Text

Final Outlines (Folder 3
	ML15043A678
Person / Time
Site:	Ginna
Issue date:	12/01/2014
From:	D'Antonio J; Operations Branch I
To:	; Exelon Generation Co
Shared Package
ML14174A998	List: ML14177A718; ML14339A320; ML15043A673; ML15043A675; ML15043A678; ML15043A679; ML15043A680; ML15043A681; ML15156A552;
References
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	Download: ML15043A678 (42)
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ES-401 PWR Examination Outline FORM ES-401-2 Facility Name: Robert E Ginna Date of Exam:12/03/14 RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 2 3 4

1 3 3 3 3 3 3 18 3 3 6

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

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

3. Generic Knowledge and Abilities 10 7 Categories 3 2 2 3 1 2 2 2 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 (i.e., 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 NRG 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 Olltline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems 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) KlAs 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 KlAs.

8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (I Rs) 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 KlAs that are linked to 10 CFR 55.43.

ES-401, 21 of 33 NRC EXAM MAT'L

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)

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

1 2 3 1 0 Reactor trip (scram): verification that the control and safety rods 000007 Reactor Trip - Stabilization - Recovery/ 1 4.4 6 are in after the trip 0

2 00008 Pressurizer Vapor Space Accident I 3 Controllers and positioners 2.5 3

3 00009 Small Break LOCA I 3 Letdown isolation valve position indication 2.9 0

4 00011 Large Break LOCA I 3 Natural circulation and cooling, including reflux boiling 4.1 1

00015 RCP Malfunctions I 4 0

00017 RCP Malfunctions (Loss of RC Flow)/ 4 000022 Loss of Rx Coolant Makeup I 2 0 5 00025 Loss of RHR System I 4 Knowledge of system purpose and/or function. 3.9 0 The automatic actions (alignments) within the CCWS resulting 6 000026 Loss of Component Cooling Water I 8 from the actuation of the ESFAS 3.6 2

000027 Pressurizer Pressure Control System 0

Malfunction I 3 00029 A TWS / 1 0 7 00038 Steam Gen. Tube Rupture I 3 Radiation levels (MREM/hr) 3.9 00040 Steam Line Rupture - Excessive Heat Transfer 0 8 Sensors and detectors 2.6 4 2 E12 Uncontrolled Depressurization of all Steam enerators I 4 0

9 000054 (CE/E06) Loss of Main Feedwater / 4 Matching of feedwater and steam flows 3.4 2

10 000055 Station Blackout I 6 Knowledg** of EOP mitigation strategies. 3.7 0

11 000056 Loss of Off-site Power I 6 Definition of saturation conditions, implication for the systems 3.1 4

The indicator, valve, breaker, or damper position which will occur 12 00057 Loss of Vital AC Inst. Bus/ 6 3.1 on a loss of power 00058 Loss of DC Power I 6 0 0

13 000062 Loss of Nuclear Svc Water I 4 Loads on the SWS in the control room 3.2 2

0 Knowing effects on plant operation of isolating certain equipment 14 000065 Loss of Instrument Air I 8 2.9 3 from instrument air 0 Annunciators and conditions indicating signals, and remedial 15 /E04 LOCA Outside Containment I 3 3.5 3 actions associated with the LOCA Outside Containment Ability to evaluate plant performance and make operational 16 /E11 Loss of Emergency Coolant Recirc. / 4 judgments based on operating characteristics, reactor behavior, 4.4 and instrument interpretation.

Facility's heat removal systems, including primary coolant, BW/E04; W/E05 Inadequate Heat Transfer - Loss of 0 emergency coolant, the decay heat removal systems, and 17 3.9 econdary Heat Sink I 4 2 relations between the proper operation of these systems to the operation of the facility 00077 Generator Voltage and Electric 0 18 Reactor controls 4.1 rid Disturbances I 6 4 KIA Category Totals: 3 3 3 3 Group Point Total: 18 ES-401, 22 of 33 NRC EXAM MAT'L

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)

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

2 3 1 00001 Continuous Rod Withdrawal/ 1 0 00003 Dropped Control Rod I 1 0 00005 Inoperable/Stuck Control Rod I 1 0 00024 Emergency Boration / 1 0 19 00028 Pressurizer Level Malfunction I 2 PZR level indicators and alarms 3.4 00032 Loss of Source Range NI I 7 0 00033 Loss of Intermediate Range NI I 7 0 00036 Fuel Handling Accident I 8 0 Ability to interpret control room indications to verify the status and 20 00037 Steam Generator Tube Leak I 3 operation of a system, and understand how operator actions and 4.2 directives affect plant and system conditions.

00051 Loss of Condenser Vacuum/ 4 0 00059 Accidental Liquid RadWaste Rel. I 9 0 00060 Accidental Gaseous Radwaste Rel. I 9 0 21 00061 ARM System Alarms/ 7 01 Detectors at each ARM system location 2.5 00067 Plant Fire On-site I 8 0 00068 Control Room Evac. I B 0 00069 Loss of CTMT Integrity I 5 22 /E14 High Containment Pressure/ 5 02 Operating behavior characteristics of the facility 3.3 00074 lnad. Core Cooling / 4 Manipulation of controls required to obtain desired operating 23 /E06 Degraded Core Cooling I 4 03 results during abnormal, and emergency situations 4.0

/E07 Saturated Core Cooling I 4 00076 High Reactor Coolant Activity I 9 0

/E01 Rediagnosis I 3 Components, and functions of control and safety systems, 24 /E02 SI Termination I 3 01 including instrumentation, signals, interlocks, failure modes, and 3.4 automatic and manual features Annunciators and conditions indicating signals, and remedial 25 /E13 Steam Generator Over-pressure/ 4 03 actions associated with the Steam Generator Overpressure 3.0 Adherence to appropriate procedures and operation within the 26 /E 15 Containment Flooding I 5 2.9 limitations in the facility's license and amendments

/E 16 High Containment Radiation I 9 0

/E03 LOCA Cooldown - Depress. I 4 0

/E09 Natural Circulation Operations I 4

/E10 Natural Circulation with Steam Voide in 27 02 Operating behavior characteristics of the facility 3.6 essel with/without RVLIS. / 4

/EOB RCS Overcooling - PTS I 4 0 KIA Category Totals: 2 2 Group Point Total:

ES-401, 23 of 33 NRC EXAM MAT'L

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

Q#

28 System # I Name 003 Reactor Coolant Pump K K 1 2 4 K

o~

K 6 AF A 1 3 4 KIA Topic(s)

Effects of RCP shutdown on secondary parameters, such IR 3.2 1

as steam pressure, steam flow, and feed flow 29 004 Chemical and Volume Control 0

II RCPS 3.7 1 30 31,32 005 Residual Heat Removal 006 Emergency Core Cooling I- ,..._

Pressure transient protection during cold shutdown Conditions requiring actuation of ECCS: Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

3.5 4.5; 4.5 1

2 0

33,34 007 Pressurizer Relief/Quench Tank RCS: Abnormal pressure in the PRT 3; 2.6 2 3

0 35 008 Component Cooling Water CCW pressure 2.7 1 3

0 36 010 Pressurizer Pressure Control

~ 2.8 1 Indicator for PORV position 3

0 Trip logic circuits; Ability to prioritize and interpret the 3.1; 37,38 012 Reactor Protection significance of each annunciator or alarm. 2 3 i 4.1 013 Engineered Safety Features 1 39 MFW isolationlreset 3.7 1 Actuation 3 ll!!'~i1 40 022 Containment Cooling

~ llit Cooling water flow 3.2 1 025 Ice Condenser !;

0 y 0 Containment sump level; Pump starts and correct MOV 3.5; 41,42 026 Containment Spray positioning 2 3 1 4.3 43 039 Main and Reheat Steam Flow paths of steam during a LOCA 3.1 1 0 0 ;;

44,45 059 Main Feedwater S/Gs; Feed regulating valve controller 3.5; 3 2 8

0 46 061 Auxiliary/Emergency Feedwater ~ "l Relationship between AFW flow and RCS heat transfer 3.6 1 1

0 0 Major system loads; Operation of inverter (e.g., precharging 3.5; 47,48 062 AC Electrical Distribution synchronizing light, static transfer) 2 1 'I 4 2.7 0

~-**-~-,-**-a- 2.9; I-49,50 063 DC Electrical Distribution parameters that are entry-level conditions for Technical 2 3 Specifications. 3.9 0

51 064 Emergency Diesel Generator t* Fuel oil storage tanks 3.2 1 8

52 073 Process Radiation Monitoring 0

1 I Release termination when radiation exceeds setpoint 4.0 1 1~1 0 !ft 53 076 Service Water RHR system 3.5 1 8

54 078 Instrument Air 0

2 ; ~

Emergency air compressor 3.3 1 55 103 Containment 0

4 f r.'~

Personnel access hatch and emergency access hatch 2.5 1 3 2 3 3 2 2 3 4 2 1 3 i Group Point Total:

0 28 ES-401, Page 24 of 33

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

K K K K K K Q# System # I Name KIA Topic(s) IR #

1 2 3 4 5 6 0

56 01 Control Rod Drive 3

One-line diagram of power supplies to logic circuits 2.7 57 02 Reactor Coolant Reactor coolant leak detection system 3.7 Reasons for selecting "manual" on letdown control valve 58 11 Pressurizer Level Control controller 2.5 0

59 14 Rod Position Indication 3.2 Knowledge of the bases in Technical Specifications for limiting 60 15 Nuclear Instrumentation conditions for operations and safety limits.

3.2 0

61 16 Non-nuclear Instrumentation 3

Input to control systems 2.8 17 In-core Temperature Monitor 0 27 Containment Iodine Removal 0 28 Hydrogen Recombiner and Purge 0

ontrol 29 Containment Purge 0 33 Spent Fuel Pool Cooling 0 62 34 Fuel Handling Equipment Radiation monitoring systems 2.6 35 Steam Generator 0 63 41 Steam Dumpffurbine Bypass Control T-ave., verification above low/low setpoint 2.9 5 Main Turbine Generator 0 55 Condenser Air Removal 0 56 Condensate 0 68 Liquid Radwaste 0 64 71 Waste Gas Disposal ' The double verification required before waste gas release 2.9 65 72 Area Radiation Monitoring *' Erratic or failed power supply 2.7 75 Circulating Water 0 79 Station Air 0 86 Fire Protection 0 KIA Category Totals: 0 10 ES-401, Page 25 of 33 NRC EXi\M MAT'L

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 (SRO)

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

1 2 3 1 000007 Reactor Trip - Stabilization - Recovery/ 1 0 00008 Pressurizer Vapor Space Accident I 3 0 00009 Small Break LOCA I 3 0 000011 Large Break LOCA I 3 0 00015 RCP Malfunctions/ 4 76 Ability to interpret and execute procedure steps. 4.6 00017 RCP Malfunctions (Loss of RC Flow)/ 4 Ability to verily system alarm setpoints and operate controls 77 000022 Loss of Rx Coolant Makeup I 2 identified in the alarm response manual.

4.0 000025 Loss of RHR System I 4 0 00026 Loss of Component Cooling Water I 8 0 00027 Pressurizer Pressure Control System Knowledge of annunciator alarms, indications, or response 78 procedures.

4.1 Malfunction I 3 79 00029 A TWS / 1 Main turbine trip switch position indication 3.9 00038 Steam Gen. Tube Rupture I 3 0 000040 Steam Line Rupture - Excessive Heat Transfer 4

E12 Uncontrolled Depressurization of all Steam Adherence to appropriate procedures and operation within the 80 limitations in the facility's license and amendments 3.9 Generators I 4 00054 (CE/E06) Loss of Main Feedwater I 4 0 000055 Station Blackout I 6 0 000056 Loss of Off-site Power I 6 0 000057 Loss of Vital AC Inst. Bus/ 6 0 81 000058 Loss of DC Power I 6 125V de bus voltage, low/critical low, alarm 3.6 000062 Loss of Nuclear Svc Water I 4 0 000065 Loss of Instrument Air I 8 0 0

/E 11 Loss of Emergency Coolant Recirc. I 4 0 BW/E04; W/E05 Inadequate Heat Transfer - Loss of 0

econdary Heat Sink I 4 0

0 0 0 0 ES-401, 22 of 33 NRC EJCAM MAT'L

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 (SRO)

.-~~~~~~~~~~~~~~~~~~~~ .......~.......~.......~.......~ ~~~~~~~~~~~~~~~~~~~-r-~~-r-~~~*

K K K A KIA Topic(s) IR #

Q# E/APE #I Name I Safety Function 1 2 3 1 000001 Continuous Rod Withdrawal/ 1 0 000003 Dropped Control Rod / 1 0 00005 Inoperable/Stuck Control Rod / 1 0 00024 Emergency Boration I 1 0 00028 Pressurizer Level Malfunction I 2 0 000032 Loss of Source Range NI I 7 0 00033 Loss of Intermediate Range NI I 7 0 Occurrence of a fuel handling incident 4.1 000037 Steam Generator Tube Leak/ 3 0 000051 Loss of Condenser Vacuum/ 4 0 000059 Accidental Liquid RadWaste Rel. I 9 0 000060 Accidental Gaseous Radwaste Rel. I 9 0 00061 ARM System Alarms 17 0 00067 Plant Fire On-site I 8 0 00068 Control Room Evac. I 8 0 00069 Loss of CTMT Integrity I 5 0

/E14 High Containment Pressure/ 5

/E06 Degraded Core Cooling I 4 0

/E07 Saturated Core Cooling I 4 00076 High Reactor Coolant Activity I 9 0

/E01 Rediagnosis I 3 0

/E02 SI Termination I 3

/E13 Steam Generator Over-pressure/ 4 0

/E15 Containment Flooding/ 5 0 Facility ccnditions and selection of appropriate procedures during 83 /E16 High Containment Radiation/ 9 abnormal and emergency operations 3.3 Knowledge of the operational implications of EOP warnings, 84 /E03 LOCA Cooldown - Depress. I 4 cautions, and notes.

4.3

/E09 Natural Circulation Operations I 4 0

/E10 Natural Circulation with Steam Voide in essel with/without RVLIS. / 4 85 /E08 RCS Overcooling - PTS I 4 Knowledge of the specific bases for EOPs. 4.0 KIA Category Totals: 0 0 0 0 Group Point Total: 4 ES-401, 23 of 26 NRC EXAM MAT'L

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2

. - - - l t - - - - - - - - - - - - - - r - - r - - r - - r - - r - - r - - r - . . . . . . . , P l a n t Systems - Tier 2/Group 1 (SRO)

Q# System #I Name K1K2K3K4K5K6A1 KIA Topic(s) IR #

003 Reactor Coolant Pump 0 86 004 Chemical and Volume Control Improper RWST boron concentration 4.2 87 005 Residual Heat Removal Ability to apply Technical Specifications for a system. 4.7 006 Emergency Core Cooling 0 007 Pressurizer Relief/Quench Tank 0 008 Component Cooling Water 0 010 Pressurizer Pressure Control 0 012 Reactor Protection 0 013 Engineered Safety Features 88 Rapid depressurization 4.7 ctuation 022 Containment Cooling 0 025 Ice Condenser 0 026 Containment Spray 0 39 Main and Reheat Steam 0 059 Main Feedwater 0 061 Auxiliary/Emergency Feedwater 0 062 AC Electrical Distribution 0 89 063 DC Electrical Distribution Grounds 3.2 064 Emergency Diesel Generator 0 073 Process Radiation Monitoring 0 076 Service Water 0 078 Instrument Air 0 Ability to interpret reference materials, such as graphs, 90 103 Containment curves, tables, etc. 4.2 0

KIA Category Totals: 0 0 0 0 0 0 0 Group Point Total: 5 ES-401, Page 24 of 33

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

K K K K K K Q# System # I Name 1 2 5 6 KIA Topic(s) IR #

3 4 01 Control Rod Drive 0 02 Reactor Coolant 0 11 Pressurizer Level Control 0 14 Rod Position Indication 0 15 Nuclear Instrumentation 0 16 Non-nuclear Instrumentation 0 91 17 In-core Temperature Monitor Thermocouple open and short circuits 3.5 27 Containment Iodine Removal 0 28 Hydrogen Recombiner and Purge 0

ontrol 29 Containment Purge 0 92 33 Spent Fuel Pool Cooling Abnormal spent fuel pool water level or loss of water level 3.5 34 Fuel Handling Equipment 0 Knowledge of limiting conditions for operations and safety 93 35 Steam Generator limits.

4.7 1 Steam DumpfTurbine Bypass Control 0 5 Main Turbine Generator 0 55 Condenser Air Removal 0 56 Condensate 0 68 Liquid Radwaste 0 71 Waste Gas Disposal 0 72 Area Radiation Monitoring 0 75 Circulating Water 0 79 Station Air 0 86 Fire Protection 0 KIA Category Totals: 0 0 0 0 0 0 0 2 0 0 1 Group Point Total: 3 ES-401, Page 25 of 33

.i' NRC EXAMMAT'L

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

"- ,,,._ ~* -- ,.., ""t E Ginna Date of Exam:12/03/14 KU SRO-Only

-Q# Category KIA# Topic IR # IR #

66 2.1. 08 Ability to coordinate personnel activities outside the control room. 3.4 1 Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor 67 2.1. 14 trips, mode changes, etc. 3.1 1 1.

- Conduct of Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high 68 Operations 2.1. 26 pressure, caustic, chlorine, oxygen and hydrogen). 3.4 1 94 2.1. 23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. 4.4 1

- Subtotal 3 1 69 2.2. 37 Ability to determine operability and/or availability of safety related equipment. 3.6 1 70 2.2. 41 Ability to obtain and interpret station electrical and mechanical drawings. 3.5 1 2.

- Equipment 95 2.2. 06 Knowledge of the process for making changes to procedures. 3.6 1

- Control Knowledge of the process for managing maintenance activities during shutdown operations, such as risk 96 2.2. 18 assessments, work prioritization, etc. 3.9 1

- Subtotal 2 2 71 2.3. 07 Ability to comply with radiation work permit requirements during normal or abnormal conditions. 3.5 1

- Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency 72 2.3. 14 conditions or activities. 3.4 1

- 3.

Radiation Ability to use radiation monttoring systems, such as fixed radiation monitors and alarms, portable survey 97 2.3. 05 instruments, personnel monitoring equipment, etc. 2.9 1 Control 98 2.3. 11 Ability to control radiation releases. 4.3 1 Subtotal 2 2 Knowledge of the organization of the operating procedures network for normal, abnormal, and 73 2.4. 05 emergency evolutions. 3.7 1

- Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity 74 2.4. 21 control, core cooling and heat removal, reactor coolant system integrity, containment conditions, 4.0 1 radioactivity release control, etc.

- 4.

75 Emergency 2.4. 41 Knowledge of the emergency action level thresholds and classifications. 2.9 1

- Procedures I

- 99 Plan 2.4. 08 Knowledge of how abnormal operating procedures are used in conjunction with EOPs. 4.5 1 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate 100 2.4. 47 control room reference material. 4.2 1 Subtotal 3 2 Tier 3 Point Total 10 7 ES-401, Page 26 of 33

ES-401 Record of Rejected K/As Form ES-401-4 Tier I Randomly Reason for Rejection Group Selected KIA T2,G1 006 A2.12 Per 06/12/14 telecom with Chief Examiner. Original KIA (4.5/4.8) A2.09 does not reflect the fact that Ginna's RWST is normally (RO) vented to atmosphere. "Reselect based upon the inability to write a discriminatory question." Randomly reselected.

T2,G1 012 K6.03 Per 06/12/14 telecom with Chief Examiner. Original KIA (3.1/3.5) K6.08 refers to a system not recognized at Ginna (Core (RO)

Operating Limit Support System). "Reselect if the system doesn't apply to Ginna." Randomly reselected.

T1, G1 APE 022 2.4.50 Per 06/12/14 telecom with Chief Examiner. Original KIA (4.2/4.0) 2.4.49. Ginna does not have any Abnormal Procedures which (SRO) have Immediate Operator Actions. "Two choices: (1) Reselect or (2) On Loss of Charging, what do you expect the operators to do?" Randomly reselected.

T1, G2 APE 037 2.2.44 Per 07/29/14 telecom with Chief Examiner. Original KIA (4.2/4.4) 2.2.36. Unable to write a discriminating question on this (RO) subject in this abnormal procedure. "Reselect based upon the inability to write a discriminating question."

T1, G1 2.4.31 Per 08/18/14 telecom with Chief Examiner. Original KIA (4.2/4.1) 2.4.35. The given APE does not have any local operator (SRO) actions associated with it. Randomly reselected.

T2,G1 007 A2.02 Per 10/6/14 telecom with Chief Examiner. Original KIA 007 (2.6/3.2) A4.01. Subject KIA is not relevant to the Ginna facility. Ginna (RO) does not have a PRT Supply Spray valve as stated in the KIA. "Reselect a KIA" T1, G1 057 AA2.18 Per 10/6/14 telecom with Chief Examiner. Original KIA 057 (3.1/3.1) AA2.11. Subject KIA is not relevant to the Ginna facility.

(RO)

Ginna does not have Feed Pump controllers and are constant speed motors that use Feed Regulating valves to control flow.

"Reselect KIA" 2014 ILT NRC Exam Rev. 07/29/14

ES-301 Administrative Topics Outline Form ES-301-1 (Rev 090314)

Facility: Ginna Date of Examination: 12/2014 Examination Level: SRO Operatin9 Test Number: 2014-301 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 ~Jraphs, curves, tables, etc.

D, P, R JPM: Determine Time to Boil for a Loss of Shutdown Cooling 2.1.8 (4.1) Abillity to coordinate personnel activities Conduct of Operations outside the control room.

M,R JPM: Cold Weather Walkdown 2.2.12 (4.1) Knowledge of Surveillance Procedures.

Equipment Control N,R JPM: Perform RCS Leakrate Surveillance 2.3.15 (3.1) Knowledge of radiation monitoring systems, Radiation Control such as fixed radiation monitors and alarms, portable survey instruments, M,R personnel monitoring equipment, etc.

JPM: Evaluate the Impact of R-14 Failure on a Planned Gaseous Decay Tank Release 2.4.41 (4.6) Knowledge of emergency action level Emergency thrE~sholds and classifications.

Procedures/Plan M,S JPM: Classify an Emergency Event NOTE: All items (5 total) are required for SROs. RO 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, (1) or Class(R)oom (4)

(D)irect from bank (:o; 3 for ROs; :o; 4 for SROs & RO retakes) (1)

(N)ew or (M)odified from bank(~ 1) (4)

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

NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 (Rev 090314)

SRO Admin JPM Sumn!!!Y A 1a This is a Bank JPM randomly selected from the last two NRC Exams. The operator will be provided with two sets of shutdown conditions; one current, and one projected to exist within four days. The operator will be directed to determine the Time to Boil given a Loss of RHR for each of the two sets of conditions. The operator will be expected to use IP-OUT-2, Outage Risk Management, and determine that the Time to Boil for the present plant conditions is 2.46 hours5.324074e-4 days 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months , and for the projected conditions four days from now is .43 hours4.976852e-4 days 0.0119 hours 7.109788e-5 weeks 1.63615e-5 months (25.96 +0.1/-0.0 minutes).

A 1b This is a modified Bank JPM. The operator will b1e told that the weekly Cold Weather Walkdown is in progress, and the Primary, Secondary and Outside Auxiliary Operators are about ready to start their Walkdown Tours; and provided with a recently completed Attachment 1, Cold Weather Temperature Log, of 0-22, Cold Weather Walkdown Procedure. They will also be told that a severe lake-effect storm is occurring with 24 inches of snow having accumulated over the last four hours. The operator will be directed to review Attachment 1 and perform section 6.1, General Instructions, of 0-22, and prioritize any areas of concern for the Auxiliary Operator Walkdown Tours; and then to document discrepancies, if any, and any required action(s). The operator will be expected to review the most recent plant area temperature data and then use 0-22 to determine that two areas of the plant are too low in temperature. Then the operator will prioritize the actions of the Auxiliary Operators to tour these areas first, identify discrepancies and actions required in accordance with the attached KEY.

A2 This is a new JPM. The operator will be told that the plant is at 100% power with 40 gpm Letdown flow, that the VCT Level point on the PPCS is OOS, that at 1020 the VCT level was 25% when the HCO started the Day Shift RCS Leakage Monitoring Surveillance, intending to monitor for about 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and that at 1125 a VCT Level instrument failure occurred which l&C corrected and restored to operation within 10 minutes. The operator will be provided with a current Attachment 3, RCS Leakage Surveillance Record, with recent Auto Makeup information, and directed to record the data for the recent Auto Makeup on Attachment 3 AND to determine the current RCS Leakrate. The operator will also be directed to identify any Tech Spec LCOs that have been exceeded, and if any LCOs are exceeded, identify any required Tech Spec ACTION. The operator will be expected to complE3te Attachment 3 in accordance with the KEY provided, and identify the LCO 3.4.13, RCS Operational Leakage, is NOT met, and that the required ACTION is to reduce leakage to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (A.1).

A3 This is a modified Bank JPM. The operator will be told that the plant is at 100% power, that an A Gas Decay Tank release is being planneid, that the HCO reports that R-14A, Plant Vent Gas, is de-energized for preventative maintenance that will require 2 more hours, and must be completed before it can be returned to service, and that R-14, Plant Vent Gas, has just failed LOW and l&C has indicated that it will take at least four hours to troubleshoot and correct the failure. The opera1tor will be directed to determine the required actions based on the failure of R-14, and to determine the earliest time at which the A Gas Decay Tank release can start, and to ide!ntify any actions that will be required to start the release. The operator will be expected to determine that ('1) on-going releases (i.e. ventilation) may continue provided grab samples are taken and analyzed for isotopic activity at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , that (2) at least one of either R-'14 or R-14A must be restored to OPERABLE status within 30 days or it must be explained in the next NUREG-1021, Revision 9

ES-301 Administrative Topics Outlline Form ES-301-1 (Rev 090314)

Annual Radioactive Effluent Release Report, and that (3) the earliest time that the A Gas Decay Tank release can be initiated is 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from now when the maintenance on R-14A is completed, and ONLY if at least two independent samples of the tank's contents are analyzed, and at least two technically quallified members of the facility staff independently verify the release rate calculations and discharge valve lineup.

A4 This is a modified Bank JPM. This JPM will be given immediately following the associated Simulator Scenario. Consequently, there are five separate versions of this JPM, one each associated with the submitted Simulator Scenarios. A portion of this JPM is Time Critical. After each Simulator Scenario, the operator will be told that the sequence of events that they have just performed on the recently completed Simulator Scenario must be evaluated for any Emergency Classifications. The operator will be directed to determine the Emergency Classification that would have been made first in the recently completed scenario, and Identify thE~ basis for this classification. The operator will be given a time that all the indications needed to classify the first event requiring an Emergency Classification were apparent in the Control Room and then directed to identify the latest time at which the Emergency Classification must have been made. Finally, the operator will be directed to identify any subsequent emergency classifications that would have been made in the Scenario, and the basis for this classification. The operator will be expected to make the appropriate classifications and identify the time frames as appropriate.

NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 (Rev_091114)

Facility: Ginna Date of Examination: 12/2014 Examination Level: RO Operatin!~ Test Number: 2014<101 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 !~raphs, curves, tables, etc.

D,R JPM: Calculate SDM for an Operatini~ Reactor witll an Untripable Rod 2.1.43 (4.1) Ability to use procedures to determine the Conduct of Operations effocts on reactivity of plant changes, such as reactor coolant system temperature, N,R secondary plant, fuel depletion, etc.

JPM: Perform an Independent Verification of a Post-Trip Xenon Reactivity Calculation 2.2.12 (3. 7) Knowledge of Surveillance Procedures.

Equipment Control N,R JPM: Perform RCS Leakrate Surveillance 2.3.5 (2.9) Ability to use radiation monitoring systems, Radiation Control such as fixed radiation monitors and alarms, portable survey instruments, N,R personnel monitoring equipment, etc.

JPM: Evaluate Steam Generator Tube Leakage from R-47 Reading NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are re~quired.

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

(D)irect from bank(::; 3 for ROs;::; 4 for SROs & RO retakes) (1)

(N)ew or (M)odified from bank (~ 1) (3)

(P)revious 2 exams (::; 1; randomly selected) (0)

NRC EJLW Mi\T'L NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 (Rev 091114)

RO Admin JPM Summm A 1a This is a Bank JPM. The operator will be provided with a set of data relating to normal power plant operation, and told that Rod K-7 has be!en determined to be untripable. The operator will be directed to calculate Shutdown Mar!~in (SOM), and determine if sufficient SOM exists for the current plant conditions. The operator will be expected to calculate Shutdown Margin in accordance with the 0-3.2, Shutdown Margin for an Operating Reactor, and determine that sufficient SOM does NOT exist.

A 1b This is a new JPM. The operator will be told that the plant has tripped from 50% power, that the reactor had been operating at this power for the last 3 days prior to the trip, that the XENON PREDICT program on the PPCS is NOT available, that the Reactor Engineer has completed the Xenon Reactivity Calc:ulation, and that it is currently 0900 on 12/3/14. The operator will be directed to independently verify the Xenon Reactivity Calculation in accordance with Step 6.1.3 of 0-3, Hot Shutdown With Xenon Present.

The operator will be expected to complete an independent verification of the Xenon Reactivity Calculation and determine that Xenon will equal the value at time of shutdown at 2200 on 12/3/14.

A2 This is a new JPM. The operator will be told that the plant is at 100% power with 40 gpm Letdown flow, that the VCT Level point on the PPCS is OOS, that at 1020 the VCT level was 25% when the HCO started the Day Shift RCS Leakage Monitoring Surveillance, intending to monitor for about 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and that at 1125 a VCT Level instrument failure occurred which l&C corrected and restored to operation within 1O minutes. The operator will be provided with a ciurrent Attachment 3, RCS Leakage Surveillance Record, with recent makeup information, and directed to record the data for the recent Auto Makeup on Attachment 3 AND to determine the current RCS Leakrate.

The operator will also be directed to identify any Tech Spec LCOs that have been exceeded. The operator will be expected to compl1ete Attachment 3 in accordance with the KEY provided, and identify the LCO 3.4.13, RCS Operational Leakage, is NOT met.

A3 This is a new JPM. The operator will be told that the plant is operating at 100% power, that the crew has been notified that RMS-R47, R-47 AIR EJECTOR NOBLE GAS MONITOR, has alarmed in the TSC, provided with current readings, and told that the crew is evaluating the need to enter AP-SG.1, Steam Generator Tube Leak. The operator will be directed to determine if entry into AP-SG.1 is appropriate by (1) determining if the R-47 alarm setpoint is set properly and (2) comparing the local reading to Curve #06-004 in accordance with PPCS-R47AR. The operator will be expected to determine that the R-47 alarm setpoint is set lower than identified by P-9, and that the local R-47 reading is too low to support entry into AP.SG.1.

NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (REV 082014)

Facility: Ginna Da1te of Examination: 12/2014 Exam Level: RO Operating Test No.: 2014-301 Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)

Type Code* Safety System I JPM Title Function A. 005 Residual Heat Removal System [005 A2.03 (2.9/3.1)]

S,D,L 4P Lineup RCDT Pump for Core Cooling B. EPE E14 High Containment Pressure [EPE E14 EA1.1 (3.7'/3.7)]

S,N,A,EN 5 Verify Containment Isolation and Heat Removal C. 004 Chemical and Volume Control System [004 A4.08 (3.8/3.4)] 1' S,N,A 2 CVCS Leak Isolation D. EPE 038 Steam Generator Tube Rupture [EPE 038 EA 1.04 (4.3/4.1 )]

S,N,A 3 Depressurize the RCS During a SGTR E. 045 Main Turbine Generator System [045 A2.17 (2. 7*/2.9*)]

S,P,D,A 4S Synchronize Generator On-Line with Improper Load Pickup F. 001 Control Rod Drive System [001 A4.06 (2.9/3.2)]

S,N,A 1 Continuous Rod Motion During Re-alignment of Misaligned Control Rod G. 062 AC Electrical Distribution [062 A 1.03 (2.5/2.8)]

S,N,A 6 Transfer Instrument Bus B to Normal Supply H. 016 Non-Nuclear Instrumentation System [016 A4.01 (2.9/2.8)]

S,D 7 Defeat Failed RCS Temperature Channel In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

I. APE 068 Control Room Evacuation [APE 068 AA 1.01 (4.3/4.5)]

P,D,E 8 Locally Operate the ARVs J. EPE 055 Station Blackout [EPE 055 EA2.03 (3.9/4.7)]

D,R,E 6 Isolate RCP Seals K. APE 062 Loss of Nuclear Service Water [APE 062 AA 1.07 (2.9/3.0)]

D,R,E 4P Fire Water Cooling to CCW Heat Exchanger A NUREG-1021, Revision 9 NRC EXAM MAT'L

ES-301 Control Room/In-Plant System:s Outline Form ES-301-2 (REV_082014)

@ 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 (A)lternate path 4-6 (6)

(C)ontrol room (D)irect from bank ~ 9 (6)

(E)mergency or abnormal in-plant ~ 1 (3)

(EN)gineered Safety Feature -

(L)ow-Power I Shutdown ~ 1 (1)

(N)ew or (M)odified from bank including 1(A) ~ 2 (5)

(P)revious 2 exams ~ 3 (2) (Randomly Selected)

(R)CA ~ 1 (2)

(S)imulator JPM Summary JPM A This is a Bank JPM. The operator will be told that the plant was performing a normal cooldown when both RHR Pumps tripped, that the S/G's are unavailable, that AP-RHR.1, Loss of RHR, was entered and attempts to restart the pumps were unsuccessful, and that the crew is in progress of establishing Containment Integrity and all personnel are clear of Containment. The operator will be directed to lineup the RCDT pumps to provide core cooling per ER-RHR.1, Section 6.2, Closed Loop RCS to RHR Cooling.

The operator will be expected to line up the RCDT pumps to provide core cooling per Section 6.2 of ER-RHR.1.

JPM B This is a new JPM. The operator will be told that th1e plant tripped from 100% power and safety injection has actuated, that the crew entered E-0, Reactor Trip or Safety Injection, and then transitioned to E-2, Faulted Steam Generator Isolation, and that due to a degrading transient, an Orange Path now exists on the Containment Critical Safety Function Status Tree. The operator will be directed to verify Containment Isolation and Heat Removal systems are operating as expected by performing FR-Z.1, Response to High Containment Pressure, starting from Step 1. During the course of this action, the operator will recognize that two Containment Isolation Valves have failed to automatically close, and that the Containment Spray ESFAS Signal has failed to function (Alternate Path). The operator will be expected to verify Containment Isolation and Heat Removal systems are operating in accordance with Steps 1-3 of FR-Z.1. When the operator discovers that two Containment Isolation Valves have failed to close as expected the operator will close or direct that alternative valves be closed in accordance with ATT-3.0, Attachment Cl/CVI. When the operator discovers that the Containment Spray ESFAS Signal has failed to function, the operator will take action to initiate Containment Spray. The operator will also start the D CNMT Recirc Fan.

JPM C This is a new JPM. The operator will be told that 1the Plant is at 100% power, that the crew has entered AP-CVCS.1, "CVCS Leak," completing steps 1-5, that the Pressurizer level is stable, and that the leak is believed to be in the Containment. The operator will be directed to take appropriate actions to isolate the leak starting with Step 6 of AP-CVCS.1, CVCS Leak, and control RCS inventory Ito maintain Pressurizer level stable.

During the course of this action, the operator will recognize that both operating Charging NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (REV 082014)

Pumps have tripped (Alternate Path). The operator will be expected to isolate Normal Letdown and Charging, recognize that both operating Charging Pumps have tripped, use AP-CVCS.3 to restore Charging flow, and then placie Excess Letdown in service.

JPM D This is a new JPM. The operator will be told that the plant has experienced a Steam Generator Tube Rupture, that the crew has completed E-3, Steam Generator Tube Rupture, through Step 17, however an attempt to re-establish Instrument Air to the Containment has been unsuccessful, that the RCS has been cooled to target temperature, and that the crew is ready to commence RCS depressurization. The operator will be directed to depressurize the RCS to minimize break flow starting with Step 18 of E-3. During the course of this action, the operator will recognize that ATT-12.0 must be used to open the Pressurizer PORV and that when the operator attempts to close the PORV following RCS depressurization, the valve fails OPEN (Alternate Path). The operator will be expected to determine that Pressurizer Spray and the normal use of the Pressurizer PORVs is not possible, and use ATT-12.0 to open one Pressurizer PORV; and then close the PORV Bloc~: Valve when it is recognized that the previously opened Pressurizer PORV has failed OPEN.

JPM E This is a Bank JPM randomly selected from the last two NRC JPM Exams. The operator will be told that the generator is being started following a refueling outage, that the generator is at 1800 rpm and the turbine is fully warmed up, and that the generator output voltage is 19 KV. The operator will be directed to synchronize the generator on-line per 0-1.2, Plant Startup From Hot Shutdown to Full Load, steps 6.1:3.1 through 6.13.10. During the course of this action, the operator will recognize that Automatic Load Pickup feature of the Turbine Control System has failed (Alternate Path). The operator will be expected to synchronize the Main Generator to the Electrical Grid, and when it is recognized that the Automatic Load Pickup has failed to function, the operator will manually load the Turbine to 40 to 60 MW, without reverse powering the Main Generator.

JPM F This is a new JPM. The operator will be told that while performing control rod testing, Shutdown Bank Rod 111 failed to move inward with the rest of the Shutdown Bank rods, that l&C has successfully repaired the problem, and that ER-RCC.2, Restoring a Misaligned RCC, has been entered to restore the misaligned rod. The operator will be directed to continue rod restoration per Section 6.2 of ER-RCC.2. During the course of this action, the operator will recognize that a continuous rod insertion event is on-going (Alternate Path). The operator will be expected to move the misaligned rod to a position in the core lower than its associated bank, and then realign the Shutdown Bank with the misaligned rod. While aligning the Shutdown Bank with the misaligned rod, the operator will diagnose a continuous rod insertion accident and manually trip the reactor.

The operator will perform the Immediate Actions of E-0 and determine that Safety Injection is NOT required.

JPM G This is a new JPM. The operator will be told that the B Instrument Bus is on Maintenance Power supply, that maintenance has been completed and the Hold Clearance has been removed, that It is desired to restore the B Instrument Bus to its normal power supply, and that there are no failed or defeated protection channels. The operator will be directed to transfer the "B" lnstrumEmt Bus from the Maintenance power supply to the normal power supply per Section 6.3 of ER-INST.3, Instrument Bus Power Restoration. During the course of this action, the operator will recognize that Pressurizer Pressure Controller has failed to control in AUTO (Alternate Path). The operator will be NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (REV 082014) expected to restore Instrument Bus B to its normal power supply in accordance with ER-INST.3, and take manual control of the Pressurizer Pressure Controller (431 K) to stabilize Pressurizer pressure, when it is recogni:z:ed that the controller has failed in AUTO.

JPM H This is a Bank JPM. The operator will be told that the plant was operating at 100%

power when Tl-402 failed high, and that the crew took all the appropriate actions and the plant is stable at current plant conditions. The operator will be directed to defeat the affected RCS temperature channel per Attachment 2 of ER-INST.1, Reactor Protection Bistable Defeat After Instrumentation Loop Failure. The operator will be expected to defeat the failed channel in accordance with ER-INST.1.

JPM I This is a Bank JPM randomly selected from the last two NRC JPM Exams. The operator will be told that the plant has experienced an uncontrollable Control Room Complex fire, that the crew performed the required actions of AP-CR.1, Control Room Inaccessibility, and transitioned to ER-FIRE.1, Alternate Shutdown For Control Complex Fire, that they are the Head Control Operator (HCO) and are equipped with a radio, that the Shift Manager is stabilizing the plant in MODE 3, and that all the required Appendix R equipment has been retrieved from the Appendix R locker outside the Control Room.

The operator will be directed to locally open the A Atmospheric Relief Valve, V-3411, three turns per P-15.2, Dump Steam Through ARVs Locally. The operator will be expected to locally open the A Atmospheric Relief Valve, V-3411, three turns per P-15.2, and then isolate this valve when the handwheel will not rotate in the close direction.

JPM J This is a Bank JPM. The operator will be told that the plant was operating at 100%

power when it suffered a loss of all AC power, and that the crew entered ECA-0.0, Loss of All AC Power, and is at the step that directs local isolation of RCP Seals. The operator will be directed to isolate RCP seals in accordance with the applicable portions of ATT-21.0, Attachment RCS Isolation. The opera1tor will be expected to locally isolate the RCP Seals per ATT-21.0.

JPM K This is a Bank JPM. The operator will be told that the plant was operating a power when a loss of all Service Water pumps occurred, that the crew responded using the emergency operating procedures and Attachment 2.4, Attachment No Service Water Pumps, as required, that the crew is ready to align alternate cooling to one CCW heat exchanger, that a normal SW discharge alignment is desired, and that SW isolations MOV-4616 and MOV-4735 have been closed from the Control Room. The operator will be directed to line up Fire Water Cooling from the plant fire header to the A CCW Heat Exchanger per ER-CCW.1, Fire Water Cooling to CCW and A SFP Heat Exchangers.

The operator will be expected to locally align fire water to CCW HX A in accordance with ER-CCW.1.

NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 (Rev 090314)

Facility: Ginna Date of Examination: 12/2014 Examination Level: SRO Operatin!~ Test Number: 2014-301 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 9raphs, curves, tables, etc.

D,P, R JPM: Determine Time to Boil for a Loss of Shutdown Cooling 2.1.8 (4.1) Ability to coordinate personnel activities Conduct of Operations outside the control room.

M,R JPM: Cold Weather Walkdown 2.2.12(4.1) Knowledge of Surveillance Procedures.

Equipment Control N,R JPM: Perform RCS Leakrate Surveillance 2.3.15(3.1) Knowledge of radiation monitoring systems, Radiation Control such as fixed radiation monitors and alarms, portable survey instruments, M,R personnel monitoring equipment, etc.

JPM: Evaluate the Impact of R-14 Failure on a Planned Gaseous Decay Tank Release 2.4.41 (4.6) Knowledge of emergency action level Emergency thre!sholds and classifications.

Procedures/Plan M,S JPM: Classify an Emergency Event NOTE: All items (5 total) are required for SROs. RO 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, (1) or Class(R)oom (4)

(D)irect from bank(~ 3 for ROs; ~ 4 for SROs & RO retakes) (1)

(N)ew or (M)odified from bank(~ 1) (4)

(P)revious 2 exams (~ 1; randomly se,lected) (1)

NRC EXAl\1 N[AT'L NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 (Rev 090314)

SRO Admin JPM Summ~

A 1a This is a Bank JPM randomly selected from the last two NRC Exams. The operator will be provided with two sets of shutdown conditions; one current, and one projected to exist within four days. The operator will be directed to determine the Time to Boil given a Loss of RHR for each of the two sets of conditions. The operator will be expected to use IP-OUT-2, Outage Risk Management, and determine that the Time to Boil for the present plant conditions is 2.46 hours5.324074e-4 days 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months , and for the projected conditions four days from now is .43 hours4.976852e-4 days 0.0119 hours 7.109788e-5 weeks 1.63615e-5 months (25.96 +0.1/-0.0 minutes).

A 1b This is a modified Bank JPM. The operator will be told that the weekly Cold Weather Walkdown is in progress, and the Primary, Secondary and Outside Auxiliary Operators are about ready to start their Walkdown Tours; and provided with a recently completed Attachment 1, Cold Weather Temperature Log, of 0-22, Cold Weather Walkdown Procedure. They will also be told that a severe lake-effect storm is occurring with 24 inches of snow having accumulated over the last four hours. The operator will be directed to review Attachment 1 and perform section 6.1, General Instructions, of 0-22, and prioritize any areas of concern for the Auxiliary Operator Walkdown Tours; and then to document discrepancies, if any, and any required action(s). The operator will be expected to review the most recent plant area temperature data and then use 0-22 to determine that two areas of the plant are too low in temperature. Then the operator will prioritize the actions of the Auxiliary Operators to tour these areas first, identify discrepancies and actions required in accordance with the attached KEY.

A2 This is a new JPM. The operator will be told that the plant is at 100% power with 40 gpm Letdown flow, that the VCT Level point on the PPCS is OOS, that at 1020 the VCT level was 25% when the HCO started the Day Shift RCS Leakage Monitoring Surveillance, intending to monitor for about 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and that at 1125 a VCT Level instrument failure occurred which l&C corrected and restored to operation within 10 minutes. The operator will be provided with a current Attachment 3, RCS Leakage Surveillance Record, with recent Auto Makeup information, and directed to record the data for the recent Auto Makeup on Attachment 3 AND to determine the current RCS Leakrate. The operator will also be directed to identify any Tech Spec LCOs that have been exceeded, and if any LCOs are exceeded, identify any required Tech Spec ACTION. The operator will be expected to complete Attachment 3 in accordance with the KEY provided, and identify the LCO 3.4.13, RCS Operational Leakage, is NOT met, and that the required ACTION is to reduce leakage to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (A.1).

A3 This is a modified Bank JPM. The operator will be told that the plant is at 100% power, that an* A Gas Decay Tank release is being planned, that the HCO reports that R-14A, Plant Vent Gas, is de-energized for preventative maintenance that will require 2 more hours, and must be completed before it can be returned to service, and that R-14, Plant Vent Gas, has just failed LOW and l&C has indicat,ed that it will take at least four hours to troubleshoot and correct the failure. The operator will be directed to determine the required actions based on the failure of R-14, and to determine the earliest time at which the A Gas Decay Tank release can start, and to identify any actions that will be required to start the release. The operator will be expected to determine that ('I) on-going releases (i.e. ventilation) may continue provided grab samples are taken and analyzed for isotopic activity at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , that (2) at least one of either R-114 or R-14A must be restored to OPERABLE status within 30 days or it must be explained in the next NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 (Rev 090314)

Annual Radioactive Effluent Release Report, and that (3) the earliest time that the A Gas Decay Tank release can be initiated is 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from now when the maintenance on R-14A is completed, and ONLY if at least two independent samples of the tank's contents are analyzed, and at least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.

A4 This is a modified Bank JPM. This JPM will be given immediately following the associated Simulator Scenario. Consequently, the~re are five separate versions of this JPM, one each associated with the submitted Simulator Scenarios. A portion of this JPM is Time Critical. After each Simulator Scenario, the operator will be told that the sequence of events that they have just performed on the recently completed Simulator Scenario must be evaluated for any Emergency Classifications. The operator will be directed to determine the Emergency Classification that would have been made first in the recently completed scenario, and Identify this basis for this classification. The operator will be given a time that all the indications needed to classify the first event requiring an Emergency Classification were apparent in the Control Room and then directed to identify the latest time at which the Emergency Classification must have been made. Finally, the operator will be directed to identify any subsequent emergency classifications that would have been made in the Scenario, and the basis for this classification. The operator will be expected to make the appropriate classifications and identify the time frames as appropriate.

NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (REV 082014)

Facility: Ginna Date of Examination: 1:2/2014 Exam Level (circle one): SRO(I) Operating Test No.: 2014-301 Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)

Type Code* Safety System I JPM Title Function A. 005 Residual Heat Removal System [005 A2.03 (2.9/3.1 )]

S,D,L 4P Lineup RCDT Pump for Core Cooling B. EPE E14 High Containment Pressure [EPE E14 EA1.1 (3.7'/3.7)]

S,N,A,EN 5 Verify Containment Isolation and Heat Removal C. 004 Chemical and Volume Control System [004 A4.08 (3.8/3.4))

S,N,A 2 CVCS Leak Isolation D. EPE 038 Steam Generator Tube Rupture [EPE 038 EA1 .04 (4.3/4.1))

S,N,A 3 Depressurize the RCS During a SGTR E. 045 Main Turbine Generator System [045 A2.17 (2. 7*/2.9*)]

S,P,D,A 4S Synchronize Generator On-Line with Improper Load Pickup F. 001 Control Rod Drive System [001 A4.06 (2.9/3.2))

S,N,A 1 Continuous Rod Motion During Re-alignment of Misaligned Control Rod G. 062 AC Electrical Distribution [062 A 1.03 (2.5/2.8)]

S,N,A 6 Transfer Instrument Bus B to Normal Supply In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

I. APE 068 Control Room Evacuation [APE 068 AA 1.01 (4.3/4.5))

P,D,E 8 Locally Operate the ARVs J. EPE 055 Station Blackout [EPE 055 EA2.03 (3.9/4.7)]

D,R,E 6 Isolate RCP Seals K. APE 062 Loss of Nuclear Service Water [APE 062 AA1 .07 (2.9/3.0))

D,R,E 4P Fire Water Cooling to CCW Heat Exchanger A NRC EXAM Mi\T'L NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (REV 082014)

@ 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 SRO(I)

(A)lternate path 4-6 (6)

(C)ontrol room (D)irect from bank ~ 8 (5)

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

(EN)gineered Safety Feature -

(L)ow-Power I Shutdown 2 1 (1)

(N)ew or (M)odified from bank including 1(A) 2 2 (5)

(P)revious 2 exams ~ 3 (2) (Randomly Selected)

(R)CA 2 1 (2)

(S)imulator JPM Summary JPM A This is a Bank JPM. The operator will be told that the plant was performing a normal cooldown when both RHR Pumps tripped, that the S/G's are unavailable, that AP-RHR.1, Loss of RHR, was entered and attempts to restart the pumps were unsuccessful, and that the crew is in progress of establishing Containment Integrity and all personnel are clear of Containment. The operator will be directed to lineup the RCDT pumps to provide core cooling per ER-RHR.1, Section 6.2, Closed Loop RCS to RHR Cooling.

The operator will be expected to line up the RCDT pumps to provide core cooling per Section 6.2 of ER-RHR.1.

JPM B This is a new JPM. The operator will be told that the plant tripped from 100% power and safety injection has actuated, that the crew entered E-0, Reactor Trip or Safety Injection, and then transitioned to E-2, Faulted Steam Generator Isolation, and that due to a degrading transient, an Orange Path now exists on the Containment Critical Safety Function Status Tree. The operator will be directed to verify Containment Isolation and Heat Removal systems are operating as expected by performing FR-Z.1, Response to High Containment Pressure, starting from Step 1. During the course of this action, the operator will recognize that two Containment Isolation Valves have failed to automatically close, and that the Containment Spray ESFAS Signal has failed to function (Alternate Path). The operator will be expected to verify Containment Isolation and Heat Removal systems are operating in accordance with Steps 1-3 of FR-Z.1. When the operator discovers that two Containment lsolatie>n Valves have failed to close as expected the operator will close or direct that alternative valves be closed in accordance with ATT-3.0, Attachment Cl/CVI. When the operator discovers that the Containment Spray ESFAS Signal has failed to function, the operator will take action to initiate Containment Spray. The operator will also start the D CNMT Recirc Fan.

JPM C This is a new JPM. The operator will be told that the Plant is at 100% power, that the crew has entered AP-CVCS.1, "CVCS Leak," completing steps 1-5, that the Pressurizer level is stable, and that the leak is believed to be in the Containment. The operator will be directed to take appropriate actions to isolate the leak starting with Step 6 of AP-CVCS.1, CVCS Leak, and control RCS inventory to maintain Pressurizer level stable.

During the course of this action, the operator will recognize that both operating Charging NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (REV 082014)

Pumps have tripped (Alternate Path). The operator will be expected to isolate Normal Letdown and Charging, recognize that both operating Charging Pumps have tripped, use AP-CVCS.3 to restore Charging flow, and then placie Excess Letdown in service.

JPM D This is a new JPM. The operator will be told that the plant has experienced a Steam Generator Tube Rupture, that the crew has completed E-3, Steam Generator Tube Rupture, through Step 17, however an attempt to re-establish Instrument Air to the Containment has been unsuccessful, that the RCS has been cooled to target temperature, and that the crew is ready to commence RCS depressurization. The operator will be directed to depressurize the RCS to minimize break flow starting with Step 18 of E-3. During the course of this action, the operator will recognize that ATT-12.0 must be used to open the Pressurizer PORV and that when the operator attempts to close the PORV following RCS depressurization, the valve fails OPEN (Alternate Path). The operator will be expected to determine that Pressurizer Spray and the normal use of the Pressurizer PORVs is not possible, and use ATT-12.0 to open one Pressurizer PORV; and then close the PORV Bloc~~ Valve when it is recognized that the previously opened Pressurizer PORV has failed OPEN.

JPM E This is a Bank JPM randomly selected from the last two NRC JPM Exams. The operator will be told that the generator is being started following a refueling outage, that the generator is at 1800 rpm and the turbine is fully warmed up, and that the generator output voltage is 19 KV. The operator will be directed to synchronize the generator on-line per 0-1.2, Plant Startup From Hot Shutdown to Full Load, steps 6.13.1 through 6.13.10. During the course of this action, the operator will recognize that Automatic Load Pickup feature of the Turbine Control System has failed (Alternate Path). The operator will be expected to synchronize the Main Generator to the Electrical Grid, and when it is recognized that the Automatic Load Pickup has failed to function, the operator will manually load the Turbine to 40 to 60 MW, without reverse powering the Main Generator.

JPM F This is a new JPM. The operator will be told that while performing control rod testing, Shutdown Bank Rod 111 failed to move inward with the rest of the Shutdown Bank rods, that l&C has successfully repaired the problem, and that ER-RCC.2, Restoring a Misaligned RCC, has been entered to restore the misaligned rod. The operator will be directed to continue rod restoration per Section 6.2 of ER-RCC.2. During the course of this action, the operator will recognize that a continuous rod insertion event is on-going (Alternate Path). The operator will be expected to move the misaligned rod to a position in the core lower than its associated bank, and then realign the Shutdown Bank with the misaligned rod. While aligning the Shutdown Bank with the misaligned rod, the operator will diagnose a continuous rod insertion accident and manually trip the reactor.

The operator will perform the Immediate Actions of E-0 and determine that Safety Injection is NOT required.

JPM G This is a new JPM. The operator will be told that the B Instrument Bus is on Maintenance Power supply, that maintenance has been completed and the Hold Clearance has been removed, that It is desired to restore the B Instrument Bus to its normal power supply, and that there are no failed or defeated protection channels. The operator will be directed to transfer the "B" lnstrumEmt Bus from the Maintenance power supply to the normal power supply per Section 6.3 of ER-INST.3, Instrument Bus Power Restoration. During the course of this action, the operator will recognize that Pressurizer Pressure Controller has failed to control in AUTO (Alternate Path). The operator will be NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (REV 082014) expected to restore Instrument Bus B to its normal power supply in accordance with ER-INST.3, and take manual control of the Pressurizer Pressure Controller (431 K) to stabilize Pressurizer pressure, when it is recognized that the controller has failed in AUTO.

JPM I This is a Bank JPM randomly selected from the last two NRC JPM Exams. The operator will be told that the plant has experienced an uncontrollable Control Room Complex fire, that the crew performed the required actions of AP-CR.1, Control Room Inaccessibility, and transitioned to ER-FIRE.1, Alternate Shutdown For Control Complex Fire, that they are the Head Control Operator (HCO) and are equipped with a radio, that the Shift Manager is stabilizing the plant in MODE 3, and that all the required Appendix R equipment has been retrieved from the Appendix R locker outside the Control Room.

The operator will be directed to locally open the A Atmospheric Relief Valve, V-3411, three turns per P-15.2, Dump Steam Through AIRVs Locally. The operator will be expected to locally open the A Atmospheric Relief Valve, V-3411, three turns per P-15.2, and then isolate this valve when the handwheel will not rotate in the close direction.

JPM J This is a Bank JPM. The operator will be told that the plant was operating at 100%

power when it suffered a loss of all AC power, and that the crew entered EC.A-0.0, Loss of All AC Power, and is at the step that directs local isolation of RCP Seals. The operator will be directed to isolate RCP seals in acc:ordance with the applicable portions of ATT-21.0, Attachment RCS Isolation. The operator will be expected to locally isolate the RCP Seals per ATT-21.0.

JPM K This is a Bank JPM. The operator will be told that the plant was operating a power when a loss of all Service Water pumps occurred, that the crew responded using the emergency operating procedures and Attachment 2.4, Attachment No Service Water Pumps, as required, that the crew is ready to align alternate cooling to one CCW heat exchanger, that a normal SW discharge alignment is desired, and that SW isolations MOV-4616 and MOV-4735 have been closed from the Control Room. The operator will be directed to line up Fire Water Cooling from the plant fire header to the A CCW Heat Exchanger per ER-CCW.1, Fire Water Cooling to CCW and A SFP Heat Exchangers.

The operator will be expected to locally align fire water to CCW HX A in accordance with ER-CCW.1.

NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2

.' DRAFT {REV 082014)

Facility: Ginna Date of Examination: 12/2014 Exam Level (circle one): SRO(U) Operating Test No.: 2014-301 Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)

Type Code* Safety System I JPM Title Function A. 005 Residual Heat Removal System [005 A2.03 (2.9/3.1)]

S,D,L 4P Lineup RCDT Pump for Core Cooling B. EPE E14 High Containment Pressure [EPE E14 EA1.1 (3.7/3.7)]

S,N,A,EN 5 Verify Containment Isolation and Heat Removal C. 004 Chemical and Volume Control System [004 A4.08 (3.8/3.4)]

S,N,A 2 CVCS Leak Isolation In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

I. APE 068 Control Room Evacuation [APE 068 AA 1.01 (4.3/4.5)]

P,D,E 8 Locally Operate the ARVs J. EPE 055 Station Blackout [EPE 055 EA2.03 (3.9/4.7)]

D,R,E 6 Isolate RCP Seals NRC EXA~1 MAT'L NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 DRAFT (REV 082014)

@ 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 SRO(U)

(A)lternate path 2-3 (2)

(C)ontrol room (D)irect from bank s 4 (3)

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

(EN)gineered Safety Feature 2 1 (1) (Control Room System)

(L)ow-Power I Shutdown 2 1 (1)

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

(P)revious 2 exams S 2 ( 1) (Randomly Selected)

(R)CA 2 1 (1)

(S)imulator JPM Summary JPM A This is a Bank JPM. The operator will be told that the plant was performing a normal cooldown when both RHR Pumps tripped, that the S/G's are unavailable, that AP-RHR.1, Loss of RHR, was entered and attempts to restart the pumps were unsuccessful, and that the crew is in progress of establishing Containment Integrity and all personnel are clear of Containment. The operator will be directed to lineup the RCDT pumps to provide core cooling per ER-RHR.1, Section 6.2, Closed Loop RCS to RHR Cooling.

The operator will be expected to line up the RCDT pumps to provide core cooling per Section 6.2 of ER-RHR.1.

JPM B This is a new JPM. The operator will be told that the plant tripped from 100% power and safety injection has actuated, that the crew entered E-0, Reactor Trip or Safety Injection, and then transitioned to E-2, Faulted Steam Generator Isolation, and that due to a degrading transient, an Orange Path now exists on the Containment Critical Safety Function Status Tree. The operator will be directed to verify Containment Isolation and Heat Removal systems are operating as expected by performing FR-Z.1, Response to High Containment Pressure, starting from Step 1. During the course of this action, the operator will recognize that two Containment Isolation Valves have failed to automatically close, and that the Containment Spray ESFAS Signal has failed to function (Alternate Path). The operator will be expected to verify Containment Isolation and Heat Removal systems are operating in accordance, with Steps 1-3 of FR-Z.1. When the operator discovers that two Containment Isolation Valves have failed to close as expected the operator will close or direct that alternative valves be closed in accordance with ATT-3.0, Attachment Cl/CVI. When the operator discovers that the Containment Spray ESFAS Signal has failed to function, the operator will take action to initiate Containment Spray. The operator will also start the D CNMT Recirc Fan.

JPM C This is a new JPM. The operator will be told that the Plant is at 100% power, that the crew has entered AP-CVCS.1, "CVCS Leak," completing steps 1-5, that the Pressurizer level is stable, and that the leak is believed to be in the Containment. The operator will be directed to take appropriate actions to isolate lthe leak starting with Step 6 of AP-CVCS.1, CVCS Leak, and control RCS inventory to maintain Pressurizer level stable.

During the course of this action, the operator will recognize that both operating Charging NUREG-1021, Revision 9

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 DRAFT (REV 082014)

Pumps have tripped (Alternate Path). The operator will be expected to isolate Normal Letdown and Charging, recognize that both operating Charging Pumps have tripped, use AP-CVCS.3 to restore Charging flow, and then plac,e Excess Letdown in service.

JPM I This is a Bank JPM randomly selected from the last two NRC JPM Exams. The operator will be told that the plant has experienced an uncontrollable Control Room Complex fire, that the crew performed the required actions of AP-CR.1, Control Room Inaccessibility, and transitioned to ER-FIRE.1, Alternate Shutdown For Control Complex Fire, that they are the Head Control Operator (HCO) and are equipped with a radio, that the Shift Manager is stabilizing the plant in MODE 3, and that all the required Appendix R equipment has been retrieved from the Appendix R locker outside the Control Room.

The operator will be directed to locally open the A Atmospheric Relief Valve, V-3411, three turns per P-15.2, Dump Steam Through ARVs Locally. The operator will be expected to locally open the A Atmospheric Relief Valve, V-3411, three turns per P-15.2, and then isolate this valve when the handwheel will not rotate in the close direction.

JPM J This is a Bank JPM. The operator will be told that the plant was operating at 100%

power when it suffered a loss of all AC power, and that the crew entered ECA-0.0, Loss of All AC Power, and is at the step that directs local isolation of RCP Seals. The operator will be directed to isolate RCP seals in accordance with the applicable portions of ATT-21.0, Attachment RCS Isolation. The operator will be expected to locally isolate the RCP Seals per ATT-21.0.

NUREG-1021, Revision 9

Appendix D Scenario Outline Form ES-D-1 Draft (110414)

Facility: Ginna Scenario No.: 2 Op Test No.: N2014-301 Examiners: Operators: (SRO)

(RO)

(BOP)

Initial Conditions: The plant is at 100% power (EOL), and has been at full power for 145 days. The area has experienced thunderstorms over the last 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and this is expected to continue for the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Turnover: The following equipment is Out-Of-Ser1ice: The B CNMT Spray Pump has been declared INOPERABLE due to excessive inboard vertical vibration during STP-0 COMP-B sur1eillance test yesterday. Engineering is evaluating the data. A-52.4 submitted for ITS 3.6.6, 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Action.

Event Malf. Event Type* Event No. No. Description 1 CLG05 C-RO Leak on the CCW System/B CCW Pump Trips CLG02A C(TS)-SRO 2 PZR03B I-BOP PZR Level Channel 427 fails LOW I-RO l(TS}-SRO 3 TUR11B R-RO Turbine Control Valve CV-L4 Drifts Closed N-BOP N-SRO 4 FDW07C C-BOP B FRV fails AS-IS (Manual Control Available)

C(TS)-SRO 5 STM03 M-RO MSLB Downstream of MSIVs with MSIVs Stuck OPEN STM05A M-BOP STM05B M-SRO 6 RPS05A C-RO RPS fails to Trip Rx in AUTO/Manual Pushbutton RPS05B C-SRO 7 STM03 NA Steam Break Degrades Creating PTS Concern 8 EDS01A C-BOP Loss of Offsite Power EDS01B C-SRO

(N)ormal, ( R)eactivity, (I )nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 Draft (110414)

Ginna 2014 NRC Scenario #2 The plant is at 100% power (EOL), and has been at full power for 145 days. The area has experienced thunderstorms over the last 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and this is expected to continue for the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The following equipment is Out-Of-Service: The B CNMT Spray Pump has been declared INOPERABLE due to excessive inboard vertical vibration during STP-0 COMP-B surveillance test yesterday. Engineering is evaluating the data. A-52.4 submitted for ITS 3.6.6, 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Action.

Shortly after taking the watch, a CCW System Supply Relief Valve will lift and fail to reseat causing a 30 gpm CCW System leak. Approximately two minutes afterwards the B CCW Pump will trip, and the B CCW Pump will automatically start. The operator will respond in accordance with AR-A-17, MOTOR OFF RCP CCWP, and enter AP-CCW.2, Loss of CCW During Power Operation. The operator will address Te:lchnical Specification 3.7.7, Component Cooling Water System.

After this, Pressurizer Level Channel 427 will fail LOW, resulting in letdown isolation and de-energizing the pressurizer heaters. The crew will n::lspond per AR-F-11, PZR LOW LEVEL 13%, and ER-INST.1, Reactor Protection Bista1ble Defeat After lnstrumi::lntation Loop Failure. They will defeat the failed channel, reset PZR heaters, reduce charging to a single charging pump, and re-establish letdown per S-3.2.E, Placing In or Removing From Service Normal Letdown/Excess Letdown. The crew will start a second charging pump and slowly restore PZR level to program (52%). The operator will address Technical Specification 3.3.1, Reactor Trip System {RTS) Instrumentation and 3.4.9, Pressurizer.

Then, turbine control valve CVL-4 will drift closed. The crew will respond per AP-TURB.2, Turbine Load Rejection, begin a load reduction to less than 50% powE::lr using AP-TURB.5, Rapid Load Reduction.

During the load reduction, a failure of the B FRV to control in AUTO. The operator will respond per AR-G-5, SIG/ B LEVEL DEVIATION +/-7%, or upon observing an abnormally high level in the B Steam Generator and control the B FRV manually. The operator will address Technical Specification 3.7.3, Main Feedwater Isolation Valves (MFIVs), Main Feedwater Regulating Valves (MFRVs ), and Associated Bypass Valves.

Subsequently, a major steamline break will occur in the! Turbine Building downstream of the MSIVs. Both MSIVs will fail to close automatically in response to manual attempts from the MCB. The reactor will fail to trip automatically and manual pushbutton will not work, requiring opening of normal supply breakers for Bus 13 and Bus 15 to de-energize rod drive MG sets to trip the reactor. The crew will enter E-0, Reactor Trip or Safety Injection.

The crew will ultimately transition to E-2, Faulted SIG Isolation. Since both MSIVs are failed open, and with a downstream break, both SGs will be affected and their pressures will lower concurrently. Eventually, the crew will transition to ECA-2.1, Uncontrolled Depressurization of Both Steam Generators. After the operator throttles AFW flow in ECA-2.1, the leak will increase in magnitude, resulting in an Orange Path on RCS Appendix D Scenario Outline Form ES-D-1 Draft (110414)

Integrity. Simultaneously the event will be complicated by a Loss of Offsite Powe,r. Both Emergency Diesel Generators will start and automatically repower the respective Emergency Susses, and the operator will need to re-establish AFW flow of 50 gpm to each Steam Generator.

The crew will implement FR-P.1, Response to Imminent Pressurized Thermal! Shock Condition, when RCS cold leg temperatures reach 311°F. In FR-P.1 the crew will stop the SI pumps and the RHR pumps.

The scenario will be terminated after the SI and RHR pumps are secured.

Critical Tasks:

1. Manually trip the reactor from the control room before transition to FR-S.1 Safety Significance: Failure to manually trip the reactor causes a challenge to the subcriticality CSF beyond that irreparably introduced by the postulated conditions.

Additionally, it constitutes an "incorrect performance that necessitates the crew taking action which complicates the event mitigation strategy that demonstrates the inability by the crew to recognize a failure of the automatic actuation of the RPS.

2. Control the AFW flowrate to 50 gpm per SG iin order to minimize the RCS Cooldown rate Safety Significance: Failure to control the AFW flow rate to the SGs when able to do so causes a challenge to the subcriticality and RCS Integrity CSFs beyond that irreparably introduced by the postulated conditions. Additionailly, it constitutes an "incorrect performance that may necessitate the crew taking action which complicates the event mitigation strategy (Transition to FR-S.1 on inability to maintain a negative SUR).

Appendix D Scenario Outline Form ES-D-1 Draft (110414)

Facility: Ginna Scenario No.: 3 Op Test No.: N2014-301 Examiners: Operators: (SRO)

(RO)

(BOP)

Initial Conditions: The plant is at 48% power (MOL). The plant power was reduced several days ago due to a malfunction on the A MFW Pump. Corrective Maintenance has been completed, and the pump is running for retest. RG&E Energy Control Center has requested that the electric plant be aligned to a 0/100 configuration on circuit 7T to allow the RG&E personnel to perform an insulator inspection on thE~ 767 Line. The 767 line will remain OPERABLE throughout the inspection.

Turnover: The following equipment is Out-Of-Service: The B AFW Pump is OOS for Bearing Replacement. A-52.4 submitted for ITS 3.7.5, 7 day Action. MRPI for Rod G3 indicates on the bottom, however, this is a confirmed instrumentation problem (l&C is working).

Event Malf. Event Type* Event No. No. Description 1 NA N-BOP Shift Electric Plant N-SRO 2 EDS07B C-RO Loss of B Instrument Bus C-BOP C(TS)-SRO 3 PZR02D I-RO Pressurizer Pressure (PT-449) fails HIGH I-BOP l(TS)-SRO 4 TUR05E R-RO Main Turbine High Vibration/Downpower N-BOP N-SRO 5 CND03A C-BOP Hotwell Level Transmitter fails HIGH C-SRO 6 FDW09A M-RO Feed Line Rupture Inside Containment/Turbine fails to Auto Trip M-BOP M-SRO 7 SIS02A C-RO Safety Injection fails to actuate Automatically SIS02B C-SRO 8 RPS07K C-BOP A AFW Pump to Auto Start/Trips after Manual Start/TDAFW Pump REM trips on Overspeed/ Standby AFW fails to function C-SRO FDW32 OVR FDW42A FDW15B

(N)ormal, (R)eactivity, (I )nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 Draft (110414)

Ginna 2014 NRC Scenario #3 The plant is at 48% power (MOL). The plant power was reduced several days ago due to a malfunction on the A MFW Pump. Corrective Maintenance has been completed, and the pump is running for retest. RG&E Energy Control Center has requested that the electric plant be aligned to a 0/100 configuration on circuit 7T to allow the RG&E personnel to perform an insulator inspection on the 76"7 Line. The 767 line will remain OPERABLE throughout the inspection.

The following equipment is Out-Of-Service: The B AFW Pump is OOS for Bearing Replacement. A-52.4 submitted for ITS 3.7.5, 7 day Action. MRPI for Rod G3 indicates on the bottom, however, this is a confirmed instrumentation problem (l&C is working).

Shortly after taking the watch, the operator will shift the Electric Line-up from 50/50 to 0/100 in accordance with 0-6.9.2, Establishing and/or Transferring Offsite Power to Bus 12A/12B.

Shortly afterwards, a loss of the B Instrument Bus will occur. The operator will respond in accordance with AR-E-14, LOSS B INSTR BUS. Power will be restored to the bus per guidance in ER-INST.3, Instrument Bus Power Restoration, which will include the isolation and re-establishment of Normal Letdown in accordance with S-3.2E, Placing In or Removing From Service Normal Letdown/Excess Letdown. The operator will address AR-K-32, CV AIR DRYER LOW PRESS SG B/D TANK HIGH LEVEL, while restoring from the transient. The operator will address Technical Specification 3.B. 7, AC Instrument Bus Sources Modes 1-4, and 3.8.9, Distribution Systems - Modes 1, 2, 3 and 4.

After this, the controlling Pressurizer Pressure Transmitter will fail HIGH, causing the Spray Valves to open. The operator will respond in accordance with AR-F-2, PRESSURIZER HIGH PRESS 2310 PSI and AR-F-110, PRESSURIZER LO PRESS 2205 PSI, and enter AP-PZR.1, Abnormal PZR Pressure. AP-PZR.1 will re~fer the operator to ER-INST.1, Reactor Protection Bistable DE~feat After Instrumentation Loop Failure, for the defeat of PT-449. The operator will address Technical Specification 3.4.1, RCS Pressure, Temperature, and Flow Departure From Nucleate Boiling (DNB)

Limits; 3.3.1, Reactor Trip System (RTS) Instrumentation; and TRM-3.4.3, Anticipated Transient Without Scram (ATWS) Mitigation.

Following this, a turbine high vibration condition on Bearing #5 will develop within about 3 minutes. The operator will respond in accordance with AR-1-27, ROTOR ECCENTRICITY OR VIBRATION; and enter AP-TURB.3, Turbine Vibration; and then AP-TURB.5, Rapid Downpower. The operator will need to lower the Turbine Load to stabilize the vibrations. During the load reduction Hotwell Level Transmitter will fail HIGH. The operator will need to manually control the Hotwell Level Controller.

Subsequently, a feed line rupture inside Containment will occur. The Reactor will trip however, the Turbine will fail to trip automatically and Safety Injection will fail to actuate automatically. The operator will enter E-0, Reactor Trip or Safety Injection. On the reactor trip the A AFW Pump will fail to Autostart, then trip after it is manually started; and the TDAFW Pump will trip on overspeed. The operator will transition from E-0 to FR-H.1, Response to a Loss of Secondary Heat Sink.

Appendix D Scenario Outline Form ES-D-1 Draft (110414)

The operator will unsuccessfully attempt to place the Standby AFW System in service, and then attempt to restore a Secondary Heat Sink using the MFW System. Once the Secondary Heat Sink is re-established using MFW, the operator will transition back to E-0, and then transition to E-2, Faulted Steam Generator Isolation.

The scenario will terminate at Step 9 of E-2, after the crew has determined that a transition to E-1, Loss of Reactor or Secondary Coolant, is required.

Critical Tasks:

1. Manually actuate Safety Injection before transition out of E-0 into ES-0.1 Safety Significance: Failure to actuate Safety Injection when it is required to be actuated, and can be actuated, violates the assumptions of the Safety Analysis and constitutes incorrect performance that could lead to misdiagnosis of the event, implementation of an incorrect mitigation strategy and ultimately degradation of the RCS and/or fuel cladding fission product barriers.

2. Establish feedwater flow into at least one Steam Generator before RCS Bleed and Feed is required.

Safety Significance: Failure to establish feedwater flow into at least one Steam Generator results in the crew having to rely upon the lower-priority action of having to initiate RCS Bleed and Feed to minimize the possibility of core uncovery. Failure to perform this task, when able to do so, constitutes incorrect performance that leads to degradation of the RCS and/or fuel cladding fission product barriers.

Appendix D Scenario Outline Form ES-D-1 Draft (110714)

Facility: Ginna Scenario No.: 4 Op Test No.: N2014-301 Examiners: Operators: (SRO)

(RO)

(BOP)

Initial Conditions: The plant is at 48% power (MOL). Power has been held at this level for approximately four days while corrective maintenance is performed on the A MFW Pump. Maintenance has just been completed. Chemistry has requested that Letdown flow be raised from 40 gpm to 60 gpm. Also, it is expected to swap EHC Pumps per P-17 for routine equipment rotation. The area has experienced severe weather over the last 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and this is expected to continue for the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Turnover: The following equipment is Out-Of-Service: The B AFW Pump is OOS for Bearing Replacement. Steam Flow channel FT-475 is OOS. The channel has been defeated per ER-INST.1.

Event Malf. Event Type* Event No. No. Description 1 NA N-RO Swap Letdown Orifice (40 gpm to 60 gpm)

N-SRO 2 NA N-BOP Alternate EHC Supply Pumps N-SRO 3 NIS06A C-RO Power Range N41 Upper Detector fails LOW/Letdown Pressure CVC07A I-BOP Controller Failure CVC23 l(TS)-SRO 4 PZROSB C-RO PORV Leak/Block Valve Failure C(TS)-SRO 5 TUR09D R-RO Downpower/Failure of Turbine Control/EHC C-BOP C-SRO 6 PZR07 M-RO Pzr Steam Space Break M-BOP M-SRO 7 RPS11- C-RO Seal Water Return Isolation CIV Fails in AUTO A3 C-SRO

(N)ormal, (R)eactivity, (I )nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 Draft (110714)

Ginna 2014 NRC Scenario #4 The plant is at 48% power (MOL). Power has been held at this level for approximately four days while corrective maintenance is performed on the A MFW Pump. Maintenance has just been completed. Chemistry has requested that Letdown flow be raised from 40 gpm to 60 gpm. Also, it is expected to swap El-IC Pumps per P-17 for routine equipment rotation. The area has experienced severe weather over the last E> hours, and this is expected to continue for the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The following equipment is Out-Of-Service: The B AFW Pump is OOS for Bearing Replacement. Steam Flow channel FT-475 is OOS. The channel has been defeated per ER-INST.1.

Shortly after taking the watch, the operator will swap Letdown Orifice Control valves in accordance with S-3.2P, Swapping CVCS Letdown Orifices.

After this, the operator will alternate the EHC Supply Pump alignment in accordance with P-17, Operations Control Room Operating Instructions, Attachment 12, CROl-12 Swapping Electro-Hydraulic Pumps.

Then, the N41 Power Ranger Upper Detector will fail Low. The operator will respond in accordance with AR-E-26, POWER RANGE CHANNEL DEV +/-2%, and/or AR-E-28, POWER RANGE ROD DROP ROD STOP -5%/5SEC, and then enter ER-NIS.3, PR Malfunction. The operator will address Technical Specification 3.3.1, Reactor Trip Instrumentation. While the CO is defeating the failed Power Range Channel, a failure of the Letdown Pressure Controller will cause PCV-135 to fail OPEN. The operator will need to take manual control of the failed controller,. and control Letdown pressure manually.

Subsequently, Pressurizer PORV-431 C will fail partially open. The operator will respond in accordance with AR-F-18, PRZR PORV OUTLET HI TEMP 145°F, and enter AP-PZR.1, Abnormal Pressurizer Pressure. When the operator attempts to isolate the PORV, the Block Valve will fail to shut fully resulting in a 2-5 gpm leak into the PRT. The crew may implement AP-RCS.1, Reactor Coolant Leak, and prepare to make a Containment entry. Ultimately, the crew will be directed to take the unit off-line. The operator will address Technical Specification 3.4.11, Pressurizer PORVs, 3.4.1 RCS Pressure, Temperature, and Flow Departure from NuclE:~ate Boiling (DNB) Limits, as well as TRM 3.4.3, Anticipated Transients Without Scram (ATWS) Mitigation. Although the crew will address Technical Specification 3.4.13, RCS Leakage, it is likely that at the current leakrate, this specification will be met.

The operator will take the unit off line in accordance with AP-TURB.5, Rapid Load Reduction During the downpower, the Main Turbine will fail in automatic control, and shift to manual control. The operator will identify that the load reduction has been stopped, and use manual control of the turbine to restart and continue the downpower.

The remainder of the downpower will need to be accomplished using manual control of the Turbine.

When the plant is at approximately 40% power, the piping leading to the failed PORV will fail so that a vapor space Small Break LOCA occurs. The plant will trip and Safety Appendix D Scenario Outline Form ES-D-1 Draft (110714)

Injection will be actuated, and the operator will entE~r E-0, Reactor Trip or Safety Injection. On the plant trip Containment Isolation Valve MOV-313, the Seal Water Return Line CIV, will fail to close automatically, requiring that the operator manually close this valve.

The operator will transition from E-0 to E-1, Loss of Reactor or Secondary Coolant, and then into ES-1.2 Post-LOCA Cooldown and Depressurization. While in ES-1.2, an Orange Path may occur on the RCS Integrity Critical Safety Function Status Tree.

The scenario will terminate at Step 14 of ES-1.2, after the crew has demonstrated the ability to evaluate/perform the SI flow reduction sequence, or upon transition to FR-P.1, Response to Imminent Pressurized Thermal Shock Condition.

Critical Tasks:

1. Trip all RCPs within 5 minutes of reaching trip criteria Safety Significance: Failure to trip all RCPs when required can lead to core uncovery and to fuel temperatures in excess of 2200°F. Analys1as have shown that if the RCPs are tripped within 5 minutes of the trip criteria being met, PCT will remain below 2200 °F, and if this action is delayed beyond 5 minutes, this PCT will be exceeded. It is a management expectation that the RCPs be tripped as quickly as possible, but within 5 minutes when the trip criteria is met. Failure to take this action represents mis-operation by the operator which leads to degradation of the fuel cladding fission produce barrier, and a violation of a license condition.

2. Close the Seal Water Return Containment Isolation Valve before transition out of E-0.

Safety Significance: Failure to close at least one Containment Isolation Valve on each critical penetration under the postulated conditions when it is possible to do so, constitutes mis-operation leading to degradation of the Containment Barrier. Failure to take this action leads to an unnecessary release of fission products to the auxiliary building, increasing the potential for release to tile environment, and reducing accessibility to vital equipment within the Auxiliary Building. Higher radiation levels within the Auxiliary Building will result in a degradation of ALARA principles.

Appendix D Scenario Outline Form ES-D-1 Draft (110414)

Facility: Ginna Scenario No.: 5 Op Test No.: N2014-301 Examiners: Operators: (SRO)

(RO)

(BOP)

Initial Conditions: The plant is at 1 x 10-8 amps (BOL). The plant ran at 100% power for 12 days, and then tripped four days ago due to a MIFW Pump failure. The repairs have been made and the plant is ready to be started back up. The crew will be directed to pull rods to the point of adding heat (POAH), and start the A MFW Pump in accordance with 0-1.2, Plant Startup From Hot Shutdown to Full Power Load, Step 6.3.4 and beyond. The area has experienced thunderstorms over the last 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and this is expected to continue for the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Turnover: The following equipment is Out-Of-Service: The B Condensate Pump is OOS for Bearing Replacement.

Event Malf. Event Type* Event No. No. Description 1 NA R-RO Raise Power to POAH N-SRO 2 NA N-BOP Testing of the MFW Oil Pumps N-SRO 3 CLG10 C-RO 480VAC Ground/A CCW Pump trips w/B CCW Pump failure to CLG02A C(TS)-SRO start in AUTO A-EDS33 4 EDS04D C-BOP Fault on 480V Bus 18/SW Pump D fails to start CLG01D C(TS)-SRO 5 CVC07A C-RO PCV-135 fails CLOSED C-SRO 6 NIS05A C-BOP Loss of Compensating Voltage to Intermediate Range N35 C(TS)-SRO 7 STM04D M-RO B ARV fails CLOSED/B SG Safety Valve Lifts and fails OPEN STM09A M-BOP M-SRO 8 RPS05A C-RO Failure of Reactor to Trip in AUTO only RPS05B C-SRO 9 SGN04B C-BOP B Steam Generator Tube Rupture/1% Failed Fuel RCS16 C-SRO

(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 Draft (110414)

Ginna 2014 NRC Scenario #5 The plant is at 1 x 10-8 amps (BOL). The plant ran at 100% power for 12 days, and then tripped four days ago due to a MFW Pump failure. The repairs have been made and the plant is ready to be started back up. The crew will be directed to pull rods to the point of adding heat (POAH), and start the B MFW Pump in accordance with 0-1.2, Plant Startup From Hot Shutdown to Full Power Load, Step 6.3.4 and beyond. The area has experienced thunderstorms over the last 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and this is expected to continue for the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The following equipment is Out-Of-Service: The B Condensate Pump is OOS for Bearing Replacement.

Shortly after taking the watch, the operator will raise power to the POAH in accordance with Step 6.3.4 of 0-1.2, Plant Startup From Hot Shutdown to Full Load. The operator will raise reactor power and stabilize reactor power at 0.5-1.0%; and then control Tavg at 547°F.

After this, the operator will conduct the oil pump checks on the A Main Feedwater (MFW)

Pump accordance with Attachment 5, Main Feed Pump A Startup of 0-1.2, Plant Startup From Hot Shutdown to Full Load, in preparation for starting the A MFW Pump.

Subsequently, a 480VAC ground will occur on Bus 14 and the A CCW Pump will trip, and the B CCW Pump will fail to automatically start. The operator will either start the A CCW pump manually per A-503.1, Emergency and Abnormal Operating Procedures Users Guide; or respond in accordance with AR-A-22:, CCW PUMP DISCHARGE LO PRESS 60 PSI, and enter AP-CCW.2, Loss of CCW During Power Operation. The operator will evaluate Technical Specification 3.7.7, CCW System.

Shortly after this, a fault on 480V Bus 18 will occur, resulting in Bus 18 de-ene~rgizing.

The operator will respond in accordance with AR-L**23, BUS18 UNDER VOLTAGE SAFEGUARDS and/or AR-L-5, SAFEGUARD BUS MAIN BREAKER OVERCURRENT TRIP and enter AP-ELECT.17/18, Loss of Safeguards Bus 17/18. The D Service Water Pump has failed to start, leaving only the B SW Pump running. The operator may leave the A EOG running or trip it within AP-ELECT.17 /18, but in either case align Alternate Cooling to the EOG. The operator will enter AP-SW.2, Loss of Service Water, and take actions to isolate non-essential SW loads. The operator will address TBchnical Specification 3.8.1, AC Sources - Modes 1, 2, 3, and 4; 3.8.9, Distribution Systems -

Modes 1, 2, 3,and 4; and 3.7.8, Service Water System.

During the recovery, Letdown Pressure Control Valve PCV-135 will fail closed causing the Letdown Line Relief valve to lift to the PRT. The operator will respond in accordance with AR-A-11, LETDOWN LINE HI PRESS 400 PSI, and take manual control of the valve.

Shortly afterwards, the compensating voltage power supply for the Intermediate~ Range Nuclear Instrument N35 will fail Low. The operator will respond in accordance with AR-E-9, IR N-35 LOSS OF COMPENSATING VOLTAGE and enter ER-NIS.2, IR MALFUNTION. The operator will address Technical Specification 3.3.1, Reactor Trip Instrumentation.

Appendix D Scenario Outlirn:~ Form ES-D-1 Draft (110414)

After this, the B ARV will inadvertently close, and one SG Safety valve will open and stick open. The reactor will fail to trip automatically, and the operator will need to manually trip the reactor and enter E-0, Reactor Trip or Safety Injection. On the trip approximately 1% failed fuel will occur. The operator will proceed through E-0, and then transition to E-2, Faulted Steam Generator Isolation.

While in E-2 a large Steam Generator Tube Rupture will develop on the B Steam Generator, and the operator will transition to E-3, Steam Generator Tube Rupture.

Because the B Steam Generator is both Ruptured and Faulted, the operator will transition to ECA-3.1, SGTR with Loss of Reactor Coolant, Subcooled Recovery Desired.

Also, while in E-2, it is expected that high temperatures will occur on the RCPs due to low service water flow, requiring the crew to stop these pumps. Because of this the NC condition of the RCS will tend to lower loop Tcold tempE~ratures causing an Orange/Red condition on RCS Integrity. If so, the operator may need to transition to FR-P.1, Response to Imminent Pressurized Thermal Shock Condition.

The scenario will terminate at Step 15 of ECA-3.1, after the crew has monitored conditions for continuing with the Subcooled Recovery procedure, or upon transition to FR-P.1.

Critical Tasks:

1. Manually trip the reactor prior to transition to FR-S.1, "Response to Nuclear Generation/A TWS."

Safety Significance: Failure to trip the reactor when required causes a challengH to the Subcriticality Critical Safety Function that otherwise would not exist. This mis-operation by the operator necessitates the crew taking compensating action which complicates the event mitigation strategy and demonstrates an inability by the operator to reco1~nize a failure of the automatic actuation of the RPS.

2. Isolate the Faulted Steam Generator before transitioning out of E-2.

Safety Significance: Failure to isolate a Faulted SG that can be isolated causes challenges to the Critical Safety Functions that would not otherwise occur. Failure to isolate flow could result in an unwarranted Orange or Red Path condition on RCS Integrity, Subcriticality (if cooldown is allowed to continue uncontrollably).

ML15043A678

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