ML14024A480

Initial Exam 2013-301 Draft Administrative Documents

ML14024A480
Person / Time
Site:	Mcguire, McGuire
Issue date:	01/23/2014
From:	NRC/RGN-II
To:	Duke Energy Carolinas
References
50-369/13-301, 50-370/13-301
Download: ML14024A480 (24)

Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: McGuire 2013 Date of Exam: December 2013 RO K/A CapoyPoints SRO-Only_Points Tier Group K K K K K K A A A A G A2 G* Total 1234561 234 Total

1. 1 333 33 318 3 3 6 Emergency & 2 1 2 2 1 2 1 9 2 2 4 Abnormal Plant Tier Totals 4 5 5 N/A 4 5 N/A 4 27 5 5 10 Evolutions

2. 2 11011111111 10 Plant Tier Totals 3 4 3 4 4 3 3 4 3 4 3 38 5 2 8 Systems

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 3 3 2 1 2 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 outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A 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 0.1 .b of ES-401 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-401 for the applicable K/As.

8. On the following pages, enter the K/A 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 K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

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

E/APE # / Name / Safety Function K K K A A2 G K/A Topic(s) IR #

1 231 007EK1 .03, Knowledge of the operational 000007 (BW/E02&E1 0; CE/E02) Reactor X 3 implications of the trip reasons for closing the Trip Stabilization Recovery / 1 main turbine governor valve and the main turbine stop valve after a reactor trip.

000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 X 009EK2.03, Knowledge of the interrelations 3.0 between the small break LOCA and S/Gs.

X 4.7 009EG2.4.4, Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

000011 Large Break LOCA / 3 01 5AG2. 1.7, Ability to evaluate plant 000015/17 RCP Malfunctions /4 X 44 performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation.

022AG2.2.3, (multi-unit license) Knowledge of 000022 Loss of Rx Coolant Makeup / 2 X 39 the design, procedural and operational differences between units.

025AA2.07, Ability to determine and interpret 000025 Loss of RHR System / 4 X 3 pump cavitation as it applies to the Loss of Residual Heat Removal System.

026AA1 .07, Ability to operate and / or 000026 Loss of Component Cooling X 2.9 monitor Flow rates to the components and Water / 8 systems that are serviced by the CCWS and interactions among the components as they apply to the Loss of Component Cooling Water.

027AG2.49, Knowledge of low power or 000027 Pressurizer Pressure Control X 3.8 shutdown implications in accident (e.g. LOCA System Malfunction / 3 or loss of RHR) mitigation strategies.

000029 ATWS / 1 038EA1 .27, Ability to operate and monitor 000038 Steam Gen. Tube Rupture / 3 X 39 Steam dump valve status lights and indicators as they apply to a SGTR.

040AA2.04, Ability to determine and interpret 000040 (BW/E05; CE/E05; W/E12) X 47 the conditions requiring ESFAS initiation as Steam Line Rupture Excessive Heat they apply to the Steam Line Rupture.

Transfer / 4

054AK1 .02, Knowledge of the operational 000054 (CE/E06) Loss of Main X 3.6 implications of the effects of feedwater Feedwater I 4 introduction on dry SIG as they apply to Loss of Main Feedwater (MFW).

055EK1 .02, Knowledge of the operational 000055 Station Blackout / 6 X 4.1 implications of Natural circulation cooling as it applies to the Station Blackout.

056AG2.4.20, Knowledge of operational 000056 Loss of Off-site Power I6 X implications of EOP warnings, cautions and 3.8 notes.

057AK3.01, Knowledge of the reasons for 000057 Loss of Vital AC Inst. Bus / 6 X 4.1 Actions contained in EOP for loss of vital ac electrical instrument bus as they apply to the Loss of Vital AC Instrument Bus: Actions 3.8 contained in EOP for loss of vital ac electrical instrument bus.

X 057AA2.02, Ability to determine and interpret core flood tank pressure and level indicators as they apply to the Loss of Vital AC Instwment_Bus.

058AA2.03, Ability to determine and interpret 000058 Loss of DC Power / 6 X 35 DC loads lost; impact on ability to operate and monitor plant systems as they apply to the Loss of DC Power.

062AA2.O1, Ability to determine and interpret 000062 Loss of Nuclear Svc Water! 4 X 2.9 the location of a leak in the SWS as t applies to the Loss of Nuclear Service Water.

062AA2.03, Ability to determine and interpret X the valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition as they apply to the Loss of Nuclear Service Water.

065AK3.04, Knowledge of the reasons for the 000065 Loss of Instrument Air! 8 X 3.0 Cross-over to backup air supplies responses as they apply to the Loss of Instrument Air.

WEO4EK2.1, Knowledge of the interrelations WIEO4 LOCA Outside Containment! 3 X 35 between the (LOCA Outside Containment) and components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features.

WEI 1 EK3.4, Knowledge of the reasons for W!E1 1 Loss of Emergency Coolant X 3.6 the RO or SRO function within the control Recirc. I 4 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.as they apply to the Loss of Emergency Coolant Recirculation.

WEO5EK22, Knowledge of the interrelations BWIEO4; WIEO5 Inadequate Heat X between the Loss of Secondary Heat Sink Transfer Loss of Secondary Heat Sink! 4 and the facilitys 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.

WEO5G2.4.37, Knowledge of the lines of X authority during implementation of the 4.1 emergency plan.

077AA1 .05, Ability to operate and/or monitor 000077 Generator Voltage and Electric X 39 the Engineered Safety Features as they apply Grid Disturbances ! 6 to IGenerator Voltage and Electric Grid Disturbances.

CD CD 0

I 0

0 0

0 m

cn cn m

m Co 0

0 0) a)

CO C-)

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)

E/APE # / Name / Safety Function K K K A A2 G K/A Topic(s) IR #

1231 OOIAAI .07, Ability to operate and 000001 Continuous Rod Withdrawal / 1 X 3.3 I or monitor RPI as it applies to the Continuous Rod Withdrawal.

000003 Dropped Control Rod / 1 005AA2.01 Ability to determine 000005 lnoperablelStuck Control Rod I 1 X 4.1 and interpret Stuck or inoperable rod from in-core and ex-core NIS, in-core or loop temperature measurement as it applies to the Inoperable / Stuck Control Rod 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction /2 000032 Loss of Source Range NI /7 000033 Loss of Intermediate Range NI /7 X 033AG2.1 .20, Ability to execute 4.6 procedure steps.

036AK3.02, Knowledge of the 000036 (BW/A08) Fuel Handling Accident /8 X 2.9 reasons for the Interlocks associated with fuel handling equipment as they apply to Fuel Handling Incidents.

000037 /3 037AK1 .01, Knowledge of the Steam Generator Tube Leak X 2.9 operational implications of the Use of steam tables as it applies to Steam Generator Tube Leak.

000051 051AA2.02, Ability to determine Loss of Condenser Vacuum /4 X 3.9 and interpret the conditions requiring reactor and/or turbine trip as they apply to the Loss of Condenser Vacuum.

000059 Accidental Liquid RadWaste Rel. I9 000060 Accidental Gaseous Radwaste Rel. /9 000061 ARM System Alarms I7 000067 Plant Fire On-site /8

068AA2.05, Ability to determine 000068 (BW/A06) Control Room Evac. I 8 X

and interpret the availability of heat sink as it applies to the Control Room Evacuation.

000069 (W/E14) Loss of CTMT Integrity! 5 WEO7EG2.4.41, Knowledge of the 000074 IE06&E07) mad. Core Cooling / 4 X emergency action level thresholds and classifications.

/9 076AK2.01, Knowledge of the 000076 High Reactor Coolant Activity X 26 interrelations between the High Reactor Coolant Activity and the process radiation monitors.

& E02 & SI WEO1EA2.1, Ability to determine 4.0 WIEO1 Rediagnosis Termination /3 X and interpret the facility conditions and selection of appropriate procedures during abnormal and emergency operations as they apply to the Reactor Trip or Safety Injection Rediagnosis.

W/E13 Steam Generator Over-pressure! 4 W/E15 Containment Flooding /5 WE15EG2.4.8, Knowledge of how X

abnormal operating procedures are used in conjunction with EOPs.

W/E16 High Containment Radiation I9 BW/E08; W/E03 LOCA /4 WEO3EK2.2, Knowledge of the Cooldown - Depress. X interrelations between the LOCA Cooldown and Depressurization and the facilitys 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.

BW/E09; CE/A13; W/E09&E10 Natural Circ. /4 CE/Al 1; W/E08 RCS PTS / 4 WEO8EK3.3, Knowledge of the Overcooling - X reasons for the manipulation of controls required to obtain desired operating results during abnormal and emergency situations as they apply to the (Pressurized Thermal Shock).

KIA Category Point Totals: 1 2 2 1 2/2 1/2 j Group Point Total: 9/2

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

System #1 Name K K K K K K A A2 A A G K/A Topic(s) IR #

1 234561 34 003A1 .03, Ability to predict 003 Reactor Coolant Pump X and/or monitor changes in 2.6 parameters (to prevent exceeding design limits) associated with operating the RCPS controls including RCP motor stator winding temperatures.

004A1 .10, Ability to predict 004 Chemical and Volume X and/or monitor changes in Control parameters (to prevent X exceeding design limits) associated with operating the X CVCS controls including reactor power.

3.8 004A4.08, Ability to manually operate and/or monitor in the control room: Charging 004G2.2.22, Knowledge of limiting conditions for operations and safety limits.

005K5.05, Knowledge of the 005 Residual Heat Removal X operational implications of the 2.7 plant response during solid plant: pressure change due to the relative incompressibility of water as they apply the RHRS 006K1 .14, Knowledge of the 006 Emergency Core Cooling X physical connections and/or 3.0 cause-effect relationships between the ECCS and the lAS 007K3.01, Knowledge of the 007 Pressurizer Relief/Quench X effect that a loss or malfunction 33 Tank of the PRTS will have on the Containment.

008K3.03, Knowledge of the 008 Component Cooling Water X effect that a loss or malfunction 4.1 of the CCWS will have on the RCP.

01 0K5.01, Knowledge of the 010 Pressurizer Pressure Control X operational implications of the 35 determination of condition of fluid in PZR, using steam tables as it applies to the PZR PCS.

012K6.10, Knowledge of the 012 Reactor Protection X effect of a loss or malfunction of 33 the permissive circuits will have on the RPS.

X 012G2.1.2, Knowledge of 4.4 operator responsibilities during all modes of plant operation.

013 Engineered Safety Features X 01 3K2.01 Knowledge of bus 3.6 Actuation power supplies to the x ESFAS/safeguards equipment control.

4.5 013A2.02, Ability to (a) predict the impacts of excess steam demand on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations.

022 Containment Cooling X 022K2.01 Knowledge of power 3.0 supplies to the containment X cooling fans.

022K4.03, Knowledge of CCS 3.6 design feature(s) and/or interlock(s) which provide for automatic containment isolation.

025K4.02, Knowledge of ice 025 Ice Condenser X condenser system design 2.8 feature(s) and/or interlock(s) which provide for system control.

026 Containment Spray X 026A4.01, Ability to manually 4.5 operate and/or monitor in the control room CSS controls.

039 Main and Reheat Steam X 039A3.02, Ability to monitor 3.1 automatic operation of the MRSS, including isolation of the MRSS.

059G2. 1.23, Ability to perform 059 Main Feedwater X specific system and integrated 4.3 plant procedures during all modes of plant operation.

X 059K3.02, Knowledge of the 3.6 effect that a loss or malfunction X of the MFW will have on the AFW system.

3.1 059A2.03, Ability to (a) predict the impacts an overfeeding event on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations.

061 K5.01, Knowledge of the 061 Auxiliary/Emergency X operational implications of the 36 Feedwater relationship between AFW flow and RCS heat transfer as it applies to the AFW.

x 061 K6.02, Knowledge of the effect of a loss or malfunction of 2.6 the pumps will have on the AFW components.

062G2.4.47, Ability to diagnose 062 AC Electrical Distribution X and recognize trends in an 4.2 accurate and timely manner utilizing the appropriate control room reference material.

x 062A2.06, Ability to (a) predict 3.4 the impacts keeping the safeguards buses electrically separate on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations 063K1 .03, Knowledge of the 063 DC Electrical Distribution X physical connections and/or 2.9 cause-effect relationships between the DC electrical system and the battery charger and battery.

064 Emergency Diesel Generator X 064A4.03, Ability to manually 3.2 operate and/or monitor the synchroscope in the control room.

073 Process Radiation X 073A2.02, Ability to (a) predict 2.7 Monitoring the impacts of detector failure on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences 076 Service Water X 076A3.02, Ability to monitor 3.7 automatic operation of the x SWS, including emergency heat loads.

076 K2.04, Knowledge of bus power supplies to reactor 2.5 building closed cooling water.

078K4.03, Knowledge of lAS 078 Instrument Air X design feature(s) and/or 3.1 interlock(s) which provide for securing of SAS upon loss of cooling water.

x 078A2.01, Ability to (a) predict the impacts of air dryer and 2.9 filter malfunctions On the lAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences.

103A2.04, Ability to (a) predict 103 Containment X the impacts of containment 3.5 evacuation (including recognition of the alarm) on the containment system and (b) based on those predictions, use procedures to correct, control, or mitiaate the consecuences.

K/A Category Point Totals: 2 3 3 3 3 2 2 3/3 2 3 2/2 Group Point Total: 28/5

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 2/Group 2 (RO I SRO)

System #1 Name K K K K K K A A2 A A G K/A Topic(s) IR #

1 234561 34 001 Control Rod Drive X 001A4.1 1 Ability to manually 3.5 operate and/or monitor in the control room: Determination of SDM.

002G2.1.19, Ability to use plant 002 Reactor Coolant X computer to evaluate system or 3.8 component status.

011 Pressurizer Level Control X 011K2.01, Knowledge of bus 3.1 power supplies to the charging x pumps.

01 1A2.06, Ability to (a) predict 3.9 the impacts of an Inadvertent PZR spray actuation on the PZR LCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences.

014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017K4.03, Knowledge of TM 017 In-core Temperature Monitor X system design feature(s) and/or 3.1 interlock(s) which provide for the range of temperature indication 027 Containment Iodine Removal 028K6.01, Knowledge of the 028 Hydrogen Recombiner X effect of a loss of hydrogen 2.6 and Purge Control recombiners on CRDS components.

029 Containment Purge 033 Spent Fuel Pool Cooling

034 Fuel Handling Equipment 035A2.04, Ability to (a) predict 035 Steam Generator X the impacts Steam flow/feed 3.8 mismatch on the GS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences 041A1.01, Ability to predict 041 Steam Dump/Turbine X and/or monitor changes in 2.9 Bypass Control parameters (to prevent exceeding design limits) associated with operating the SDS controls including T-ave.

verification above low/low setpoint 045K5.23, Knowledge of the 045 Main Turbine Generator X operational implications of the 2.7 relationship between rod control and RCS boron concentration during T/G load increases as it applies to the MT/B System 055 Condenser Air Removal X 055G2A.4.1 8, Knowledge of the 3.3 specific bases for EOP5.

056K1 .03, Knowledge of the 056 Condensate X physical connections and/or 2.6 cause-effect relationships between the Condensate System and the MFW 068A202 Ability to (a) predict the 068 Liquid Radwaste X impacts of a, tack of tank 2.7 recirculation prior to release on the Liquid Radwaste System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences 071 Waste Gas Disposal 072A3.01, Ability to monitor 072 Area Radiation Monitoring X automatic operation of the ARM 2.9 system, including changes in ventilation alignment 075 Circulating Water 079 Station Air 086 Fire Protection K/A Category Point Totals: 1 1 0 1 1 1 1 1/2 1 1 1/i. Group Point Total: 10/3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Date of Exam:

Category KIA# Topic RD SRO-Only lR IR #

G2.1.36, Knowledge of procedures and limitations involved in 2.1. core alterations 30 G2.1.45, Ability to identify and interpret diverse indications to 2.1. validate the response of another indication 2.1. G2.1.1, Knowledge of conductof operations requirements. 4.2 Conduct of Operations G2.1.45, Ability to identify and interpret diverse indications to 2.1. validate the response of another indication 43 2.1.

2.1.

Subtotal 2 2 G2.2.18, Knowledge of the process for managing maintenance 2.2. activities during shutdown operations. 26 G2.2.38, Knowledge of conditions and limitations in the facility 2.2. license. 36 G2.2.21 Knowledge of pre- and post-maintenance operability 2.2. requirements. 4 1 2.

Equipment 2.2.

Control 2.2.

2.2.

Subtotal 2 G2.3.12, Knowledge of radiological safety principles pertaining to 2.3. licensed operator duties 32 G2.3.13, Knowledge of radiological safety procedures pertaining 2.3. to licensed operator duties 3.4

G2.3.4, Knowledge of radiation exposure limits under normal and 2.3. emergency conditions 32 G2.3.14, Knowledge of radiation or contamination hazards that

• 2.3. may arise during normal, abnormal, or emergency conditions or 3.8 Radiation activities Control 2.3. G2.3.6, Ability to approve release permits 3.8 2.3.

Subtotal 3 2 2.4. G2.4.1 .1 Knowledge of abnormal condition procedures. 4.0 2.4. G2.4.25, Knowledge of fire protection procedures. 3.3 4.

G2.4.31, Knowledge of annunciators alarms, indications or 2.4. response procedures 4.2 Emergency Procedures I G2.4.16, Knowledge of EOP implementation hierarchy and Plan 2.4. coordination with other support procedures or guidelines. 4.4 G2.4.5, Knowledge of the organization of the operating 2.4. procedures network for normal, abnormal and emergency 4.3 evolutions.

2.4.

Subtotal 3 2 Tier3PointTotal 10 10 7

ES-401 Record of Rejected KIAs Form ES-401-4 Tier I Randomly Reason for Rejection Group Selected K/A Tier Oversampled. WEO2EA2.1 was also originally randomly selected.

WEO2EA2 2 1/Group 2 Both were SRO questions. Randomly chose EA 2.1 to be replaced.

Replaced with 005AA2.O1.

All selections randomly selected using Random.org website.

Tier 2/ 073A2.O1 Oversampled. 073A2.O1 and 073A2.02 both randomly selected.

Group 1 Both RO questions. 073A2.O1 randomly rejected. Replaced with 062A2 .06.

Tier 1/ 027G2.4.34 Unable to write to SRO level.

Group 1 Tier 2/ 061G2.1.28 Unable to write to SRO level Group 1

ES-301 AdministraUve Topics Outline Form ES-301-1 DRAFT (Rev_052913)

FacHity: McGuire Date of Examination: 12/2013 Examination Level: P0 Operating Test Number: N13-1 Administrative Topic Type Code* Describe activity to be performed (see Note>

2.1 .25 (3.9) Ability to interpret reference materials, such Conduct of Operations as graphs, curves, tables, etc.

M P JPM: Determine Boric Acid Addition to FWST 2.1.7 (4,4) Ability to evaluate plant performance and Conduct of Operations make operational judgments based on operating characteristics, reactor behavior, M, R and instrument interpretation JPM: Manual AFD Calculation 2.2.43 (3.0) Knowledge of process used to track Equipment Control inoperable alarms.

N, P JPM: Partial Loss of Annunciators 2.3.14 (3.4) Knowledge of radiation or contamination Radiation Control hazards that may arise during normal, abnormal, or emergency conditions or M, P. R activities.

JPM: Predict Radiation Levels While Responding to a Damaged Spent Fuel Pool NOTE: All items (5 total) are required for BROs. P0 applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

Type Codes & Criteria: (C)ontrol room, (0) (S)imulator, (0) or Class(R)oom (4)

(D)irect from bank ( 3 for ROs; 4 for BROs & RO retakes) (0)

(N)ew or (M)odif led from bank (> 1> (4)

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

NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-l DRAFT (Revj)52913)

RO Admin JPM Summary Al a This is a modified bank JPM. The operator will be told that a leak, which is now isolated has lowered the FWST level to 440 inches, and that it has been decided to use the Recycle Holdup Tank (RHT) to refill the FWST. The operator will be told that Enclosure 4.4, (FWST Makeup Using the RHT), of OP/l/A16200/014 (Refueling Water System) is in progress and completed through Step 3.10, and provided with Chemistry Data for the BAT and RHT. The operator will then be directed to determine the amount of Boric Acid needed to raise the FWST level to 480 using the RHT in accordance with Step 3.11 of Enclosure 4.4 of 0P111A162001014 (Refueling Water System). The operator will be expected to calculate the amount of Boric Acid that must be added from the BAT to refill the FWST in accordance with the attached KEY.

Aib This is a modified JPM using Bank JPM ADM-NRC-A1-021 as its basis. The operator will be told that Unit 1 is at 94% power, the QAC has been out of service for 30 minutes, that PT/1/A/4600/021 A (Loss of Operator Aid Computer While in Mode 1) is being performed, and that the Main Control Board AFD meters are INOPERABLE. The operator will be given the present current values for the Power Range upper and lower detectors and directed to calculate AFD per PT/l/A4600/021A (Loss of Operator Aid Computer while in Mode 1), Section 12.12 for current plant conditions and verify that AFD is within the limits specified in the COLR. The operator will be expected to manually calculate AFD, and determine that the AFD calculated for N41 and N42 is in excess of the limits allowed by the COLR.

A2 This is a New JPM. The operator will be told that while Unit 1 was operating at 100% power, a lightning strike caused several of the Unit 1 Control Room Annunciators to fail requiring entry into PT/1/A4600/033 (Loss of Control Room Annunciators>. The operator will be provided with a list of failed annunciators; and directed to continue with Enclosure 13.2 (Partial Loss of Annunciator Panels), and identify any Alternate Methods for Surveillance that are applicable.

The operator will be expected to determine that 11% of the Annunciators have failed in accordance with the attached Key, and identify five (5) specific annunciators that have an identified Alternative Method for Surveillance.

A3 This is a modified Bank JPM used on the 2012 NRC Exam. The operator will be given a set of conditions reflecting a damaged and leaking Spent Fuel Pool with a full core off-loaded, where attempts of makeup have failed, but are expected to be successful within four hours. The operator will also be given a present Spent Fuel Pool leak rate and level. The operator will be directed to refer to Enclosure 13 (Spent Fuel Pool Radiation Level vs. Water level Above Fuel) of AP/1/AJ5500/41 (Loss of Spent Fuel Pool Cooling or Level), and determine the expected radiation levels one hour, two hours, three hours and four hours from now, based on the last known leak rate. The operator will be expected to determine the expected dose rate within +/-50%.

NUREG-1 021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 DRAFT (Rev_05291 3)

Facility: McGuire Date of Examination: 12/2013 Examination Level: SRO Operating Test Number: N 13-1 Administrative Topic Type Code* Describe activity to be performed (see Note>

2.1.25 (4,2> Ability to interpret reference materials, such Conduct of Operations as graphs, curves, tables, etc.

M R JPM: Determine Boric Acid Addition to FWST__

2.1.37 (4.6) Knowledge of procedures, guidelines or Conduct of Operations limitations associated with reactivity M, A management JPM: Perform an ECP 2.2.12 (4.1) Knowledge of Surveillance Procedures.

Eouipment Control

, N,R JPM: Determine Procedure Sections that Must be Performed 2.3.14 (3.8) Knowledge of radiation or contamination Radiation Control hazards that may arise during normal, abnormal, or emergency conditions or activities.

M,P,R JPM: Calculate Spent Fuel Pool Boiloff rate and predict when Spent Fuel Pool Radiation levels will exceed 100 mrem/hour 2.4.44 (4.4) Knowledge of emergency plan protective Emergency action recommendations.

Procedures/Plan M, A JPM: Provide an updated PAR NOTE: All items (5 total) are required for SAOs. AC applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

Type Codes & Criteria: (C)ontrol room, (0) (S)imulator, (0) or Class(R)oom (5)

(D)irect from bank ( 3 for ADs; 4 for SACs & RD retakes) (0)

(N)ew or (M)odified from bank ( 1) (5)

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

NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301 -1 DRAFT (Rev_05291 3)

SRO Admin JPM Summary Ala This is a modified Bank JPM. The operator will be told that a leak, which is now isolated has lowered the FWST level to 440 inches, below the Technical Specification Limit 1 and that it has been decided to use the Recycle Holdup Tank (RHT) to refill the FNST. The operator will be told that Enclosure 4.4 (FWST Makeup Using the RHT), of OP/l/A16200/014 (Refueling Water System) is in progress and completed through Step 3.1 1, and provided with Chemistry Data for the BAT and RHT. The operator will then be directed to perform the Independent Verification (SRO aspect) of the calculation in Step 3.11 of Enclosure 4.4 to determine the amount of Boric Acid that must be added from the Boric Acid Tank (BAT), in order to raise the FWST Level to 480 using the RHT. The operator will discover two errors within the previous calculation, and determine the correct volume of Boric Acid to add in accordance with the attached KEY. Following this, the operator will be given a makeup flowrate to the FWST and asked to identify the impact on the Technical Specification ACTION. The operator will be required to identify that ACTION C is applicable after one hour.

Al b This is a modified Bank JPM. The operator will be told that Reactor Startup is an hour away, and provided with a set of initial conditions. The operator will be asked to perform an Estimated Critical Position (ECP) in accordance with OP/0/A/61 00/06, Reactivity Balance Calculation, Enclosure 4.2, Estimated Critical Rod Position, During the course of the ECP, the operator will be given a set of power history conditions, and asked to perform a Shutdown Fission Product Correction calculation in accordance with OP/0/A/61 00/06, Reactivity Balance Calculation, Enclosure 4.8, Shutdown Fission Product Correction Calculation, in support of the ECP. The operator will be expected to calculate the Estimated Critical Rod Position Bank for No and Peak Xenon at time of Criticality per the attached KEY.

A2 This is a New JPM. The operator will be told that a plant startup is in progress in accordance with OP/1/A16100/003 (Controlling Procedure for Unit Operation),

and that the crew has just stabilized the plant at 3.6% power and in a 10 minute hold. The operator will be provided with a listing of failed OAC Alarms; and directed to assess the QAC Points using PT/l/A14600/021 B (Loss of Operator Aid Computer while in Mode 2); specifically to identify all procedure sections that must be performed, all personnel and/or organizations that must be notified, and all procedure Enclosures or other procedures that must be performed as power is raised. The operator will be expected to identify the five procedure Sections that must be performed as 12.2, 12.5, 12.6, 12.9 and 12,13; identify that the Engineering OAC Group must be notified, and that Enclosure 13.2 Part A needs to be performed.

A3 This is a modified Bank JPM used on the 2012 NRC Exam. The operator will be given a set of conditions reflecting a loss of level and cooling to the Spent Fuel Pool with a full core offloaded, where attempts of makeup have failed, The NUREGlO21, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 DRAFT (Rev05291 3) operator will also be given a present Spent Fuel Pool level and time since shutdown. The operator will be directed to use Enclosure 5 (Spent Fuel Pool Boiloff rate) of AP/1/A15500/41 (Loss of Spent Fuel Pool Cooling or Level), and estimate how fast the Spent Fuel Pool level will go down as the pool boils; and then use this boiloff rate and Enclosure 13 (Spent Fuel Pool Radiation Level vs.

Water level Above Fuel) of AP/1/A15500/41 (Loss of Spent Fuel Pool Cooling or Level), and predict when the radiation levels in the Spent Fuel Pool area will be greater than 100 mrem/hour if no makeup is added. The operator will be expected to determine that the Spent Fuel Pool level will go down at a rate of 3.1 inches/hour, and that the expected time that the Dose Rate in the Spent Fuel Area exceeds 100 mrem/hour is 41.8+/-2 hours.

A4 This is a modified Bank JPM. The operator will be placed in a post-accident condition with a Large Break LOCA with a release from the Containment. The operator will be told that a General Emergency has been declared, and provided with the initial Protective Action Recommendation (PAR). The operator will be given a subsequent set of plant conditions and meteorological data, and asked to provide an updated PAR in accordance with Enclosure 4.4 (Offsite Protective Recommendations) of RPIO/615700/029 (Notifications to Offsite Agencies from the Control Room). The operator will be expected to determine the Updated PAR for the subsequent conditions.

NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems OutHne Form ES-301-2 ORAFT (REVJ73013)

Facility: McGuire Date of Examination: 12/2013 Exam Level (circle one): RD (only) / SRO(I) I SRO (U) Operating Test No.: NI 3-I Control Room Systemse (8 for RO: 7 for SRO-l; 2 or 3 for SRO-U, including 1 ESF)

, Type Code Safety System I JPM Title Function A. EPE 007 Reactor Trip/Stabilization/Recovery (007 EA1 .08 (4.4/4.3)]

S,D,A,EN I CA Suction Source Realignment B. E02 SI Termination [E02 EAI.1 (4.013.9)]

S,N,A,EN 3 Isolate the NV S1l Flowpath While Terminating Safety Injection C. APE 022 Loss of Rx Coolant Makeup [022 AA1 .04 (3.3/3.2)]

SMA 2 Establish Charging Flow Using the PD Pump D. 003 Reactor Coolant Pump System [003 A4.01 (3.313.2)]

S,M,L 4P Start and Stop the lB NCP for NCS Venting E. 027 Containment Iodine Removal System [027 A4.01 (3.3/3.3)]

S,D,EN 5 Perform the 1 A Annulus Ventilation Operability Test F. APE 067 Plant Fire On Site [067 AA2.04 (3.1/4.3)]

S,P,D,A 8 Restore from a Fire in the Unit I Cable Spreading Room G. 015 Nuclear Instrumentation System [015 A2.01 (3.5/3.9)]

S,D,L,A 7 Respond to a Source Range Nuclear Instrumentation Failure H. APEO56 Loss of Off-Site Power [056 AA 1.02 (4.0/3.9)]

SD,EN $

Restore Normal Power to 1ETB and Unload the 18 EDO In-Plant Systems@ (3 for RO; 3 for SRQ-l; 3 or 2 for SRO-U)

I. APE 054 Loss of Main Feedwater [054 AM .02 (4.414.4)]

D,R,E 4S Reset Unit 2 Turbine Driven CA Pump Stop Valve J. APE 069 Loss of Containment Integrity [069 AA1 .03 (2.8/3,0)]

D,R,E 5 Start the Hydrogen Analyzers K. EPE 055 Station Blackout [055 EK3.02 (4.314.6)]

DE 6 Establish NC Pump Seal Injection From the SSF NUREG-1 021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 DRAFT (REV_073013)

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SAC-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

Type Codes Criteria for AC / SRO-l / SAC-U (A)ltemate path 4-6 (5) /4-6 (5) / 2-3 (2)

(C)ontrol room (D)irect from bank 9 (8) /<; 8 (7) / 4 (2)

(E)mergency or abnormal in-plant 1 (3) / 1 (3) / 1 (2)

(EN)gineered Safety Feature / / 1 (1) (ConhoIRornSystm)

(L)ow-Power/ Shutdown 1 (2) k 1 (2) / 1 (1)

(N)ew or (M)odified from bank including I (A) 2 (3) / 2 (3) / 1 (3)

(P)revious 2 exams 3 (1) / 3 (1)1 2 (0) (RandmySIected)

(A)CA 1 (2)/1 (2)/1 (1) jS)imulator mmar JPM A This is a Bank JPM. The operator will be told that Unit 1 has just tripped from 100%

power, due to seismic activity, that the crew is now implementing EP/1/A/5000/ES-O.1 (Reactor Trip Response), that the CAST has developed a leak, and that level has lowered to 1. feet. The operator will be directed to perform EPJ1/N5000/G-1, Generic Enclosure 20 (CA Suction Source Realignment), while the crew continues with ES-0.1.

The operator will be expected to realign the suction of the CA Pumps from the non-safety related to the safety-related source (RN). During the course of this action, the operator will recognize that RN Supply to the lB MDCA Pump cannot be established (Alternate Path), and stop the pump.

JPM B This is a New JPM. The operator will be told that a plant transient at Unit 1 has resulted in a reactor trip from power and a Safety Injection actuation, that Safety Injection Termination criteria has been met and that the crew is implementing EP/1/A15000/ES-1.1 (Safety Injection Termination). The operator will be directed to isolate the NV S/I flowpath in accordance with Step 6 of EP/1/N5000/ES-1 .1 (Safety Injection Termination). While performing this task the operator will recognize that one NV Pump Recirculation Valve cannot be opened (Alternate Path). The operator will be expected to attempt to isolate NV S/I flow per EP/1/A1500/ES-l .1, recognize that this cannot be accomplished and complete Generic Enclosure 18 (Aligning Normal Charging With NV Recirc Path Isolated) to isolate NV S/I flow.

JPM C This is a modified Bank JPM. The operator will be told that Unit I is at 100% power when lB NV Pump breaker trips on Overcurrent, and that AP/1/N5500/12 (Loss of Letdown, Charging, or Seal Injection) is implemented and completed through step 16.

The operator will be directed to re-establish charging, beginning with step 20 of AP/1/A/5500/12 (Loss of Letdown, Charging, or Seal Injection). When the operator attempts to start the 1 A NV Pump, this Pump will NOT start (Alternate Path), The operator will be expected to place the PD Pump in service and establish Charging and Seal Injection flow.

JPM D This is a modified Bank JPM. The operator will be told that a plant startup is in progress per OP/1/A/6100/001 (Controlling Procedure For Unit Startup), that the crew is NUREG-1 021, Revision S

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 DRAFT (REV_073013) implementing Enclosure 4.2 (Venting the NC System (Control Room Activities)) of OP/1/A/6100/SU-6 (Venting the NC System), that the NC System is water solid, and that NC System pressure is being maintained between 330-370 psig. The operator will also be told that the crew is ready to conduct a 60 second run of the 1 B NC Pump, and that Enclosure 4.1 (Startup and Operation) of OP/1/A/6150/002 A (Reactor Coolant Pump Operation) has been marked up for place-keeping through step 3.3 to support NC Pump operation. The operator will be directed to start the 1 B NCP per Section 3,5 of Enclosure 4.1 (Startup and Operation) of OP/1/A/6150/002 A (Reactor Coolant Pump Operation); and then stop the 1 B NCP after 60 seconds of operation, or if a low temperature condition develops. The operator will be expected to conduct a 60 second run of the 1 B NC Pump in accordance with Enclosure 4.1 of OP/1/A/61 50/002 A.

JPM 2 This is a Bank JPM. The operator will be told that Unit 1 is operating at 100% power, that Unit 1 VE System is aligned for Engineered Safeguards Operation, and that PT/1/A14450/003 A (Annulus Ventilation System Train A Operability Test) is on the Operations schedule for today. The operator will be directed to perform PT/1/AJ4450/003 A (Annulus Ventilation System Train A Operability Test). The operator will be expected to place the 1A VE Fan in Recirculation Mode with the cross connect from B Train closed. The 1A VE Fan will be shut down after flow verification and returned to normal alignment.

JPM F This is a Bank JPM. The operator will be told that Unit 1 is at 100% power and that a fire has been reported in the Unit I Cable Spreading Room. The operator will be told that the crew has implemented AP/0/A/5500/45 (Plant Fire) and is presently in Enclosure 17 (AB 750 Unit 1 Cable Spreading Room Fire Unit 1 and Unit 2 Actions). The operator will also be told that several control room switch manipulations have been made, that the Fire Brigade has reported that the fire is no longer active, and that Station Management has indicated that the crew may return Control Room controls to normal as identified within Enclosure 17. The operator will be directed to restore the Control Room controls to normal by performing Step 21.a through e of Enclosure 17 (AB 750 Unit I Cable Spreading Room Fire Unit 1 and Unit 2 Actions) of AP/0/AJ5500/45 (Plant Fire). The operator will be expected to determine that one Pzr PORV has inadvertently opened, and take action to isolate it by ensuring that its isolation valves is closed, and by directing that its motor breaker be opened (Alternate Path). The operator will then open the remaining Pzr PORV isolation valves, direct that the motor breaker for ICA-7A be closed, and open the manual loaders for the Main Steam Line PORVs while the valves remain closed.

3PM G This is a Bank 3PM. The operator will be told that Unit 1 is in the process of performing a Unit Start-Up, that Control Bank A has been pulled to 50 steps withdrawn, that Mode 2 has just been declared, that Source Range count rate is approximately 60 CPS, that SR channel N-31 has failed low, but that SR channel N-32 is still OPERABLE. The operator will also be told that AP/1/A/5500/1 6 Case I (Source Range Malfunctions) has been entered. The operator will be directed to remove the failed SR channel (N-31) from service by performing AP/1/A5500/1 6 (Malfunction of Nuclear Instrumentation), Case I (Source Range Malfunctions). The operator will be expected to place the Source Range N31 Level Trip Bypass Switch to Bypass, and ensure that Source Range N32 is selected on the NIS Recorder. When the only other channel of SR Nuclear instrumentation fails high and the reactor does not trip (Alternate Path), the operator will recognize that a reactor trip is required and then manually trip the Reactor.

NUREG-1 021, Revision 9