ML14246A233

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Draft - Outlines (Folder 2)
ML14246A233
Person / Time
Site: Beaver Valley
Issue date: 07/01/2014
From:
FirstEnergy Nuclear Operating Co
To: Brian Fuller
Operations Branch I
Shared Package
ML14058A120 List:
References
U01884
Download: ML14246A233 (33)
v • d • e


Text

ES-401 PWR Examination Outline Form ES-401-2 Facil Date of Exam 7/7 thru 7118 2014 RO KIA Catego Points Tier Group K K K K K K A A

A A

G TOTAL 1

2 3

4 5

6 1

2 3

4 1

3 4

2 3

18

1.

Emergency 2

1 2

1 1

9 Abnormal Plant Tier 4 6 3 4

27 Evolutions Totals 1

3 2

3 3

2 2

3 3

2 2

3 28

2.

Plant 2

Systems 0

0 1

1 1

1 1

2 1

1 1

10 Tier 3 2 4 4 3 3 4 s,

3 3 4 38 Totals 1

2 3

4

3. Generic Knowledge and 10 Abilities Category 3

3 2

2 Note:

1.

Ensure that at least two topics from every applicable KIA cate~Jory 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 NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total :~5 points.

3.

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

4.

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

5.

Absent a plant-specific priority, only those KIAs having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6.

Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.

7.*

The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KIAs.

8.

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

9.

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

NUREG-1021, Revision 9 Supplement 1 RO Page 1 of 13 FENOC Facsimile Rev. 1

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

~=======================F=r=F=F9 EIAPE #I Name I Safety Function 000007 Reactor Tripl1

[Question 1]

000008 Pressurizer (PZR) Vapor Accident (Relief Valve Stuck Open)l3

[Question 2]

000011 Large Break LOCA/3

[Question 3]

0000151000017 Reactor (RCP) Malfunctionsl4

[Question 4]

Pump 000022 Loss of Reactor Coolant Makeupl2

.:stion 5]

000025 Loss of Residual Heat Removal System (RHRS)I4

[Question 6]

000027 Pressurizer Pressure Control System (PZR PCS) Malfunctionl3

[Question 7]

000029 Anticipated Transient Without Scram (A TWS)I1

[Question 8]

[Question 9]

NUREG-1021, Revision 9 Supplement 1 RO Page 2 of 13 KIA Topic(s)

EK2 Knowledge of the interrelations between a reactor trip and the following:

EK2.02 Breakers, relays and disconnects (CFR 41.7 145.7)

AK2 Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the followi,ng:

AK2.01 Valves (CFR 41.7 145.7)

EK3 Knowledge the reasons for the g

responses as the apply to the Large Break LOCA:

EK3.09 Maintaining DIGs available to provide standby power (CFR 41.5 141.10 145.61 45.13)

AK2 Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following:

AK2.07 RCP seals (CFR 41.7 145.7) 2.4.41' Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

(CFR: 41.10 143.5145.12)

AK2. Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following:

AK2.05 Reactor building sump IR 2.6 2.7*

4.2 2.9 4.2 2.6 Ability to operate and or monitor the following 3.1

  • as the~y apply to the Pressurizer Pressure Control Malfunctions:

M 1.02 SCR-controlled heaters in manual mode (CFR 41.7 145.5145.6)

EK1 Knowledge of the operational implications of the following concepts as they apply to the A TWS:

EK1.05 definition of negative temperature coefficient as applied to large PWR coolant systems (CFR 41.8 141.10 145.3)

AK1 Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture:

AK1.07 Effects of feedwater introduction on dry SIG (CFR 41.8 I 41.1 0 I 45.3) 2.8 3.4 FENOC Facsimile Rev. 1

ES-401 EIAPE #I Name I Safety Function 000054 Loss of Main Feedwater (MFW)I4

[Question 1 0]

000056 Loss of Offsite Powerl6

[Question 11]

000057 Loss of Vital AC Electrical Instrument Busl6

[Question 12]

000058 Loss of DC Powerl6

[Question 13]

IJvJ062 Loss of Nuclear Service Waterl4

[Question 14]

000065 Loss of Instrument Airl8

[Question 15]

000077 Generator Voltage and Electric Grid Disturbancesl6

[Question 16]

WIE04 LOCA Outside ContainmenU3

[Question 17]

NUREG-1 021, Revision 9 Supplement 1 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions-Tier 1/Group 1(RO) Continued RO Page 3 of 13 KIA Topic(s)

IR AA 1 Ability to operate and I or monitor the following 3.5 as they apply to the Loss of Main Feedwater (MFW):

AA 1.03 AFW auxiliaries, including oil cooling water supply (CFR 41.7 I 45.5 I 45.6) 2.4.34 Knowledge of RO tasks performed outside 4.2 the main control room during an emergency and the resultant operational effects.

(CFR: 41.10143.5 I 45.13)

AK3 Knowledge the reasons the following responses as they apply to the Loss of Vital AC Instrument Bus:

AK3.01 Actions contained in EOP for loss of vital ac electrical instrument bus (CFR 41.5 I 41.10 I 45.6 I 45.13)

AK1 Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power:

AK1.01 Battery charger equipment and instrumentation (CFR 41.8 I 41.10 I 45.3) 4.1 2.8 AA2 Ability to determine and interpret the ng 2.8*

as they apply to the Loss of Nuclear Service Water:

AA2.06 The length of time after the loss of SWS flow to a component before that component may be dama!led (CFR: 43.5 I 45.13)

AA2 Ability to determine and interpret the following as they apply to the Loss of Instrument Air:

AA2.03 Location and isolation of leaks (CFR: 43.5 I 45.13)

AA2 Ability to determine and interpret the ng as they apply to Generator Voltage and Electric Grid Disturbances:

AA2.09 Operational status of emergency diesel generators (CFR: 41.5 and 43.5 I 45.5, 45.7, and 45.8) 2.4.47 Ability to diagnose and recognize trends n an acc:urate and timely manner utilizing the appropriate control room reference material.

(CFR: 41.10 I 43.5 I 45.12) 2.6 3.9 4.2 FENOC Facsimile Rev. 1

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

~=======================r=r9F9F9 EIAPE #I Name I Safety Function WIE11 Loss of Emergency Coolant Recirculationl4

[Question 18]

KIA Category Point Totals:

NUREG-1 021, Revision 9 Supplement 1 K K K 1 2 3

RO Page 4 of 13 KIA Topic(s)

IR EA 1 Ability to operate and I or monitor the following

3. 7 as they apply to the (Loss of Emergency Coolant Recirculation)

EA 1.3 Desired operating results during abnormal and emergency situations.

(CFR: 41.7 145.5 I 45.6)

Group Point Total:

FENOC Facsimile Rev. 1 18

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

IF===========================~9F=r=T~~

E/APE #I Name I Safety Function 000003 Dropped Control Rod/1 (Question 19]

000005 Inoperable/Stuck Control Rod/1

[Question 20]

(PZR) Level 000060 Accidental Gaseous Release/9

[Question 22]

161 Area

<>rto<>toron Monitoring (ARM)

~ Jtem Alarms/7

[Question 23]

Plant fire on site/9

[Question 24]

W/E06 Degraded

[Question 25]

Cooling/4 W/E08 Pressurized Thermal Shock/4 (Question 26]

NUREG-1021, Revision 9 Supplement 1 KIA Topic(s)

IR Knowledge the operational implications of the 2.6 following concepts as they apply to Dropped Control Rod:

AK1.1 0 Definitions of core quadrant power tilt (CFR 41.8 /41.10 /45.3)

AK2 Knowledge of the interrelations between the Inoperable I Stuck Control Rod and the following:

AK2.02 Breakers, relays, disconnects, and control room switches M 1 Ability to operate and I or monitor lowing as the'y apply to the Pressurizer Level Control Malfunctions:

M 1.04 Regenerative heat exchanger and tempe!rature limits (CFR 41.7 /45.5/45.6) to determine ret the following as the'y apply to the Accidental Gaseous Radwaste:

M2.04 The effects on the power plant of isolating a given radioactive-gas leak (CFR: 43.5/ 45.13)

M 1 Ability to operate and I or monitor the following as the'y apply to the Area Radiation Monitoring (ARM)System Alarms:

2.5 2.7 2.6 3.6 E-'I'YY""~ M 1.01 Automatic actuation RO Page 5 of 13 (CFR 41.7 /45.5 I 45.6) to determine ret the following as thE!Y apply to the Plant Fire on Site:

M2.12 Location of vital equipment within fire zone (CFR: 43.5/ 45.13) 2.1.2~i Ability to interpret reference materials, such as graphs, curves, tables, etc.

(CFR: 41.10 /43.5/45.12) 2.9 3.9 EK3 Knowledge of the reasons for the following 3.4 responses as they apply to the (Pressurized Thermal Shock)

EK3.1 Facility operating characteristics during transi,ent conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity chan~1es and operating limitations and reasons for these operating characteristics.

(CFR: 41.5/41.10 /45.6/ 45.13)

FENOC Facsimile Rev. 1

ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions -Tier 1/Group 2(RO) Continued IF===========================~9F=r=r==

E/APE #I Name I Safety Function W/E 13 Steam Generator Overpressure/4

[Question 27]

KIA Category Point Totals:

NUREG-1021, Revision 9 Supplement 1 RO Page 6 of 13 KIA Topic(s)

EK2 Knowledge of the interrelations between the (Stearn Generator Overpressure) and the following:

EK2.2 Facility's heat removal systems, including prima1y coolant, emergency coolant, the decay heat removal systems, and relations between the proper operaltion of these systems to the operation of the facility.

(CFR: 41.7 /45.7)

Group Point Total:

IR 3.0 FENOC Facsimile Rev. 1 9

ES-401 System # I Name K

003 Reactor Coolant Pump System (RCPS)

[Question 28]

003 Reactor Coolant Pump System (RCPS)

[Question 29]

004 Chemical and Volume Control System (CVCS)

[Question 30]

005 Residual Heat Removal System (RHRS)

[Question 31]

006 Emergency Core Cooling em (ECCS)

L'-"uestion 32]

007 Relief Tank/Quench Tank System (PRTS)

[Question 33]

007 Pressurizer Relief Tank/Quench Tank System (PRTS)

[Question 34]

008 Component Cooling Water System (CCWS)

[Question 35]

010 Pressurizer Pressure Control System (PZR PCS) ff"'*Jestion 36]

NUREG-1021, Revision 9 Supplement 1 1

PWR Examination Outline Form ES-401-2 Plant Systems

-Tier 2/Group 1 (RO)

KIA Topic(s)

IR K6 Knowledge of the effect of a loss or 2.6 malfunction on the following will have on the RCPS:

K6.14 Starting requirements (CFR: 41.7/45.5)

K3 Knowledge of the effect that a loss or 3.5 malfunction of the RCPS will have on the following:

K3.02 S/G (CFR: 41.7 /45.6)

K4 Knowledge of CVCS design feature(s) 2.8 and/or interlock(s) which provide for the following:

K4.01 Oxygen control in RCS (CFR: 41.7)

K5 Knowledge of the operational 2.9*

of the following concepts as they apply the RHRS:

K5.03 Reactivity effects of RHR fill water (CFR: 41.5/45.7)

K3 Knowledge of the effect that a loss or 4.2 malfunction of the ECCS will have on the following:

12 Knowledge surveillance procedures.

3.7 (CFR: 41.10 /45.13)

A2 Ability to (a) predict the impacts of the 3.9 following malfunctions or operations on the PRTS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.01 Stuck-open PORV or code safety (CFR: 41.5/43.5 /45.3/45.13)

K3 Knowledge of the effect that a loss or 3.4 malfunction of the CCWS will have on the following:

K3.01 Loads cooled by CCWS (CFR: 41.7 /45.6)*

A4 Ability to manually operate and/or monitor 4.0 in the control room:

A4.03 PORV and block valves (CFR: 41.7 /45.5 to 45.8)

RO Page 7 of 13 FENOC Facsimile Rev. 1

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems IF==================F=F9F9F9F9F=F=

System # I Name 010 Pressurizer Pressure Control System (PZR PCS)

[Question 37]

012 Reactor Protection ""'"r..,m

[Question 38]

012 Protection System

[Question 39]

013 Engineered Safety Features Actuation System (ESFAS)

[Question 40]

' Containment Cooling

.em (CCS)

[Question 41]

026 Containment Spray System (CSS)

[Question 42]

026 Containment Spray System (CSS)

[Question 43]

039 Main and Reheat Steam System (MRSS)

[Question 44]

NUREG-1021, Revision 9 Supplement 1 RO Page 8 of 13

-Tier 2/Group 1 (RO) Continued KIA Topic(s)

K4 Knowledge of PZR PCS design feature(s) and/or interlock(s) which provide for the following:

K4.01 Spray valve warm-up (CFR: 41.7)

K2 Knowledge of bus power supplies to the following:

K2.01 RPS channels, components, and interconnections (CFR: 41.7)

K6 Knowledge of the effect of a loss or malfunction of the following will have on the RPS:

K6.01 Bistables and bistable test equipment (CFR: 41.7/45.7)

K1 Knowledge of the physical connections and/or cause effect relationships between the ESFAS and the following systems:

K1.03 CCS (CFR: 41.2 to 41.9/45.7 to 45.8)

IR 2.7 3.3 2.8 3.8 A3 Ability to monitor automatic operation the 4.1 CCS, including:

A3.01 Initiation of safeguards mode of operation (CFR: 41.7/45.5)

K1 Knowledge of the physical connections and/or cause/effect relationships between the CSS and the following systems:

K1.01 ECCS (CFR: 41.2 to 41.9/45.7 to 45.8)

A4 Ability to manually operate and/or monitor in the control room:

A4.05 Containment spray reset switches (CFR: 41.7 /45.5 to 45.8) 4.2 3.5 2.2.39 Knowledge of less than or equal to one 3.9 hour1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> Technical Specification action statements for systems.

(CFR: 41.7/41.10 /43.2/45.13)

FENOC Facsimile Rev. 1

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems

~=================r=r=F=F9F9F=F=

System # I Name 039 Main and Reheat Steam System (MRSS)

[Question 45) 059 Main Feedwater (MFW)

System

[Question 46) 061 Auxiliary I Emergency Feedwater (AFW) System

[Question 47) 062 AC Electrical Distribution

~"stem

~stion 48) 063 DC Electrical Distribution System

[Question 49) 064 Emergency Generator (ED/G) System

[Question 50]

073 Process Radiation Monitoring (PRM) System

[Question 51)

NUREG-1021, Revision 9 Supplement 1 K K 1 2 RO Page 9 of 13

-Tier 2/Group 1 (RO) Continued KIA Topic(s)

A 1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including:

A 1.09 Main steam line radiation monitors (CFR: 41.5 I 45.5)

IR 2.5*

A2 Ability to (a) predict the impacts of the 3.0*

following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.11 Failure of feedwater control system (CFR: 41.5/43.5/45.3/ 45.13)

K5 Knowledge of the operational implications of the following concepts as the apply to the AFW:

K5.03 Pump head effects when control valve is shut (CFR: 41.5/ 45. 7)

A 1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including:

A 1.01 Significance of DIG load limits (CFR: 41.5/ 45.5) 2.6 3.4 A3 Ability to monitor automatic operation of the

2. 7 DC electrical system, including:

A3.01 Meters, annunciators, dials, recorders, and indicating lights (CFR: 41.7 /45.5)

K1 Knowledge of the physical connections 3.6 and/or cause/effect relationships between the ED/G system and the following systems:

K1.04 DC distribution system (CFR: 41.2 to 41.9 /45.7 to 45.8)

A 1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRM system controls including:

A 1.01 Radiation levels (CFR: 41.5 I 45.7) 3.2 FENOC Facsimile Rev. 1

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems

~==================~==r=r=T=~=F=F~

System # I Name 076 Service Water System (SWS)

(Question 52]

078 Instrument Air System (lAS)

(Question 53]

1 03 Containment System

[Question 54]

103 nment System (Question 55]

KIA Category Point Totals:

NUREG-1021, Revision 9 Supplement 1 RO Page 10 of 13

-Tier 2/Group 1 (RO) Continued KIA Topic(s)

K4 Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following:

K4.01 Conditions initiating automatic closure of closed cooling water auxiliary building header supply and return valves (CFR: 41.7)

K2 Knowledge of bus power supplies to the following:

K2.01 Instrument air compressor (CFR: 41.7)

IR 2.5*

2.7 2.1.19 Ability to use plant computers to 3.9 evaluate system or component status.

(CFR: 41.10/45.12)

A2 Ability to (a) predict the impacts of the 3.5*

following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.03 Phase A and B isolation (CFR: 41.5/43.5 /45.3/45.13)

Group Point Total:

FENOC Facsimile Rev. 1 28

ES-401 System # I Name 001 Control Rod Drive System

[Question 56]

011 System (PZR LCS)

[Question 57]

014 Rod Position Indication System (RPIS)

[Question 58]

015 Nuclear Instrumentation System (NIS)

[Question 59]

016 Non-Nuclear ln~trumentation System (NNIS) stion 60]

034 Fuel Handling Equipment System (FHES)

[Question 61]

041 Steam Dump System (SDS)/Turbine Bypass Control

[Question 62]

056 Condensate

[Question 63]

NUREG-1021, Revision 9 Supplement 1 K K 1 2 PWR Examination Outline Form ES-401-2 RO Page 11 of 13

- Tier 2/Group 2{RO)

KIA Topic(s)

K6 Knowledge of the effect of a loss or malfunction on the following CRDS components:

K6.02 Purpose and operation of sensors feeding into the CRDS (CFR: 41.7/45.7)

IR 2.8 A 1 Ability to predict and/or monitor changes in 3.1 parameters (to prevent exceeding design limits) associated with operating the PZR LCS controls including:

A1.04 T-ave (CFR: 41.5/45.5) 2.2.37 Ability to determine operability and/or availability of safety related equipment.

(CFR: 41.7 /43.5/45.12)

K3 of the effect that a loss or malfunction of the NIS will have on the following:

K3.01 RPS (CFR: 41.7 /45.6)

A3 Ability to monitor automatic operation of the NNIS, including:

A3.02 Relationship between meter readings and actual parameter value (CFR: 41.7 /45.5)

K4 Knowledge of design feature(s) and/or interlock(s) which provide for the following:

K4.01 Fuel protection from binding and dropping (CFR: 41.7)

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; and (b) based on those predictions use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.03 Loss of lAS (CFR: 41.5/43.5 /45.3 /45.13)

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.04 Loss of condensate pumps (CFR: 41.5/ 43.5/ 45.3/ 45.13) 3.6 3.9 2.9*

2.6 2.8 2.6 FENOC Facsimile Rev. 1

ES-401 PWR Examination OIUtline Form ES-401-2 Plant Systems

- Tier 2/Group 2(RO) Continued System # I Name K

KIA Topic(s) 1 IR 075 Circulating Water System A4 Ability to manually operate and/or monitor 3.2*

[Question 64]

in the control room:

A4.01 Emergency/essential SWS pumps (CFR: 41.7 /45.5 to 45.8) 086 Fire Protection System K5 Knowledge of the operational 3.1 (FPS) the following concepts as they apply to the Fire

[Question 65]

Protection System:

K5.03 Effect of water spray on electrical components (CFR: 41.5/45.7)

KIA Category Point Totals:

0 Group Point Total:

10 NUREG-1 021, Revision 9 Supplement 1 RO Page 12 of 13 FENOC Facsimile Rev. 1

ES 401 Generic Knowledge and Abilities C>utline (Tier 3)

Form ES-401-3

.l "cility: BVPS UNIT 1 RO Date ofExam 7/7 thru 7/18 2014 Category

1.

Conduct of Operations

2.

Equipment Control

3.

Radiation Control

4.

Emergency Procedures/

Plan KJA#

Topic IR 2.1.5 Ability to use procedures related to shift staffing, :such as minimum crew 2.9*

complement, overtime limitations, etc.

(CFR: 41.10 I 43.5 I 45.12)

[Question 66]

2.1.20 Ability to interpret execute procedure steps.

4.6 (CFR: 41.10 I 43.5 I 45.12)

[Question 67]

2.1.27 Knowledge of system purpose and/or function.

3.9 (CFR: 41.7)

[Question 68]

Subtotal 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

(CFR: 41.10 I 43.2 I 45.13)

[Question 69]

2.2.37 Ability to determine operability and/or availability of safety related 3.6 equipment.

(CFR: 41.7 I 43.5 I 45.12)

[Question 70]

2.2.42 Ability to recognize system parameters that are entry-level conditions for 3.9 Technical Specifications.

(CFR: 41.7 I 41.10 I 43.2 I 43.3 I 45.3)

[Question 71]

Subtotal 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(CFR: 41.12 I 45.9 I 45.10)

[Question 72]

2.3.14 Knowledge of radiation or contamination hazards that may arise during 3.4 normal, abnormal, or emergency conditions or activities.

(CFR: 41.12 I 43.4 I 45.10)

[Question 73]

Subtotal 2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

(CFR: 41.7 I 45.7 I 45.8)

[Question 74]

2.4_31 Knowledge of annunciator alarms, indications, or response procedures.

4.2 (CFR: 41.10 I 45.3)

[Question 75]

Subtotal Tier 3 Point Total RO NUREG-1021, Revision 9 Supplement 1 RO Page 13 of 13 FENOC Facsimile Rev. 1

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

SRO ONLY Points Tier Group A2 G*

TOTAL 1

3 3

6

1.

Emergency 2

2 2

4 Abnormal Plant Tier 5

5 10 Evolutions Totals 1

3 2

5

2.

Plant 2

Systems 0 2 1

3 Tier 5

3 8

Totals

3. Generic Knowledge and 1 2 3 4 7

Abilities Category 2 1 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 NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3.

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

4.

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

5.

Absent a plant-specific priority, only those KIAs having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6.

Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.

7.*

The generic (G) KIAs 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 0.1.b of ES-401 for the applicable KIAs.

8.

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

9.

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

NUREG-1021, Revision 9 Supplement 1 SRO Page 1 of 6 FENOC Facsimile Rev. 1

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

~=======================F=F=F=F=

EIAPE #I Name I Safety Function 000009 Small Break LOCA/3

[Question 76]

000026 Loss of Component Cooling Water (CCW)I8

[Question 77]

000038 Steam Generator Tube Rupture (SGTR)I3

[Question 78]

000055 Loss Offsite and Onsite Power (Station Blackout)l6

[Question 79]

WIE05 Loss

[Question 80]

..:12 Steam Generatorsl4

[Question 81]

KIA Category Point Totals:

NUREG-1021, Revision 9 Supplement 1 of all K K K 1

2 3

KIA Topic(s)

IR 2.4.8 Knowledge of how abnormal operating 4.5 procedures are used in conjunction with EOPs.

(CFR: 41.10 I 43.5 I 45.13)

M2 Ability to determine and interpret the following 3.6 as they apply to the Loss of Component Cooling Wate1r:

M2.02 The cause of possible CCW loss (CFR: 43.5 I 45.13) 2.1.32 Ability to explain and apply system limits and 4.0 precautions.

(CFR: 41.10 I 43.2 I 45.12)

EA2 Ability to ne or interpret the following as 3.7 they apply to a Station Blackout:

EA2.01 Existing valve positioning on a loss of instrument air system (CFR 43.5 I 45.13)

EA2 Ability to determine and interpret the following 4.4 as they apply to the (Loss of Secondary Heat Sink)

EA2:1 Facility conditions and selection of appropriate procedures during abnormal and emer1~ency operations.

(CFR: 43.5 I 45.13) 2.4.4l Ability to diagnose and recognize trends in an 4.2 accurate and timely manner utilizing the appropriate control room reference material.

(CFR: 41.10 I 43.5 I 45.12)

Group Point Total:

SRO Page 2 of 6 FENOC Facsimile Rev. 1 6

ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal P~ant Evolutions -Tier 1/Group 2(SRO) l~=======================r=r9r9F=

E/APE #I Name I Safety Function 000037 Steam Generator (S/G) Tube Leak/3

[Question 82]

000059 Release/9 Liquid Radwaste

[Question 83]

000076 High Reactor Activity/9

[Question 84]

W/E1 0 Natural Circulation with Steam Void in Vessel with/without RVLIS/4

[Question 85]

gory Point Totals:

NUREG-1021, Revision 9 Supplement 1 K K K A 1 2 3

1 0

0 0

KIA Topic(s)

AA2 Ability to determine and interpret the following as thElY apply to the Steam Generator Tube Leak:

AA2.09 System status, using independent readings from redundant Condensate air ejector exhaust monitor (CFR: 43.5/45.13) 2.4.4'1 Knowledge of the emergency action level thresholds and classifications.

(CFR: 41.10 I 43.5/45.11)

AA2 Ability to determine and interpret the following as thElY apply to the High Reactor Coolant Activity:

AA2.06 Response of PZR LCS to changes in the letdown flow rate (CFR: 43.5/45.13) 2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

(CFR: 41.10 /43.5 /45.3)

Group Point Total:

IR 3.4*

4.6 2.5 4.0 SRO Page 3 of 15 FENOC Facsimile Rev. 1 4

ES-401 PWR Examination 01LJtline Form ES-401-2 Plant Systems

~=================r=r=F=F9=9=9F=K System # I Name 004 Chemical and Volume Control System

[Question 86) 022 Containment Cooling System (CCS)

[Question 87) 063 Electrical Distribution System

[Question 88]

064 Emergency Diesel Generator (ED/G) System

[Question 89) 078 Instrument Air System (lAS)

[Question 90]

KIA Category Point Totals:

NUREG-1021, Revision 9 Supplement 1 K K 1 2 SRO Page 4 of,6

-Tier 2/Group 1 (SRO)

KIA Topic(s)

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the eves; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.26 Low VCT pressure (CFR: 41.5/43.5/45.3/45.5)

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.01 Fan motor over-current (CFR: 41.5/ 43.5 /45.3/ 45.13)

IR 3.0 2.7 2.2.37 Ability to determine operability and/or 4.6 availability of safety related equipment.

(CFR: 41.7 /43.5/45.12)

A2 Ability to (a) predict the impacts of the 2.9 following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.02 Load, VARS, pressure on air compressor, speed droop, frequency, voltage, fuel oil level, temperatures (CFR: 41.5/43.5 /45.3 I 45.13) 2.2.44 Ability to interpret control room 4.4 indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5/43.5/45.12)

Group Point Total:

FENOC Facsimile Rev. 1

ES-401 System # I Name K

015 Nuclear Instrumentation System (NIS)

[Question 91]

071 Waste Disposal System (WGDS)

[Question 92]

079 Station Air System

[Question 93]

KIA Category Point Totals:

NUREG-1021, Revision 9 Supplement 1 1

PWR Examination Outline Form ES-401-2 Plant Systems

-Tier 2/Group 2{SRO)

KIA Topic(s)

IR A2 Ability to (a) predict the impacts of the 3.9 following malfunctions or operations on the NIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.01 Power supply loss or erratic operation (CFR: 41.51 43.5 I 45.3145.5)

A2 Ability to (a) predict the impacts of the 3.6 following malfunctions or operations on the Waste Gas Disposal System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.02 Use of waste gas release monitors, radiation, gas flow rate, and totalizer (CFR: 41.5143.5145.3145.13) 2.2.25 Knowledge of the bases in Technical 4.2 Specifications for limiting conditions for operations and safety limits.

(CFR: 41.5141.7143.2)

Group Point Total:

3 SRO Page 5 of (j FENOC Facsimile Rev. 1

ES 401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 F'"-~cility: BVPS UNIT 1 SRO Date ofExam 7/7 thru 7118 2014 Category KIA#

Topic SROOnly IR

1.

2.1.39 Knowledge of conservative decision making practices.

4.3 Conduct of Operations (CFR: 41.10 /43.5/45.12)

[Question 94]

2.1.31 Ability to locate control room switches, controls, and indications, and to 4.3 determine that they correctly reflect the desired plant lineup.

(CFR: 41.10 /45.12)

[Question 95]

Subtotal 2

2.

2.2.44 Ability to interpret control room indications to verify the status and Equipment operation of a system, and understand how operator actions and Control directives affect plant and system conditions.

(CFR: 41.5/43.5/45.12)

[Question 96]

Subtotal 1

3.

2.3.5 Ability to use radiation monitoring systems, such as fixed radiation Radiation monitors and alarms, portable survey instrumen1ts, personnel monitoring trol equipment, etc.

(CFR: 41.11 /41.12 /43.4/45.9)

[Question 97]

2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation 3.1 monitors and alarms, portable survey instrumen1ts, personnel monitoring equipment, etc.

(CFR: 41.12 I 43.4/45.9)

[Question 98]

Subtotal 2

4.

2.4.3 Ability to identify post-accident instrumentation.

Emergency (CFR: 41.6/45.4)

Procedures/

Plan

[Question 99]

2.4.44 Knowledge of emergency plan protective action recommendations.

4.4 (CFR: 41.10 /41.12/43.5/45.11)

[Question 1 00]

Subtotal 2

Tier 3 Point Total 7

NUREG-1021, Revision 9 Supplement 1 SRO Page 6 of 15 FENOC Facsimile Rev. 1

ES-401 R ecor d f R. t d KIA 0

ejec e s

F orm ES-401 4 -

Facility: BVPS Unit 1 Date of Exam 7/7 thru 7/18 2014 Operating Test No.: BV1LOT14 NRC

<:tr I Randomly Reason for Rejection

_.oup Selected KIA RO OUTLINE 1/2 000067 AA2.06 Question #24; At Beaver Valley the need for pressurizing the control room during a plant fire on site is an SRO decision based on the SRO's judgment of the control room environment. Therefore we are unable to construct a question that will have definitive criteria for the need to pressurize the control room. Randomly selected 000067 AA2.12 as a replacement.

2/1 012 K6.07 Question #39; Beaver Valley does not have a core protection calculator.

Randomly selected 012 K6.01 as a replacement.

2/1 039 2.2.25 Question #44; Knowledge of TS bases is an SRO knowledge lAW the NRC Clarification guidance for SRO-only questions Rev. 1 (3/11/1 0). Randomly selected 039 2.2.39 as a replacement.

2/1 059 A2.07 Question #46; Beaver Valley does not have MFW pump turbines. Randomly selected 059 A2. 11 as a replacement.

2/1 103 A2.05 Question #55; Beaver Valley does not have any emergency containment entry procedures. Randomly selected 103 A2.03 as a replacement.

2/2 015 K3.04 Question #59; Beaver Valley does not have ICS. Randomly selected 015 K3.01 as a replacement.

2/2 041 K2.02 Question #62; Beaver Valley does not have ICS. Randomly selected 041 A2.03 as a replacement.

3 2.3.11 Question #71; Ability to control radiation releases has already been addressed in this exam with RO and SRO JPMs as well as Q# 92 of the written exam. In order to maintain exam balance, randomly selected 2.2.42 as a replacement.

1/2 000060 AA2.04 Question #22; Beaver Valley only has a single ARM that has automatic actions and this is already being addressed in Question #23. Randomly selected 000060 AA2.04 as a replacement.

1/1 000025 AK3.02 Question #6; Beaver Valley no longer isolates RHR low pressure piping on increasing pressure. Rather, this function is accomplished with the low pressure over pressurization protection system. The system employs the PORVs to relieve pressure without isolating the RHR low pressure piping. Therefore the KIA is not applicable to Beaver Valley. Discussed with the chief examiner and the decision was made tore-select a different KIA. Randomly selected 000025 AK3.03 as a replacement.

NUREG-1021, Revision 9 Supplement 1 Page 1 of 2 FENOC Facsimile Rev. 1

ES-401 R ecor d f R. t d KIA 0

eJec e s

F orm ES-401-4 ity: BVPS Unit 1 Date of Exam 7/7 thru 7/18 2014 Operating Test No.: BV1LOT14 NRC SRO OUTLINE 1/1 W/E12 2.4.1 Question #81; Knowledge of EOP entry conditions and immediate actions is RO knowledge lAW the NRC Clarification guidance for SRO-only questions Rev.

1(3/11/10). Randomly selected W/E12 2.4.47 as a replacement.

2/2 027 A2.01 Question #91; Beaver Valley has retired the containment iodine removal system and replaced it with a passive s1odium tetraborate system. Randomly selected 015 A2.01 as a replacement.

3 2.1.44 Question #95; Knowledge of RO duties in the control room is RO knowledge lAW the NRC Clarification guidance for SRO-only questions Rev. 1 (3/11/1 0).

Randomly selected 2.1.31 as a replacement.

3 2.4.8 Question #99; Overlap with question 76 identical KIA creates overlap and potential double-jeopardy issues. Randomly selected 2.4.3 as a replacement.

2/1 064 A2.09 Question # 89; Unable to develop a discriminatory question for the second part of the KIA which requires procedure use to address a malfunction regarding an EDG that fails to synchronize. Other than to call maintenance to effect repairs, there is no operational procedural guidance for this condition. Discussed with the chief examiner and the decision was made to re-select a different KIA Randomly selected 064 A2.02 as a replacement.

NUREG-1021, Revision 9 Supplement 1 Page 2 of2 FENOC Facsimile Rev. 1

ES-301 BVPS Unit 1 I Facility

~amination Level RO I:R1 Administrative Topic (See Note)

Conduct of Operations (RO A.1.1)

Conduct of Operations (RO A.1.2)

..:juipment Control (RO A.2)

Radiation Control (RO A.3)

Emergency Procedures/Plan (RO A.4)

Type Code*

M,R M,R N,R D,R Administrative Topics Outline Form ES-301-1 Date of Examination:

717 thru 7118 2014 SROD Operating Test Number BV1LOT14 NRC Des<:ribe activity to be performed 2.1.23 (4.3)

Ability to perform specific system and integrated plant procedures during all modes of plant operation.

(1AD-007) Perform an ECP Calculation 2.1.43 (4.1)

Ability to use procedw*es to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

(1AD-016) Plot and Evaluate 1/M Data 2.2.13 (4.1)

Knowledge of tagging and clearance procedures.

(1AD-047) Prepare a CIE~arance Tagout for FW-P-38 2.3.11 (3.8)

Ability to control radiation releases.

(1AD-004) Perform Decay Tank Discharge Pre-Release Verification NOT EVALUATED II

..,()T!::: 11.:: *

'~

or SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • Type Codes & Criteria NUREG-1021, Revision 9 Supplement 1 (C)ontrol Room, (S)imulator, or Class(R)oom (D)irect from bank (~ 3 for ROs; ~ 4 for SROs & RO retakes)

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

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

FENOC Facsimile Rev. 1

ES-301 Administrative Topics Outline Form ES-301-1 Facility:

BVPS Unit 1 Date of Examination: 7/7 thru 7/18 2014

<amination Level RO D SRO lXI Operating Test Number BV1LOT14 NRC Administrative Topic Type Describe activity to be performed Code*

(See Note)

Conduct of Operations M, R 2.1.23 (4.4)

(SRO A.1.1)

Conduct of Operations (SRO A.1.2)

Equipment Control

,.:>RO A.2)

Radiation Control (SRO A.3)

Emergency Procedures/Plan (SRO A.4)

N, R N,R N,R N,S Ability to perform specific system and integrated plant procedures during all modes of plant operation.

(1AD-011) Review an ECP Calculation 2.1.3 (3.9)

Knowledge of shift or short-term relief turnover practices.

(1AD-048) Determine Availability for Call-in (3 ROs) 2.2.17 (3.8)

Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.

(1AD-041) Perform a Risk Assessment [1 FW-P-3A] Maintenance 2.3.11 (4.3)

Ability to control radiation releases.

(1AD-042) Determine Compensatory Actions for RM-P-1GW-108 and 02A-1 GW-11 0-1 Being OOS 2.4.41 (4.6)

Knowledge of the emerfrency action level thresholds and classifications.

(1AD-046) Classify an Emergency Event (Scenario Specific)

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

  • Type Codes & Criteria NUREG-1021, Revision 9 Supplement 1 (C)ontrol Room, (S)imulator, or Class(R)oom (O)irect from bank (5. 3 for ROs; ~ 4 for SROs & RO retakes)

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

(P)revious 2 exams (5. 1 ; randomly selected)

FENOC Facsimile Rev. 1

ES 301 C

t I R on ro

/1 PI t S t oom n-an iys ems o tr u me F orm ES 301 2 Facility:

BVPS Unit 1 Date of Examination:

7171 thru 7/18 2014 I Exam Level: RO 1:&1 SRO(I) D SRO(U) D Operating Test No.:

BV1LOT14 NRC Control Room Systems© (8 for RO); (7 for SR0-1); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code*

Safety Function S1 -Shift From Main F/W Reg. Valves to Bypasses (1 CR-072)

S,D 4S S2-Discharge Laundry and Contaminated Shower Drain Tank- (High S,N,A 9

radiation alarm received - No Auto Actions occur) (1 CR-658)

S3-Rod Control Assembly Partial Movement Test- (Rod fails to move S,D,A 1

during performance of the test) (1 CR-596)

S4 - Drain SIS Accumulator - (Low pressure alarm receiv~3d during S,N,A,EN 3

draining) (1 CR-659)

S5-Place Excess Letdown in Service (1 CR-056)

S,D 2

S6-Respond to High PRT Temperature- (High PRT Ieveli received during S,N,A 5

spray down of PRT) (1 CR-661)

S7 - X-Fer 4KV Emer Bus from EDG to Normal Feed - (EDG overcurrent S,N,A 6

occurs and EDG auto actions fails to occur) (1 CR-6,60)

SS - Respond to a Loss of the RHR System (RHR pump trips requiring S, D, A, L, E 4P RCP start) (1CR-594)

In-Plant Systems© (3 for RO); (3 for SR0-1); (3 or 2 for SRO-U)

P1 - Startup a Rod Drive MG (1 PL-003)

D,L 1

P2-Reset the Terry Turbine Trip Throttle Valve (1 PL-004)

D,R 4S P3-Place the Diesel Air Compressor in Service (1PL-031)

D,E 8

All RO and SRO control room (and in-plant) systems m1Jst be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO I SR0-1 I SRO-U (A)Iternate Path 4-6 14-6 12-3 (C)ontrol room (D)irect from bank S91S81S4 (E)mergency or abnormal in-plant

~11~11~1 (EN)gineered safety feature

-I-I~ 1 (Control room system)

(L)ow-power I Shutdown

~11~11~1 (N)ew or (M)odified from bank including 1 (A)

~21~21~1 (P)revious 2 exams s 31 s 3 Is 2 (randomly selected)

(R)CA

~11~11~1 II (S)imulator NUREG-1 021, Revision 9 Supplement 1 FENOC Facsimile Rev.1

ES 301 C

t I R on ro

/1 PI t S t oom n-an iys ems o tr u me F orm ES 301 2 Facility:

BVPS Unit 1 Date of Examination:

7/7/thru 7/18 2014 Exam Level: RO D SRO(I) I2II SRO(U) D Operating Test No.:

BV1LOT14 NRC I Control Room Systems@ (8 for RO); (7 for SR0-1); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code*

Safety Function 51 - Shift From Main FIW Reg. Valves to Bypasses (1 CR-072)

S,D 4S 52-Discharge Laundry and Contaminated Shower Drain Tank- (High S,N,A 9

radiation alarm received - No Auto Actions occur) (1 CR-658) 53-Rod Control Assembly Partial Movement Test- (Rod fails to move S,D,A 1

during performance of the test) (1 CR-596) 54 - Drain SIS Accumulator- (Low pressure alarm received during S, N,A,EN 3

draining) (1 CR-659) 56-Respond to High PRT Temperature- (High PRT leve~l received during S,N,A 5

spray down of PRT) (1CR-661) 57 - X-Fer 4KV Emer Bus from EDG to Normal Feed - (EDG overcurrent S,N,A 6

occurs and EDG auto actions fails to occur) (1 CR-660) 58-Respond to a Loss of the RHR System (RHR pump trips requiring S,D,A,L,E 4P RCP start) (1CR-594)

In-Plant Systems@ (3 for RO); (3 for SR0-1); (3 or 2 for SRO-U)

P1 -Startup a Rod Drive MG (1 PL-003)

D, L 1

P2-Reset the Terry Turbine Trip Throttle Valve (1 PL-004)

D,R 4S P3-Place the Diesel Air Compressor in Service (1PL-03>1)

D,E 8

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

  • Type Codes Criteria for RO I SR0-1 I SRO-U (A)Iternate Path 4-6 14-6 /2-3 (C)ontrol room (D)irect from bank
s;9/S8/:s;4 (E)mergency or abnormal in-plant

~1/~1/~1 (EN)gineered safety feature 1 ~ 1 (Control room system)

(L)ow-power I Shutdown

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

~2/~2/~1 (P)revious 2 exams s 3/ s 3/ :s; 2 (randomly selected)

(R)CA

~1/~1/~1 11 (S)imulator NUREG-1 021, Revision 9 Supplement 1 FENOC Facsimile Rev.1

A d"

D

~ppen 1x Scenario Outline 1L14N1 Facility:

BVPS Unit 1 Scenario No. 1 Op Test No.:

BV1LOT14 NRC Examiners:

Candidates:

SRO ATC BOP Initial IC 68(29): 100% power, EOL, Equ. XE Conditions, CB "D" @ 227 steps, RCS boron - 100 Conditions:

ppm.

Turnover:

Maintain 100% power.

"A" Train Priority, 1FW-P-3B is OOS with 1FW-P-2 aligned to the "B" header lAW TS 3.7.5, Condition B.

Critical Tasks:

1. CT-50 (FR-S.l.A) Crew isolates the main turbine
2. CT -3 (E-O.E) Crew manually actuates CIB
3. CT-11 (E-0.0) Crew closes cnmt isolation valves
4. CT -33 (ECA-2.1.A) Crew controls the AFW flow rate Event Event Type Event Description No.

1 (TS) SRO RCS loop 1, ch III, flow transmitter, FT-1RC-416 fails low.

2 (C) ATC, SRO River water pump, 1 WR-P-1A trips, requiring manual start of (TS) SRO 1WR-P-1B.

3 (C)ATC, SRO "B" RCP thermal barrier leak/ TV -1 CC-1 07B fails to auto close on high flow, requires manual closure.

4 (R) ATC "B" S/G feedwater leak inside CNMT requiring an unplanned (N) BOP, SRO power reduction per AOP 5

"B" feedwat1;!r leak increases once power is reduced to 94%.

(M)ALL Reactor fails to trip from the control room, requires entry into FR-S.l.

6 (M)ALL Upon Rx trip, all 3 steam lines fault.

7 (I) BOP, SRO Automatic main steam line isolation failure, requires manual main steam line isolation.

8 Train "B" CIA fails to actuate along with train "A" valve (I) ATC, SRO MOV-1CH-378 failing to automatically close, ATC must manually actuate CIA or close MOV-1CH-378.

9 (I)ATC, SRO Automatic CIB actuation failure, requires manual actuation.

10 (C) BOP, SRO Control room dampers, 1VS-D-40-1A, 1B, 1C and 1D fail to automatically close, requires manual closing.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Revision 9 Supplement 1 FENOC Facsimile Rev. 1

Appendix D Scenario Outline 1L14N1

  • fter taking the shift at 100% power, RCS loop flow transmitter, FT-1RC-416 fails low. The crew will

.agnose the indications and the SRO will address Tech Specs for the failed transmitter.

The 1A Reactor Plant River Water Pump, 1 WR-P-1A then trips with an auto start failure of the standby pump, 1 WR-P-1B. The SRO will direct the crew to perform the actions of AOP 1.30.2, River Water/ Main Intake Structure Loss, to start 1 WR-P-1B and restore river water. The SRO will address Tech Specs.

A leak will then occur in the "B" RCP thermal barrier heat exchanger. The isolation valve, TV -1 CC-1 07B, will fail to auto close on high flow and will require manual closure. The BOP will refer to the alarm response procedure and the SRO will direct closure of the valve.

A feed water leak will occur inside of the containment, the SRO will direct the crew to commence a rapid power reduction IA W AOP 1.51.1, "Unplanned Power Reduction".

The feedwater leak will increase in size during the power reduction, due to degrading conditions, the SRO will direct the A TC to trip the reactor.

The reactor will fail to trip from the control room. The SRO will direct the ATC and BOP to perform the lOA's ofFR-S.1, "Response to Nuclear Power Generation-ATWS. The turbine will also fail to automatically trip and must be manually tripped. The reactor will be tripped via a local operator after being dispatched and the crew will return to E-0.

When the reactor is tripped, all three main steam lines will fault inside of containment.

'T'l:le main steam line isolation signal will fail to occur and the valves must be manually aligned.

The safety injection that occurred as a result of the MSLB will fail to actuate the train "B" CIA signal, and train "A" CIA valve, MOV-1CH-378 will fail to automatically close. The ATC will recognize the failure and isolate the containment penetration via either manually actuating Train "B" CIA or manually closing MOV-1CH-378.

A CIB signal will fail to actuate on high containment pressure requiring the ATC to manually initiate CIB actuation. The control room ventilation dampers 1 VS-D-40-ll A through 1 D will not close on the CIB signal requiring the BOP to manually close the dampers.

The SRO will transition to EOP E-2, "Faulted Steam Generator Isolation" based upon the rapid depressurization of the Steam Generators, then transition to ECA-2.1, "Uncontrolled Depressurization of All Steam Generators".

The scenario will be terminated after the crew evaluates if th~~ LHSI pumps are to be secured in ECA-2.1.

Expected procedure flow path is E-0._ FR-S.1._ E-0._ E-2._ ECA-2.1 NUREG-1021, Revision 9 Supplement 1 FENOC Facsimile Rev. 1

A d" D

~ppen IX s

  • o r cenar1o utme 1L14N2 Facility:

BVPS Unit 1 Scenario No.:

2 Op Test No.:

BV1LOT14 NRC Examiners:

Candidates:

SRO ATC BOP Initial IC 67(17): 65% power, MOL, Equ. XE Conditions, CB "D" @ 177 steps, RCS boron -

Conditions:

1102 ppm.

Turnover:

Maintain current power level.

1FW-P-3B OOS for maintenance, with lFW-P-2 aligned to "B" header lAW TS 3.7.5, Condition B.

Control rods are in manual due to I&C MSP in progress.

Critical Tasks:

1. CT -4 (E-O.F) Establish AFW flow
2. CT -18 (E-3.A) Isolate ruptured SG
3. CT -20 (E-3.C) Crew depressurizes RCS
4. CT-21 (E-3.D) Crew terminates SI Event Malf. No.

Event Type Event Description No.

1 (C) ATC, SRO lA charging pump trip, requires manual starting of lB.

(TS) SRO 2

(C) ATC, SRO 200 gpd "B" steam generator tube leak (TS) SRO 3

(R) ATC Main feed pump, lFW-P-lB trips, requires rapid power (N) BOP, SRO reduction (AOP 1.24.1) 4 (M)ALL "B" Main feed regulating valve, FCV-lFW-488 fails shut, leads to a Rx trip with a loss of offsite power.

5 Turbine driven aux feed pump, lFW-P-2 trips on overspeed (C) BOP, SRO and motor driven aux feed pump, 1FW-P-3A fails to auto start, requiring manual start of 1FW-P-3A.

6 (M)ALL 500 gpm SGTR in the "B" steam generator.

7 Main steam line isolation valve on the ruptured SG, TV-lMS-(C) BOP, SRO 101 B failed open, requires the crew to perform an alternate MSLI alignment.

8 (C) ATC, SRO PRZR PORV fails open during depressurization requires closing ofblock valve.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Revision 9 Supplement 1 FENOC Facsimile Rev. 1

Appendix D Scenario Outline 1L14N2

  • fter taking the shift at 65% power, the running charging pum.p will trip. The SRO will enter AOP 1.7.1 to

,drt the standby pump. The standby pump must be started and the backup pump will be aligned to replace the failed pump. Charging and letdown will be isolated per the AOP. The SRO will address Technical Specifications for the failed charging pump.

Subsequently, a 200 gpd SG tube leak will develop on the "B'" SG. AOP 1.6.4 will be implemented. AOP 1.6.4 will provide direction to enter Mode 3 within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The SRO will address Technical Specifications which also will require Mode 3 entry.

1FW-P-1B will trip, requiring the crew to rapidly reduce power to< 52% lAW AOP 1.24.1.

The "B" SG main feedwater regulating valve fails closed, resulting in a reactor trip.

Following the reactor trip, offsite power is lost.

Additionally, the 3A motor driven aux feed pump fails to automatically start but may be started manually, The turbine driven aux feed pump will start but trips during startup and can be recovered if requested.

As a result of the reactor trip, a 500 gpm steam generator tube rupture will occur in the "B" SG.

When isolating the "B" SG, the "B" main steam line isolation valve will fail open requiring the crew to close the "A" and "C" MSIV's.

.tbsequently, the PRZR PORV used to depressurize the RCS will fail open and require the motor operated block valve be closed to stop the PORV leakage.

The scenario will be terminated when the crew has terminated safety injection.

Expected procedure flow path is EO ~ E3.

NUREG-1 021, Revision 9 Supplement 1 FENOC Facsimile Rev. 1

A d' D

~_pen 1x s

  • o r cenano utme 1L14N3 Facility:

BVPS Unit 1 Scenario No.:

3 Op Test No.:

BV1LOT14 NRC Examiners:

Candidates:

SRO ATC BOP Initial IC 211: 5% power, BOL, Equ. XE Conditions, CB "D"@ 114 steps, RCS boron-1851 Conditions:

ppm.

Turnover:

Raise Power to 10-14% to place turbine online.

Critical Tasks:

1. CT -12 (E-O.P), Manually actuate SLI
2. CT -17 (E-2.A), Isolate faulted SG
3. CT -48 (FR-P.l.A), Terminate ECCS flow I challenge to the integrity CSF Event Malf. No.

Event Type Event Description No.

1 (R) ATC Raise Power to 1 0-14%

(N) SRO 2

(C) BOP, SRO Leak collection exhaust fan, 1VS-F-4A trips, requiring manual (TS) SRO start of 1 VS.. f-4B 3

(C)ATC, SRO Charging flow control valve, FCV-1CH-122 fails shut in AUTO, requires manual control ofPRZR level.

4 (I) BOP, SRO Main steam header pressure transmitter, PT-1MS-464 fails low, requiring manual control of condenser steam dumps.

(C) BOP, SRO SG atmospheric reliefvalve, PCV-1MS-101B fails open, 5

(TS) SRO requires manual control of condenser steam dumps during failure and local operator isolation.

6 (M) ALL Steam Break inside cnmt on "B" SG 7

(I) BOP, SRO SLI auto actuation failure, requires manual MSLI.

8 (C) BOP, SRO Aux feedwater control valve failed open, requires alternate isolation method.

9 (I) BOP, SRO Control room emergency ventilation system fails to actuate on CIB signal, requiring manual initiation.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1 021, Revision 9 Supplement 1 FENOC Facsimile Rev 1

Appendix D Scenario Outline 1L14N3

~1.e crew will assume the shift at 5% power with instructions to raise power to 10-14% to place the turbine

.1line lAW the reactivity plan and 1 OM-52.4.A. The ATC will initiate a dilution and withdraw control rods.

After the power has raised to 6.4%, 1 VS-F-4A will trip, the crew will respond using the ARP which will direct the BOP to manually start 1 VS-F-4B. The SRO will address applicable TS.

After the power has raised to 8.0%, FCV -1 CH-122 will fail closed in Auto, the ATC will be required to identify the failure and manually control FCV-1CH-122 to maintain Pzr level.

2 minutes after the crew blocks the Power Range Low Overpower Reactor trip, the main steam header pressure transmitter, MS-1PT-464 will fail low causing the condenser steam dumps to close in response. The ensuing RCS heatup will cause SG atmospheric steam relief valve, PCV-1MS-101B to fail open, causing Rx power to rise and Tavg to drop due to the increased steam flow. The BOP will restore Rx power and Tavg by manually controlling the condenser steam dumps. The crew will unsuc,cessfully attempt to close the valve from the control room and then dispatch an operator to locally isolate. The BOP will control the condenser steam dumps while a local operator isolates the failed open SG atmospheric: valve.

A large steam break will then occur on the "B" SG inside of CNMT.

The crew will enter E-0 and transition to E-2 after diagnosing a faulted SG. The steam break is of a magnitude that will cause conditions for an Integrity Red Path and require the crew to transition to FR-P.l.

"' additional malfunctions occur during the event;

1. Automatic SLI fails to automatically actuate
2. "A" header AFW control valve will fail to close from the control room.
3. The control room emergency ventilation system, CREV's, fails to automatically actuate.

The scenario will be terminated when the crew has established normal charging flow in FR-P.1.

Expected procedure flow path is E E FR-P.l.

NUREG-1021, Revision 9 Supplement 1 FENOC Facsimile Rev 1

A d"

D

.ppen IX s cenano Outline 1L14N4 Facility:

BVPS Unit 1 Scenario No.:

4 Op Test No.:

BV1LOT14 NRC Examiners:

Candidates:

SRO ATC BOP Initial IC 62(18): 81% power, MOL, Equilibrium Xe, CB "D"@ 206 steps, RCS boron-1012 Conditions:

ppm.

Turnover:

Rx is at 81% power due to isolating the condenser "D" Waterbox for cleaning.

"A" Train Priority, 1FW-P-3B is OOS with 1FW-P-2 aligned to the "B" header lAW TS 3.7.5, Condition B.

Critical Tasks:

1. CT -1 (E-O.A) - Manually Trip Rc~actor
2. CT -6 (E-0.1) - Establish 1 train of HHSI
3. CT-16 (E-l.C)- Manually trip RCPs Event Malf. No.

Event Type Event Description No.

1 (C)ATC, SRO The 1A bori1;; acid transfer pump trips requiring aligning the (TS) SRO standby train for service.

2 (TS) SRO Ch 2, cnmt pressure transmitter, PT-1LM-100B, fails high.

3 (R) ATC Cooling tower pump, CT -P-1 C trips, causing degraded (N) BOP, SRO condenser vacuum, requiring power reduction.

4 (I) ATC, SRO Auto rod insertion fails on load reduction, A TC manually controls Tavg.

5 (M)ALL Condenser Low-Low vacuum causes turbine trip.

6 (I)ATC, SRO Auto Rx trip fails to occur, requires ATC to manually trip the reactor from BB-A.

7 Exciter circuit breaker fails to auto open on Rx trip, requires (C) BOP, SRO BOP to manually trip the exciter circuit breaker.

8 (M)ALL 1500 GPM LOCA one minute after Rx trip.

9 "A" charging pump, 1CH-P-1A trips, and "B" charging pump, (C) ATC, SRO 1CH-P-1B f::tils to auto start on Safety Injection, requiring manual start.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Revision 9 Supplement 1 FENOC Facsimile Rev. 1

Appendix D Scenario Outline 1L14N4

~er taking the shift, boric acid transfer pump, 1 CH-P-2A will trip. SRO will direct placing the 2B boric acid

.~dllsfer pump in service IA W 1 OM-7.4.R, "Transfer of Inservice Boric Acid Tank". SRO will also address Technical Specifications.

After the crew has reestablished a boric acid flowpath, containment pressure ch 2 transmitter, PT-1LM-100B fails high. The SRO will enter 10M-1.4.IF and review the Technical Specifications. The SRO will then contact I&C to trip the applicable bistables.

The "C" cooling tower pump will trip causing condenser vacuum to degrade. The crew will enter AOP 1.26.2, "Loss of Condenser Vacuum". While working through the Loss of Condenser Vacuum AOP, they will recognize the need to reduce turbine load to prevent exceeding hotwell temperature of 130 °F. At this point the SRO will enter AOP - 1.51.1, "Unplanned Power Reduction" and reduce turbine load.

The control rods will fail to auto insert during the turbine load reduction. The A TC will have to insert control rods in manual to maintain Tavg to Tref.

During the load reduction the turbine will trip due to low condenser vacuum, the plant will be greater than 49%

power (P-9), but the Rx will fail to auto trip. The ATC will successfully trip the reactor from Bench board "A".

During the immediate operator actions ofE-0, the BOP will be required to manually open the exciter circuit breaker due to an auto open failure.

"ne minute after the Rx is manually tripped a 1500 GPM LOCA will occur on the "B" RCS loop. The LOCA

.11 cause a Safety Injection actuation due to low RCS pressure. Upon the SI actuation the "A" charging pump will trip, and the "B" charging pump will fail to auto start, requiring the ATC to manually start the "B" charging pump to restore high head safety injection flow.

The crew will enter E-0, transition to E-1 due to containment parameters, and then to ES-1.2 to cooldown the RCS.

The scenario will be terminated when RCS cooldown is commenced in ES-1.2.

Expected procedure flow path is E E ES-1.2.

NUREG-1021, Revision 9 Supplement 1 FENOC Facsimile Rev. 1

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