ML22136A132

2A-1 Exam Draft Items- Delayed

ML22136A132
Person / Time
Site:	Vogtle
Issue date:	05/16/2022
From:	NRC/RGN-II
To:	Southern Nuclear Operating Co
References
Download: ML22136A132 (13)

Text

ES-401N AP-1000 Examination Outline Form ES-401N-2 Facility: VOGTLE UNITS 3 & 4 Date of Exam: MARCH 2022 Tier Group RO K/A Category Points SRO-Only Points K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G*

Total A2 G*

Total

1.

Emergency and Abnormal Plant Evolutions 1

3 3

4 N/A 4

2 N/A 2

18 4

2 6

2 2

2 1

2 1

1 9

2 2

4 Tier Totals 5

5 5

6 3

3 27 6

4 10

2.

Plant Systems 1

3 2

3 2

2 2

3 3

3 3

2 28 3

2 5

2 1

1 1

1 1

1 1

1 1

1 0

10 2

1 3

Tier Totals 4

3 4

3 3

3 4

4 4

4 2

38 5

3 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

2 3

2 3

2 2

1 2

Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline section (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

2.

The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points, and the SRO-only exam must total 25 points.

3.

Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401N for guidance regarding the elimination of inappropriate K/A statements.

4.

Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.

5.

Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6.

Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7.

The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401N for the applicable K/As.

8.

On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 above does not apply.) Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401N-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As ML22136A132

ES-401N 2

Form ES-401N-2 ES-401N AP-1000 Examination Outline Form ES-401N-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G*

K/A Topic(s)

IR E-0, Reactor Trip or Safeguards Actuation / 1, 2, 3, 4 R

EK 2.06 Unavailability of either the startup feedwater pumps or passive residual heat removal system 3.6 ES-0.1, Reactor Trip Response / 1, 2, 3, 4

ES-1.3, ADS Stage 1-3 Actuation Response / 3 S

EA 2.04 Fuel pool cask loading pit level 2.6 ES-1.4, ADS Stage 4 Actuation Response / 3 R

EK 3.01 Proper operation of ADS Stage 4 3.7 A-313, Uncontrolled Cooldown / 4 R

AK 2.03 Effect of changing feedwater temperature on reactor power (nuclear instrumentation system, calorimetric, and T) 3.7 A-336, Malfunction of Protection and Safety Monitoring System / 7 R

AA 1.14 Reactor trip system 4.0 E-1, Loss-of-Coolant Accident / 2, 3 R

EA 1.16 Containment recirculation cooling system 2.7 A-342, Reactor Coolant Pump Malfunction / 1, 2, 3, 4 S

EA 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation 4.7 A-337, Passive RHR Heat Exchanger Leak / 4 A-343, Loss of Normal Residual Heat Removal / 4 S

AA 2.03 Component cooling water system to normal residual heat removal system heat exchanger flow 2.7 A-317, Loss of Component Cooling Water

/ 8 R

AK 3.02 Stopping all reactor coolant pumps and variable frequency drives 3.4 ES-0.2, Natural Circulation Cooldown / 4 R

EK 1.03 Chemical and volume control system 2.8 FR-S.1, Response to Nuclear Power Generation / 1 R

EA 1.05 Passive residual heat removal system actuation, reset, and/or flow control 3.6 E-3, Steam Generator Tube Rupture / 3 R

EG 2.4.14 Knowledge of general guidelines for emergency and abnormal operating procedures usage 3.8 E-2, Faulted Steam Generator Isolation /

4 R

EK 1.04 Main steam system 3.4 A-301, Rapid Power Reduction / 1 R

AK 3.02 Energizing pressurizer backup heaters 2.6 A-307, DAS Operations at Local Cabinets

/ 7 R

AK 3.02 Operation of diverse actuation system manual enable switch 3.3 FR-C.1, Response to Inadequate Core Cooling / 4 R

EA 2.02 Reactor coolant system subcooling 3.7 A-323, Loss of 6.9-kV, 4,160-V, or 480-V Bus Power / 6 R

AK 2.05 Loss of power to non-Class IE DC and UPS system 2.7 ES-1.1, Passive Safety System Termination / 3 R

EA 2.01 Reactor coolant system pressure, temperature, and pressurizer level 3.2 A-345, Loss of Nuclear Service Water / 4 S

AG 2.2.23 Ability to track TS limiting conditions for operation 4.6 A-329, Loss of Instrument Air / 8 R

AG 2.1.19 Ability to use available indications to evaluate system or component status 3.9 ECA-1.1, Loss-of-Coolant Accident Outside Containment / 3 R

EA 1.02 Diverse actuation system 3.6 FR-H.1, Response to Loss of Heat Sink /

4 S

EA 2.01 Passive residual heat removal system flow 3.7 SDP-1, Response to Loss of RCS Inventory During Shutdown / 2 S

EA 2.04 Normal residual heat removal system flow and/or pump amps 3.3 SDP-2 Response to Loss of RNS During Shutdown / 4 R

EK 1.04 Diverse actuation system (OE related) 3.6 K/A Category Totals:

3 3

4 4

2 4

2 2

Group Point Total:

18/

6

ES-401N 3

Form ES-401N-2 ES-401N AP-1000 Examination Outline Form ES-401N-2 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G*

K/A Topic(s)

IR A-311, Rod Control System Malfunction / 1 R

AK 1.01 Chemical and volume control system 2.8 A-308, Loss of Control Room AC / 8 S

AA 2.02 Nuclear island nonradioactive ventilation system flow 2.8 A-320, Loss of Circulating Water / 8 A-302, Emergency Boration / 1 A-327, Startup Feedwater System Malfunction / 4 A-328, Malfunction of Feedwater Heaters and Extraction Steam / 4 FR-I.1 Response to High Pressurizer Level / 2 A-314, Fuel-Handling Incident / 8 A-304, Steam Generator Tube Leak / 3 A-333, Main Turbine Malfunction / 4 FR-Z.1, Response to High Containment Pressure /

5 SDP-4, Response to Rising Nuclear Flux During Shutdown / 1 SDP-5, Response to RCS Cold Overpressure During Shutdown / 3 SDP-6 Response to Unexpected RCS Temperature Changes During Shutdown / 4 A-306, Evacuation of Control Room / 8 A-318, Condensate System Malfunctions / 4 R

AK 2.02 Deaerator storage tank level outside of normal operating band 2.7 FR-C-2, Response to Degraded Core Cooling / 4 FR-C.3, Response to Saturated Core Cooling / 4 R

EK 3.02 Stopping the reactor coolant pumps 3.2 FR-H.2, Response to Steam Generator Overpressure / 4 S

EG 2.4.17 Knowledge of emergency and abnormal operating procedures terms and definitions 4.3 FR-Z.2, Response to Containment Flooding / 5 FR-Z.3, Response to High Containment Radiation /

9 FR-Z.4, Response to Low Containment Pressure /

5 R

EA 1.01 Passive containment cooling system 2.8 A-332, Turbine Trip Without Reactor Trip / 4 R

AA 2.02 Steam generator level and/or pressure 2.8 ES-1.2, Post LOCA Cooldown and Depressurization / 4 A-321, Loss of Data Display and Processing System / 7 R

AG 2.1.19 Ability to use available indications to evaluate system or component status 3.9 FR-P.1, Response to Imminent Pressurized Thermal Shock Condition / 3 R

EK 1.07 Reactor coolant system (head vents, pressurizer normal spray, and/or heaters) 2.9 A-340, Reactor Coolant Leak / 2 FR-1.2, Response to Low Pressurized Level / 2 FR-1.3, Response to Voids in Reactor Vessel / 2 S

EA 2.03 Chemical and volume control system makeup flow 2.7 A-326, Feedwater System Malfunctions / 4 R

AK 2.03 Main feed line break 2.6

ES-401N 4

Form ES-401N-2 A-331, Loss of Plant DC Power or Batteries / 6 R

AA 1.01 250-V DC switchboards Class 1E 3.2 A-348, Degraded Grid / 6 S

AG2.4.16 Knowledge of emergency and abnormal operating procedures implementation hierarchy and coordination with other support procedures or guidelines, such as operating procedures, abnormal operating procedures, or severe accident management guidelines 4.4 K/A Category Totals:

2 2

1 2

1 1

Group Point Total:

9/4 2 2

ES-401N 5

Form ES-401N-2 ES-401N AP-1000 Examination Outline Form ES-401N-2 Plant SystemsTier 2/Group 1 (RO/SRO)

System Name / Safety Function K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G*

K/A Topic(s)

IR Reactor Coolant / 2, 4 R

S SF2 K 1.04 Containment system SF4 A 2.23 Failure of PZR heaters 3.4 3.6 Steam Generator / 4 R R K 2.08 Main feedwater control valve control power K 3.03 Main steam system 2.5 3.5 Normal Residual Heat Removal /

4 R

K 3.04 Spent fuel pool cooling system 2.8 Passive Residual Heat Removal

/4 R

K 4.05 Passive residual heat removal system flow control 3.6 Passive Core Cooling / 2 R

K 5.03 Failure of reactor coolant pump to trip 4.3 Component Cooling Water / 8 R

K 6.17 Normal residual heat removal system heat exchanger tube leak 3.1 Pressurizer Pressure Control / 3 R

A 1.14 Makeup flow 3.0 Automatic Depressurization / 3 R

R A 2.01 Diverse actuation system G2.2.22 Knowledge of limiting conditions for operations and safety limits 4.0 4.0 Reactor Trip System / 7 R

A 3.13 Steam generator narrow-range water levelhigh-2 reactor trip 4.1 Engineered Safeguards Actuation

/ 2 R

A 4.03 Steamline isolation actuation 4.1 Diverse Actuation / 7 R

R G2.4.25 Knowledge of fire protection procedures A 4.11 In-containment refueling water storage tank injection line valve actuation 3.3 4.0 Passive Containment Cooling / 5 R

S A 4.02 Makeup to passive containment cooling water storage tank from passive containment cooling ancillary water storage tank G2.4.40 Knowledge of SRO Responsibilities in emergency plan implementing procedures 3.0 4.5 Main Steam / 4 R

A 3.05 Steamline isolation actuation 3.5 Main and Startup Feedwater / 4 R

A 2.13 Steam flow 3.0 AC Electrical Distribution / 6 R

A 1.02 Ancillary diesel generator electrical operating parameters 3.0 Class 1E and Non 1E DC and UPS / 6 R

S K 6.11 Non-Class 1E inverter failure (OE related)

A 2.06 Class 1E Inverter failure (OE related) 2.5 3.8 Onsite Standby Power System / 6 R

R K 5.03 Number of diesel starts from the available volume of starting air A 3.01 Standby diesel generator starting and loading 3.2 3.2 Service Water / 4 R

K 4.06 Service water system tower makeup 2.7 Compressed Air / 8 R

R K 3.20 Service water system (OE related)

A 2.14 Containment isolation 2.5 3.5 Containment System / 5 R

R K 1.12 Class 1E DC and UPS system A 1.04 Containment water level 2.9 3.9

ES-401N 6

Form ES-401N-2 Reactor Coolant Pump / 4 R

S K 1.06 Steam generator system G2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator 2.7 3.8 Chemical and Volume Control / 1, 2

R S

SF1 K 2.04 Containment isolation valves SF2 A 2.15 Mixed bed demineralizer 3.7 2.6 K/A Category Point Totals:

3 2

3 2

2 2

3 3

3 3

3 2

2 Group Point Total:

28/

5

ES-401N 7

Form ES-401N-2 ES-401N AP-1000 Examination Outline Form ES-401N-2 Plant SystemsTier 2/Group 2 (RO/SRO)

System Name / Safety Function K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G*

K/A Topic(s)

IR Digital Rod Control / 1 R

K 1.10 Reactor trip system 4.0 Pressurizer Level Control / 2 R

K 2.03 Pressurizer level channels 3.0 Rod Position Indication / 1 S

A 2.11 Failed digital rod position indication coil 3.2 In-Core Instrument System / 7 R

K 3.04 Protection and safety monitoring system 3.6 Containment Air Filtration / 8 R

K 4.05 Containment pressure control during normal operation 2.8 Containment Hydrogen Control /

5 S

G2.1.25 Ability to interpret reference materials, such as graphs, curves, and tables 4.2 Main Control Room HVAC / 8 Spent Fuel Pool Cooling / 8 R

K 5.01 P between containment and fuel handling buildings 2.8 Condensate / 4 R

K 6.06 Condenser air removal system 2.5 Condenser Air Removal / 4 Main Turbine and Main Turbine Control / 4 Fuel Handling / 8 R

A 2.16 Dropped or damaged fuel assembly 3.7 Gaseous Radwaste / 9 R

A 1.02 Carbon bed and/or vault temperatures 2.5 Radiation Monitoring / 7 S

A 2.02 Reactor coolant system leakage into containment 3.8 Circulating Water / 8 Fire Protection / 8 Steam Dump Control System / 4 R

A 3.02 Load rejection control in Tavg mode 3.1 Nuclear Instrumentation System /

7 R

A 4.03 Block power range high neutron flux low setpoint reactor trip 3.7 Liquid Radwaste System / 9 K/A Category Point Totals:

1 1

1 1

1 1

1 1

1 1

0 Group Point Total:

10/3 2 1

ES-401N Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401N-3 Facility: VOGTLE AP1000 UNITS 3 & 4 Date of Exam: MARCH 2022 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.15 Admin Reqts for temp mgmt direction 2.7 2.1.29 How to conduct system lineups, vlvs, bkrs, switches 4.1 2.1.

2.1.

2.1.20 Interpret/execute procedure steps 4.6 2.1.41 Knowledge of refueling process 3.7 Subtotal 2

2

2. Equipment Control 2.2.2 Manipulate console controls between SD & power levels 4.6 2.2.13 Tagging & clearance procedures 4.1 2.2.44 Interpret CR indications to verify status/operation 4.2 2.2.

2.2.23 Track TS LCOs 4.6 2.2.36 Effect of maint activities (power sources) on LCOs 4.2 Subtotal 3

2

3. Radiation Control 2.3.5 Use radiation monitoring systems 2.9 2.3.11 Control rad releases 3.8 2.3.

2.3.

2.3.

2.3.6 Approve liquid or gas release permits 3.8 Subtotal 2

1

4. Emergency Procedures/Plan 2.4.19 EOP & AOP layout, symbols, icons 3.4 2.4.22 Bases for prioritizing safety functions..abnormal/emerg 3.6 2.4.42 Emergency response facilities 2.6 2.4.

2.4.3 Identify post accident instrumentation 3.9 2.4.37 Lines of authority during EPs 4.1 Subtotal 3

2 Tier 3 Point Total 10 10 7

7

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Vogtle Units 3&4 Date of Examination:

March 2022 Examination Level:

RO

SRO

Operating Test Number:

Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations R, M Determine if watch stander can relieve the watch.

G2.1.5 RO 2.9 Conduct of Operations R, N Given that 2 rods did not fully insert during a reactor trip; determine the volume to be added to the RCS and time needed to complete the boration in accordance with applicable procedures.

G2.1.20 RO 4.6 Equipment Control R, D Review completed surveillance ECS-OTS-17-001 for meeting acceptance criteria.

G2.2.12 RO 3.7 Radiation Control R, D, P Determine best route to limit personnel exposure and not to exceed exposure limits.

G 2.3.12 RO 3.2 Emergency Plan NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

• Type Codes and Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from EDQN 3 for ROs; 4 for SROs and RO retakes)

(N)ew or (M)odified from bank  1)

(P)revious 2 exams  1, randomly selected)

ES-301, Page 22 of 33

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Vogtle Units 3&4 Date of Examination:

March 2022 Examination Level:

RO

SRO

Operating Test Number:

Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations R, M Determine if watch stander can relieve the watch and determine actions required for call-out during time off.

G2.1.5 SRO 3.9 Conduct of Operations R, N Determine requirements for Reactivity Management SRO.

G2.1.37 SRO 4.6 Equipment Control R, D, P Review completed surveillance PMS-OTS-017-001.

Determine TS requirements.

G 2.2.12 SRO 4.1 Radiation Control R, M Determine ODCM actions for failed VFS Rad Detector and Flow Detector.

G 2.3.11 SRO 4.3 Emergency Plan R, N Initial PAR Development.

G2.4.44 SRO 4.4 NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

• Type Codes and Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from EDQN 3 for ROs; 4 for SROs and RO retakes)

(N)ew or (M)odified from bank  1)

(P)revious 2 exams  1, randomly selected)

ES-301, Page 22 of 33

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Vogtle Units 3&4 Date of Examination:

March 2022 Exam Level: RO

SRO-I

SRO-U

Operating Test No.: NRC 2022 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function a.

Perform AOP-104 (Rod Control Malfunction) Attachment 5

==

Description:==

Rx Startup in progress, SD 4 Group 1 and 2 rods are at 267 steps. The operator uses AOP-104 Attachment 5 (ROD POSITION WITHDRAWAL ABOVE ARO POISTION) to restore SD 4 Group 1 and 2 rods to 264 steps.

SF1 A-311 AA2.02 RO/SRO 2.8 L, N, S 1

b.

Perform AOP-112 (Reactor Coolant Leak

==

Description:==

The operator is directed to perform AOP-112 (Reactor Coolant Leak) due to an uncontrolled reduction in RCS inventory. Initial RCS leak rate is less than 100 gpm. CVS M/U Pump A will not start, operator will have to start CVS M/U Pump B. RCS leak rate rises to 300 gpm, and operator will perform AOP-112 RNO Step 3 to Trip the Reactor and perform E-0.

SF2 A-340 AA2.04 RO/SRO 3.4 A, D, EN, S, L 2

c.

Perform AOP-103 (Steam Generator Tube Leak) to reduce RCS pressure

==

Description:==

The unit has been shutdown due to a steam generator tube leak. The operator is directed to perform AOP-103 starting with Step 34.

While attempting to lower pressure to less than 1920 psig, RCP spray valve fails to open so Auxiliary spray must be used to lower pressure.

SF3 A-304 AA1.01 RO/SRO 2.8 A, EN, L, N, S 3

(SRO-U) d.

Perform FR-H.4 (Loss of Normal Steam Release Capabilities)

==

Description:==

The unit has tripped following a manual Reactor trip. The operator has been directed to perform FR-H.4. The operator will be unable to open the steamline drain valves in Step 2 and will be forced to depressurize the affected SG using its associated PORV.

SF4P SGS A4.07 RO/SRO 3.7 A, N, EN, S 4P (SRO-U) e.

Perform AOP-207 (Loss of Turbine Load)

==

Description:==

The Unit is in Mode 1 at 50% Reactor Power and the Main Generator output breaker spuriously trips and the turbine failed to trip.

The operator is directed to perform AOP-207. The operator must trip the turbine manually at the PDSP. Once the turbine is tripped, after a slight delay, a Main Steam Safety Valve will lift which will require the operator to trip the Reactor using guidance from previous CA Step 5.

SF4S A-332 AA1.07 RO/SRO 3.8 A, D, P, S 4S (SRO-U)

f.

Perform AOP-702 (Loss of Component Cooling Water)

==

Description:==

The Unit is at 75% Reactor Power with CCS surge tank level lowering. The operator is directed to perform AOP-702 to address the lowering surge tank level. Procedure steps are unsuccessful requiring the operator to perform previous CA Step 1 to trip the reactor, secure RCPs, and secure Both CCS Pumps.

SF8 A-317 AA1.01 RO/SRO 3.4 A, N, S 8

g.

Respond to a loss of reactive load control in accordance with AOP-301 (Main Generator Malfunction)

==

Description:==

The unit is at 75% Reactor Power. The operator has been directed to perform AOP-301 beginning with Step 10 due to reactive load changing without operator action. When RNO step 10 is unsuccessful using SP control, manual voltage control, and reducing MWe, the operator will trip the turbine.

SF6 ECS K6.08 RO/SRO 3.1 A, D, S 6

h.

Pump the Containment Sump to Waste Holdup Tanks

==

Description:==

The Unit is operating in MODE 5. The candidate is directed to perform 3-WLS-SOP-001 (Liquid Radwaste System) (Containment Sump Operations) Section 4.3, Manual Transfer of Containment Sump, to lower containment sump level to the bottom of the indicating range.

Pre-JPM brief required.

SF9 WLS A 4.01 RO/SRO 2.6 D, S 9

(RO ONLY)

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U i.

Perform AOP-602 (DAS Operations at Local Cabinets)

==

Description:==

The MCR has been evacuated to the RSR. Control from the RSR could NOT be established. The operator is directed to perform AOP-602 and must trip the Reactor, actuate CMTs, actuate PRHR, and actuate Containment Isolation from the local DAS processor cabinets.

SF7 DAS A2.02 RO/SRO 3.3 E, N, R 7

(SRO-U) j.

Perform local actions of FR-P.2 (Response to Anticipated Pressurized Thermal Shock)

==

Description:==

The operator is directed to perform local actions of FR-P.2 Step 20.b to energize Accumulator Discharge MOVs to allow isolating accumulators.

SF3 FR-P.1 EA1.14 RO/SRO 3.2 E, N, R 3

k.

Perform local actions of AOP-702 (Loss of CCS)

==

Description:==

The operator is directed to perform the local actions of AOP-702 Step 10 d. RNO to locally isolate RNS HX B.

SF8 A-317 AA1.07 RO/SRO 2.9 E, N, R 8

(SRO-U)

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for R /SRO-I/SRO-U

(A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3

  4

  1

 (control room system)

  1

  1

  2 (randomly selected)

  1 ES-301, Page 23 of 33