leakage paths other than PORV or code safety Proposed Answer: C Justification: A reference leg break on the non-selected Pressurizer level control channel would cause its indication to rise (dP between LREF and LVAR trending to zero) and RCS pressure to lower. RCS pressure lowers due to the leak. The selected Pressurizer level control channel would not see a noticeable change (in the short term) since the mass being lost is steam (no change in level)

Plausibility:

A. Plausible since a secondary transient that increases steam demand would cause RCS pressure to lower.

B. A secondary load reject transient would cause pressurizer level to increase. The student may see the non-selected channel increasing and focus on that.

C. Correct.

D. A variable leg break on the non-selected Pressurizer level control channel would cause its indication to lower.

Technical Reference(s):

GFES Chapter 7, Sensors, Instructor Guide. Pg 83 Provided reference(s): None Learning Objective: 283279 Given a set of plant conditions, determine AOP 2568, Reactor Coolant Leak, and AOP 2568, RCS Leak, MODE 4, 5, 6, and DEFUELED, applicability Source: New Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics Comments:

Question #10 The plant is performing AOP 2575, Rapid Downpower from 100% power.

Reactor power is 50% and lowering.

According to Step 18, Evaluate High Load Valve and Low Load Valve Function, what is the expected status of the MSR second stage reheat steam supply valves?

1. The Low Load valves are 1..

AND

2. The High Load valves are 2..

A. 1.

throttling

2.

fully open B. 1.

throttling

2.

fully closed C. 1.

fully open

2.

fully open D. 1.

fully open

2.

fully closed

Question #10 RO SRO Tier # 2 Group # 1 K/A # 039 A3.01 Importance Rating: 2.8 K/A Statement: Ability to monitor automatic features of the MAIN AND REHEAT STEAM SYSTEM (MSS), including: Moisture separator reheater steam supply Proposed Answer: B Justification: At normal 100% power, both the High Load and Low Load valves are fully open.

The low load valves ramp closed in response to turbine load decrease below 65%. The high load valves stroke closed around 65% turbine load.

Plausibility:

A. The student may forget when the valves start repositioning. This valve status would be observed at the transition point around 65% during a power increase.

B. Correct answer C. The student may forget when the valves start repositioning. This would be the normal valve status at greater than 65%.

D. Valve status that has been observed at the 65% transition point where the Low Load valves open when the High Load valves initially close.

Technical Reference(s):

AOP 2575, Rapid Downpower

OP 2317, Moisture Separator Reheaters Provided reference(s): None Learning Objective: ILT With a decreasing Turbine Load, describe the automatic valve operations that occur within the MSR for the following turbine loads. [281701]

a. 65% Turbine Load

b. 60% Turbine Load

c. 15% Turbine Load

d. 10% to 5% Turbine Load Source: Modified Bank [419248]

Cognitive Level: Comprehension 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #11 Compare adding boric acid to the RCS during natural circulation conditions versus forced circulation conditions.

1. Adding 10 gallons of boric acid during natural circulation requires 1. time to achieve complete mixing in the RCS when compared to adding 10 gallons of boric acid with RCPs running.

AND

2. Increasing RCS boron 1 ppm during natural circulation will cause a/an 2. change in reactivity for a given reactor coolant temperature when compared to forced circulation.

A.

1. more

2. smaller B.

1. more

2. equal C.

1. less

2. smaller D.

1. less

2. equal

Question #11 RO SRO Tier # 4 Group #

K/A # 192007 K1.05 Importance Rating: 3.2 K/A Statement: FUEL DEPLETION AND BURNABLE POISONS: Describe the effects of boration/dilution on reactivity during forced-flow and natural circulation conditions Proposed Answer: B, more; equal Justification: Natural circulation (NC) has significantly lower flowrates, therefore more time is required to achieve complete mixing of the boron added to the RCS. Once the boron is mixed in the RCS a 1 ppm change in the RCS boron concentration during NC operations will cause the same change in core reactivity that it would during forced circulation.

Plausibility:

A. Incorrect. The more is correct. Smaller is not correct but is plausible because the candidate could reason that since delta temperature is higher (and therefore average temperature is higher) in NC that boron has less of an effect.

B. Correct.

C. Incorrect. Both less and smaller are not correct. This is plausible because the candidate may not understand that greater flow enhances mixing and could reason that since delta temperature is higher (and therefore average temperature is higher) in NC that boron has less of an effect.

D. Incorrect. The less is incorrect and equal is correct. This is plausible because the candidate may not understand that greater flow enhances mixing.

Technical Reference(s): INPO PWR GFE Solution, P3364.

Provided reference(s): None Learning Objective:

Source: Bank 391099 (GFES Bank P3364)

Cognitive Level: Analysis 10CFR55.41(b)(1): Fundamentals of reactor theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects Comments:

Question #12 The plant is at 100% power when it experiences a Main Steam Line Break inside Containment.

The faulted Steam Generator has blown dry.

The faulted Steam Generator is isolated.

1. What system(s) are operating to prevent Containment pressure from exceeding 54 psig?

AND

2. What is the expected condition of Containment pressure after 15 minutes?

A.

1. Containment Spray system ONLY.

2. Lowering B.

1. Containment Air Recirculation system ONLY.

2. Stable C.

1. BOTH Containment Spray & Containment Air Recirculation systems.

2. Lowering D.

1. BOTH Containment Spray & Containment Air Recirculation systems.

2. Stable

Question #12 RO SRO Tier # 2 Group # 1 K/A # 026 A1.01 Importance Rating: 4.0 K/A Statement:) Ability to predict and/or monitor changes in parameters associated with operation of the CONTAINMENT SPRAY SYSTEM (CSS), including: Containment pressure Proposed Answer: C Justification: Both the Containment Spray (CS) and Containment Air Recirculation (CAR) coolers reduce the pressure in Containment. CS and CAR coolers remove heat and lowering pressure. Containment pressure is lowering since heat is no longer entering the Containment atmosphere because the faulted S/G has blown dry. Therefore, the heat removal by CS and CARs causes Containment pressure to lower.

Plausibility:

A. Incorrect. Containment Spray (CS) ONLY is not correct. Lowering is correct. Both CS and CAR coolers lower Containment pressure. This is plausible because the candidate could reason that only CS lowers pressure similar to the sprays in the Pressurizer, and not realize that CAR coolers lower temperature and therefore pressure.

B. Incorrect. Both CAR coolers ONLY and Stable are not correct. This is plausible because the candidate could reason that CS functions to strip radionuclides out of the Containment atmosphere to limit the radiation release and not lower pressure and that pressure would be stable for some time because CAR coolers will take time to lower pressure.

C. Correct.

D. Incorrect. Both CTMT Spray and CAR coolers is correct. This is plausible because the candidate could reason that pressure would be stable for some time before lowering.

Technical Reference(s):

Containment and Containment Systems Lesson Text CCS-00C, Rev. 10/03, page 5.

Containment Spray System Lesson Text CSS-00C, Rev. 7/0, pages 3 & 4.

FSAR Figure 14.8.2-1.

Provided reference(s): None Learning Objective: 281136. ILT Describe the relationships of the Containment Spray and Containment Air Recirculation Systems in limiting Containment Building temperature and pressure during the design-basis loss-of-coolant accident and the main steam line break accident.

Source: New Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics Comments:

Question #13

1. Where are the Inadequate Core Cooling (ICC) temperatures used for subcooling sensed?

AND

2. What combination of RCS pressure and temperature is used to determine Subcooling Margin?

A. 1. Above the active fuel region.

2. Highest pressure and lowest temperature.

B. 1. Above the active fuel region.

2. Lowest pressure and highest temperature.

C. 1. In the active fuel region.

2. Highest pressure and lowest temperature.

D. 1. In the active fuel region.

2. Lowest pressure and highest temperature.

Question #13 RO SRO Tier # 2 Group # 2 K/A # 017 K4.01 Importance Rating: 3.9 K/A Statement: Knowledge of IN CORE TEMPERATURE MONITOR SYSTEM design features and/or interlocks that provide for the following: Input to subcooling monitors Proposed Answer: B Justification: The temperature indication for the Inadequate Core Cooling (ICC) system are generated from the highest of the two available THOT and TCOLD signals, the second highest CET (CET High), the highest CET (CET Max), and the highest of the top three Unheated Sensors (UHTC Max). All these temperatures are above the core region (above the fuel). The input to the subcooling monitor uses the lowest pressure combined with the highest temperature to determine the minimum subcooled margin.

Plausibility:

A. Incorrect. Above the core region (above the fuel assemblies) is correct. The highest pressure and lowest temperature is plausible because the examinee may not understand the concept of subcooling and what would provide the least subcooled margin.

B. Correct.

C. Incorrect. In the core region is not correct. Highest pressure and lowest temperature is not correct. This is plausible because the examinee may think that temperature is monitored in the core region (around the fuel assemblies). A common misconception is that the Reactor Vessel Level Monitoring System (RVLMS) measures level in the core region when it actually monitors level and temperatures from just above the core to the top of the reactor vessel head. The highest pressure and lowest temperature is plausible because the examinee may not understand the concept of subcooling and what would provide the least subcooled margin.

D. Incorrect. In the core region is not correct. This is plausible because the examinee may think that temperature is monitored in the core region (around the fuel assemblies). A common misconception is that the RVLMS measures level in the core region when it actually monitors level and temperatures from just above the core to the top of the reactor vessel head.

Technical Reference(s): OP 2387G, Inadequate Core Cooling System Provided reference(s): None Learning Objective: 281500 - State the purpose of the Inadequate Core Cooling System Source: Bank 2022 Q58 Cognitive Level: Memory or Fundamental Knowledge 10CFR55.41(b)(7): Design, components, and functions of safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Question #14 The plant is operating at 100% power.

The charging line ruptures at the Containment penetration.

What is the condition of CVCS after conditions have stabilized?

1. Letdown flow will be 1..

AND

2. VCT level will be 2..

A.

1. at the limiter (28 gpm)

2. lowering B.

1. at the limiter (28 gpm)

2. constant C.

1. isolated (0 gpm)

2. lowering D.

1. isolated (0 gpm)

2. constant

Question #14 RO SRO Tier # 1 Group # 1 K/A # 022 AK3.04 Importance Rating: 3.7 K/A Statement: Knowledge of the reasons for the following responses and/or actions as they apply to LOSS OF REACTOR Coolant Makeup: Isolating letdown Proposed Answer: C Justification: The loss of charging flow through the regenerative heat exchanger will cause letdown to isolate on high temperature. Letdown will isolate very shortly (within 1 minute).

Per ARP 2590B-032, LETDOWN REGEN HX OUTLET TEMP HI (window B-8),

this occurs at 470°F outlet temperature. The charging pump will continue to pump water out the break causing VCT level to lower.

Plausibility:

A. Candidate may forget the L/D high temperature isolation and reason the leak will cause PZR level to lower with a resultant lowering of L/D flow to the limiter.

B. Candidate may forget the L/D high temperature isolation and reason the leak will cause PZR level to lower with a resultant lowering of L/D flow to the limiter. Candidate may also reason that the leak lowers VCT level to the point that Charging pumps will shift suction to the RWST (No VCT level change).

C. Correct.

D. Candidate may think the leak lowers VCT level to the point that Charging pumps will shift suction to the RWST (No VCT level change).

Technical Reference(s):

ARP 2590B-032, LETDOWN REGEN HX OUTLET TEMP HI (B-8)

Provided reference(s): None Learning Objective: Describe the effects, including a loss or malfunction, of the Chemical and Volume Control System on the following: (281166)

A) RCS B) Pressurizer Level C) Pressurizer Pressure (Aux Spray)

D) Safety Injection System E) Shutdown Cooling System Source: Bank 451513, 2001 NRC Q46 Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Question #15 EOP 2525, Standard Post-Trip Actions, verifies proper operation of the Control Element Assembly Drive System (CEADS) by observing what parameters?

A.

1. CEA position(s)

2. Reactor Power level B.

1. Reactor Power level

2. Control Valve position(s)

C.

1. Control Valve position(s)

2. MG set supply breaker position D.

1. MG set supply breaker position

2. CEA position(s)

Question #15 RO SRO Tier # 1 Group # 1 K/A # CE02 EA1.04 Importance Rating: 4.1 K/A Statement: Ability to operate and/or monitor the following as they apply to STANDARD POST-TRIP ACTIONS AND REACTOR TRIP RECOVERY: Rod control system Proposed Answer: A Justification: EOP 2525 directs the operator to verify all CEAs fully inserted, Reactor power lowering and a negative Start Up Rate to ensure the Reactivity Safety function is being met.

Plausibility:

A. Correct.

B. EOP 2525 has the operator verify Control Valve position to ensure the Turbine trip as part of Verifying Reactivity Control function met, the student may confuse Reactivity Control function with CEADS operation.

C. EOP 2525 has the operator verify Control Valve position to ensure the Turbine trip as part of Verifying Reactivity Control function met, the student may confuse Reactivity Control function with CEADS operation. The MG supply breakers are opened by the operator if the CEAs do NOT insert on a trip. The student may believe opening the MG set breakers is the same as verifying CEAs inserting D. The MG supply breakers are opened by the operator if the CEAs do NOT insert on a trip.

The student may believe opening the MG set breakers is the same as verifying CEAs inserting.

Technical Reference(s):

EOP 2525, Standard Post Trip Actions Provided reference(s): None Learning Objective: 283649 Outline the Instruction and Contingency Actions for the Immediate Actions in EOP 2525, Standard Post Trip Actions Source: New Cognitive Level: Memory 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics Comments:

Question #16 The plant is in MODE 5 on Shutdown Cooling (SDC).

How are the following affected by a loss of Instrument Air?

SI-306, SDC Total Flow Control Valve.

SI-657, SDC HX Flow Control Valve.

T351Y, SDC to RCS Temperature.

A.

SI-306 fails closed.

SI-657 fails open.

T351Y rises.

B.

SI-306 fails closed.

SI-657 fails open.

T351Y lowers.

C.

SI-306 fails open.

SI-657 fails closed.

T351Y rises.

D.

SI-306 fails open.

SI-657 fails closed.

T351Y lowers.

Question #16 RO SRO Tier # 2 Group # 1 K/A # 005 K1.15 Importance Rating: 3.3 K/A Statement: Knowledge of the physical connections and/or cause and effect relationships between the RESIDUAL HEAT REMOVAL SYSTEM (SDC) and the following systems: IAS Proposed Answer: C.

Justification: SI-306 fails open on a loss of air and SI-657 fails closed. This would cause a loss of SDC based on no flow going through the SDC HX. The result would be a heatup of the RCS as seen on T-351Y.

Plausibility:

A. Examinee may recall that a loss of IA results in a loss of SDC and an RCS temperature rise but not recall the valve failure positions.

B. Based on given valve failure positions (if true) RCS temperature would lower. Examinee may choose based a more conservative or fail-safe condition.

C. Correct response.

D. Examinee may recall the correct valve failure positions but not the impact to the RCS return temperature.

Technical Reference(s):

BKG AOP 2572, Loss of Shutdown Cooling Background Document Provided reference(s): None Learning Objective:

ILT Describe the effects on the Shutdown Cooling System of a loss or malfunction of the following: (MB-03179) [282215]

D) Instrument Air System Source: New Cognitive Level: Comprehensive 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Question #17 A reactor startup is in progress.

Regulating Group 4 CEAs are at 62 steps.

Channel B Wide Range (WR) Power circuit fails HIGH.

1. What is the effect on the Control Element Drive System (CEDS)

AND

2. Why would the reactor startup be terminated?

A.

1. CEA Withdrawal Prohibit (CWP) occurs.

2. MODE change would not be permitted with the CWP bypassed.

B.

1. CEA Motion Inhibit (CMI) occurs.

2. MODE change would not be permitted with the CMI bypassed.

C.

1. CEA Withdrawal Prohibit (CWP) occurs.

2. MODE change would not be permitted with only 3 WR Monitors.

D.

1. CEA Motion Inhibit (CMI) occurs.

2. MODE change would not be permitted with only 3 WR Monitors.

Question #17 RO SRO Tier # 1 Group # 2 K/A # 033 AK3.01 Importance Rating: 3.6 K/A Statement: Knowledge of the reasons for the following responses and/or actions as they apply to LOSS OF INTERMEDIATE RANGE Nuclear Instrumentation: Termination of startup following loss of intermediate range instrumentation Proposed Answer: B Justification: When the Wide Range (WR) Power circuit failed HIGH it cleared its associated level 2 bistable which arms the CMI from PDIL (Power Dependent Insertion Limit). Since CEAs are below the PDIL floor value of 72 steps on Group 4 a CMI occurs. Continuation of the reactor startup would require bypassing the CMI which would make the CMI inoperable. CMI is required to be OPERABLE in MODEs 1 & 2. The plant transitions from MODE 3 to MODE 2 when Group 4 CEAs are above 72 steps. It would not be permissible to enter MODE 2 with the conditions of the LCO not met and no clear path of conforming to the action requirements.

Plausibility:

A. Examine recalls that WR Power circuit failing HIGH will clear its associated level 2 bistable which arms the CWP. The CWP is initiated by the Reactor Protection System on a high-power condition but requires 2 of 4 High Power pretrips on Linear Power Range instruments not WR. RPS High Power trips can be bypassed but pretrips cannot be bypassed.

B. Correct Response.

C. Examine recalls that WR Power circuit failing HIGH will clear its associated level 1 bistable which arms the CWP. Examine recalls that a CWP is initiated by the Reactor Protection System on a high-power condition. Examinee recalls that WR instruments are required to be OPERABLE in MODE 3 but not the fact that only 2 are required.

D. Examinee correctly determines that a CEA Motion Inhibit (CMI) is inserted. Examinee recalls that WR instruments are required to be OPERABLE in MODE 3 but not the fact that only 2 are required.

Technical Reference(s):

ARP 2590C-110, CEA WITHDRAW PROHIBIT

ARP 2590C-136, CEA MOTION INHIBIT

UNIT 2 TECHNICAL SPECIFICATIONS Provided reference(s): None Learning Objective: 284047 LOIT Given any of the following CEA motion interlocks, identify the purpose of the interlock, the CEA motion which is inhibited, the conditions which result in the interlock, the CEAs or CEA group(s) affected, the CEA control mode affected, and the control room indication of the interlock:

A) CEA Motion Inhibit (CMI)

G) Power Dependent Insertion Limit (PDIL)

I) CEA Withdrawal Prohibit (CWP)

Source: Modified [415748]

Cognitive Level: Comprehension 10CFR55.41(b)(6): Design, components, and functions of reactivity control mechanisms and instrumentation.

Comments:

Question #18 Facility 2 (TWO) Control Room Air Conditioning (CRAC) is operating in NORMAL mode.

BOTH Facilities of Smoke Purge actuate.

How will the Facility 2 (TWO) CRAC system respond?

A. Column A.

B. Column B.

C. Column C.

D. Column D.

A B

C D

F-31B Z2 Exhaust Fan ON OFF ON OFF HV-497 Z2 Exhaust to Cable vault OPEN OPEN CLOSED CLOSED HV-496 Z2 Exhaust to Outside CLOSE CLOSE OPEN OPEN

Question #18 RO SRO Tier # 2 Group # 2 K/A # 050 A3.02 Importance Rating: 3.0 K/A Statement:) Ability to monitor automatic features of the CONTROL ROOM VENTILATION, including: Initiation/failure of FPS (fire protection system)

Proposed Answer: C Justification: When either Facility smoke detector located within the CRAC ductwork activates, the associated system goes into Smoke Purge (unless the system is operating in Recirculation mode). Smoke Purge secures the supply fan and closes the supply damper. The exhaust fan remains in service (or starts if secured) and the smoke-filled exhaust is isolated from the cable vault by closing BOTH Facility 1 & 2 dampers (dampers in series) and redirected to the outside environment by opening BOTH Facility 1 & 2 dampers (dampers in series).

Plausibility:

A. The student may think that smoke detection shifts CRAC system to RECIRC mode (the opposite happens: a RECIRC signal will override smoke purge).

B. The student may think that since Facility 2 exhaust secures to prevent spreading the fire.

C. Correct.

D. The student may think that since Facility 2 exhaust secures to prevent spreading the fire..

Technical Reference(s):

ARP 2590I, ZONE 2, CONTROL ROOM DUCT Z2 Provided reference(s): None Learning Objective: 281107 As given in CRA-00-C:

1.Identify the condition which automatically initiates Control Room Smoke Purge Source: Modified 412342 Cognitive Level: Comprehension 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments: Changed stem showing both facilities of smoke purge actuate to be more operationally valid. DF 12/7

Question #19 The plant experienced a Main Turbine trip at 100% power.

The Diverse Scram System (DSS) trips the reactor.

What Reactor Protection System (RPS) trips failed?

A.

S/G High Level and Turbine Trip.

B.

TM/LP and S/G Low Level.

C.

Turbine Trip and High RCS Pressure.

D.

High RCS Pressure and TM/LP.

Question #19 RO SRO Tier # 2 Group # 1 K/A # 013 K5.12 Importance Rating: 4.1 K/A Statement: Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (ESFAS): Reactor trip actuation Proposed Answer: C.

Justification: As part of the ESFAS the DSS is a redundant automatic reactor trip on high pressurizer pressure (2400 psia) that is electrically independent of the Reactor Protection System (RPS). If 2 of 4 pressurizer pressure channels exceed 2400 psia the DSS system will open both CEDM MG Set output contactors. The RPS High RCS Pressure trip (2397 psia) is the credited trip for a load reject in the accident analysis.

One of the RPS trips is a Turbine Trip (Low EHC pressure at Control Valve).

Plausibility:

A. The RPS Turbine Trip did fail. The student may think the S/G High Level trip is a RPS Trip not a Turbine Trip.

B. RPS S/G Low Level Trip failure is plausible due to the expected S/G level shrink. The student may think a turbine trip causes RCS pressure to lower causing a TM/LP trip.

C. Correct response.

D. The RPS High RCS Pressure trip (2397 psia) did fail. The student may think a turbine trip causes RCS pressure to lower causing a TM/LP trip.

Technical Reference(s):

ARP 2590C-017, PZR HI PRES TRIP CH A

ARP 2590C-033, TURBINE TRIP CH A

ARP 2590C-101, DIVERSE RX TRIP ACTUATED Provided reference(s): None Learning Objective: 281380 Describe each of the 11 Engineered Safety Features Actuation System (ESAS) actuations or trips with respect to:

A. Conditions sensed and setpoints B. Purpose C. Functions Source: Bank #413446 Cognitive Level: Comprehension 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

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

Comments: Changed D to remove Low S/G Level and add TM/LP because S/G low level is correct and adding TM/LP balances answers DF 12/8

Question #20 The plant is operating at 100% power.

An Aerated Liquid Waste discharge is in progress.

Unit 3 is operating at 100% power.

The A Circulating Water (CW) pump trips.

AOP 2517, Circulating Water Malfunctions, is entered.

1. What (if any) action is required regarding the waste discharge?

AND

2. Why?

A.

1.

Secure the discharge.

2.

Plant conditions dont meet the Discharge Permit.

B.

1.

Secure the discharge.

2.

Unit 2 CW flow is now less than 400,000 gpm.

C.

1.

No action required.

2.

Unit 3 CW flow makes up for the lost pump flow.

D.

1.

No action required.

2.

Unit 2 CW flow is still greater than 100,000 gpm.

Question #20 RO SRO Tier # 1 Group # 2 K/A # 059 AK1.06 Importance Rating: 3.2 K/A Statement: Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to ACCIDENTAL LIQUID Radwaste Release: Loss of the CWS during discharge Proposed Answer: A Justification: A loss of a Circulating Water pump at 100% requires securing any liquid radioactive waste discharges in progress. This is documented in AOP 2517, Circulating Water Malfunctions. The basis is that the discharge no longer meets the specifics described in the Discharge Permit for the discharge in progress.

Plausibility:

A. Correct.

B. The candidate may think 400,000 gpm is required to discharge to Long Island Sound.

C. The candidate may recall that Unit 3 Circulating Water flow is sufficient to allow proper mixing of the discharged liquid. SP 2617A allows discharging using Unit 3 flow (>100,000 gpm).

D. The candidate may recall that SP 2617 only requires 100,000 gpm to conduct a discharge.

SP 2617 also does not have provisions to secure a discharge if a Circulator is lost, just minimum flow requirements Technical Reference(s):

AOP 2517, Circulating Water Malfunctions Provided reference(s): None Learning Objective:287869 Describe the relationship between radioactive liquid waste discharges and the following:

A) Circulating Water System Source: New Cognitive Level: Memory 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics Comments:

Question #21 The plant is operating at 100% power.

Intake temperature is 65 F.

The 'B' Circulating Water pump trips off-line.

1. Plant efficiency will 1. due to this casualty.

AND

2. The plant efficiency changes are due to 2. increasing.

A.

1. increase

2. condensate depression.

B.

1. increase

2. condenser backpressure.

C.

1. decrease.

2. condensate depression.

D.

1. decrease.

2. condenser backpressure.

Question #21 RO SRO Tier # 2 Group # 2 K/A # 075 K3.06 Importance Rating: 3.2 K/A Statement: Knowledge of the effect that a loss or malfunction of the CIRCULATING WATER SYSTEM will have on the following systems or system parameters: Plant efficiency Proposed Answer: D Justification: With intake temperature at 65 F, a loss of a circulator will cause condenser backpressure to increase due to losing the ability to condense all the steam going through the turbine. The increased backpressure will reduce the work being done by the turbine (higher condenser pressure/temperature higher enthalpy in condenser smaller enthalpy drop across turbine stage of steam cycle less work performed by turbine) while the heat being produced by the S/Gs will remain constant; thus, plant efficiency decreases.

Plausibility:

A. The student may deduce efficiency increases because there is less subcooling in the hotwell (condensate depression). Condensate depression would decrease on a loss of a Circulator.

B. The student may deduce efficiency increases because there is less subcooling in the hotwell (condensate depression). This is not the case due to elevated intake temperature.

C. Condensate depression would decrease on a loss of a Circulator.

D. Correct.

Technical Reference(s):

GFES Chapter5, Thermodynamic cycles, pg 36 Provided reference(s): None Learning Objective: GFES Explain how changes in secondary system parameters affect plant efficiency.

Source: New Cognitive Level: Comprehension 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #22 The trip two/leave two RCP trip strategy when a SIAS occurs during a LOCA is meant to reduce A.

RCP seal wear and subsequent seal failure.

B.

the amount of heat being added to the RCS.

C.

pressure in the cold legs to enhance SI flow.

D.

the amount of water mass inventory lost through the break.

Question #22 RO SRO Tier # 1 Group # 1 K/A # 011 EK3.14 Importance Rating: 4.0 K/A Statement: Knowledge of the reasons for the following responses and/or actions as they apply to LARGE-Break LOCA: RCP tripping requirement Proposed Answer: D Justification: During a Large Break LOCA, 2 RCPs are secured when a SIAS signal is processed.

When RCS pressure lowers to below the NPSH of the RCPs, the remaining 2 RCPs are secured. The pumps are secured to reduce the amount of mass being pumped out the break, allowing the maximum amount of inventory to remain within the RCS for core cooling..

Plausibility:

A. Plausible since extended pump run time at pressures below NPSH will adversely affect the RCP seals.

B. Plausible because the RCPs are secured to reduce heat input on a Loss of All Feedwater.

The student might think reducing heat input from the RCPs overrides their heat removal capabilities.

C. If the RCPs are running with good NPSH, their heat removal capability is better than the SI system. The student may think SI flow has superior heat removal capability due to its lower temperature and is preferred over RCPs D. Correct Technical Reference(s):

Millstone Unit 2 EOP Technical Guide, EOP 2532 Loss of Coolant Accident Provided reference(s): None Learning Objective: 283789 Outline and explain the bases for the major actions in EOP 2532, Loss of Coolant Accident Source: Bank #453437 NRC 2001 Q6 Cognitive Level: Memory 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments: Clarified in the stem that it is asking about Trip 2/Leave 2. SM 12/7

Question #23 The plant is in MODE 5.

RCS is in Reduced Inventory Operations (RIO).

Draindown is in progress IAW OP 2301E, Draining the RCS (ICCE).

The following plant conditions exist:

o A LPSI pump running.

o Total Shutdown Cooling Flow F306 is 1600 GPM.

o RCS level by L122, Narrow Range RCS Level, is +0.5 inches.

As the draindown continues, the following occur:

o annunciator LPSI PUMP A SUCTION PRESSURE LO (C-01, A-8) alarms.

o 'A' LPSI Pump current starts oscillating.

1. What is the status of the A LPSI pump?

AND

2. Why?

A.

1. Cavitating.

2. Vortexing in Loop 2 Hot Leg.

B.

1. Operating at shutoff head.

2. Vortexing in Loop 2 Hot Leg.

C.

1. Cavitating.

2. SI-306, SDC SYS TOTAL FLOW VALVE, is full open.

D.

1. Operating at shutoff head.

2. SI-306, SDC SYS TOTAL FLOW VALVE, is full closed.

Question #23 RO SRO Tier # 1 Group # 1 K/A #025 AA1.20 Importance Rating: 3.7 K/A Statement: Ability to operate and/or monitor the following as they apply to LOSS OF RESIDUAL Heat Removal System: ECCS Proposed Answer: A Justification: Precaution 3.10 in OP 2301E says to limit SDC flow to 1600 GPM in RIO to avoid vortexing. The caution prior to step 4.2.10 of OP 2301E says risk of vortexing increases significantly when level is reduced below hot leg centerline.

Plausibility:

A. Correct.

B. Pump operating at shutoff head would indicate steady current and static head at pump suction. Plausible: an applicant may assume that vortexing will cause pump to operate at shutoff head conditions.

C. SI-306 failing closed would tend to reduce flow rate, thereby reducing the chance of cavitation. SI-306 failing open would have no effect on flowrate because the LPSI Loop Injection Valves were previously throttled to limit total flow to less than or equal to 1600 gpm. Plausible: an applicant may think that SI-306 failing open in this situation would result in a high flow.

D. Decay heat load requires flow through SDC HX via SI-657. System flow may be reduced by failure closed of SI-306. However, flow would continue through the HX, ensuring pump does not run at shutoff head. Plausible: an applicant may assume that closure of SI-306 will force pump to run at shutoff head conditions.

Technical Reference(s):

OP 2301E, Draining the RCS (ICCE)

Provided reference(s): None Learning Objective: Given a set of plant conditions concerning a Loss of Shutdown Cooling, determine if a LPSI pump should be started. (283361)

Source: Bank 451655, NRC 2005 Q7 Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Question #24 The plant is performing a downpower in accordance with OP 2204, Load Changes.

Power is 65% and lowering.

Both SGFPs are in service.

Annunciator SGFP A TURBINE TRIP (C05, D-4) annunciates.

A SGFP speed is 4100 RPM and stable.

Both Steam Generators are 70% and stable.

1. What is the status of the A SGFP?

AND

2. What is the required response?

A.

1. Running.

2. Trip the reactor and go to EOP 2525, Standard Post Trip Actions.

B.

1. Running.

2. Troubleshoot the pressure switch that drives the annunciator.

C.

1. Tripped.

2. Trip the reactor and go to EOP 2525, Standard Post Trip Actions.

D.

1. Tripped.

2. Troubleshoot the pressure switch that drives the annunciator.

Question #24 RO SRO Tier # 1 Group # 1 K/A # E06 G2.2.44 Importance Rating: 4.2 K/A Statement: Loss of Main Feedwater EQUIPMENT CONTROL: 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 Proposed Answer: B Justification: IAW ARP 2590D-013, 2 of 3 trip header pressure switches (any 2 of PS-15, 16 or 17) open will trip the A SGFP. PS-16 drives the alarm. If it is the only switch to open, the pump will not trip. The ARP directs requesting I&C troubleshoot PS-16 if the alarm occurs with no pump trip. The student should recognize, through diverse indications (stable A SGFP speed, stable SG water levels), that the pump has not tripped.

Plausibility:

A. Plausible: The student may believe the alarm response does not give leeway to consider other indications that the pump is running, and the student may believe that one SGFP is inadequate at 65% power.

B. Correct.

C. Plausible: The student may not recognize that the A SGFP is running, and the student may believe that one SGFP is inadequate at 65% power.

D. Plausible: The student may not recognize that the A SGFP is running.

Technical Reference(s):

ARP 2590D-013, SGFP A TURBINE TRIP (C05 A-4)

OP 2204, Load Changes Provided reference(s): None Learning Objective: Given a set of plant conditions and any input failure with the FWCS in automatic mode of control:

1. Predict the response of the FWCS to the failure and

2. Identify the manual manipulations required to stabilize the FWCS in accordance with OP 2585.

Source: New Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments: Rewrote distractors to make question more straightforward.

Question #25 The plant is in MODE 3 with source range monitors indicating 20 cps and stable.

A LOCA occurs.

RVLMS = 19% and lowering.

Due to inadequate Safety injection flow two-phase natural circulation flow requirements are NOT satisfied.

If voiding occurs in the reactor vessel downcomer,

1. Count rate will 1.

2. Because of 2.

A.

1. increase

2. more neutron leakage is occurring B.

1. decrease

2. less neutron leakage is occurring C.

1. increase

2. boron exiting the core D.

1. decrease

2. boron entering the core

Question #25 RO SRO Tier # 2 Group # 1 K/A # 007 191002 K1.17 Importance Rating: 3.5 K/A Statement: PRESSURIZER RELIEF/QUENCH TANK SYSTEM. SENSORS AND DETECTORS: (NUCLEAR INSTRUMENTATION) Effects of core voiding on neutron detection Proposed Answer: A Justification: From GFES P1612: Initially, excore source/startup range neutron indications will increase. This is because steam in the downcomer region (which is much less dense than subcooled water) significantly lowers the fast-non-leakage probability (Lf).

Therefore, due to less neutron moderation occurring in the downcomer, more neutrons reach the excore source range neutron level, causing an increase in detector signal.

EOP 2532 checks for adequate two-phase natural circulation with CET temperatures less than 700° F. It is also evident that the core remains covered with RVLMS at 17%.

Plausibility:

A. Correct response.

B. Examinee may not understand the location of the downcomer region but relate the RVLMS level indication with core voiding resulting in a loss of moderator and decrease in neutron population.

C. Examinee may not understand the location of the downcomer region but relate the RVLMS level indication with core voiding resulting in movement of boron out of the core causing count rate to increase.

D. Examinee may relate the void formation with movement of boron into the core, or two phase natural circulation concentrating boron in the core, causing count rate to decrease.

Examinee may also relate boron increase with SI flow but this would refill the downcomer and eliminate a void.

Technical Reference(s):

EOP 2532 Loss of Coolant Accident

OE INPO 22-008 TMI Unit 2 Case Study Provided reference(s): None Learning Objective: GFES 191002 3.2 State the effect core voiding, core loading pattern, and environmental effects could have on neutron detection and power indication.

Source: Modified GFES QID: P1612 Cognitive Level: Memory 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #26 The plant tripped from 100% power.

24C and 24D are dead.

The crew has transitioned to EOP 2530, Station Blackout.

Which of the following describes a required action and the reason for the action's time restriction?

A.

If Vital AC is not expected to be restored within 30 minutes of event initiation, take actions to align supplemental cooling to mitigate the effects of exceeding equipment design limits, possibly causing a loss of instrumentation.

B.

If Vital AC is not restored within 90 minutes of event initiation, open PPC UPS breakers to prevent reversing a computer battery cell due to low battery voltage.

C.

Within 60 minutes of commencing a cooldown, and every 50°F thereafter, sample the RCS for boron to ensure Shutdown Margin is met.

D.

If Vital AC is not restored within 30 minutes of event initiation, ensure backup air is aligned to applicable valves to preclude loss of valve position control.

Question #26 RO SRO Tier # 2 Group # 1 K/A # 063 A2.05 Importance Rating: 4.2 K/A Statement: Ability to (a) predict the impacts of the following on the DC ELECTRICAL DISTRIBUTION SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Loss of all AC Proposed Answer: A.

Justification: The requirement to align supplemental cooling is to limit the heat loading in the Vital DC Switchgear Rooms. This requirement is one of the most important time-critical operator actions in 2530 as it could result in the loss of ALL remaining Safety Channels of instrumentation.

Plausibility:

A. Correct answer B. The examinee may only recall that the PPC is removed from service during a Station Blackout but not recall that it is part of the alignment for supplemental cooling.

Protecting the PPC UPS battery is a reasonable assumption for the action and the 90 minutes is representative of the battery life.

C. There is a requirement in 2530 to verify RCS boron concentration meets shutdown margin every 50 °F on a cooldown, but the concentration MUST be verified BEFORE the cooldown is started and for the NEXT 50 °F temperature.

D. There are steps to restore Instrument Air by cross tying with Unit 3 and RNO actions to align backup air which will prevent valve repositioning. The step is not time critical and is preceded by a note which states that RNO actions may be delayed until after power is restored.

Technical Reference(s):

EOP 2530 Station Blackout

EOP 2541 Appendix 14 Supplemental Cooling

SP-EE-0362, Millstone Unit 2 Station Blackout Safe Shutdown Scenario Document Provided reference(s): None Learning Objective: ILT Predict how operator action or inaction affects plant and system conditions concerning a Station Blackout. [283759]

Source: Bank 415257 Cognitive Level: Memory 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics Comments:

Question #27 The plant has just entered MODE 5.

Numerous control board annunciators alarm.

Several Facility 1 components change position.

Several Facility 1 component indicating lights are out.

Indicating lights for all Bus 24C breakers are out.

Which one of the following is the appropriate procedure to enter for this event?

A.

AOP 2501, Diagnostic for Loss of Electrical Power.

B.

AOP 2502C, Loss of Vital 4.16kv Bus 24C.

C.

AOP 2506A, Loss of 125 VDC Instrument Panel DV10.

D.

AOP 2507A, Loss of 125 VDC Instrument Panel D-11.

Question #27 RO SRO Tier # 2 Group # 1 K/A # 062 K2.02 Importance Rating: 3.5 K/A Statement: Knowledge of electrical power supplies to the following: AC ELECTRICAL DISTRIBUTION SYSTEM Breaker control power Proposed Answer: C Justification: The indications are associated with a loss of either Vital Bus 201A or Vital DC Panel, DV10.

Plausibility:

A. Examinee decides the diagnostic AOP is a sure bet, but AOP 2501 is not used for a loss of DC only AC.

B. Examinee may relate the indications with a loss of Vital AC Bus 24C which would be lost if the Unit was in MODE 1, but in MODE 5 the bus would remain energized from the RSST.

C. Correct response D. Examinee recalls that D-11 is a Facility 1 DC power supply, but it does not the supply the control power for Vital AC Bus 24C.

Technical Reference(s):

AOP 2506A, Loss of 125 VDC Instrument Panel DV10 Provided reference(s): None Learning Objective: ILT Given a set of plant conditions, determine applicability of AOP 2506A, B, C, or D, "Loss of Vital 125 VDC Instrument Panel DV10, DV20, DV30, or DV40". [282905]

Source: Bank [451794], NRC-2014, Q-73, R-1, RO Cognitive Level: Comprehension 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments: Changed two distractors to make more plausible.

Question #28 The plant is operating at 100% power.

RCP CONTROL BLEED OFF PRES HI HI (C-02/3, B-10) alarms and locks in.

RCP Bleed-Off Pressure = 60 psig and stable.

Bleed-Off is verified aligned to the VCT.

ALL 4 RCP Bleed-Off flow rates are 1.0 gpm and stable.

1. What procedure would address this issue?

AND

2. What procedurally directed actions are taken?

A.

1. OP 2387A, Annunciator System Operation and Control.

2. Clear the alarm by isolating its SECTION.

B.

1. OP 2387A, Annunciator System Operation and Control.

2. Remove the annunciator from service.

C.

1. AOP 2519, Loss of Annunciators.

2. Clear the alarm by isolating its SECTION.

D.

1. AOP 2519, Loss of Annunciators.

2. Remove the annunciator from service.

Question #28 RO SRO Tier # 2 Group # 1 K/A # 003 G2.4.31 Importance Rating: 4.2 K/A Statement: REACTOR COOLANT PUMP SYSTEM. EMERGENCY PROCEDURES/PLAN: Knowledge of annunciator alarms, indications, or response procedures Proposed Answer: B Justification: The conditions listed show an annunciator that is locked in with normal system conditions. For an individual annunciator that is failed, it is removed from service iaw OP 2387A, Annunciator System Operation and Control.

Plausibility:

A. The student may think the annunciator is grounded, causing it to be locked in. Grounded annunciators are isolated by SECTION. OP 2387A does not address grounded annunciators.

B. Correct C. AOP 2519, Loss of Annunciators, is entered when multiple annunciators are failed or acting erratically. The student may think the annunciator is grounded, causing it to be locked in.

Grounded annunciators are isolated by SECTION.

D. AOP 2519, Loss of Annunciators, is entered when multiple annunciators are failed or acting erratically.

Technical Reference(s):

OP 2387A, Annunciator System Operation and Control Reference 2

ARP 2590B-038, RCP CONTROL BLEED-OFF PRES HI HI Provided reference(s): None Learning Objective: 280946 Describe the four (4) conditions that meet the criteria to remove an individual annunciator from service with OP 2387A Source: New Cognitive Level: Comprehension 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #29 The plant has been operating at 100% power for 2 months.

Power is reduced to 10% power in one (1) hour (step change).

What actions are taken over the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to compensate for Xe-135 effects in the reactor?

A.

Withdrawal CEAs and dilute the RCS during the entire time period.

B.

Insert CEAs and borate the RCS during the entire time period.

C.

Initially withdrawal CEAs and dilute the RCS, then insert CEAs and borate the RCS.

D.

Initially insert CEAs and borate the RCS, then withdrawal CEAs and dilute the RCS.

Question #29 RO SRO Tier # 4 Group #

K/A # 192006 K1.14 Importance Rating: 3.3 K/A Statement: FISSION PRODUCT POISONS: Explain the methods and reasons for the reactor operator to compensate for the time-dependent behavior of xenon-135 concentration in the reactor Proposed Answer: C Justification: On a negative step change in power, the Xenon production term is greater than the removal term due to a lower neutron flux. After a period of time, the production term is less than the removal term due to a lower number of fission product daughters being produced.

Plausibility:

A. May think Xenon production remains greater than the removal term for a longer period of time.

B. May think Xenon production is lower than the removal term for the whole period period.

C. Correct.

D. May confuse whether Xenon is initially building in or burning out on the step change.

Technical Reference(s):

GFES 192006 - Fission Product Poisons, slides 42&43 Provided reference(s): None Learning Objective: GFES Explain how Xe-135 concentration reacts during the following nuclear power plant operations: xenon free initial reactor start up, reactor shutdown, decrease in reactor power, increase in reactor power, and reactor start up with xenon present in the core.

Source: Bank 499550 Cognitive Level: Comprehension 10CFR55.41(b)(1): Fundamentals of reactor theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects Comments:

Question #30 The plant tripped from 100% power when off-site power was lost.

EOP 2528, LOOP/LOFC has been entered.

The "B" Emergency Diesel Generator (EDG) is unavailable.

Unit 3 is preparing to make the Station Blackout (SBO) Diesel Generator available.

Which of the following conditions, taken by itself, will necessitate declaration of an ELAP?

A.

If the LOOP was the result of a hurricane and storm surge requires the Facility 1 Service Water pump be secured.

B.

If the "A" EDG trips on a ground fault while Bus 24D is being prepared to be energized from Unit 3.

C.

If while attempting to restore Bus 24D from Unit 3, the SBO Diesel Generator is found to be unavailable.

D.

If the "A" EDG is damaged and lost and a ground fault on Bus 24E is not expected to be repaired for ~ 90 minutes.

Question #30 RO SRO Tier # 1 Group # 1 K/A # 055 G2.4.20 Importance Rating: 3.8 K/A Statement: Station Blackout EMERGENCY PROCEDURES/PLAN: Knowledge of the operational implications of emergency and abnormal operating procedures warnings, cautions, and notes.

Proposed Answer: D Justification: EOP 2530, Note states that "An ELAP is declared when it is predicted that AC power will NOT be restored within ONE hour of the event initiation". When the "A" EDG was lost, the plant is now in a Station Blackout event, requiring transition to EOP 2530. Power cannot be restored from Unit 3 Station Blackout EDG until 24E is available, which is predicted to be 90 minutes away.

Plausibility:

A. The examinee may assume that the loss of the Facility 1 SW pump results in a loss of "A" EDG and a loss of all AC. AOP 2560, Storms, High Winds and High Tides, gives guidance to ensure one EDG would still be available by aligning it to be cooled by fire water before all Service Water is secured due to storm surge.

B. The examinee may determine that this would cause a loss all AC, it does not however meet the ELAP criteria since the SBO diesel is available.

C. There has not been a loss of all AC but the examinee may relate an ELAP to the inability to energize a vital bus from the SBO diesel.

D. Correct answer.

Technical Reference(s):

EOP 2530 Station Blackout Provided reference(s): None Learning Objective: Outline and explain the bases for the major actions in EOP 2530, "Station Blackout." [283760]

Source: Bank [458961] Modified Cognitive Level: Memory 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #31 A reactor startup is in progress IAW OP 2202, Reactor Startup.

Reactor power indicates 200 CPS on all channels.

RO takes CEA control switch to WITHDRAW.

RE calls out a doubling.

The RO releases the CEA control switch and pushes the CEDS OFF button.

CEAPDS and PPC indicate Group 5 at 104 steps and withdrawing.

What diverse indication can be used to confirm this is a continuous rod withdrawal?

A.

Rising Tcold.

B.

Wide range counts rise slowly and level off.

C.

Rising letdown flow.

D.

Rising SUR.

Question #31 RO SRO Tier # 1 Group # 2 K/A # 001 AA2.09 Importance Rating: 3.8 K/A Statement: Ability to determine and/or interpret the following as they apply to CONTINUOUS Rod Withdrawal: RPI Proposed Answer: D Justification: If CEAs are continuing to withdraw, they are continuing to add positive reactivity.

This will cause SUR to rise.

Plausibility:

A. Plausible, if reactor was critical above the POAH, temperature would rise with a positive reactivity addition.

B. Plausible, normal rod withdrawals on a subcritical reactor cause counts to rise and level off.

C. Plausible, if reactor was critical above the POAH, letdown would rise to compensate for expanding RCS, due to the temperature rise from adding positive reactivity.

D. Correct.

Technical Reference(s):

OP 2202, Reactor Startup Provided reference(s): None Learning Objective: Evaluate control board indications as specified in OP 2202, Reactor Startup.

(282480)

Source: New Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #32 The plant tripped from 100% power.

Letdown flow is being diverted to Clean Liquid Radwaste (CLRW).

A CLRW Receiver tank level = 85% and increasing.

B CLRW Receiver tank level = 20% and stable.

Where will letdown be directed to when the A CLRW Receiver tank gets full?

A.

B CLRW Receiver tank.

B.

A CLRW Monitor tank.

C.

Volume Control tank (VCT).

D.

Aerated Waste Monitor tank.

Question #32 RO SRO Tier # 2 Group # 2 K/A # 068 K4.03 Importance Rating: 3.0 K/A Statement: Knowledge of LIQUID RADWASTE SYSTEM design features and/or interlocks that provide for the following: Automatic system realignments Proposed Answer: A Justification: The CLRW system is designed such that, when the in-service Receiver tank gets full, the system re-aligns to the other Receiver tank.

Plausibility:

A. Correct.

B. Plausible since the system re-aligns to the Monitor tank if both Receiver tanks are full.

C. Plausible since the VCT is the normal flowpath for letdown.

D. Plausible since may think Coolant Waste will automatically cross-tied to Aerated Waste (can be manually done).

Technical Reference(s):

OP 2335A, Clean Liquid Radwaste, Section 4.1 Provided reference(s): None Learning Objective: 287449 Describe the functions of the following CLRWS controls, including how the controlled component(s) is/are affected by each mode or position of the control:

D) Receiver Tank Auto-Bypass E) Monitor and Receiver Tank Inlet and Outlet valves Source: Modified 451589 NRC 2002 Q55 Cognitive Level: Memory 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #33 The plant is operating at 100% power.

Channel X Pressurizer pressure control is selected.

Channel X Pressurizer level control is selected.

The plant experiences a loss of VA-10.

Immediate Operator Actions have been completed.

1. What is the effect on Pressurizer Pressure?

AND

2. Why?

A.

1. Rising.

2. Charging flow exceeds letdown.

B.

1. Rising.

2. All Pressurizer heaters are energized.

C.

1. Lowering.

2. PORVs have opened.

D.

1. Lowering.

2. All Pressurizer heaters are de-energized.

Question #33 RO SRO Tier # 2 Group # 1 K/A # 010 K2.05 Importance Rating: 3.3 K/A Statement: Knowledge of electrical power supplies to the following: PRESSURIZER PRESSURE CONTROL SYSTEM (PZR PCS) Pressure channels Proposed Answer: D Justification: Channel X of pressurizer level fails to zero, causing heaters to cut out if heaters are selected to X or X+Y.

Plausibility:

A. Plausible: Student may fail to heed the bullet that IOAs have been completed (meaning charging and letdown are secured).

B. Plausible: Channel X level fails to zero, which would call for heaters.

C. Plausible: Channel A of RPS is lost. Two lost channels would cause PORVs to open.

D. Correct.

Technical Reference(s):

AOP 2501, Millstone 2 Diagnostic for Loss of Electrical Power

AOP 2585, Immediate Operator Actions Provided reference(s): None Learning Objective: Given a set of plant conditions during a loss of Non-Vital Instrument Panels VR-11 & VR-21 and-Vital Instrument Panels VA-10, VA-20, VA-30, & VA-40, determine equipment limitations caused by those conditions (282852).

Source: New Cognitive Level: Memory 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments: Question rewritten based on validation feedback. SM 12/8

Question #34 The plant experienced a SBLOCA.

RVLMS is 80% on both trains.

Both S/Gs are 55% and stable on Main Feed.

Pressurizer level is 17% and rising.

CETs are 520°F and slowly lowering.

Pressurizer Pressure is 1100 psia and stable.

CET subcooling is 36°F and slowly rising.

1. Which parameter remains to be fulfilled to satisfy HPSI throttle/stop criteria?

AND

2. Why is this parameter important in the context of a SBLOCA?

A.

1. Pressurizer level raised to greater than 20%.

2. It ensures the fluid surrounding the core is subcooled.

B.

1. Pressurizer pressure raised above 1150 psia.

2. It ensures the fluid surrounding the core is subcooled.

C.

1. Pressurizer level raised to greater than 20%.

2. It is an indication that RCS inventory control is established.

D.

1. Pressurizer pressure raised above 1150 psia.

2. It is an indication that RCS inventory control is established.

Question #34 RO SRO Tier # 1 Group # 1 K/A # 009 EK3.24 Importance Rating: 3.9 K/A Statement: Knowledge of the reasons for the following responses and/or actions as they apply to SMALL-Break LOCA: ECCS throttling or termination criteria Proposed Answer: C Justification: Per EOP 2532, Loss of Coolant Accident, 20% and not dropping is the minimum pressurizer level to meet HPSI Throttle/Stop criteria. Per Millstone Unit 2 Emergency Operating Procedure Technical Guide for EOP 2532, Loss of Coolant Accident: Pressurizer level is greater than the minimum level for inventory control and not dropping. A pressurizer level greater than the minimum level for inventory control and not dropping, in conjunction with adequate RCS subcooling, is an indication that RCS inventory control is established.

Plausibility:

A. Plausible, adequate inventory will aid in maintaining adequate subcooling by helping maintain pressure.

B. Plausible, the P/T curves on MON1 usually also include a RCP NPSH curve and the candidate may believe that NPSH may need to be met.

C. Correct.

D. Plausible, the P/T curves on MON1 usually also include a RCP NPSH curve and the candidate may believe that NPSH may need to be met. Additionally, adequate pressure may be an alternate indication that inventory control is established.

Technical Reference(s):

EOP 2532, Loss of Coolant Accident

MP2 EOPe Tech Guide for EOP 2532, Loss of Coolant Accident Provided reference(s): None Learning Objective: Given a set of plant conditions concerning a Loss of Coolant Accident determine if criteria for the following are met: (283792)

Trip 2/Leave 2 RCPs HPSI/LPSI Termination Initiation of SRAS Source: New Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Question #35 The BOP is performing a quarterly surveillance.

The procedure calls for the data to be forwarded to the system engineer.

The name and extension provided in the procedure are for a retired system engineer.

The Shift Manager has the name and extension of the current system engineer.

In accordance with AD-AA-100, Technical Procedure Process Control,

1. Can an administrative correction be used for this change?

AND

2. What is required to implement the change?

A.

1. Yes.

2. Handwrite change in procedure and have supervisor initial and date.

B.

1. Yes.

2. Complete Administrative Correction Process form and obtain approvals.

C.

1. No.

2. Handwrite change in procedure and have supervisor initial and date.

D.

1. No.

2. Complete FIP form and obtain approvals.

Question #35 RO SRO Tier # 3 Group #

K/A # G2.2.6 Importance Rating: 3.0 K/A Statement: EQUIPMENT CONTROL: Knowledge of the process for making changes to procedures Proposed Answer: A Justification: AD-AA-100, step 5.3.5 (definition of Administrative Correction) lists several items which are non-intent changes where administrative corrections are allowed. Updating names and phone numbers is listed in the first bullet. Step 3.11.1 of AD-AA-100 describes the process used for admin corrections when the procedure is in use, and substeps c and d direct handwriting the change with initial and date of supervisor.

Plausibility:

A. Correct.

B. Plausible: If the procedure was not already in progress, a pen and ink change to the working copy would not be appropriate.

C. Plausible, a FIP is always allowed, however, the question asks whether an admin correction can be used.

D. Plausible, a FIP is always allowed, however, the question asks whether an admin correction can be used.

Technical Reference(s):

AD-AA-100, Technical Procedure Process Control Provided reference(s): None Learning Objective: Describe how a procedure change may be made when needed immediately (251837)

Source: Modified Unit 2 NRC 2020 Q70 Cognitive Level: Memory 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments: Shortened part 1 to a simple yes or no. DF 12/9

Question #36 The unit is operating at 100% power.

A tube rupture causes a main steam line radiation monitor to alarm.

What parameter is used to validate the alarm?

A.

Steam Jet Air Ejector radiation monitor is in alarm.

B.

Steam flow rate is greater than feed flow rate.

C.

N-16 radiation monitors are alarming.

D.

Pressurizer level lowering with no change in TCOLD.

Question #36 RO SRO Tier # 3 Group #

K/A # G2.4.46 Importance Rating: 4.2 K/A Statement: EMERGENCY PROCEDURES/PLAN: Ability to verify that the alarms are consistent with the plant conditions Proposed Answer:

D.

FI-202 LTDN FLOW indicates 28 gpm, LR-110 PZR LVL indication lowering 1%/min.

Justification: ARP directs validating the alarm prior to tripping the reactor but is not specific on what to validate. AOP 2569 Steam Generator Tube Leak checks that the alarm is valid based on other changing RCS indications. AOP 2569 Basis Document describes validation based on other RCS indications show a loss of RCS inventory.

Plausibility:

A. Examinee recalls that the SJAE Rad Monitor is used in triggering required actions of AOP 2569, S/G Tube Leak B. Examinee recalls that steam flow/feed flow mismatch is used in determining the most affected SG during a SGTR.

C. Examinee recalls that N-16 rad monitors are used in triggering required actions of AOP 2569, S/G Tube leak.

D. Correct response.

Technical Reference(s):

ARP 2590A-117, MAIN STEAMLINE HI RAD/INST. FAIL

AOP 2569, Steam Generator Tube Leak

AOP 2569, Background (Bases) Document Provided reference(s): None Learning Objective: Evaluate abnormal system operation and determine the cause and course of action in accordance with AOP 2569, "Steam Generator Tube Leak". [283303]

Source: New Cognitive Level: Comprehension 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #37 During a Steam Generator Tube Rupture (SGTR), what is the

1. lower limit of the band for affected Steam Generator level?

AND

2. reason for this limit?

A.

1. 33%.

2. Iodine scrubbing.

B.

1. 40%.

2. Iodine scrubbing.

C.

1. 33%.

2. Minimize thermal stress on tubes.

D.

1. 40%.

2. Minimize thermal stress on tubes.

Question #37 RO SRO Tier # 1 Group # 1 K/A # 038 EK2.20 Importance Rating: 3.8 K/A Statement: (000) Knowledge of the relationship between (EPE 38) STEAM GENERATOR Tube Rupture and the following systems or components: S/Gs Proposed Answer: B Justification: Per OP 2260, attachment 1 step 6, the BOP is expected to maintain the affected generator 40-45% to aid in iodine scrubbing. Top of tube bundle is 33%.

Plausibility:

A. Plausible: top of tube bundle is 33%.

B. Correct.

C. Plausible: maintaining at or above top of tube bundle will maintain the tubes covered and minimize thermal stress.

D. Plausible: maintaining at or above top of tube bundle will maintain the tubes covered and minimize thermal stress.

Technical Reference(s):

OP 2260, Unit 2 EOP Users Guide Provided reference(s): None Learning Objective: Given a set of plant conditions, predict how operator action or inaction affects plant and system conditions concerning SPTAs (283651)

Source: New Cognitive Level: Memory 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #38 An approach to criticality is being performed by means of CEA withdrawal.

The RO stops CEA motion when the reactor is close to criticality, but still subcritical.

How will the Wide Range Nuclear Instruments respond from the moment CEA motion stops?

1. Source Range Count Rate (CR) will 1..

AND

2. Startup Rate (SUR) will 2..

A.

1. continue to increase for a short time and then stabilize

2. gradually decrease to zero B.

1. continue to increase but at a slower rate

2. stabilize at a slightly positive value C.

1. immediately stop increasing and stabilize at its present value

2. immediately decrease to zero D.

1. immediately begin to decrease but at a slower rate

2. gradually decrease to a slightly negative value

Question #38 RO SRO Tier # 4 Group #

K/A # 192008 K1.03 Importance Rating: 4.0 K/A Statement: (K1.03) REACTOR OPERATIONAL PHYSICS: (STARTUP AND APPROACH TO CRITICALITY) Describe count rate and instrument response that should be observed for rod withdrawal during the approach to criticality Proposed Answer:

A.

1. continue to increase for a short time and then stabilize

2. gradually decrease to zero Justification: From GFES 192008, ELO 1.2, When an operator stops rod motion in a subcritical reactor, the source range count rate will achieve a new equilibrium level. SUR decays to zero indicating subcritical multiplication has reached equilibrium. Each control rod withdrawal requires a longer amount of time for count rate to reach equilibrium.

Plausibility:

A. Correct response.

B. These conditions describe what is observed when the reactor is critical.

C. These conditions describe what would be observed if there were no subcritical multiplication.

D. These conditions describe what could be observed during CEA insertion, reactor shutdown or reactor trip.

Technical Reference(s):

OP 2202, Reactor Startup

GFES 192008, Reactor Operational Physics Provided reference(s): None Learning Objective:

ILT As described in OP 2202, "Reactor Startup", give the general sequence of events followed for a reactor startup. (MB-05372) [282474]

LOIT List the parameters that the Reactor Operator should monitor during reactivity manipulations involving changes in CEA height or RCS boron concentration.

(MB 02126) [314072]

GFES 192008, ELO 1.2, Describe the nuclear instrumentation response during a reactor startup to criticality.

Source: Bank [413841]

Cognitive Level: Memory 10CFR55.41(b)(1): Fundamentals of reactor theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects Comments:

Question #39 The unit is operating at 100% power.

The following alarms annunciate:

PROCESS MON RAD HI HI/FAIL (C-06/7, DB-24)

CH `C' CTMT AIR GASEOUS RADIATION HI (C-01, C-28)

CTMT RAD ACTUATION SIG CH 1 TRIP (C-01, C-36) (RED WINDOW)

CTMT RAD ACTUATION SIG CH 2 TRIP (C-01, D-36) (RED WINDOW)

These alarms indicate a failure of what Radiation Monitor?

A.

RM-8123B, Facility 1 CTMT Atmosphere Rad Monitor.

B.

RM-8240A, Facility 1 CTMT HI Range Rad Monitor.

C.

RM-8240B, Facility 2 CTMT HI Range Rad Monitor.

D.

RM-8262B, Facility 2 CTMT Atmosphere Rad Monitor.

Question #39 RO SRO Tier # 2 Group # 1 K/A # 073 A1.02 Importance Rating: 3.2 K/A Statement: Ability to predict and/or monitor changes in parameters associated with operation of the PROCESS RADIATION MONITORING SYSTEM, including: Lights and alarms Proposed Answer: A Justification: RM-8123B, Facility 1 Containment Gaseous Radiation monitor inputs to ESAS Sensor Cabinet C, a failure of the RM input to the Sensor Cabinet will place its associated bistable in the tripped condition resulting in the Ch C CTMT Air Gaseous Rad Hi alarm. This will also cause modules in both Actuation Cabinets 5 & 6 to actuate resulting in CTMT Rad Actuation signal trip alarms for Ch1 and Ch2.

Plausibility:

A. Correct.

B. Might think the alarms are associated with isolating the Hydrogen Purge valves on CTMT High Radiation.

C. Might think the alarms are associated with isolating the Hydrogen Purge valves on CTMT High Radiation.

D. May think Channel C is tripped from RM-8262.

Technical Reference(s):

ARP 2590A-111, CH C" CTMT AIR RADIATION HI

ARP 2590A-143, CTMT RAD ACTUATION SIG CH 1 TRIP

ARP 2590A-144, CTMT RAD ACTUATION SIG CH 2 TRIP Provided reference(s): None Learning Objective: Describe the automatic protective actions/functions associated with the following Area and Process radiation monitors (MB-00619) [287855]:

C) RM-8123A/B & RM-8262A/B, Z1 & Z2 Containment Particulate/Gaseous Source: New Cognitive Level: Memory 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics Comments: Changed answers to be more operationally valid. Previous answers were just numbers DF 12/7

Question #40 The plant is operating at 100% power.

The selected Pressurizer Pressure controller output fails to a value of 52%.

The non-selected Pressurizer Pressure controller output is 30% and lowering.

1. What is the effect on pressurizer pressure?

AND

2. How does the Pressurizer Pressure control system respond?

A.

1. increases.

2. proportional heaters at minimum.

B.

1. increases.

2. spray valves open.

C.

1. decreases.

2. backup heaters energize.

D.

1. decreases.

2. spray valves close.

Question #40 RO SRO Tier # 1 Group # 1 K/A #027 AA2.11 Importance Rating: 4.0 K/A Statement: Ability to determine and/or interpret the following as they apply to PRESSURIZER PRESSURE Control System Malfunction: RCS Pressure Proposed Answer:

C.

1. decreases

2. backup heaters energize Justification: When the controller output rises to >50%, the proportional heaters are at minimum and spray valves are beginning to open, causing pressurizer pressure to decrease. This response is confirmed by the output of the non-controlling channel. Backup heaters will automatically energize at 2200 psia lowering.

Plausibility:

A. Proportional heaters going to minimum is an expected response to this failure and an expected response if pressure was increasing.

B. Spray valves opening is an expected response to this malfunction and an expected response if pressure was increasing.

C. Correct response.

D. Pressure decreasing is the expected effect of the malfunction, spray valves closing is normally the expected response to a pressure decrease.

Technical Reference(s):

OP 2204, Load Changes Provided reference(s): None Learning Objective: ILT Given the plant with a steam bubble in the pressurizer, and given a pressurizer pressure deviation from setpoint, describe the response of the Pressurizer Pressure Control System including setpoints. [281901]

Source: Modified [413193]

Cognitive Level: Comprehension 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features..

Comments:

Question #41 The plant is operating at 100% power.

Containment parameters are as follows:

Pressure = 1 psig and rising.

Temperature = 110 F and rising.

Normal sump level leak rate is rising.

Gaseous Rad monitors = 5000 cpm and rising What Containment parameter would differentiate between a RCS leak and a steam leak?

A.

Pressure.

B.

Temperature.

C.

Normal Sump level.

D.

Gaseous Rad monitors.

Question #41 RO SRO Tier # 2 Group # 1 K/A # 103 K6.11 Importance Rating: 3.8 K/A Statement: Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the CONTAINMENT SYSTEM: RCS Proposed Answer: D Justification: The only difference between a RCS leak and a steam leak is a RCS leak will cause atmospheric gaseous activity to rise.

Plausibility:

A. Plausible if the student thinks the steam leak has a lower enthalpy and wont cause CTMT pressure to increase as much as a RCS leak.

B. Plausible if the student thinks the steam leak has a lower enthalpy and wont cause CTMT temperature to increase as much as a RCS leak.

C. May think the steam leak wont fill the sump because its steam that is leaking D. Correct.

Technical Reference(s):

AOP 2568, Reactor Coolant Leak

AOP 2589, Excess Steam Load Provided reference(s): None Learning Objective: 283279 Given a set of plant conditions, determine AOP 2568, Reactor Coolant Leak applicability Source: New Cognitive Level: Memory 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #42 Per EOP 2528, Loss Of Offsite Power/Loss Of Forced Circulation, what is the MAXIMUM acceptable loop T when checking for single phase natural circulation?

A.

24°F B.

34°F C.

44°F D.

54°F

Question #42 RO SRO Tier # 1 Group # 1 K/A # 056 AA2.88 Importance Rating: 4.3 K/A Statement: Ability to determine and/or interpret the following as they apply to Loss of Offsite Power: Conditions necessary for natural circulation Proposed Answer: D Justification: EOP 2528 step 12c gives four items to check in order to verify natural circulation:

Loop delta-T less than 55°F, TH and TC constant or dropping, CET subcooling above 30°F and TH and CET temperatures within 10°F of each other.

Plausibility:

A. Plausible: 24°F is near the normal post-trip delta-T in natural circulation.

B. Plausible: 34°F is slightly above the normal post-trip delta-T in natural circulation.

C. Plausible: 44°F is slightly above the normal post-trip delta-T in natural circulation.

D. Correct.

Technical Reference(s):

EOP 2528, Loss of Offsite Power/ Loss of Forced Circulation Provided reference(s): None Learning Objective: Describe the indications that are available to the operator to determine the effectiveness of natural circulation (283715).

Source: Bank Vision #451894, NRC 2014 Question 1 Cognitive Level: Memory 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #43 Why do the Containment Air Recirculation (CAR) fans shift to slow speed on a Loss of Coolant Accident (LOCA)?

A.

Prevent lifting CAR Cooler thermal relief valves due to high heat transfer rate.

B.

Prevent collapse of CAR inlet ducting due to high differential pressure.

C.

Prevent over-pressurization of CAR ducting with the fans running in fast speed.

D.

Prevent CAR fan high motor current due to higher Containment air density.

Question #43 RO SRO Tier # 2 Group # 1 K/A # 022K5.03 Importance Rating: 3.1 K/A Statement: Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the CONTAINMENT COOLING SYSTEM (CCS): Containment equipment subject to damage by high or low temperature, humidity, and pressure Proposed Answer: D Justification: Precaution 3.2 of OP 2313A states: To prevent damage to motors, CAR fans must not be operated in fast speed during Containment ILRT (increased pressure may overload fan motors). ILRT simulates LOCA conditions (without the moisture content)

Plausibility:

A. When removing or returning CAR fans from service, out of sequence manipulation of inlet and outlet valves may result in lifting the heat exchanger thermal reliefs. Plausible but with RBCCW flow, the thermal reliefs would not lift due to high heat transfer rate.

B. There is no Inlet ducting to collapse, the CAR fans are connected to the base of the CAR coolers. Plausible: the student may not recall this configuration.

C. To prevent overpressurizing CAR fan ducting, no more than three will be operated simultaneously in fast speed. Plausible but not reason for shifting to SLOW during LOCA.

D. Correct.

Technical Reference(s):

OP 2313A, Containment Air Recirculation and Cooling System.

CCS-00-C, Containment and Containment Systems Lesson text.

Provided reference(s): None Learning Objective: 280972 Describe the relationships between the Containment Air Recirculation System and the Containment Spray System with respect to limiting Containment pressure and temperature during a design-basis LOCA.

Source: Bank 412123 Cognitive Level: Memory 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Question #44 AOP 2580, Degraded Voltage, has been entered based on notification by CONVEX.

Main generator output is 740 MWe.

Main generator hydrogen pressure is 58 psig.

Given AOP 2580, Attachment 1, Estimated Capability Curves, what is the maximum amount of excitation (MVARs) that the generator can produce and stay within its limits?

A.

420 MVARs B.

435 MVARs C.

540 MVARs D.

560 MVARs

Question #44 RO SRO Tier # 1 Group # 1 K/A # 077 AA2.01 Importance Rating: 3.7 K/A Statement: Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC Grid Disturbances: Operating point on the generator capability curve Proposed Answer: C.

Justification: AOP 2580, Attachment 1, with a lagging power factor (overexcited) at 740 MWe and 58# hydrogen, the maximum MVAR loading is 540 MVARs Plausibility:

A. 420 MVARs equates to the leading power factor limit (underexcited) and also equates to the normal 100% (900 MWe) lagging limit lagging with 58 psig.

B. 435 MVARS equates to the leading power factor limit with 60 psig (Examinee uses normal value or does not interpolate). It is also very close to the normal 100% (900 MWe, 60 psig) lagging power factor limit.

C. Correct response.

D. 560 MVARS equates to the lagging power factor limit with 60 psig (Examinee uses normal value or does not interpolate)

Technical Reference(s):

AOP 2580, Degraded Voltage Provided reference(s): AOP 2580, Attachment 1 Estimated Capability Curves Learning Objective: ILT Outline the major actions for a degraded voltage condition.

Source: Bank [414978]

Cognitive Level: Analysis 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #45 The plant is in MODE 1.

24E is aligned to 24D.

C Service Water pump trips on overload.

RBCCW HX TEMP HI (C06/07, D-8) alarms.

C RBCCW HX outlet temperature is 88°F and rising.

1. What is the effect of this Loss of Service Water on the RBCCW system?

AND

2. What actions are required by AOP 2565, Loss of Service Water?

A.

1. A RBCCW HX has lost cooling.

2. Start the B Service Water Pump on Facility 2.

B.

1. A RBCCW HX has lost cooling.

2. Trip the reactor and affected RCPs, Go to EOP 2525.

C.

1. C RBCCW HX has lost cooling.

2. Start the B Service Water Pump on Facility 2.

D.

1. C RBCCW HX has lost cooling.

2. Trip the reactor and affected RCPs, Go to EOP 2525.

Question #45 RO SRO Tier # 1 Group # 1 K/A # 062 AA1.05 Importance Rating: 3.4 K/A Statement: Ability to operate and/or monitor the following as they apply to LOSS OF SERVICE WATER: CCWS Proposed Answer: C Justification: The C Service Water Pump supplies the B Service Water Header, which normally cools the Facility 2 RBCCW HX. Only B or C RBCCW heat exchangers are allowed to be aligned to Facility 2 Service Water in MODE 1. AOP 2565, Loss of Service Water, directs the swing (B) Service Water pump to be aligned and started when a running Service Water pump is lost and the swing pump is available.

Plausibility:

A. Plausible: The student could confuse the RBCCW heat exchanger layout (which has A on Z1 and B on Z2) with the TBCCW heat exchanger outlet (which has C on Z1 and A on Z2).

B. Plausible: The student could confuse the RBCCW heat exchanger layout (which has A on Z1 and B on Z2) with the TBCCW heat exchanger outlet (which has C on Z1 and A on Z2). Additionally, if Facility 2 Service Water couldnt be restored prior to C RBCCW HX outlet reaching 120°F, AOP 2565 would direct a trip.

C. Correct.

D. Plausible: If Facility 2 Service Water couldnt be restored prior to C RBCCW HX outlet reaching 120°F, AOP 2565 would direct a trip.

Technical Reference(s):

AOP 2565, Loss of Service Water Provided reference(s): None Learning Objective: Given a set of plant conditions, determine the section within AOP 2565, Loss of Service Water, that best mitigates the abnormal condition (283251).

Source: Modified 453289 Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

Question #46 The Unit is operating at 100% power.

Instrument Air (IA) is lost.

IA receiver pressure falls below 85 psig.

What is the status of the following Station Air (SA) valves?

2-SA-10.1, Cross Tie from Station Air to Instrument Air

2-SA-11.1, Station Air Header Isolation Valve SA-10.1 SA-11.1 A.

Open Open B.

Closed Open C.

Closed Closed D.

Open Closed

Question #46 RO SRO Tier # 2 Group # 1 K/A # 078 A1.05 Importance Rating: 2.9 K/A Statement: Ability to predict and/or monitor changes in parameters associated with operation of the INSTRUMENT AIR SYSTEM (IAS), including: Service air pressure Proposed Answer: D. Open, Closed, 85 psig and lowering Justification: If SA/IA X-TIE, SA-10.1, handswitch is in AUTO and Instrument Air header pressure falls below 85 psig, then SA-10.1 will automatically open and SA-11.1 will automatically close isolating the SA header. With the SA header isolated PI-7099 pressure will decrease.

Plausibility:

A. Examinee recalls SA-10.1 opening but not SA-11.1 going closed. Determines that the SA compressor is now suppling both SA and IA.

B. Examinee recalls the normal position of both valves with the handswitch in AUTO.

Determines that there is no change in the state of any of the components.

C. Examinee recalls SA-11.1 going closed but not SA10.1 going open.

D. Correct response.

Technical Reference(s):

OP 2332A, Station Air System

AOP 2563, Loss of Instrument Air

25203-26009 SH-8 P&ID Instrument and Station Air Systems Provided reference(s): None Learning Objective: Describe the functions of the following Station Air and Instrument Air Systems Control Room controls at Panel C-06, including how the controlled components are affected by each mode or position of the control: (MB-02635) [281596]

A. Cross Tie from Station Air to Instrument Air 2-SA-10.1 and Station Air Header Isolation Valve 2-SA-11.1, Describe how the Station Air and Instrument Air Systems components automatically function to cross-connect the Station Air System with the Instrument Air System when Instrument Air System pressure falls below setpoint. (MB-02642) [281600]

Source: New Cognitive Level: Comprehension 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics Comments:

Question #47 A fire in which location will cause the Unit 2 Electric Fire Pump to start?

A.

West DC Switchgear Room.

B.

Phase C GSU Transformer.

C.

Generator Exciter Doghouse.

D.

Loop 1, B RCP Cubicle.

Question #47 RO SRO Tier # 1 Group # 2 K/A # 067 AK2.06 Importance Rating: 3.7 K/A Statement: Knowledge of the relationship between PLANT Fire On Site and the following systems or components: Fire pumps.

Proposed Answer: B Justification: The Main Transformer is protected by a separate deluge system for each phase.

Plausibility:

A. Plausible: The West DC Switchgear room has fire suppression, but it is halon.

B. Correct.

C. Plausible: The Generator Exciter Doghouse has fire suppression, but it is carbon dioxide.

D. Plausible: The student may believe that there is fire suppression in Containment, because it is not accessible at power.

Technical Reference(s):

PowerPoint for lesson FPS-04-C, Fire Protection System Provided reference(s): None Learning Objective: State the purpose and describe the operating characteristics of the following major Fire Protection System components: (287608)

B) Fire Water Pumps Source: New Cognitive Level: Memory 10CFR55.41(b)(8): Components, capacity, and functions of emergency systems.

Comments:

Question #48 A subcooled liquid exists 1. its boiling point. Any additional heat addition will 2..

A.

1. below

2. cause boiling B.

1. below

2. raise the fluids temperature C.

1. at

2. cause boiling D.

1. at

2. raise the fluids temperature

Question #48 RO SRO Tier # 4 Group #

K/A # 193003 K1.16 Importance Rating: 2.7 K/A Statement: STEAM: Define the following term: -- subcooled and compressed liquids Proposed Answer: B Justification: The definition of a sub-cooled (compressed) liquid is one that exists below its boiling temperature and any heat added to it is sensible Plausibility:

A. The student may confuse a saturated liquid with subcooled since they are similar.

B. Correct.

C. The student may confuse a saturated liquid with subcooled since they are similar.

D. The student may think a subcooled liquid exists at the boiling point.

Technical Reference(s):

GFES Basic Energy Concepts Student Guide, page 5 Provided reference(s): None Learning Objective: GFES Explain the difference between the state and phase of a working substance.

Source: New Cognitive Level: Memory 10CFR55.41(b)(14): Principles of heat transfer thermodynamics and fluid dynamics Comments:

Question #49 An Excess Steam Demand Event (ESDE) occurred in #1 Steam Generator.

SIAS/CIAS/EBFAS/MSI actuated.

CETs = 472 °F and stable.

RVLMS = 100%.

Pressurizer Pressure = 1350 psia and slowly rising.

CET subcooling = 110 °F and slowly rising.

#2 S/G level = 28% and rising.

Pressurizer level = 64% and rising.

1. What action is taken to control of pressurizer level?

AND

2. When is the earliest this action may be taken?

A.

1. Restore Letdown flow.

2. When #2 S/G reaches 40%.

B.

1. Restore Letdown flow.

2. Immediately.

C.

1. Throttle HPSI flow.

2. When #2 S/G reaches 40%.

D.

1. Throttle HPSI flow.

2. Immediately.

Question #49 RO SRO Tier # 1 Group # 1 K/A # 040 AK2.18 Importance Rating: 3.4 K/A Statement: Knowledge of the relationship between STEAM LINE RUPTURE and the following systems or components: PZR LCS Proposed Answer: B Justification: Stopping the uncontrolled rise in Pressurizer Level will aid in maintaining pressure below the upper end of the P/T limit. Step 19 of EOP 2536 allows controlling charging and letdown and/or throttling HPSI to accomplish this, once HPSI throttle criteria are met. Step 26 allows restoration of letdown once HPSI throttle criteria are met. Both steps may be pulled forward as needed, and both refer to step 14b to list the HPSI throttle criteria, which are as follows: RCS subcooling above minimum on the P/T curve, Pressurizer Level >20% and stable or rising, RVLMS at least 43% and at least one steam generator 40%-70% or being restored. Throttling HPSI would be no help at 1350 psia, as this is above HPSI shutoff head.

Plausibility:

A. Plausible: candidate may not recall that 40%-70% S/G level is not necessary as long as level is being restored.

B. Correct.

C. Plausible: Throttling HPSI is a method of reducing rise in pressurizer level when pressurizer pressure is below ~1250 psia.

D. Plausible: Throttling HPSI is a method of reducing rise in pressurizer level when pressurizer pressure is below ~1250 psia, and candidate may not recall that 40%-70% S/G level is not necessary as long as level is being restored.

Technical Reference(s):

EOP 2536, Excess Steam Demand Event

EOP 2541 Appendix 2, Figures Provided reference(s): None Learning Objective: List the plant conditions used to determine if HPSI flow may be stopped as specified in EOP 2536, "Excess Steam Demand." (283853)

Source: New Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #50 The plant is operating at 100% power. Containment radiation levels rise significantly.

1. What Radiation monitors will isolate the Hydrogen Purge valves (EB-91, 92, 99, and 100)?

AND

2. How are the vent paths configured to isolate?

A. 1. Containment Atmosphere Particulate and Gaseous.

2. One radiation monitor isolates only one vent path.

B. 1. Containment High Range.

2. One radiation monitor isolates only one vent path.

C. 1. Containment Atmosphere Particulate and Gaseous.

2. Either radiation monitor isolates both vent paths.

D. 1. Containment High Range.

2. Either radiation monitor isolates both vent paths.

Question #50 RO SRO Tier # 1 Group # 2 K/A # 061AK1.02 Importance Rating: 3.6 K/A Statement: Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to AREA RADIATION MONITORING (ARM) System Alarms: Adverse containment conditions Proposed Answer: D Justification: The Containment High Range radiation monitors (RM 8240 and 8241) close the Hydrogen Purge valves (EB-91, 92, 99, and 100) which isolate the purge flow paths and stops the venting of Containment. Either radiation monitor will isolate both flow paths. Radiation monitor 8240 closes EB-99 which is in one flow path and EB-92 which is in the other flow path (outside ctmt valves). Radiation monitor 8241 closes EB-100 which is in one flow path and EB-91 which is in the other flow path (inside ctmt valves).

Plausibility:

A. Incorrect. The Containment Atmosphere Particulate and Gaseous radiation monitors (RM 8123A/B and RM8262A/B) do not close the Hydrogen Purge valves (EB-91, 92, 99, and 100). This is plausible because the Containment Atmosphere Particulate and Gaseous radiation monitors do close the Containment Purge valve (AC-4, 5, 6, and 7). It is reasonable to think that the Containment Atmosphere Particulate and Gaseous radiation monitors isolate venting of Containment on high radiation. It is also reasonable that one radiation monitor will isolate only one vent path (Facility dependent) since many systems operate equipment only in one train and are Facility dependent.

B. Incorrect. One Containment High Range radiation monitor isolates only one vent path (Facility dependent) is not correct. Either Containment High Range radiation monitor will isolate both flow paths. This is plausible because it is reasonable that one radiation monitor will isolate only one vent path (Facility dependent) since many systems operate equipment only in one train and are Facility dependent.

C. Incorrect. Containment Atmosphere Particulate and Gaseous radiation monitors is not correct. The Containment High Range radiation monitors (RM 8240 and 8241) close the Hydrogen Purge valves (EB-91, 92, 99, and 100) which isolate the purge flow paths and stops the venting of Containment. This is plausible because the Containment Atmosphere Particulate and Gaseous radiation monitors do close the Containment Purge valve (AC-4, 5, 6, and 7). It is reasonable to think that the Containment Atmosphere Particulate and Gaseous radiation monitors isolate venting of Containment on high radiation.

D. Correct.

Technical Reference(s):

1. Drawing 25203-26028 SH 3, P&ID Containment and Enclosure Building Ventilation, rev. 26.

2. ARP 2590A-119, HYDROGEN PURGE ISO VLVS HI RAD, rev. 000.

3. ARP 2590A -109, CH A CTMT AIR PARTICULATE RADIATION HI, rev. 000-02.

4. OP 2314B, Containment and Enclosure Building Purge, rev. 027-00. Pages: cover, 46-48.

Provided reference(s): None Learning Objective: 282105 ILT Describe the automatic actions and logic associated with the Containment High Range Radiation Monitoring Subsystem.

Source: Bank NRC 2022 Q23 History: 2022 ILT NRC final Q23 Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #51 Describe the actions taken if your Self-Reading Dosimeter (SRD) alarms (dose rate).

1. What actions are taken regarding the task being performed?

AND

2. Who is notified first?

A.

1. Complete the task at hand.

2. Control Room.

B.

1. Complete the task at hand.

2. Health Physics.

C.

1. Place the work in a safe condition.

2. Control Room.

D.

1. Place the work in a safe condition.

2. Health Physics.

Question #51 RO SRO Tier # 3 Group #

K/A # G2.3.5 Importance Rating: 2.9 K/A Statement: RADIATION CONTROL: Ability to use RMSs, such as fixed radiation monitors and alarms or personnel monitoring equipment Proposed Answer: D Justification: Dosimeters are programmed to the RWP being performed which takes in account the dose rates in the area(s) where the work is being performed. An alarming dosimeter indicates a change in plant conditions since the last survey or the worker is in the wrong area. In any case, if the dosimeter alarms on dose rate, the work is placed in a safe condition and HP is notified.

Plausibility:

A. The student may reason that completing the task is the safest action to take. In a large majority of cases, the Control Room is notified first.

B. The student may reason that completing the task is the safest action to take.

C. In a large majority of cases, the Control Room is notified first.

D. Correct.

Technical Reference(s):

Generic Radiation Worker Training Lesson Plan, pg. 47 Provided reference(s): None Learning Objective: Generic Radiation Worker: #33 State the action(s) to be taken if dosimetry is lost, damaged, or alarming Source: New Cognitive Level: Memory 10CFR55.41(b)(11): Purpose and operation of radiation monitoring system, including alarms and survey equipment.

Comments:

Question #52 The plant is at 100% power.

PRESSURIZER CH Y LEVEL HI/LO (C02/03, A-39) annunciates.

L110Y is 80% and rising.

L110X is 64% and lowering.

Pressurizer Pressure reads 2240 psia and lowering on all 4 safety channels.

Immediate Operator Actions were completed to stabilize the plant.

1. Which pressurizer level channel has failed?

AND

2. Which action will be procedurally required prior to restoring level control to automatic?

A.

1. Channel X is failing low.

2. Select channel Y on Pressurizer Level Control.

B.

1. Channel X is failing low.

2. Position CHG PP OVERRIDE to match running pumps.

C.

1. Channel Y is failing high.

2. Select channel X on Pressurizer Level Control.

D.

1. Channel Y is failing high.

2. Position CHG PP OVERRIDE to match running pumps.

Question #52 RO SRO Tier # 1 Group # 2 K/A # 028 AA1.08 Importance Rating: 3.8 K/A Statement: Ability to operate and/or monitor the following as they apply to PRESSURIZER (PZR) Level Control Malfunction: Selection of an alternate PZR level channel if one has failed.

Proposed Answer: C Justification: Step 9.7 of ARP 2590B-217, PRESSURIZER CH Y LEVEL HI/LO, directs selecting Pressurizer Level Control to Channel X prior to restoring automatic level control in Step 10. Manipulating the CHG PP OVERRIDE switch is NOT required per step 9.6 as no charging pumps are expected to start or stop with this failure (One charging pump would be expected to be running pre-event and L110Y is failing high, which would not start additional charging pumps).

Plausibility:

A. Plausible: examinee may miss or misinterpret the cue that pressure is lowering.

B. Plausible: examinee may miss or misinterpret the cue that pressure is lowering. A channel failing low (if selected) would eventually start a backup charging pump, and the CHG PP OVERRIDE switch would be used to maintain it running.

C. Correct.

D. Plausible: A channel failing low (instead of high) would eventually start a backup charging pump, and the CHG PP OVERRIDE switch would be used to maintain it running.

Technical Reference(s):

ARP 2590B-217, PRESSURIZER CH Y LEVEL HI/LO Provided reference(s): None Learning Objective: Given the plant with a steam bubble in the pressurizer, and given a pressurizer level or pressure transmitter failure (high or low) on either control channel (selected or non-selected), describe (281907):

a) The system response that would result from this failure, b) The actions necessary to mitigate this failure, c) The plant response if no operator actions are taken.

Source: New Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Question #53 Plant is operating at 100% power.

PROCESS MON RAD HI HI/FAIL alarms on C-07.

RM-4262, SG Blowdown Radiation Monitor Instrument Fail light is lit.

Which of the following contains ONLY those valves which will receive a CLOSE signal due to the instrument failure?

A.

2-MS-15, Blowdown Tank discharge valve.

2-MS-135, Quench Tank discharge valve.

2-MS-191A & B, SG Sample isolation valves.

B.

2-MS-15, Blowdown Tank discharge valve.

2-MS-135, Quench Tank discharge valve.

2-MS-220A & B, SG Blowdown isolation valves.

C.

HV-4287 & 4288, Secondary Sample Sink isolation valves.

MS-117A & B, Blowdown to Quench Tank isolation valves.

2-MS-220A & B, SG Blowdown isolation valves.

D.

HV-4287 & 4288, Secondary Sample Sink isolation valves.

MS-117A & B, Blowdown to Quench Tank isolation valves.

2-MS-191A & B, SG Sample isolation valves.

Question #53 RO SRO Tier # 2 Group # 1 K/A # 073 K3.05 Importance Rating: 3.4 K/A Statement: Knowledge of the effect that a loss or malfunction of the PROCESS RADIATION MONITORING SYSTEM will have on the following systems or system parameters: S/GB (steam generator blowdown)

Proposed Answer: B Justification: Failure of RM-4262 will result in closure of all potential discharge paths to prevent an unmonitored release of activity. Not closing the SG sample isolations allows for continued monitoring of SG activity. The following will close when RM-4262 fails:

2-MS-15, Blowdown Tank discharge valve 2-MS-135, Quench Tank discharge valve 2-MS-220A & B, SG Blowdown isolation valves HV-4287 & 4288, Secondary Sample Sink isolation valves Plausibility:

A. Examinee may confuse the blowdown isolation signal with a CIAS signal that closes MS-191A & B SG Sample isolation valves.

B. Correct response.

C. Examinee may think the Quench Tank gets isolated to prevent an unmonitored release.

D. Examinee may think the Quench Tank gets isolated to prevent an unmonitored release and confuse the blowdown isolation signal with a CIAS signal. Examinee may confuse the blowdown isolation signal with a CIAS signal that closes MS-191A & B SG Sample isolation valves.

Technical Reference(s):

ARP 2590E-135, PROCESS MON RAD HI HI/FAIL

ARP 2590H-005A, RM-4262 S/G BLOWDOWN GROSS ACTIVITY Provided reference(s): None Learning Objective: ILT Describe the automatic actions and logic associated with the SJAE and/or SG Blowdown Radiation Monitoring Subsystem(s). (MB 03134) [282110]

Source: Modified [452113]

Cognitive Level: Memory 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #54 The plant is operating at 100% power.

The main generator Automatic Voltage Regulator control switch is placed in the RAISE position.

Which of the following Main Generator parameters would INCREASE?

A.

MVAR output.

B.

Frequency.

C.

MWe output.

D.

Power Factor.

Question #54 RO SRO Tier # 2 Group # 2 K/A # 045K5.12 Importance Rating: 2.8 K/A Statement: Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the MAIN TURBINE GENERATOR SYSTEM: Role of field excitation in generator Proposed Answer: A.

Justification: Raising Automatic Voltage control (field excitation) while tied to the grid will increase Main Generator MVARs.

Plausibility:

A. Correct answer B. This would be correct for operation of the Load/Speed Control Switch with the unit off-line.

C. This would be correct for operation of Load/Speed Control Switch with the unit on-line.

D. This would be correct for operation of Automatic Voltage Regulator control switch but only if the unit was operating with a leading power factor which is not permitted.

Technical Reference(s):

OP 2204 Load Changes, Attachment 16 Generator Voltage Adjustment Provided reference(s): None Learning Objective: ILT Given that the Turbine Generator is synchronized to the grid, describe the reaction of the generator to: (MB 02691) [282682]

A) a change in the Main Turbine load demand, or B) a change in AC voltage regulator setpoint.

Source: Bank [412998]

Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics Comments:

Question #55 The plant is operating at 100% power.

The monthly CEA operability surveillance is in progress.

CEA #45 is inserted 5 steps

CEA #45 slips an additional 18 steps.

1. What are the CEAPDS and PPC position indications for CEA #45 AND

2. What is the required response?

A.

1.

CEAPDS and the PPC both indicate 157 steps.

2.

Refer to the ARP for CEA group deviation and realign the CEA in accordance with OP 2302A for CEA operation.

B.

1.

CEAPDS and the PPC both indicate 157 steps.

2.

Enter AOP 2556, CEA Malfunctions, and reduce power to less than 69% prior to realigning the CEA.

C.

1.

CEAPDS indicates 157 steps, the PPC indicates 175 steps.

2.

Refer to the ARP for CEA group deviation and realign the CEA in accordance with OP 2302A for CEA operation.

D.

1.

CEAPDS indicates 157 steps, the PPC indicates 175 steps.

2.

Enter AOP 2556, CEA Malfunctions, and reduce power to less than 69% prior to realigning the CEA.

Question #55 RO SRO Tier # 2 Group # 2 K/A # 014 A2.03 Importance Rating: 4.0 K/A Statement: Ability to (a) predict the impacts of the following on the ROD POSITION INDICATION SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations:

Dropped rod Proposed Answer: D Justification: CEAPDS uses reed switch position indication which would indicate the dropped rods current height while the PPC relies on pulse counting from the ACTM which does not register a drop in rod height. A dropped CEA is defined as a CEA that is >20 steps out of position from any other CEA in its respective group. AOP 2556 requires reactor power be reduced to < 69% prior to realigning the CEA with its group.

Plausibility:

A. Examinee believes that both position indication systems are capable of determining the true rod height. Examinee recalls that CEAs may be realigned without a power reduction but forgets that it only applies for CEA misalignment of <10 steps.

B. Examinee believes that both position indication systems are capable of determining the true rod height.

C. Examinee recalls that CEAs may be realigned without a power reduction but forgets that it only applies for CEA misalignment of <10 steps.

D. Correct response.

Technical Reference(s):

ARP 2590C-095 Group Deviation

ARP 2590C-140 Group Deviation Backup

AOP 2556 CEA Malfunctions Provided reference(s): None Learning Objective: 283050 ILT Outline the major actions for the following CEA malfunctions in accordance with AOP 2556, CEA Malfunctions:

b) Dropped CEA Source: Modified [452093]

Cognitive Level: Comprehension 10CFR55.41(b)(6): Design, components, and functions of reactivity control mechanisms and instrumentation Comments:

Question #56 What is the primary reason for forcing Pressurizer sprays while changing power level?

A.

Minimize stress on Pressurizer spray nozzles by limiting the temperature differential across the nozzle.

B.

Prevent the formation of an explosive hydrogen/oxygen environment in the Pressurizer steam space.

C.

Maintain RCS pressure control function by preventing thermal stratification in the Pressurizer.

D.

Enhance mixing between the RCS and the Pressurizer to equalize their boron concentrations.

Question #56 RO SRO Tier # 2 Group # 1 K/A # 004 K6.01 Importance Rating: 3.1 K/A Statement: Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the CHEMICAL AND VOLUME CONTROL SYSTEM (CVCS): Spray/heater combination in PZR to ensure uniform boron concentration Proposed Answer: D Justification: Raising or lowering power requires dilution or boration. Forcing Pressurizer sprays minimizes the difference in boron concentration between the Pressurizer and the rest of the RCS which results in minimizing a potential power transient as a result of a subsequent Pressurizer in-surge or out-surge.

Plausibility:

A. Plausible, forcing sprays will limit this temperature differential, but this is not why we force sprays.

B. Plausible, forcing sprays does promote good mixing to keep chemistry consistent throughout the RCS and pressurizer and the student may believe this helps improve scavenging of oxygen. However, this is not why we force sprays when moving the plant.

C. Plausible, forcing sprays will easily minimize stratification in the pressurizer, however this is not why we force sprays during a power maneuver.

D. Correct.

Technical Reference(s):

PowerPoint for N04-00-C, 2204 Load Changes, Rev 3 Chg 2, Slide 71 Provided reference(s): None Learning Objective: Describe the actions required to force pressurizer spray including why this evolution is required during load changes (282538).

Source: Bank 413918 Cognitive Level: Memory 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Question #57 The plant is operating at 100% power.

Loop 2 Thot (TI-121Y) input to RRS fails to 520°F.

Which of the following describes the system effect of this failure?

A.

Letdown flow goes to maximum and the B/U heaters energize.

B.

On a plant trip, the Steam Dump/Bypass Valves will NOT modulate open.

C.

Loop 2 THOT (TI-121Y) is automatically removed from the TAVE calculation.

D.

On a plant trip, the Steam Dump/Bypass valves stay open longer than normal.

Question #57 RO SRO Tier # 2 Group # 2 K/A # 011 K6.18 Importance Rating: 3.3 K/A Statement: Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the PRESSURIZER LEVEL CONTROL SYSTEM:

Reactor regulating system Proposed Answer: A Justification: When the Th input fails low RRS calculates an abnormally low Tavg (~551°F). The low Tavg will cause a corresponding drop in Pressurizer level setpoint (~54%). Pressurizer level was originally at the normal 100% power level (65%), therefore letdown flow will go to MAXIMUM in an attempt to bring Pressurizer level to the "new", abnormally low setpoint. B/U heaters will energize due to Pressurizer level is greater than 3.6% above the new programmed level.

Plausibility:

A. Correct Response B. Examinee determines that the Quick Open signal will not be armed with Tavg < 551°F but fails to recognize that the valves will still modulate open based on steam header pressure.

C. Examinee recalls the RRS capability to detect and automatically remove a failed instrument input. This is not the case when the Instrument is still within its normal range.

D. Examinee incorrectly determines that failure results in an artificially high Tavg post trip resulting in prolonged opening of the Steam Dump/Bypass valves.

Technical Reference(s):

OP 2386 RRS

OP 2204 Attachment 5 & 6 Provided reference(s): None Learning Objective: ILT Explain how the Reactor Regulating System calculates the following functions or parameters, including the inputs used and the outputs derived from those inputs: (MB-03171) [282176]

A) Tave, B) Tref, C) Pressurizer Level Setpoint, D) Area Demand (automatic throttling) signal for the Condenser Steam Dump Valves E) Quick Open Signal for the Atmospheric and Condenser Steam Dump Valves.

Source: Bank [449634]

Cognitive Level: Analysis 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #58 A procedure is being developed to test the Power Operated Relief Valves.

What would be classified as an Infrequently Conducted or Complex Evolution (ICCE)?

A.

Raising RCS pressure to 2397 psia to ensure the PORVs open.

B.

Closing a Block Valve and inserting a signal at RPS to open a PORV.

C.

Bench testing the PORVs to ensure proper operation.

D.

Closing a Block Valve and stroking a PORV to verify proper operation.

Question #58 RO SRO Tier # 2 Group # 1 K/A # 010 G2.2.7 Importance Rating: 2.9 K/A Statement: PRESSURIZER PRESSURE CONTROL SYSTEM (PZR PCS) EQUIPMENT CONTROL: Knowledge of the process for conducting infrequently performed tests or evolutions Proposed Answer: A Justification: Raising RCS pressure to test the PORV setpoint would meet the criterion of an ICCE iaw OP-AA-106: Tests designed to verify key features used in the station design basis, such as the time dependence of reactor coolant flow following the loss of reactor coolant pump. Part of the process of conducting an ICCE is the determination that the evolution is one.

Plausibility:

A. Correct.

B. The student may think opening the PORVs with the Block Valves closed may cause the PORVs not to seat properly thus putting the valves at risk for leakage and potentially be declared inoperable.

C. The student may think removing the PORVs to bench test could be considered an infrequently performed evolution requiring an ICCE.

D. The student may think opening the PORVs with the Block Valves closed may cause the PORVs not to seat properly thus putting the valves at risk for leakage and potentially be declared inoperable.

Technical Reference(s):

OP-AA-106, Infrequently Conducted or Complex Evolutions(ICCE), Definitions Provided reference(s): None Learning Objective: 251453 As outlined in OP-AA-106, Infrequently Conducted or Complex Evolutions (ICCE), describe the conditions which require categorizing an evolution an ICCE Source: New Cognitive Level: Comprehension 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #59 The plant is operating at 100% power.

MCC B-13 is lost

1. How many traveling screens are now de-energized?

AND

2. How many Service Water pumps remain OPERABLE?

A.

1. 2

2. 1 B.

1. 2

2. 2 C.

1. 4

2. 0 D.

1. 4

2. 2

Question #59 RO SRO Tier # 2 Group # 1 K/A # 076 K2.09 Importance Rating: 2.6 K/A Statement: Knowledge of electrical power supplies to the following: SERVICE WATER SYSTEM Intake screens Proposed Answer: B Justification: A loss of B-13 will affect one Screenwash pump and 2 Traveling screens (the other Screenwash pump and travelling screens are powered form B-42). Service Water pumps will remain OPERABLE since neither they nor their supporting systems will lose power. Travelling Screens are not required for Service Water pump operation.

Plausibility:

A. May think Service Water requires the traveling screens for OPERABILITY.

B. Correct.

C. May think all screens are powered from 1 MCC. May think Service Water requires the traveling screens for OPERABILITY.

D. May think all screens are powered from 1 MCC.

Technical Reference(s):

OP 2326A, Service Water, Prerequisites

AOP 2503A, Loss of Load Center 22A, B13 Load List Provided reference(s): None Learning Objective: 282312 Given any operating condition for the service Water system, state whether the condition requires entry into Technical Specifications.

Source: New Cognitive Level: Comprehension 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments: Changed answers based on feedback from validator

Question #60 The plant is operating at 30% power.

A SGFP is feeding both steam generators.

B SGFP is secured.

What is the status of the SGFP minimum flow controllers?

1. A SGFP MIN FLOW RECIRC, FIC-5237.

AND

2. B SGFP MIN FLOW RECIRC, FIC-5240.

A.

1. Manual.

2. Manual.

B.

1. Automatic.

2. Manual.

C.

1. Manual.

2. Automatic.

D.

1. Automatic.

2. Automatic.

Question #60 RO SRO Tier # 2 Group # 1 K/A # 059K4.20 Importance Rating: 3.0 K/A Statement: Knowledge of (SF4S MFW) MAIN FEEDWATER SYSTEM design features and/or interlocks that provide for the following (CFR: 41.7): Automatic feed pump recirculation flow Proposed Answer: B Justification: When placing a SGFP in service, its min flow recirc valve is open to roll the pump with condensate, then the feed pump is started with steam. As speed comes up, suction flow is raised to 4200 GPM, then the min flow recirc valve is placed in AUTO to maintain it there. This happens prior to feeding forward with the pump.

When a SGFP is secured, its min flow recirc valve is placed in MANUAL and CLOSED to avoid robbing flow from the running pump.

Plausibility:

A. Plausible, other procedure sections will place the min flow recirc in MANUAL, and when starting the pump, the min flow recirc is not placed in AUTO until near the end.

B. Correct.

C. Plausible, other procedure sections will place the min flow recirc in MANUAL, and when starting the pump, the min flow recirc is not placed in AUTO until near the end.

Additionally, student may not recall that the min flow recirc for the off-service pump is in MANUAL and closed.

D. Plausible, student may not recall that the min flow recirc for the off-service pump is in MANUAL and closed.

Technical Reference(s):

OP 2321, Main Feedwater System Provided reference(s): None Learning Objective: Given a faceplate drawing of the SGFP mini-flow recirc valve controllers (FIC-5240, FIC-5237) label each identified item and describe their operation as given in MFW-01-C. (281654)

Source: New Cognitive Level: Comprehension 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments: Minor modification to make question more generic. DF 12/7

Question #61 The plant is in MODE 6.

CORE ALTERATIONS are in progress.

Two Source Range NIs are OPERABLE.

The Refuel SRO reports that the audible counts have stopped in Containment.

What action(s) is/are required by Technical Specifications?

A.

Only suspend CORE ALTERATIONS.

B.

Only suspend positive reactivity additions to the RCS.

C.

No action required, 1 channel of Source Range NIs remains OPERABLE.

D.

Suspend CORE ALTERATIONS and positive reactivity additions to the RCS.

Question #61 RO SRO Tier # 1 Group # 2 K/A # 032 G2.2.42 Importance Rating: 3.9 K/A Statement: Loss of Source Range Nuclear Instrumentation EQUIPMENT CONTROL: Ability to recognize system parameters that are entry-level conditions for TS Proposed Answer: D Justification: Loss of audible counts in Containment constitutes a loss of one of the two required source range neutron flux monitors, because TS 3.9.2 specifies one of the two must have audible indication in Containment, therefore action a is applicable to suspend core alterations and stop positive reactivity additions.

Plausibility:

A. Student may believe that only suspension of fuel movement is required.

B. Student may believe that only suspension of positive reactivity additions is required.

C. Student may believe that one channel is sufficient to continue the on-load.

D. Correct.

Technical Reference(s):

Millstone Unit 2 Technical Specifications TSAS 3.9.2 Provided reference(s): None Learning Objective: Given a list of plant conditions and/or parameter values for the Nuclear Instrumentation System, and a copy of the Technical Specifications, determine if any LCOs or LSSSs are violated, and identify appropriate action statements. (281835)

Source: Bank Vision #413121 Cognitive Level: Memory 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Question #62 What Main Steam System valves receive a CLOSE signal on a Main Steam Isolation (MSI)?

A.

2-MS-64A/B, MSIVs.

2-MS-65A/B, MSIV Bypass valves.

2-MS-265B/266B, Main Steam Line Drain valves.

B.

2-MS-64A/B, MSIVs.

2-MS-65A/B, MSIV Bypass valves.

2-MS-201/202, Terry Turbine Steam Supply valves.

C.

2-MS-201/202, Terry Turbine Steam Supply valves.

2-MS-220A/B, S/G Blowdown Isolation valves.

2-MS-191A/B, S/G Sample Isolation valves.

D.

2-MS-220A/B, S/G Blowdown Isolation valves.

2-MS-265B/266B, Main Steam Line Drain valves.

2-MS-191A/B, S/G Sample Isolation valves.

Question #62 RO SRO Tier # 2 Group # 1 K/A # 039K4.09 Importance Rating: 2.6 K/A Statement: Knowledge of MAIN AND REHEAT STEAM SYSTEM (MSS) design features and/or interlocks that provide for the following: Main steam line drains Proposed Answer: A Justification: The following valves receive a CLOSE signal on a MSI:

2-MS-64A/B, MSIVs

2-MS-65A/B, MSIV Bypass valves

2-MS-265B/266B, Main Steam Line Drain valves Plausibility:

A. Correct answer B. MS-220A/B receive other auto close signals including CIAS but not MSI.

C. MS-191A/B and MS-220A/B receive close signal on CIAS but not MSI.

D. MS-191A/B receive close signal on CIAS but not MSI.

Technical Reference(s):

ARP 2590A-145, MSI ACTUATION SIG CH 1 TRIP

ARP 2590A-146, MSI ACTUATION SIG CH 2 TRIP Provided reference(s): None Learning Objective: ILT Describe the design feature(s) and/or component(s) of the Main Steam System that provide for Containment Isolation on a Steam Line break accident. (MB 02884) [281721]

Source: Bank [413020]

Cognitive Level: Memory 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #63 The plant has experienced a loss of Instrument Air.

AOP 2563, Loss of Instrument Air has been entered.

Station Air is aligned to supply Instrument Air.

A PEO has been directed to blowdown instrument air receivers hourly to remove moisture.

Why does AOP 2563 direct hourly blowdowns of the Instrument Air receivers when supplied with non-dried air?

A.

Small passages in air control devices are easily fouled from condensate and corrosion products.

B.

Moisture in compressed air may freeze and block lines when air is rapidly decompressed.

C.

Moisture may fill up the air accumulator, leaving no room for compressed air, causing pressure instability.

D.

Condensation can pool in various areas of the instrument air system, leading to water hammer events.

Question #63 RO SRO Tier # 1 Group # 1 K/A # 065 AK1.02 Importance Rating: 3.1 K/A Statement: Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to LOSS OF Instrument Air: Effects of water and/or particulate matter in instrument air lines (operating experience)

Proposed Answer: A Justification: Step 16 of AOP 2563 states, While Non-Dried Air Being Supplied to Instrument Air, BLOWDOWN Instrument Air Receivers Hourly To Remove Moisture. OP 2332B states in the Discussion session, Instrument air dryer(s) must be operating to prevent moisture accumulation in pneumatic controllers. The lesson text for Instrument and Station Air elaborates on this: The air dryer removes moisture from the air and thus prevents condensation, corrosion, and corrosion products from collecting in the Instrument air system.

Air control devices contain many extremely small flow passages and are easily fouled if condensate or corrosion products were allowed to form in the system. Therefore, the importance of maintaining a clean and dry air supply cannot be over stated.

Plausibility:

A. Correct.

B. Plausible, freezing and blockage of lines from rapid decompression of air that is too moist is a known problem in air systems which operate at much higher pressures (3000-4500 psig), this problem contributed to the loss of the USS Thresher.

C. Plausible because student may have had experience blowing down a receiver and getting a lot of water.

D. Plausible because it is a true statement, but it is not an answer to this question because velocities of air in the Instrument Air system are not sufficient for water hammer events.

Technical Reference(s):

AOP 2563, Loss of Instrument Air OP 2332B, Instrument Air System ISA-00-C Rev 9 change 5, Instrument and Station Air lesson text Provided reference(s): None Learning Objective: State the purpose and describe the operating characteristics of the following major Station Air & Instrument Air Systems components: (287677)

D) Air Dryers Source: New Cognitive Level: Memory 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments: Changed distractor C to one less plausible, as validation has shown that experienced ROs frequently chose the original C distractor.

Question #64 The plant tripped due to a SBLOCA.

480V Load Centers are as follows:

Which battery exhaust fan(s) may be started, given available power configuration?

1. Battery Exhaust Fan F-112A 1. be started.

AND

2. Battery Exhaust Fan F-112B 2. be started.

A.

1. can

2. cannot B.

1. can

2. can C.

1. cannot

2. cannot D.

1. cannot

2. can 22A LIVE 22B LIVE 22C DEAD 22D LIVE 22E LIVE 22F DEAD

Question #64 RO SRO Tier # 2 Group # 1 K/A # 063 K2.02 Importance Rating: 2.5 K/A Statement: Knowledge of electrical power supplies to the following: DC ELECTRICAL DISTRIBUTION SYSTEM Battery room ventilation Proposed Answer: A Justification: Battery Exhaust Fans F-112A and F-112B are powered from MCCs B52 and B62, respectively, and therefore the upstream load centers are 22E and 22F, respectively.

Plausibility:

A. Correct.

B. Plausible: Student may believe fans are powered from non-vital load centers 22A/22B.

C. Plausible: Student may believe one fan to be vital and one fan to be non-vital.

D. Plausible: Student may believe fans are powered from non-vital load centers 22C/22D.

Technical Reference(s):

OP 2315B, Non-Radioactive Ventilation System Provided reference(s): None Learning Objective: Describe how the 125 VDC System affects or is affected by the following (287307):

c)

Battery Room Ventilation Source: New Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Question #65 The plant is in MODE 1.

A Technical Specification LCO requires two OPERABLE subsystems (trains).

The LCO is APPLICABLE in MODEs 1, 2 and 3.

Both subsystems (trains) were declared NOT OPERABLE at 0000 TODAY.

TS 3.0.3 has been entered.

1. To what MODE must the plant be placed in?

AND

2. By what time does the plant need to be there?

A. 1. MODE 5.

2. Today 1300.

B. 1. MODE 4.

2. Today 1300.

C. 1. MODE 5.

2. Tomorrow 1300.

D. 1. MODE 4.

2. Tomorrow 1300.

Question #65 RO SRO Tier # 3 Group #

K/A # G2.2.40 Importance Rating: 3.4 K/A Statement: EQUIPMENT CONTROL: Ability to apply TS with action statements of less than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Proposed Answer: B Justification: TS 3.0.3 requires a condition which does not have a defined ACTION to be remedied within one hour or the unit is placed in a mode in which the equipments LCO is not applicable, on the timeline of MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , MODE 4 in the next six, and MODE 5 in the following 24. In the case of the equipment in question, it is required in MODE 3 or higher, therefore the crew must drive to MODE 4, and they have a total of 1+6+6=13 hours to get there.

Plausibility:

A. Plausible: TS 3.0.3 has provisions to go all the way to MODE 5, however, this equipment is not required in MODE 4.

B. Correct.

C. Plausible: TS 3.0.3 has provisions to go all the way to MODE 5, however, this equipment is not required in MODE 4.

D. Plausible: Student may fail to realize that the additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is not available when they are only driving to MODE 4.

Technical Reference(s):

Unit 2 Technical Specifications Provided reference(s): None Learning Objective: As given in the Technical Specifications, state the actions required or permitted: (282392)

A) When a condition exists which is less conservative than a Limiting Condition for Operation and the corresponding ACTION, and B) When conditions are restored which permit operation under the ACTION.

Source: New Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(10): Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments: Rewrote question to more closely align with K/A. SM 12/9

Question #66 The plant is operating at 100% power.

The Main Turbine tripped without a Reactor trip.

The Diverse Scram System (DSS) functioned as designed.

1. Reactor power 1. due to the Turbine trip.

AND

2. DSS opens the CEDS MG Set 2..

A.

1. rose

2. supply breakers B.

1. rose

2. output contactors C.

1. lowered

2. supply breakers D.

1. lowered

2. output contactors

Question #66 RO SRO Tier # 1 Group # 1 K/A # 029 EK1.03 Importance Rating: 4.2 K/A Statement: Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to ANTICIPATED TRANSIENT WITHOUT SCRAM (ATWS): Addition of negative reactivity Proposed Answer: D Justification: When the turbine trips and the Reactor does NOT, RCS temperature will rise due the sudden decrease in heat removal. This will also cause a rise in the Fuel temperature.

The rise in both fuel and moderator temperature will each add negative reactivity causing Reactor power to lower. The DSS is designed to de-energize the CEDMs by an alternative method (from RPS) and cause the insertion of all CEAs. This method is opening the CEDS MG Set Output Contactors.

Plausibility:

A. Incorrect. Reactor power will NOT rise. RCS Pressure will rise and Xenon production will lower slightly inserting a small amount of positive reactivity, but it will be insignificant compared to the negative reactivity inserted due to the RCS temperature rise. Also, the DSS trips the MG set output contactors not the supply breakers. Plausible: The examinee may believe that the positive reactivity inserted by the significant rise in RCS pressure and the lower Xenon production will overshadow the negative Reactivity inserted by the rise in RCS temperature. Additionally, the examinee may believe that the DSS inserts the CEAs by causing a loss of the MG sets.

B. Incorrect. Although the DSS does open the CEDs MG Set Output contactors, reactor power will NOT rise initially. As power is reduced due to the rise in temperature, Xenon production will lower, but will be negligible. RCS Pressure will rise and insert a small amount of positive reactivity, but it will be insignificant. Plausible: The examinee may believe that the positive reactivity inserted by the significant rise in RCS pressure and the lower Xenon production will overshadow the negative Reactivity inserted by the rise in RCS temperature resulting in a rise in Reactor power, which will stop rising when CEAs are inserted.

C. Incorrect. Although power will lower due to the effects of MTC and FTC, the DSS does NOT open the CEDs MG Set Supply breakers. Plausible: The examinee may believe that the DSS trips the MG set supply breaker, which is controlled by a switch just above the CEA control insert on main Control Board C-04.

D. Correct.

Technical Reference(s): Control Element Drive System Lesson Text, CED-01-C, pages 35, 36, revision 6/5.

Provided reference(s): None Learning Objective:

Source: Bank 451831. 2011 NRC Exam Q10.

Cognitive Level: Memory 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #67 The reactor was tripped due to an RCS leak.

Bus 22E de-energized on the trip.

SIAS was initiated after the loss of 22E.

What is the status of the ECCS injection valves?

1. LPSI injection valves are OPEN.

AND

2. HPSI injection valves are OPEN.

A.

1. Four (4)

2. Eight (8)

B.

1. Two (2)

2. Eight (8)

C.

1. Four (4)

2. Four (4)

D.

1. Two (2)

2. Four (4)

Question #67 RO SRO Tier # 2 Group # 1 K/A # 006 A3.06 Importance Rating: 3.7 K/A Statement: Ability to monitor automatic features of the EMERGENCY CORE COOLING SYSTEM (ECCS), including: Valve lineups Proposed Answer: B Justification: The HPSI injection valves are throttled to their open position when performing the HPSI valve lineup. Therefore, the loss of bus 22E will not affect their response to a SIAS signal. The LPSI injection valves are verified closed on their valve lineup and open in response to a SIAS signal. Two LSPI injection valves are powered from bus 22E and therefore will not open in response to a SIAS.

Plausibility:

A. Incorrect: The (4) LPSI valves are not open. The Facility 1 LPSI injection valves 2-SI-615 and 2-SI-625 will not be open because the SIAS occurred after the loss of bus 22E.

Therefore, these MOVs did not have any power to open. Plausible: Examine may not know the power supply to the LPSI injection valves. The second part is correct, all HPSI injection valves are open. Also may think there are (8) LPSI injection valves and half of them opened. Or there were problems during the 2020 refuel outage and one of the possible fixes (which wasnt done) was to leave all LPSI injection valves open like the HPSI injection valves B. Correct: The (2) Facility 2 LPSI injection valves open (since they have power) and all (8)

HPSI injection valves are open (HPSI valves are maintained open) is correct.

C. Incorrect: The (4) LPSI valves are not open. The Facility 1 LPSI injection valves 2-SI-615 and 2-SI-625 will not be open because the SIAS occurred after the loss of bus 22E.

Therefore, these MOVs did not have any power to open. All (8) HPSI injection valves are open. Plausible: Examine may not know the power supply to the LPSI injection valves.

Also may think there are (8) LPSI injection valves and half of them opened. Or there were problems during the 2020 refuel outage and one of the possible fixes (which wasnt done) was to leave all LPSI injection valves open like the HPSI injection valves. May think either half the HPSI injection valves opened or that there are only (4) HPSI injection valves D. Incorrect: All (8) HPSI injection valves are open not just (4). Plausible: Examinee might think only half the HPSI injection valves opened because of the loss of Facility 1 power similar to how the LPSI injection valves respond.

Technical Reference(s): HPSI Lesson Text HPI-00-C and LPSI Lesson Text LPI-00-C.

Provided reference(s): None Learning Objective: ILT State the purpose and describe the operating characteristics of the following major High Pressure Safety Injection System components: b) injection Valves.

Source: Bank 2020 NRC Exam Q33

Cognitive Level: Comprehension 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #68 The plant is at 100% power.

Pressurizer pressure = 2235 psia.

A Power Operated Relief Valve (PORV) is leaking.

Quench Tank pressure = 20 psig.

What is the temperature of the PORV tailpiece?

A.

220 °F.

B.

228 °F.

C.

259 °F.

D.

650 °F.

Question #68 RO SRO Tier # 4 Group #

K/A # 193004 K1.15 Importance Rating: 2.8 K/A Statement: THERMODYNAMIC PROCESS: (THROTTLING AND THE THROTTLING PROCESS) Determine the exit conditions for a throttling process based on the use of steam and/or water Proposed Answer: C Justification: The leakage from the PORV is an isenthalpic process (constant enthalpy). The temperature of the fluid is found by finding the enthalpy at 2235 psia. With saturated steam leaking from the PORV at 2,235 psia, the initial (and final) enthalpy (approximately 1118 Btu/lbm) can be found on the Mollier diagram using the 2,235 psia constant pressure line, or steam table. The constant enthalpy line intersects 20 psig (35 psia) under the dome, therefore, the fluid downstream will be saturated steam. The saturation temperature at 35 psia is 259ºF.

Plausibility:

A. Incorrect. Plausible if the candidate makes the mistake of expanding down to standard atmosphere.

B. Incorrect. Plausible if the candidate doesnt convert 20 psig to 35 psia. This is a common mistake.

C. Correct.

D. Incorrect. Plausible if the candidate doesnt understand that throttling across a valve is a constant enthalpy process and concludes the temperature would not change.

Technical Reference(s): February 2017 GFES exam bank explanation for question P1677.

Provided reference(s): Steam Tables Learning Objective: K1.15 Determine the exit conditions for a throttling process based on the use of stream and/or water.

Source: Bank P1677 (666824)

Cognitive Level: Comprehension / Analysis 10CFR55.41(b)(14): Principles of heat transfer thermodynamics and fluid dynamics Comments: This question was changed slightly to make it a little more plant specific and answer A was changed to make it more plausible.

Question #69 The A Emergency Diesel Generator (EDG) is being placed onto the offsite grid for a surveillance run. The operator will setup the Synchroscope for closing the EDG breaker as follows:

1. With the Synchroscope rotating 1..

AND

2. After the EDG breaker is closed the Synchroscope will be 2..

A.

1. 0.5 - 1 rpm in the FAST direction

2. at top dead center B.

1. 0.5 - 1 rpm in the FAST direction

2. rotating slowly in FAST direction C.

1. 0.5 - 1 rpm in the SLOW direction

2. at top dead center D.

1. 0.5 - 1 rpm in the SLOW direction

2. rotating slowly in SLOW direction

Question #69 RO SRO Tier # 2 Group # 1 K/A # 064 A4.03 Importance Rating: 3.8 K/A Statement: Ability to manually operate and/or monitor the EMERGENCY DIESEL GENERATOR SYSTEM (EDG) in the control room: Synchroscope Proposed Answer: A Justification: Rotating 0.5 - 1 rpm in the FAST direction prior to closing the breaker and at top dead center once the breaker is closed. The FAST direction indicates the EDG is rotating faster than the grid. It is setup like this so that the EDG will pick up some load. If it was operating in the SLOW direction the EDG would be driven by the grid. Once the EDG breaker is closed the EDG and grid are synchronized and the synchroscope will be stopped at top dead center.

Plausibility:

A. Correct.

B. Incorrect. Rotating 0.5 - 1 rpm in the FAST direction prior to closing the breaker is correct.

Rotating slowly in FAST direction after the breaker is closed is not correct. Plausible because the candidate may not understand the operation of the synchroscope.

C. Incorrect. Rotating 0.5 - 1 rpm in the SLOW direction prior to closing the breaker is not correct. At top dead center after the breaker is closed is correct. Plausible because the candidate may not understand the operation of the synchroscope relative to which is the incoming and which is the running indications.

D. Incorrect. Rotating 0.5 - 1 rpm in the SLOW direction prior to closing the breaker is not correct. Rotating slowly in SLOW direction after the breaker is closed is not correct.

Plausible because the candidate may not understand the operation of the synchroscope.

Technical Reference(s): SP 2613A, Diesel Generator Operability Tests, Facility 1, Rev.028, pg 14.

Provided reference(s): None Learning Objective: 281350 Manual start of Diesel Generator A or B from the control room.

Source: New Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #70 The plant is operating at 100% power.

X18C, C RBCCW Heat Exchanger, is tagged out for tube cleaning.

The A RBCCW pump trips (OVERLOAD alarm lit).

What actions are performed per AOP 2585, Immediate Operator Actions?

A.

Align the B RBCCW pump to the A header and start the B RBCCW pump.

B.

Go To AOP 2564, Loss of RBCCW.

C.

RESET and attempt ONE restart of the A RBCCW pump.

D.

Trip the reactor and the affected RCPs.

Question #70 RO SRO Tier # 2 Group # 1 K/A # 003 A2.06 Importance Rating: 3.5 K/A Statement: Ability to (a) predict the impacts of the following on the REACTOR COOLANT PUMP SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: CCWS malfunction Proposed Answer: D Justification: With the C RBCCW Heat exchanger tagged out, the B RBCCW Heat exchanger is in service on the Facility 2 header. When the A pump trips, with the B RBCCW pump cross-ties open to the Facility 2 header the B RBCCW pump CANNOT be aligned to the A header. The A RBCCW pump cannot be restarted so the reactor needs to be tripped and the A and C RCPs need to be secured.

Plausibility:

A. The student might think the B RBCCW pump can still be aligned to the Facility 1 header.

B. The procedure directs the operator to AOP 2564 after the standby pump is started. The student might not think about going to the RNO if the B RBCCW pump cant be started.

C. The student might think that one restart of the A RBCCW pump can be attempted. The RNO allows this if the pump did NOT trip on overload D. Correct.

Technical Reference(s):

AOP 2585, Immediate Operator Actions, Section 12 Provided reference(s): None Learning Objective: 283589 Given a plant condition requiring the use of AOP 2585, Immediate Operator Actions, STATE from memory the immediate operator actions.

Source: New Cognitive Level: Comprehension 10CFR55.41(b)(5): Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics Comments:

Question #71 The plant is operating at 75% power.

No Reactor Protection System (RPS) trip units are in bypass.

A Power Trip Test Interlock (PTTI) is in effect on Ch. B RPS.

The Channel C linear power range drawer fails low

The Channel C NON-OPR light is lit.

How do the Atmospheric Dump Valves (ADVs) and the Condenser Steam Dumps respond?

1. ADVs 1..

AND

2. Condenser Steam Dumps 2..

A.

1. Remain closed

2. Remain closed B.

1. Remain closed

2. Operate only on setpoint C.

1. Operate only on setpoint

2. Operate only on setpoint D.

1. Quick open, then operate on setpoint

2. Quick open, then operate on setpoint

Question #71 RO SRO Tier # 2 Group # 1 K/A # 012 K3.03 Importance Rating: 3.3 K/A Statement: Knowledge of the effect that a loss or malfunction of the REACTOR PROTECTION SYSTEM (RPS) will have on the following systems or system parameters: SDS (Steam Dump System and Turbine Bypass Control)

Proposed Answer:

D.

1. Quick open, then operate on setpoint

2. Quick open, then operate on setpoint Justification: A PTTI on RPS Ch. B causes a trip on Hi Power, TM/LP and LPD. When Ch. C linear power range drawer fails low with a Non-Operate condition a second PTTI is generated and a reactor trip / turbine trip occurs. The quick open signal is armed with TAVE > 554 °F (approximately 60% power). The turbine trip will initiate a quick open signal to the ADVs and Condenser Steam Dumps. Following the quick open, valves will continue to operate based on setpoint.

Plausibility:

A. Examinee determines that the reactor has not tripped and therefore all of the valves remain closed.

B. Based on the lower power level, the examinee determines that quick open is not armed and the condenser steam dumps alone will maintain pressure and temperature post trip.

C. Based on the lower power level, the examinee determines that quick open is not armed and that all valves will respond, but only to setpoint.

D. Correct response.

Technical Reference(s):

OP 2380, RPS and NI Safety Channel Operation Provided reference(s): None Learning Objective:

ILT Describe the conditions that actuate each of the following Reactor Protection system trips or interlocks: (MB 03138) [282144]

A) Power Trip Test Interlock (PTTI)

ILT Describe the function or use of each of the following Reactor Protection System indications:

Y) Non Opr Source: Modified [486607]

Cognitive Level: Comprehension 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #72 The plant experiences a LOCA while operating at 100% power.

The RSST failed to transfer on the trip.

SIAS fully actuated.

Instrument Air pressure = 89 psig and lowering.

1. What action(s) are directed by EOP 2541, Appendix 4A, Reactor Trip Subsequent Actions?

AND

2. Why?

A.

1. Cross-tie Instrument Air with Unit 3.

2. The Instrument Air Compressors trip off-line on a SIAS.

B.

1. Cross-tie Instrument Air with Unit 3.

2. The Instrument Air Compressors could overload the EDGs.

C.

1. Reset the Instrument Air Compressors.

2. The Instrument Air Compressors dont auto-start on a SIAS.

D.

1. Reset the Instrument Air Compressors.

2. The Instrument Air Compressors dont auto-start on a LNP.

Question #72 RO SRO Tier # 2 Group # 1 K/A # 078 K6.01 Importance Rating: 3.4 K/A Statement: Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the Instrument Air System: Air compressors Proposed Answer: B Justification: On a SIAS with an under-voltage condition, the EDGs will be supplying the Vital buses with a full complement of safety equipment operating. EOP 2541, Appendix 4 directs the Instrument Air system be cross-tied with Unit 3 to prevent overloading the EDGs with the IACs Plausibility:

A. The candidate knows air needs to be cross-tied but thinks the IACs trip off-line. The IACs are available to AUTO restart once the reset is pushed.

B. Correct.

C. The candidate thinks the IACs can be reset (they can be) because of the SIAS condition D. The candidate thinks the IACs can be reset (they can be) because of the LNP condition Technical Reference(s):

EOP 2541, Appendix 4A, Reactor Trip Subsequent Actions Provided reference(s): None Learning Objective: 283656 Outline the Subsequent Actions for EOP 2525, Standard Post Trip Actions Source: Bank Vision #414641 Cognitive Level: Comprehension 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #73 When compared to end of cycle (EOC), the thermal conductivity from the fuel pellets to the reactor coolant at beginning of cycle (BOC) is 1. due to 2..

A.

1. larger

2. a higher fuel pellet density B.

1. larger

2. less contamination of fuel rod gas from fission product gases C.

1. smaller

2. a larger gap between the fuel pellets and cladding D.

1. smaller

2. a thinner corrosion layer on the fuel rods surface

Question #73 RO SRO Tier # 4 Group #

K/A # 193009 Importance Rating: 3.3 K/A Statement: (K1.07) CORE THERMAL LIMITS: Describe factors that affect peaking and hot channel factors Proposed Answer: C Justification: Fuel assemblies are designed and installed at beginning of life with some pellet-to-clad gap in the fuel rod. This gap is pressurized with helium gas at about 300-400# to minimize internal clad strain. When a U-235 nucleus fissions, two smaller fission fragments are created. These fission fragments are usually gaseous and tend to make the fuel pellet expand (swell) over core life. Engineering this pellet-to-clad gap allows for the expansion of fuel pellets over core life and reduces fuel clad strain that would occur if this gap were not manufactured in. One byproduct of the reduction in this pellet-to-clad gap over core life is improved thermal conductivity.

Plausibility:

A. Incorrect. Thermal conductivity improves over core life and is therefore not larger at BOC.

Plausible because it is true fuel pellet density would be higher at the BOC and the candidate may not know a gap in fuel pellet to clad is designed in.

B. Incorrect. Thermal conductivity improves over core life and is therefore not larger at BOC.

Plausible because there would be less contamination of the gap fill gas and the candidate may not know a gap in fuel pellet to clad is designed in.

C. Correct.

D. Incorrect. Smaller at BOC is correct but it is not due to a smaller corrosion film on the on the surface of the fuel rods. Plausible because less corrosion film would be better for thermal conductivity.

Technical Reference(s): National Academy for Nuclear Training - Core Thermal Limits -

193009, student guide under Time in Core Effects section pages 17-18, revision 3.2.

Provided reference(s): None Learning Objective:

Source: Bank GFES P2195 (VISION 390467)

Cognitive Level: Memory 10CFR55.41(b)(14): Principles of heat transfer thermodynamics and fluid dynamics Comments: Cleaned up verbiage and formatting based on feedback from validators. DF 12/9

Question #74 The plant is in MODE 2, performing a reactor startup per OP 2202, Reactor Start Up.

The A RCP trips.

1. What MODE is the plant in?

AND

2. What procedure will the crew be in?

A.

1. 3.

2. OP 2202, Reactor Start Up.

B.

1. 3.

2. EOP 2525, Standard Post Trip Actions.

C.

1. 2.

2. EOP 2525, Standard Post Trip Actions.

D.

1. 2.

2. OP 2206 Reactor Shutdown.

Question #74 RO SRO Tier # 1 Group # 1 K/A # 015 G2.2.35 Importance Rating: 3.6 K/A Statement: Reactor Coolant Pump Malfunctions EQUIPMENT CONTROL: Ability to determine TS for mode of operation Proposed Answer: B Justification: A RCP trip will cause a reactor trip on low flow, the plant will be in MODE 3 with all CEAs inserted. The crew would then enter EOP 2525, Standard Post Trip Actions.

Plausibility:

A. The student might think the crew would still be in the Reactor Startup procedure.

B. Correct.

C. Student may not recognize a reactor trip occurred or that where the MODE changes from 2 to 3 (Group 4 at 72 steps).

D. Student may not recognize a reactor trip occurred or that where the MODE changes from 2 to 3 (Group 4 at 72 steps). The student may recognize the issue with the RCP requires the reactor to be shutdown.

Technical Reference(s):

OP 2202, Reactor Startup (ICCE)

Provided reference(s): None Learning Objective: 282466 Given a set of plant conditions, determine Technical Specification applicability during a reactor startup Source: New Cognitive Level: Comprehension or Analysis 10CFR55.41(b)(7): Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features: Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features:

Comments:

Question #75 The plant is operating at 100% power.

D0106, 125 VDC BUS 201A to INVERTER 1 (INV-1) trips open.

What affect will this have on the plant?

A.

2-FW-43A, A Auxiliary Feedwater Regulating Valve, fails open.

B.

ESAS Actuation Cabinet 5 will still process an UV signal in the event of a LNP.

C.

NO Facility 1 Main Steam Isolation components will actuate in the event of a MSI.

D.

The A EDG will start and come up to speed but will not energize Bus 24C.

Question #75 RO SRO Tier # 2 Group # 1 K/A # 062 K1.05 Importance Rating: 4.2 K/A Statement: Knowledge of the physical connections and/or cause and effect relationships between the AC ELECTRICAL DISTRIBUTION SYSTEM and the following systems: Vital AC electrical instrument buses Proposed Answer: B Justification: When breaker D0106 trips open, Vital AC instrument bus VA-10 will be powered from Inverter 5 the Static Switch VS-1. Actuation Cabinet #5 will still have power and process all ESAS signals (including a Loss of Normal Power, LNP).

Plausibility:

A. The student may think VA-10 is de-energized and 2-FW-43A will fail open B. Correct.

C. The student may think that VA-10 is lost and no Facility 1 MSI components will actuate.

D. The student may think the EDG is running unloaded because of a loss of VA-10, DV-10 will cause this phenomenon.

Technical Reference(s):

120VAC/125VDC One Line diagram (LVD-00-C, Figure 4)

Provided reference(s): None Learning Objective: 281638 From memory, draw a one-line diagram of the 120 Volt AC System showing major (components inverters, Static Switches, Manual Bypass switches, electrical panels, transformers and batteries)

Source: Modified Vision #414242 Cognitive Level: Comprehension 10CFR55.41(b)(8): Components, capacity, and functions of emergency systems. Components, capacity, and functions of emergency systems.

Comments:

v t e Letter Sequence Draft Other
EPID:L-2023-OLL-0036, (Open)
Initiation Acceptance... Administration Questions Answers Results Approval... Other: ML23059A226, ML23059A351, ML23065A068, ML23065A079, ML23072A351
MONTHYEAR2023-02-22 [Table View]

ML23059A335

Text

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