ML103490597

Final Outlines (Folder 3)

ML103490597
Person / Time
Site:	Seabrook
Issue date:	11/12/2010
From:	Momm E NextEra Energy Seabrook
To:	David Silk Operations Branch I
Hansell S
Shared Package
ML101870654	List: ML101940088 ML103190314 ML103490482 ML103490486 ML103490487 ML103490490 ML103490597 ML103490620 ML103490635 ML103490648
References
TAC U01694
Download: ML103490597 (28)

Text

ES-401 PWR Examination Outline Form ES-401-2 I Facility: SEABROOK RO Section Date of Exam: December 2010 RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

• Total 1 1 2 2 5 4 2 3 18 6 Emergency &

Abnormal 2 1 1 3 N/A 1 2 N/A 1 9 4 Plant Evolutions Tier Totals 3 3 8 5 4 27 10 1 5 2 2 6 I 2 1 2 4 2 1 28 5 2.

Plant 2 1 0 1 1 1 0 2 2 I 0 I 10 3 Systems Tier Totals 6 2 3 7 2 2 3 4 5 2 2 38 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 2 2 3 Note: 1. Ensure that at least two topics from every applicable KiA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KiA category shall not be less than two).

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 associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KiA 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 KiAs 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 KiA categories.

7.* The generiC (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the KiA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KiAs.

8. On the following pages, enter the KiA numbers, a brief description of each topic, the topics' importance ratings (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 does not apply). Use duplicate pages for RO and SRO-only exams.

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

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 EmerQencil and Abnormal Plant Evolutions - Tier 1/Group 1 (RO 1 SRO)

E/APE # 1 Name I Safety Function K K K A A G KIA Topic(s) IR #

1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor X EK2 Knowledge of the interrelationship between the 2.6 I Trip Stabilization - Recovery 11 reactor trip and the following:

EK2.02 Breakers, relays and disconnects.

CFR41.7/45.7 000008 Pressurizer Vapor Space Accident I 3 000009 Small Break LOCA 1 3 X EA2 Ability to determine or interpret the following as 4.2 2 they apply to a small break LOCA:

EA2.37 Existence of adequate natural circulation.

CFR 43.5/45.13 000011 Large Break LOCA I 3 X EK3 Knowledge of the reasons for the following 4.1 3 responses as they apply to the large break LOCA:

EK3.14 RCP tripping requirement.

i 41.5/41.10/45.6/45.13 i 000015/17 RCP Malfunctions 1 4 000022 Loss of Rx Coolant Makeup I 2 000025 Loss of RHR System 1 4 000026 Loss of Component Cooling X AK3 Knowledge of the reasons for the following 3.6 4 Water 18 responses as they apply to the Loss of Component Cooling Water:

AK3.02 The automatic actions (alignments) within the CCWS resulting from the actuation of the ESFAS.

CFR41.7/45.7 000027 Pressurizer Pressure Control X 2.1.7 Ability to evaluate plant performance and 4.4 5 System Malfunction 1 3 make operational judgments based on operating characteristics, reactor behavior and instrument interpretation.

CFR 41.5/43.5/45.12/45.13 000029 A TWS 1 1 X EA 1 Ability to operate and monitor the following as 4.2 6 they apply to an ATWS:

I EA 1.14 Driving of Control Rods into the core.

CFR 41.7/45.5/45.6 EK3 Knowledge of the reasons for the following 000038 Steam Gen. Tube Rupture 1 3 X responses as they apply to the SGTR: 4.2 16 EK3.06 Actions contained in EOP for RCS water inventory balance, SIG tube rupture, and plant shutdown procedures.

CFR 41.5/41.10/45.6/45.13

1 PWR Exat;!!inatiqn Outline ,,... Form ES-401-2 1 (~() / SRO)

E/APE # 1 Name 1 Safety Function K K K A A G KIA Topic(s) IR #

1 2 3 1 2 000040 (BW/E05; CE/E05; W/E12) X AA1 Ability to operate and/or monitor the following 4.1 7 Steam Line Rupture - Excessive Heat as they apply to Steam Line Rupture:

Transfer 14 AA1.20 Containment pressure and temperature trends.

CFR 41.7/45.5/45.6 000054 (CE/E06) Loss of Main X 2.4.21 Knowledge of the parameters and logic used 4.0 8 Feedwater 1 4 to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, reactivity release control, etc.

CFR 41.7/43.5/45.12 000055 Station Blackout 1 6 X EK3 Knowledge of the reasons for the following 4.3 9 responses as they apply to the Station Blackout:

EK3.02 Actions contained in EOP for loss of offsite and onsite power.

41.5/41.10/45.6/45.13 000056 Loss of Off-site Power 16 X AA1 Ability to operate andlor monitor the following 3.3 10 as they apply to the Loss of Offsite Power:

AA1.21 Reset of ESF load sequencers.

CFR 41.7/45.5/45.6 000057 Loss of Vital AC Inst. Bus 16 X AA2 Ability to determine and interpret the following 3.8 11 as they apply to the Loss of Vital AC Instrument Bus:

AA2.15 That a loss of ac has occurred.

CFR 43.5/45.13 000058 Loss of DC Power 1 6 X AK3 Knowledge of the reasons for the following 4.0 12 responses as they apply to the Loss of DC Power:

i AK3.02 Actions contained in EOP for loss of dc power.

CFR 41.5/41.10/45.6/45.1 000062 Loss of Nuclear Svc Water 1 4 X AA 1 Ability to operate and/or monitor the following 3.2 13

! as they apply to the Loss of Nuclear Service Water:

AA1.02 Loads on the SWS in the control room.

CFR 41.7/45.5/45.6 000065 Loss of Instrument Air 18 X 2.1.32 Ability to explain and apply system limits and 3,8 14 precautions.

CFR 41.10/43.2/45,12

ES-401 PWR Examination Outline Form ES-401 Emergencli and Abnormal Plant Evolutions - Tier 1/Group 1 (RO 1 SRO)

E/APE # 1 Name 1 Safety Function K K K A A G KIA Topic(s) IR #

1 2 3 1 2 W/E04 LOCA Outside Containment 1 3 X EK1 Knowledge of the operational implications of 3.5 15 the following concepts as they apply to the LOCA Outside Containment:

EK1.3 Annunciators and conditions indicating signals, and remedial actions associated with the LOCA Outside Containment CFR 41.8/41.10/45.3 W/E11 Loss of Emergency Coolant I Recirc./4 BW/E04; W/E05 Inadequate Heat X EK2 Knowledge of the interrelations between the 3.9 17 Transfer - Loss of Secondary Heat Sink 14 Loss of Secondary Heat Sink and the following:

EK2.2 Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operations of these systems to the operation of the facility.

CFR 41.7/45.7 000077 Generator Voltage and Electric X AK1 Knowledge of the operational implications of 3.3 18 Grid Disturbances 16 the following concepts as they apply to Generator Voltage and Electrical Grid Disturbances:

I AK1.03 Under-excitation.

CFR 41.4/41.5/41.7/41.10/45.8 KIA Category Totals: 2 J I: 181 6

ES-401 3 Form ES-401-2 1 PWR Examination Outline Form ES-401-2 E/APE # I Name I Safety Function K K K A A G KJA Topic(s) IR #

1 2 3 I 2 000001 Continuous Rod Withdrawal/1 000003 Dropped Control Rod I 1 000005 Inoperable/Stuck Control Rod 11 X AK1 Knowledge of the operational 3.1 19 implications of the following concepts as they apply to Inoperable/Stuck Control i Rod:

AK1.02 Flux tilt CFR 41.8/41.10/45.3 000024 Emergency Boration I 1 000028 Pressurizer Level Malfunction 12 X AA1 Ability to operate and/or monitor the 3.3 20 following as they apply to the Pressurizer Level Control Malfunctions:

AA1.07 Charging pump maintenance of PZR level (including manual backup).

CFR 41.7/45.5145.6 000032 Loss of Source Range Nil 7 X AA2 Ability to determine and interpret 3.1 21 the following as they apply to Loss of Source Range Nuclear Instrumentation:

AA2.04 Satisfactory source-range/intermediate-range overlap.

CFR 43.5/45.13 000033 Loss of Intermediate Range NI/7 000036 (BW/A08) Fuel Handling Accident I 8 000037 Steam Generator Tube Leak 1 3 X AK3 Knowledge of the reasons for the 3.7 22 following responses as they apply to the Steam Generator Tube Leak:

AK3.05 Actions contained in procedures for radiation monitoring, RCS water level inventory balance, S/G tube failure, and plant shutdown.

CFR 41.5/41.10/45.6/45.13 000051 Loss of Condenser Vacuum 14 X 2.4.11 Knowledge of abnormal condition 4.0 23 procedures.

i CFR 41.10/43.5/45.13 i 000059 Accidental Liquid RadWaste ReI. 19 000060 Accidental Gaseous Radwaste ReI. I 9 000061 ARM System Alarms 17 000067 Plant Fire On-site I 8 000068 (BW/A06) Control Room Evac. I 8 000069 (W/E14) Loss of CTMT Integrity 15

F1 m

PWR Examination Outline

\I E/APE # 1 Name 1 Safety Function r.

I K K 2 3 A

I A

2 G KiA Topic(s) 000074 (W/E06&E07) Inad. Core Cooling 14 X (E06) EA2 Ability to determine and 3.4 24 interpret the following as they apply to the Degraded Core Cooling:

EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency conditions.

CFR 43.5/45.13 000076 High Reactor Coolant Activity 19 W/E01 & E02 Rediagnosis & 81 Termination 1 3 X (E02) EK2 Knowledge of the 3.5 25 interrelations between the SI Termination and the following:

EK2.2 Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

CFR 41.7/45.7 W/E13 Steam Generator Over-pressure 14 X EK3 Knowledge of the reasons for the 3.2 26 following responses as they apply to the Steam Generator Overpressure:

EK3.3 Manipulation of controls required to obtain desired operating results during abnormal and emergency situations.

CFR 41.5/41.10/45.6/45.13 W/E15 Containment Flooding 1 5 i

I WlE16 High Containment Radiation 19 BW/A01 Plant Runback 11 BW/A02&A03 Loss of NNI-x/Y 1 7 BW/A04 Turbine Trip 14 BW/A05 Emergency Diesel Actuation 1 6 BW/A07 Flooding 18 BW/E03 Inadequate Subcooling Margin 14 BW/E08; W/E03 LOCA Cooldown - Depress. 14 BW/E09; CE/A13; W/E09&E10 Natural Circ. 14 BW/E13&E14 EOP Rules and Enclosures CE/A 11; W/E08 RCS Overcooling PTS/4 X EK3 Knowledge of the reasons for the 3.6 27 following responses as they apply to the Pressurized Thermal Shock:

EK3.2 Normal, abnormal and emergency operating procedures associated with Pressurized Thermal Shock.

CFR 41.5/41.10/45.6/45.13

ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1!Group 2 (RO! SRO)

E!APE # I Name! Safety Function K K K A A G KJA Topic(s) IR #

J 2 3 1 2 CE/A 16 Excess RCS Leakage! 2 i CE/E09 Functional Recovery KJA Category Point Totals: I c.::J.:ili1 2 I Group Point Total:

53

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant S stems - Tier 2/Group 1 (RO / SRO)

System # I Name K K K K K K A A A A G KIA Topic(s) IR #

I 2 3 4 5 6 I 2 3 4 003 Reactor Coolant Pump X A2 Ability to (a) predict the impacts of 2.5 28 the following malfunctions or operations on the RCP's and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2,05 Effects ofVCT pressure on RCP sealleakoff flows.

CFR 41,5/43.5/45,3/45,13 004 Chemical and Volume X X K1 Knowledge of the physical 2.7 29 Control connections andlor cause effect relationships between the CVCS and the following systems:

K 1.26 Flow path from CVCS to reactor coolant drain tank and holdup tank, CFR 41.2 to 41,9/45,7 to 45,8 3.0 30 K4 Knowledge of the CVCS design feature(s) and/or interlock(s) which provide for the following:

K4,07 Water supplies CFR41.7 005 Residual Heat Removal X X A 1 Ability to predict and/or monitor 3.3 31 changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including:

A 1,02 RHR flow rate CFR 41.5/45.5 4.0 32 2.2.22 Knowledge of limiting conditions for operations and safety limits.

CFR 41.5/43.2/45.2 006 Emergency Core Cooling X A4 Ability to manually operate and/or 4.4 33 monitor in the control room:

A4.07 ECCS pumps and valves.

CFR 41.7/45.5 to 45.8 007 Pressurizer Relief/Quench X A3 Ability to monitor automatic 2.7 34 Tank operation of the PRTS, including:

A3.01 Components which discharge to the PRT.

CFR 41.7/45.5

I ES-401 PWR Examination Outline Form E'"

Plant S stems - Tier 21Group 1 (RO / SRO)

System # / Name K K K K K Klcl A AAG KiA Topic(s) IR #

1 2 3 4 5 6 2 3 4 008 Component Cooling Water X K2 Knowledge of the bus power 3.0 35 supplies to the following:

K2.02 CCW pump, including emergency backup.

CFR 41.7 010 Pressurizer Pressure Control X X K3 Knowledge of the effect that a loss 3.8 36 or malfunction of the PZR PCS will have on the following:

K3.01 RCS CFR 41.7/45.6 3.6 37 A4 Ability to manually operate and/or monitor in the control room:

A4.02 PZR heaters.

CFR 41.7/45.5 to 45.8 012 Reactor Protection X X K4 Knowledge of the RPS design 3.2 38 feature(s) and/or interlocks(S) which provide for the following:

K4.0S Automatic or manual enable/disable of reactor trips.

CFR41.7 K6 Knowledge of the effect that a loss 3.3 39 or malfunction of the following will have on the RPS:

KS.10 Permissive Circuits CFR 41.7/45.7 013 Engineered Safety Features X X K5 Knowledge of the operational 2.9 40 Actuation implications of the following concepts as they apply to the ESFAS:

K5.02 Safety system logic and reliability.

CFR 41.5/45.7 K6 Knowledge of the effect that a loss 2.7 41 or malfunction of the following will have on the ESFAS:

K6.01 Sensors and detectors.

CFR 41.7/45.5 to 45.8 022 Containment Cooling X K4 Knowledge of the CCS design 2.8 42 features(s) andlor interlock(s) which provide for the following:

K4.04 Cooling of control rod drive motors.

CFR 41.7 025 Ice Condenser I I

ES-401 PWR Examination Outline II

ll Pla~ms - Tier 2/GroulJ 1 (RO I SRO)

System # 1 Name K K KK AA AAG KIA Topic(s) IR I 2 3 4 5 6 I 2 3 4 026 Containment Spray X A3 Ability to monitor automatic 4.3 43 operation of the CCS, including:

A3.01 Pump starts and correct MOV positioning.

CFR 41.7/45.5 039 Main and Reheat Steam X K4 Knowledge of the MRSS design 3.7 44 features and/or interlocks which provide for the following:

K4.05 Automatic isolation of steam line.

CFR 41.7 059 Main Feedwater X K1 Knowledge of the physical 3.4 45 connections andlor cause effect relationships between the MFW and the following systems:

K1.04 S/Gs water level control system.

CFR 41.2 to 41.9/45.7 to 45.8 061 Auxiliary/Emergency X X K2 Knowledge of the bus power 3.2 46 Feedwater supplies to the following:

K2.01 AFW system MOV's.

CFR41.7 A3 Ability to monitor automatic 40 47 operation of the AFW, including:

A3.02 RCS cooldown during AFW operations.

CFR 41.7/45.5 062 AC Electrical Distribution X K1 Knowledge ofthe physical 4.1 48 connections and/or cause-effect relationships between the ac distribution system and the following systems:

K1.02 ED/G CFR41.2to41.9 063 DC Electrical Distribution X K4 Knowledge of DC electrical system 2.9 49 design feature(s) and/or interlock(s) which provide for the following:

K4.02 Breaker interlocks, permissives, bypasses and cross-ties.

CFR41.7

ES-401 PWR Examination Outline Plant S stems - Tier 2/Group 1 (RO I SRO)

System # I Name K K K K K K A A A A G K/A Topic(s) IR #

I 2 3 4 5 6 1 2 3 4 064 Emergency Diesel Generator X X K 1 Knowledge of the physical 3.1 50 connections and/or cause effect relationships between the EDG system and the following systems:

K1.02 D/G cooling water system.

CFR 41.2 to 41.9/45.7 to 45.8 A3 Ability to monitor automatic 3.6 51 operation of the EDG system, including:

A3.07 Load sequencing.

CFR 41.7/45.5 073 Process Radiation X K3 Knowledge of the effect that a loss 3.6 52 Monitoring or malfunction of the PRM system will have on the following:

K3.01 Radioactive effluent releases.

CFR 41.7/45.6 076 Service Water X K4 Knowledge ofthe SWS design 2.5 53 feature(s) andlor interlock(s) which provide for the following:

K4.01 Conditions initiating automatic closure of closed COOling water auxiliary building header supply and return valves.

CFR 41.7 078 Instrument Air X K1 Knowledge of the physical 2.7 54 connections and/or cause-effect relationships between the lAS and the following systems:

K1.02 Service air CFR 1.2 to 41.9/45.7 to 45.8 103 Containment X A2 Ability to (a) predict the impacts of 3.5 55 the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.03 Phase A and B isolation.

CFR 41.5/43.5/45.3/45.13 I KIA Category Point Totals: 5 2 2 6 1 2 1 2 4 2 1 Group Point Total:

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form E Plant Systems - Tier 21Gro JD 2 RO 1 SRO)

System # 1 Name K K K K K K A A A A G KIA Topic(s) IR #

1 2 3 4 5 6 I 2 3 4 001 Control Rod Drive X K3 Knowledge of the effect that a loss or 3.4 56 malfunction of the CRDS will have on the following:

K3.02 RCS CFR 41.7/45.6 002 Reactor Coolant 011 Pressurizer Level Control X A3 Ability to monitor automatic operation 3.2 57 of the PZR LCS, including:

A3.03 Charging and Letdown CFR 41.7/45.5 014 Rod Position Indication X A 1 Ability to predict and/or monitor 3.2 58 changes in parameters (to prevent exceeding design limits) associated with operating the RPIS controls, including:

A 1.02 Control rod position indication on control room panels.

CFR 41.5/45.5 015 Nuclear Instrumentation X 2.4.4 Ability to recognize abnormal 4.5 59 indications from system operating parameters that are entry level conditions for emergency and abnormal operating procedures.

CFR 41.10/43.2/45.6 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge X A 1 Ability to predict and/or monitor 3.0 60 changes in parameters (to prevent exceeding design limits) associated with operating the Containment Purge System controls including:

A 1.03 Containment pressure, temperature, and humidity.

CFR 41.5/45.5 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment I I I 035 Steam Generator

• ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 RO 1 SRO)

System # 1 Name K K K K K K A A A A G KIA Topic(s) IR #

I 2 3 4 5 6 I 2 3 4 041 Steam OumplTurbine X K4 Knowledge of the SOS design 3.0 61 Bypass Control feature(s) andlor interlock(s) which provide for the following:

K4.09 Relationship of low/low Tavg setpoin! in SOS to primary cooldown.

CFR41.7 045 Main Turbine Generator X K5 Knowledge of the operational 2.5 62 implications of the following concepts as they apply to the MTfG System:

K5.17 Relationship between MTC and boron concentration in RCS as turbine load increases.

CFR 41.5/45.7 055 Condenser Air Removal X K1 Knowledge of the physical 2.6 63 connections andlor cause effect relationship between the CARS and the following:

K1.06 PRM system 056 Condensate X A2 Ability to (a) predict the impacts of the 2.6 64 following malfunctions or operations on the condensate system; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations:

A2.04 Loss of condensate pumps.

CFR 41.5/43.5/45.3/45.13 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water X A2 Ability to (a) predict the impacts of the 2.5 65 following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations:

A2.02 Loss of circulating water pumps.

CFR 41.5/43.5/45.3/45.13 079 Station Air 086 Fire Protection I I v ~

- 1 0 I Group Point Total: 101 3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: SEABROOK Date of Exam: December 2010 Category KIA # Topic RO SRO-Only IR # IR #

2.1.1 Knowledge of conduct of operations requirements. 3.8 1 66

1. CFR 41.10/45.13 Conduct 2.1.28 Knowledge of the purpose and function of major 4.1 1 67 of Operations system components and controls.

CFR 41.7 2.1.37 Knowledge of procedures, guidelines, or limitations 4.3 1 68 associated with reactivity management.

CFR 41.1/43.6/45.7 2.1.

i 2.1.

2.1.

Subtotal 3 2.2.12 Knowledge of surveillance procedures. 3.7 1 69 CFR41.10/45.13

2. 2.2.43 Knowledge of the process used to track inoperable 3.0 1 70 Equipment alarms.

Control CFR 41.10/43.5/45.13 2.2.

2.2.

2.2.

2.2.

Subtotal 2 2.3.11 Ability to control radiation releases. 3.8 1 71 CFR 41.11/43.4/45.10

3. 2.3.13 Knowledge of the radiological safety procedures 3.4 1 72 Radiation pertaining to licensed operator duties, such as Control response to radiation monitor alarms, containment

. entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

i CFR 41.12/43.3/45.9/45.10 2.3.

2.3.

2.3.

2.3. Facility

SEABROOK Date of Exam: December 2010 Category KIA # Topic RO SRO-Only IR # IR #

Subtotal 2 2.4.9 Knowledge of low power/shutdown implications in 3.8 1 73 I accident (e.g., loss of coolant accident or loss of

4. residual heat removal) mitigation strategies.

Emergency Procedures I CFR 41.10/43.5/45.13 Plan ,

2.4.11 Knowledge of abnormal condition procedures. 4.0 1 74 CFR 41.10/43.5/45.13 2.4.20 Knowledge of the operational implications of EOP 3.8 I 75 warnings, cautions, and notes.

CFR 41.10143.5/45.13 2.4.

2.4.

2.4.

Subtotal 3 Tier 3 Point Total 10 7

Seabrook Station 2010 NRC Written Exam ES-401 Record of Rejected KlAs Form ES-401-4 Tier I Randomly Reason for Rejection Group Selected KIA During exam outline generation the KIA's associated with Ice Tier 21 025 Ice Group 1 Condenser (025) were suppressed.

Condenser ROand This system is N/A to Seabrook.

SRO section Tier 21 027 Containment During exam outline generation theKlA's associated with Group 2 Iodine Removal Containment Iodine Removal (027) were suppressed.

ROand This system is included as part of the Containment Building SRO Spray System at Seabrook section Generic 2.2.3 (Multi-Unit During exam outline generation this KIA was suppressed.

KIA License) Seabrook is not a multi-unit facility.

Knowledge of the design, procedural, and operational differences between units.

Generic 2.2.4 (Multi-Unit During exam outline generation this KIA was suppressed.

KIA License) Ability Seabrook is not a multi-unit facility.

to explain the variations in control board/control room layouts, systems, instrumentation, and procedural actions between units at a facility.

Tier 11 040 Steam Line Originally selected KA AA 1.22, Ability to operate and/or Group 1 Rupture monitor the following as they apply to the Steam Line RO Rupture: Load Sequencer Status Light. Rejected this KA as it was difficult finding an operationally valid question topic relating the Emergency Diesel Generator load sequencer to a Steam Line Rupture.

Selected new KA AAl.20: Containment pressure and temperature trends.

Page 1 of3

Seabrook Station 2010 NRC Written Exam ES-401 Record of Rejected KJAs Form ES-401-4 Tier I Randomly Reason for Rejection Group Selected KIA Tier 056 Loss of Originally selected KA AAl.15, Ability to operate and/or I/Group I Offsite Power monitor the following as they apply to the Loss of Offsite RO Power: Service water booster pumps. Rejected KA as Seabrook Station does not have Serviee Water Booster Pumps.

Selected new KA AA 1.21: Reset of ESF load sequencers.

Tier II 062 Loss of Originally selected KA AAl.03, Ability to operate and/or Group 1 Nuclear Service monitor the following as they apply to the Loss of Nuclear RO Water Service Water: SWS as a backup to CCWS. Rejected KA as Seabrook Stations CCWS is not configured such that the SWS functions as a backup.

Selected new KA AAl.02: Loads on the SWS in the control room.

Tier 2/ 012 Reactor Originally selected KA K6.09, Knowledge of the effect ofa Group 1 Protection System loss or malfunction ofthc following will have on the RPS:

RO CEAC. Rejected the KA as Seabrook Station does not have a Control Element Assembly System. Per NUREG-1122 Tier 2/ 014 Rod Position Prior to randomly selecting an Al KA identified KA Al.Ol Group 2 Indicating System associated with "metroscope reed switch display" as not being RO applicable to Seabrook Station. KA AI.01 was suppressed from the random sampling.

Tier 075 Circulating 2/Group 2 Water Originally selected category 075 Circulating Water, A2 Ability to SRO (a) predict the impacts of the following malfunctions or operations on the Circulating Water System; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: A2.01 Loss of intake structure.

During exam validation the specific open reference question selected for KA was found to overlap questions and provide answers to two questions on RO section of exam. Per discussion with NRC lead examiner new Tier 2/Group2 category 016, Non nuclear Instrumentation, A2 Ability to (a) predict the impacts ofthe following malfunctions or operations on the NNIS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: A2.01 Detector failure, was selected.

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Seabrook Station 2010 NRC Written Exam ES401 Record of Rejected KJAs Form ES-4014 Tier I Randomly Reason for Rejection Group Selected KIA Tier II W/Ell Loss Originally selected category WlEl1 Loss of Emergency Group 1 of Emergency Coolant Recire., EK3 Knowledge of the reasons for the Coolant Recirc. following responses as they apply to the Loss of Emergency Coolant Recirculation:

EK3.4 RO and SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

Resampled per discussion with lead examiner regarding originally submitted exam question.

Selected new KA:

038 Steam Generator Tube Rupture (SGTR), EK3 Knowledge of the reasons for the following responses as they apply to the SGTR: EK3.06 Actions contained in EOP for RCS water inventory balance, SIG tube rupture, and plant shutdown procedures.

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ES-301 Administrative Topics Outline Form ES-301-1 Seabrook Station 2010 NRC Exam JPM - RO Facility: Seabrook Date of Examination: December 2010 Examination Level: RO ~ SRO 0 Operating Test Number:

Administrative Topic Type Describe activity to be performed (see Note) Code*

Manual QPTR Calculation R,M,P KA: 2.1.7 Ability to evaluate plant performance and Conduct of Operations make operational judgments based on the operating characteristics, reactor behavior, and instrument interpretation, Calculate Boron Change R,M KA: 2.1.37 Knowledge of procedures, guidelines, or Conduct of Operations limitations associated with reactivity management.

Shutdown Margin Calculation (Mode 1)

R,M KA: 2.2.12 Knowledge of Surveillance Procedures Equipment Control Establish COP Exhaust RM Setpoints Prior to Gaseous Effluent Release R,M Radiation Control KA: 2.3.11 Ability to control radiation releases.

Emergency Procedures/Plan NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S 3 for ROs; S 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (2: 1)

(P)revious 2 exams (S 1; randomly selected)

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ES-301 Administrative Topics Outline Form ES-301-1 Seabrook Station 2010 NRC Exam JPM - SRO -I Facility: ;:)l::dUI JUI\ Date of Examination: December 2010 Examination Level: RO D SRO [gJ Operating Test Number:

Administrative Topic Type Describe activity to be performed (see Note) Code*

Verify Manual QPTR Calculation R,N KA: 2,1.7 Ability to evaluate plant performance and Conduct of Operations make operational judgments based on the operating characteristics, reactor behavior, and instrument interpretation.

Verify Shutdown Margin Calculation(Mode 1)

R,N KA: 2,1.37 Knowledge of procedures, guidelines, or Conduct of Operations limitations associated with reactivity management.

Technical Specification Determination and Required Off-site Notifications.

R,M Equipment Control 2,2.40 Ability to apply Tech. Specs for a System Authorize an Emergency Dose Limit Extension R,M KA: 2.3.4 Knowledge of radiation exposure limits under Radiation Control normal or emergency conditions, Post Scenario EAL Determination and Event Classification S,D Emergency Procedures/Plan KA: 2.4.29 Knowledge of the Emergency Plan NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (::; 3 for ROs; ::; 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (<!: 1)

(P)revious 2 exams (::; 1; randomly selected)

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ES-301 Administrative Topics Outline Form ES-301-1 Seabrook Station 2010 NRC Exam ...IPM - SRO*U Facility: Seabrook Date of Examination: December 2010 Examination Level: RO 0 SRO ~ Operating Test Number:

Administrative Topic Type Describe activity to be performed (see Note) Code*

Verify Manual QPTR Calculation R,N KA: 2.1.7 Ability to evaluate plant performance and Conduct of Operations make operational judgments based on the operating characteristics, reactor behavior, and instrument

• interpretation.

I Verify Shutdown Margin Calculation(Mode 1)

R,N KA: 2.1.37 Knowledge of procedures, guidelines, or Conduct of Operations limitations associated with reactivity management.

Importance: RO 3.9 SRO 4.2 CFR: 41.10/43.5/45,12 Technical Specification Determination and Required Off-site Notifications.

R,M Equipment Control 2,2.40 Ability to apply Tech. Specs for a System Authorize an Emergency Dose Limit Extension

. Radiation Control R,M KA: 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

Post Scenario EAL Determination and Event S,D Classification Emergency Procedures/Plan 2.4.29 Knowledge of the Emergency Plan NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S 3 for ROs; S 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (:2: 1)

(P)revious 2 exams (S 1; randomly selected)

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-ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Seabrook Station 2010 NRC Exam JPM RO Facility: S eabr ook Date of Examination: December 2010 Exam Level: RO I8J SRO-I 0 SRO-U Operating Test No.:

Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code* Safety Function

a. FR H.1 Bleed and Feed A,L,M,S 4 (Primary)

b. Recover from a CRFM Actuation D,S,EN 2

c. Loss of Containment Instrument Air A,N,S g

d. Transfer SW from the Cooling Tower to the Ocean A,D,S 4 (Secondary)

e. Lower Accumulator Level N,S 3

f. Emergency Trip of EDG B A,P,S 6

g. Recover a Dropped Rod D,S 1

h. Start Hz Recombiners D,P,S 5 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

M,E,A 7

a. Local Reactor Trip D,E 5

b. Hydrogen Analyzer Local Operation D,E,R,P 1 I c. Align Alternate (Firewater) Cooling to CCP Lube Oil Cooler o and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems ,.. u;"

serve different safety functions; in-plant systems and functions may overlap those tested in the control room .

• Type Codes Criteria for RO , SRO-II SRO-U (A)ltemate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank $9/$ 81$4 (E)mergency or abnormal in-plant  ;::1/;::1/;::1 (EN)gineered safety feature - I - / ;::1 (control room system)

(L)ow-Power / Shutdown 2:1/2:1/;::1 (N)ew or (M)odifjed from bank including 1(A) 2:2/;::2/;::1 (P)revious 2 exams $ 31 S 31 $ 2 (randomly selected)

(R)CA 2:1/2:1/;::1 (S)imulator

-ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Seabrook Station 2010 NRC Exam JPM - SRO-U Facility: " Date of Examination: December 2010 Exam Level: RO 0 SRO-ID SRO-U t8l Operating Test No.:

Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code* Safety Function

a. FR H.1 Bleed and Feed A,L,M,S 4 (Primary)

b. Recover from a CRFM Actuation D,EN,S 2 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

I

a. Local Reactor Trip M,E,A 7 5

b. Hydrogen Analyzer Local Operation D,E 1

c. Align Alternate (Firewater) Cooling to CCP Lube Oil Cooler D,E,R,P I All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; In-plant systems and functions may overlap those tested in the control room.

"Type Codes Criteria for RO I SRO-II SRO-U (A)lternate path 4-6/4-6/2-3 (Clontrol room (D)irect from bank :59/:58/:54 (E)mergency or abnormal in-plant ~1/~1/~1 (EN}gineered safety feature - I - 1 ~1 (control room system)

(L}ow-Power I Shutdown ~1/~1/~1 (N)ew or (M}odified from bank including l(A) ?2/~2/~1 (P)revious 2 exams :5 31 :5 3 1 :5 2 (randomly selected)

(R)CA ~1/?1/~1 (S)imulator

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Seabrook Station 2010 NRC Exam JPM - SRO-I Facility: S eabr ook Date of Examination: December 2010 Exam Level: RO D SRO-I 181 SRO-U D Operating Test No,:

• Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code* Safety Function

a. FR H.1 Bleed and Feed A,L,M,S 4 (Primary)

b. Recover from a CRFM Actuation D,S,EN 2 C. Loss of Containment Instrument Air A,N,S 8

d. Transfer SW from the Cooling Tower to the Ocean A,D,S aryl

e. Lower Accumulator Level N,S 3

f. Emergency Trip of EDG B A,P,S 6

g. Recover a Dropped Rod D,S 1 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) 7

a. Local Reactor Trip M,E,A 5

b. Hydrogen Analyzer Local Operation D,E

c. Align Alternate (Firewater) Cooling to CCP Lube Oil Cooler 1 D,E,R,P

,,,nA

"'RO-I control room (and in-plant) systems must be different and serve different safety functions; all serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO I SRO-II SRO-U (A)lternate path 4-6 I 4-6 I 2-3

{C)ontrol room (D)irect from bank S9/S8/S4 (E)mergency or abnormal in-plant ~1/~1/~1 (EN)gineered safety feature - I - 1 2:1 (control room system)

(L)ow-Power I Shutdown ~1/;::1/;::1 (N)ew or (M)odified from bank including 1(A)  ;;;2/2:2/;;;1 (P)revious 2 exams s 31 S 3 / S 2 (randomly selected)

(R)CA ~1/;;;1/~1 (S)imulator

Appendix D Scenario Outline Form ES-D-1 Seabrook StatIon 2616 NRC Exam-SImulator Scenanos Facility: Seabrook Scenario No.: 1 Op-Test No.:

Examiners: O.Silk Operators:

Initial Conditions: MOL 50% power. Both Main feed Pumps and one Heater Drain pump is in service.

Turnover:

• Plant is at 55% power.

• FW~FK-520, "B" SG FRV controller has experienced a failure of its auto tracking driver card. This failure prevents automatic operation of the "B" FRV. The controller is currently in manual requiring operator intervention to control "B" SG water level. Additional manpower for feed station operations is not available at this time.

• Procedure OS1000.05. Power Increase is being performed and is completed to step 4.3.10.

• Maintain AFD on target.

• Increase power to 75% at 5%/hour.

• SW-P-41C tagged out for motor replacement.

Event Malf. No. Event Event No. Type* Description 1 100ZMDIF BOPC FW-FK-520, "S" SG FRV controller had previously experienced WFK520M a failure of its auto tracking driver card. This failure prevents automatic operation of the "S" FRV. The controller is in manual requiring SOP intervention to control the "S" SG level during the following At-Power events.

2 PSOR Crew begins a 5%/hr. power increase BOPN USN 3 ptFWPT505 PSO I Turbine Impulse Pressure PT-505 Fails Low causing automatic USI, TS rod insertion BOPI 4 RCLT459 PSOI RC-U-459 PZR Level, Channel 1 Fails High, requiring manual US I. TS control of charging and letdown.

5 mfRC049C PSOC Crew responds to a 20 GPM RCS Leak. Supports E Plan USC,TS Classification (End of scenario) 6 mfHF001 PSOM Crew responds to sequential loss of EHC pumps with complete mfHF002 USM loss producing an automatic Turbine/Reactor trip signal. A sv07186HF BOPM turbine trip - Reactor trip will occur when EHC pressure decays rvMSAVR36 below 1100 psig or when MANUALLY tripped by the crew.

rvMSAVR50 When the MSIVs are shut, two SG safeties will fail open causing the need for a SI.

7 avMSVSV2 BOPC One Turbine Stop and Control valve sticks Open requiring a avMSVCV2 Manual MSI.

mfRPS019 mfRPS020 8 mfSI003 mfSI004 I PSOC SI pumps A & B will fail to auto-start on the SI signal and require manual actions to start

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Note: Anticipated AOP/EOP flow-path: OS1235.05, OS1201.07, OS1201.02, E-O, E-2, ES 1.1 or E-1 1

Appendix 0 Scenario Outline Form ES-D-1 seabrook StatIon 2010 NRC Exam-Slm@ator Scenanos Facility: Seabrook Scenario No.: 2 Op-Test No.:

Examiners: D. Silk Operators:

Initial Conditions: MOL 100% Power Turnover:

• Unit at 100% power. Boron conc. 1103 ppm

• No other outstanding testing or maintenance in progress.

• SW-P-41 C tagged out for motor replacement.

• Due to grid stability issues, ISO New England has requested that the unit decrease power to 1100 MWE net, in less than 25 minutes after the crew takes the watch.

Event Malf. No. Event Event No. Type* Description 1 PSOR Crew commences 25 minute power reduction to less than 1100 USN MWE net for grid stability control.

BOPN 2 mfCS016 psoe The operating Charging Pump 2A Trips requiring manual US e,TS actions to isolate letdown and eventual restoration of charging i and letdown.

3 mfRM002 USTS The Containment Particulate, Iodine, and Gas (PIG) radiation monitor will fail (Alarms in Control Room) requiring entry into the RCS Leak Detection Tech. Spec.

Bope SCC-P-34B trips with no automatic start of the standby pump 4 csCCP34B

! use requiring operator action to start the standby pump.

mfSCC005 5 ptMSPT507 BOP I Main Steam Header Press Instrument PT-507 Fails Low. The USI speed of the main feed pumps will decrease. SG levels will decrease requiring manual operator actions.

6 mfED037 BOPM The turbine generator breaker will trip producing a large load USM rejection. Automatic reactor trip is disabled requiring manual bkCPRTA PSOM trip that fails, producing an A TWS. The main turbine #1 stop bkCPRTB valve and #3 control valves fail open requiring manual closure mfRPS001 of the Main Steam Isolation Valves.

mfRPS002 7 mfCC015 psoc When the reactor is tripped locally. CC-P-11 8 will trip, with P mfCC009 11 D failing to auto-start requiring manual actions to restore.

8 100ZMDRC usc When the Gen BKR trips, the "8" PORV fails open, requiring S45681 psoc manual isolation.

• (N)ormal, (R)eactivity. (I)nstrument, (C)omponent, (M)ajor Note: Anticipated AOP/EOP flow-path: OS1231.04, OS1202.02, OS1230.01, ON1237.01, E-O, FR-S.1, OS1201.06, E-O, ES-0.1 or ES..:1.1.

1

'Appendix D Scenario Outline Form ES-D-1 Seabrook Station 2010 NRC Exam-Simulator Scenarios Facility: .seabrook Scenario No.: Q Op-Test No.:

Examiners: Dave Silk Operators:

Initial Conditions: MOL, 100% Power.

Turnover:

• The plant is at 100% power.

• Decrease plant power at 1O%/hr to 45% power to allow for repairs of FW-P-32A oil leak. Currently at Step 1.3 of OS1000.06, Figure 6

• Main Steam Atmospheric Steam Dump (UN ASDV), MS-PV-3001 is Danger Tagged closed due to excessive seat leakage. MS-V-5 is closed and tagged. Tech. Spec. 3.7.1.6 and 3.6.3 were entered 4 days ago.

• SW-P-41 C tagged out for motor replacement.

Event Malt. No. Event Event No. Type* Description 1 PSOR Decrease plant power @ 10%/hr.

BOPN USN 2 ttRCTT411 PSO I RCS Loop 1 T cold will fail high causing inward control rod US I,TS motion.

3 ptMSPK3004 Bope "D" ASDV controlling pressure channel will fail high causing USTS "D" ASDV to open.

4 mfCS012 psoe Letdown line leak requiring letdown isolation.

use, TS 5 mfED038 PSOM Loss of Offsite Power, complicated by "Aft EDG failure to start, mfED031 BOPM "B" EDG starts then trips on low oil pressure and Steam Driven EFW pump trips on overspeed.

mfED034 USM avMSVSV2 6 mfSW013 Bope On Bus 6 power restoration from SEPs no Service Water mfSW015 Pump will start.

mfSW017 The Motor Driven EFW Pump Trips on Overcurrent.

mfFW055 7 rmvMSV129 BOpe Crew directs reset and restart of FW-P37A (terry turbine) to use establish EFW flow to the Steam Generators.

• (N)ormal, (R)eactivity , (I )nstrument, (C)omponent, (M)ajor Note: AntiCipated AOP/EOP flow-path: OS1201.08, OS 1201.02, E-O, ECA-O.O, E-O, ES-0.1 1

I. P ....

Appendix D Scenario Outline Form ES-D-1 seabrook StatIOn 2010 NRC Exam-SImulator Scenanos Facility: Seabrook Scenario No.: 4 Op-Test No.:

Examiners: D.Sllk Operators:

Initial Conditions: MOL 78% power Turnover:

• ISO directed power reduction to < 80% due to grid loading limitations. Achieved - 78% power 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 45 minutes ago. Maintaining power at -78% per OS1000.06.

• Xenon is still building in slowly at - (-)0.3 pcm I minute or (-) 0.15°F every 10 minutes

• SW-P-41 C is tagged out for motor replacement.

• • No Tech. Spec. Action Statements in effect.

Event Malf. No. Event Event i No. TYl'e* Description 1 PSOR Crew begins 5%/hr. power increase.

BOPN USN 2 mfNI002 PSO I NI 42 Fails High causing automatic rod insertion.

US I,TS Bope 3 mfSW001 Service Water Pump 41A Overcurrent Trip. With Pump 41C US C, TS tagged out and automatic actuation disabled, manual cooling cSWV22 tower alignment is required.

100ZMDISW CS61481 cSWP41C 4 mfED026 PSO I Loss of Vital Instrument Panel 1B causing multiple instruments BOP I to fail, requiring manual actions by the PSO and BOP to US I, TS stabilize.

5 mfSG001B PSOC "B SG tube leak develops at 50 gpm tube leak requiring manual US C, TS control of Charging and Letdown. Crew commences a rapid power decrease that is required due to the steam generator tube leak.

6 mfSG001B PSOM "B" SG tube ruptures at 500 gpm, requiring manual start of the BOPM second charging pump manual reactor trip and safety injection.

USM 7 mfRPS011 PSOC PSO (C) PSO recognizes automatic T signal failure on Train A and CS-V-150 and CS-V-168 remain open on Train B. Operator i

aligns in accordance with E-O Attachment "A". i

• (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Note: Anticipated AOP/EOP flow-path: OS1211.04, OS1216.01, OS1247.01, OS1227.02, OS1231.04, E-O, E-3 1