ML17003A284

Draft - Outlines (Folder 2), (Facility Letter Dtd. - 7/26/2016)

ML17003A284
Person / Time
Site:	Beaver Valley
Issue date:	07/26/2016
From:	FirstEnergy Nuclear Operating Co
To:	FirstEnergy Nuclear Operating Co, NRC Region 1
Shared Package
ML16076A416	List: IR 05000334/2016302 ML16089A007 ML16292A483 ML16342C521 ML17003A276 ML17003A277 ML17003A278 ML17003A280 ML17003A281 ML17003A282 ML17003A283 ML17003A284 ML17003A285
References
U01922
Download: ML17003A284 (32)

Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: BVPS UNIT 1 RO Date of Exam 10/31 thru 11118/2016 RO KIA Category Points SRO, ONLY Points Tier Group K K K K K K A A A A G TOTAL TOTAL 1 2 3 4 5 6 1 2 3 4

• 1 3 3 3 3 18 1.

Emergency 2

& 1 2 1 1 9 Abnormal Plant Tier Evolutions Totals 4 5 4 4 27 1

2.

3 2 3 3 2 2 2 3 2 3 3 28 Plant Systems 2 1 0 1 1 1 1 1 1 1 1 1 10 Tier Totals 4 2 4 4 3 3 3 4 3 4 4 38 1 2 3 4 10

3. Generic Knowledge and Abilities Category 2 3 3 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).

(One Tier 3 Radiation Control KIA is allowed if the KIA is replaced by a KIA from another Tier 3 Category).

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-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 (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 a category 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.

G* Generic KIAs NUREG-1021, Revision 10 RO Page 1of13 FENOC Facsimile Rev. 0

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

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

l L 1 2 3 1 2 .

000007 Reactor Trip I 1 )( *,  :,* *. EA 1 Ability to operate and monitor the following as 4.3 1

,; ; . they apply to a reactor trip:

[Question 1]

1,.,.,-" EA 1.07 MT/G trip; verification that the MT/G has been tripped (CFR 41. 7 I 45.5 I 45.6) 000008 Pressurizer Vapor Space Accident I . ' )( 2.1.20 Ability to interpret and execute procedure 4.6 1 3 steps.

• .*1:

.. (CFR: 41.10 I 43.5 I 45.12)

[Question 2] .*

I*. "" .**

L_ ,',

000009 Small Break LOCA I 3 J~

.. 1.-~<' EA2 Ability to determine or interpret the following as they apply to a small break LOCA

3.8 1

• ;* le**.* EA2.11 - Containment temperature, pressure, and

[Question 3]

. humidity

.. ,.l

.... (CFR 43.5 I 45.13) 000011 Large Break LOCA I 3 )(I EA2 Ability to determine or interpret the following as 3.2? 1 I* they apply to a Large Break LOCA:

[Question 4] *,I.< . EA2.07 - That equipment necessary for functioning I of critical pump water seals is operable J,)j' 1'.1** 1~~ (CFR 43.5 I 45.13) l~f: 1,~*i.;

1015/000017 RCP Malfunctions I 4 )( AK3. Knowledge of the reasons for the following 3.7 1

responses as they apply to the Reactor Coolant 11 Pump Malfunctions (Loss of RC Flow) :

[Question 5]

IC AK3.03 - Sequence of events for manually tripping reactor and RCP as a result of an RCP malfunction 1~; * . ,*.

l:.'L "' '

,.*,.* 1:: (CFR 41.5,41.10 I 45.6 I 45.13) 000022 Loss of Rx Coolant Makeup I 2 h*.'

)( AK1. Knowledge of the operational implications of 2.7 1 the following concepts as they apply to Loss of Iv*;; h:'* Reactor Coolant Makeup:

[Question 6]

It..*; l;c,., AK1 .02 - Relationship of charging flow to pressure differential between charging and RCS 1* '; .*

I**

(CFR 41.8 I 41.10 I 45.3) 000025 Loss of RHR System I 4 )( AK2. Knowledge of the interrelations between the 2.7 1 Loss of Residual Heat Removal System and the following:

[Question 7]

AK2.03 - Service water or closed cooling water pumps (CFR 41.7 I 45.7) 000027 Pressurizer Pressure Control )( *, ,, AK2. Knowledge of the interrelations between the 2.6 1 System Malfunction I 3 Pressurizer Pressure Control Malfunctions and the i:~ '

[Question 8]

',, " ii '.

L ..

following:

AK2.03 - Controllers and positioners (CFR 41.7 I 45.7)

NUREG-1021, Revision 10 RO Page 2of13 FENOC Facsimile Rev. O

ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions -Tier 1/Group 1(RO) Continued E/APE #I Name I Safety Function K 1

K K A A (3

.

KIA Topic(s) IR #

2 3 1 '2 000029 A TWS I 1 'j' 1: . EA2 Ability to determine or interpret the following as 3.4* 1

' they apply to a ATWS:

[Question 9] *'. EA2.05 - System component valve position I., indications (CFR 43.5 I 45.13) 000038 Steam Gen. Tube Rupture I 3 )( 2.4.34 - Knowledge of RO tasks performed outside 4.2 1 the main control room during an emergency and the resultant operational effects.

[Question 1OJ (CFR: 41.10 I 43.5 I 45.13) 000056 Loss of Off-site Power I 6 )  ;'

1:.:. AK3. Knowledge of the reasons for the following 4.4 1 responses as they apply to the Loss of Offsite

'*'*.,;

Power:

• • • ., AK3.02 - Actions contained in EOP for loss of

[Question 11]

f: offsite power

.;. [;.

l.t:\.;

~.*:, 1-. ..1 (CFR 41.5,41.10 I 45.6 I 45.13) 000058 Loss of DC Power I 6 )( /;;

l§i,. AK1. Knowledge of the operational implications of the following concepts as they apply to Loss of DC 2.8 1

[Question 12]

h' l;l~ Power:

I:,:.

I*':: AK1 .01 - Battery charger equipment and

[','j

,. 16~'} instrumentation

.I*.* *

(CFR 41.8 I 41.10 I 45.3) 1 If~'

000062 Loss of Nuclear Svc Water I 4 )( AA 1. Ability to operate and I or monitor the 2.9 1 following as they apply to the Loss of Nuclear r

Service Water (SWS):

Question 13] I ..* ;,

, j;* .* AA 1.07 - Flow rates to the components and h' systems that are serviced by the SWS; interactions among the components

..,*I> .*. (CFR 41. 7 I 45.5 I 45.6)

<<**> [;.,... .,

000065 Loss of Instrument Air I 8 x ,,., ... I:> , AA 1. Ability to operate and I or monitor the 2.9 1 i

• following as they apply to the Loss of Instrument Air:

[Question 14]

' AA 1.03 - Restoration of systems served by instrument air when pressure is regained (CFR 41.7 I 45.5 I 45.6)

W/E04 LOCA Outside Containment I 3 )( 'I* EK3. Knowledge of the reasons for the following 3.4 1

. responses as they apply to the (LOCA Outside

• ' Ii.,...' Containment)

,; .

[Question 15] '

1:*

1. . .

I*

i'*' ' EK3.2 - Normal, abnormal and emergency

• • • i., operating procedures associated with (LOCA

,...... ' ,.,. {

Outside Containment).

I.:., (CFR: 41.5 I 41.10, 45.6, 45.13)

NUREG-1021, Revision 10 RO Page 3of13 FENOC Facsimile Rev. O

ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1(RO) Continued E/APE #I Name I Safety Function K K K A A G KIA Topic(s) IR #

1 2 3 1 2 W/E05 Inadequate Heat Transfer - Loss of ~

',' 1:

EK1. Knowledge of the operational implications of 3,9 1 Secondary Heat Sink I 4 the following concepts as they apply to the (Loss of Secondary Heat Sink)

[Question 16] EK1 ,3 - Annunciators and conditions indicating

"' signals, and remedial actions associated with the I

Loss of Secondary Heat Sink (CFR: 41,8 / 41.10, 45.3)

W/E11 Loss of Emergency Coolant Recirc, I )(

EK2. Knowledge of the interrelations between the 3,9 1 4 (Loss of Emergency Coolant Recirculation) and the

[Question 17] following:

EK22 - Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility

,, (CFR: 41,7 / 45,7)

W/E12 - Uncontrolled Depressurization of I .) 2.4.2 - Knowledge of system set points, interlocks 4.5 1 all Steam Generators I 4 I >,

!',,;'. and automatic actions associated with EOP entry I'

[Question 18]

';*' conditions.

I*

I*, I~" (CFR: 41,7 / 45.7 / 45.8) 3 3 3 3 3 3 Category Point Totals: Group Point Total: 18 NUREG-1021, Revision 10 RO Page 4of13 FENOC Facsimile Rev, O

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

I*

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

1 2 3 1 : 2.

l I;* .. I*I**

000036 Fuel Handling Accident I 8 )( . I*

AA 1. Ability to operate and I or monitor the following 3.5 1

• . ' as they apply to the Fuel Handling Incidents:

(Question 19]

I*

AA 1.03 - Reactor building containment evacuation alarm enable switch (CFR 41. 7 I 45.5 I 45.6) 000061 ARM System Alarms 17 )( AA2. Ability to determine and interpret the following 2.9 1 as they apply to the Area Radiation Monitoring

[Question 20] . (ARM) System Alarms:

t: AA2.02 - Normal radiation intensity for each ARM system channel (CFR: 43.5 I 45.13)

• .': '< AA 1. Ability to operate and I or monitor the following 000068 Control Room Evac. I 8 )( . 4.0 1 as they apply to the Control Room Evacuation:

[Question 21] *...

AA 1.22 - Flow control valve for RCS charging header

,f 1:,1 .. *'.5i (CFR 41.7 / 45.5 / 45.6)

i~

000076 High Reactor Coolant Activity I 9 )( I';~ AK3. Knowledge of the reasons for the following 3.2 1 IU~t~

responses as they apply to the High Reactor Coolant

[Question 22]

Activity:

AK3.06 - Actions contained in EOP for high reactor 1n:;; "* ;i:'. coolant activity

., I*. (CFR 41.5,41.10 I 45.6 I 45.13)

W/E02 SI Termination I 3  : )( 2.4.20 - Knowledge of operational implications of 3.8 1 I**'** . EOP warnings, cautions, and notes.

(CFR: 41.10 / 43.5 / 45.13)

[Question 23]

W/E07 Saturated Core Cooling I 4 )( *:. I'* EK2. Knowledge of the interrelations between the 3.5 1 I'  : (Saturated Core Cooling) and the following:

(Question 24] c) 1* EK2.2 - Facility's heat removal systems, including

. primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper

..: operation of these systems to the operation of the facility (CFR: 41.7 /45.7)

W/E13 Steam Generator Over-pressure I 4 )( EK2. Knowledge of the interrelations between the 3.0 1 (Steam Generator Overpressure) and the following:

[Question 25] '

EK2.1 - Components, and functions of control and safety systems, including instrumentation, signals,

. interlocks, failure modes, and automatic and manual features (CFR: 41.7 / 45.7)

NUREG-1021, Revision 10 RO Page 5of13 FENOC Facsimile Rev. 0

ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions -Tier 1/Group 2(RO) Continued E/APE #I Name I Safety Function K K K A>A :G KIA Topic(s) IR #

1 2 3 1

• 2 ,,<,

W/E14 Loss of CTMT Integrity I 5 )(  ; !,/, EK1, Knowledge of the operational implications of 3,3 1

.':::,.* the following concepts as they apply to the (High

[Question 26] *'

. ,i' Containment Pressure)

. EK1 .1 - Components, capacity, and function of emergency systems I ,.

(CFR: 41.8 I 41.10, 45.3)

W/E16 High Containment Radiation I 9 ')( ' EA2. Ability to determine and interpret the following 3.0 1

., . as they apply to the (High Containment Radiation)

[Question 27] i.'

EA2.2 - Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments

,

I*"* (CFR: 43.5 I 45.13) 1 2 1 2 2 1 KIA Category Point Totals: Group Point Total: 9 NUREG-1021, Revision 10 RO Page 6of13 FENOC Facsimile Rev. 0

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

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

1 2 3 4 5 6 1 2 3 4 L

003 Reactor Coolant Pump x K2. Knowledge of bus power supplies to the following:

2.5* 1

[Question 28] 1.*

K2.02 - CCW pumps (CFR: 41.7) 003 Reactor Coolant Pump )( A4 Ability to manually operate and/or monitor 3.1 1 in the control room:

[Question 29]

I*._, A4.05 - RCP seal leakage detection I**'

instrumentation I

r I*'* . :*

(CFR: 41. 7 / 45.5 to 45.8) 004 Chemical and Volume )( K3 Knowledge of the effect that a loss or 3.6 1 Control ,, malfunction of the eves will have on the following:

[Question 30] 1:' I*

I<.'

,, K3.08 - RCP seal injection 1/***

}:,. '

(CFR: 41. 7/45/6) l..~f:

005 Residual Heat Removal )( I; . K6 Knowledge of the effect of a loss or 2.5 1 malfunction on the following will have on the

[Question 31] 1./f RHRS:

I ~' K6.03 - RHR heat exchanger 1* , ,.

1***; (CFR: 41.7 / 45. 7) 006 Emergency Core Cooling )( K4 Knowledge of ECCS design feature(s) 3.9 1 and/or interlock(s) which provide for the

[Question 32]

following:

K4.14 - Cross-connection of HPl/LPl/SIS 1~

' *.. 1*,:" (CFR: 41.7)

Ii 1:* K1 Knowledge of the physical connections 007 Pressurizer Relief/Quench Tank

)( I' I*> and/or cause effect relationships between the 2.9 1 1-.:<'

PRTS and the following systems:

[Question 33] I{'

1*

'* 1....

1.: K1 .01 - Containment system

,. (CFR: 41.2 to 41.9 / 45. 7 to 45.8) 007 Pressurizer Relief/Quench )( A3 Ability to monitor automatic operation of the 2.7* 1 1.-

Tank PRTS, including:

[Question 34] A3.01 - Components which discharge to the PRT (CFR: 41.7 / 45.5) 008 Component Cooling Water I )( '*

A2 Ability to (a) predict the impacts of the 3.0 1 I .*.

following malfunctions or operations on the

[Question 35]

CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or I* operations:

A2.03 - High/low CCW temperature

• 11 (CFR: 41.5 / 43.5 / 45.3145.13)

NUREG-1021, Revision 10 RO Page 7of13 FENOC Facsimile Rev. O

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1(RO) Continued 1* System #I Name K K K K K K A A A A G KIA Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 010 Pressurizer Pressure )( *. K5 Knowledge of the operational implications 3.5 1 Control of the following concepts as the apply to the PZR PCS:

[Question 36)

K5.01 - Determination of condition of fluid in PZR, using steam tables (CFR: 41.5 / 45. 7) 010 Pressurizer Pressure )( 2.4.31 - Knowledge of annunciator alarms, 4.2 1 Control indications, or response procedures.

[Question 37) . (CFR:41.10/45.3) 012 Reactor Protection }( *'  :::. *;

K1 Knowledge of the physical connections 3.4 1 and/or cause effect relationships between the

[Question 38] i.*

RPS and the following systems:

K1 .02 - 125V de system (CFR: 41.2 to 41.9 / 45.7 to 45.8) 013 Engineered Safety Features }( A3 Ability to monitor automatic operation of the 3.7* 1 Actuation ESFAS including:

1*

[Question 39] A3.01 - Input channels and logic

... * ... (CFR: 41. 7 1. 45.5) 022 Containment Cooling

• ..._, 1estion 40)

IJ A2 Ability to (a) predict the impacts of the following malfunctions or operations on the 2.5 1

• • ..... CCS; and (b) based on those predictions, use I**.** I>> procedures to correct, control, or mitigate the

>1 I*;;*< consequences of those malfunctions or I~ .. operations:

A2.01 - Fan motor over-current I*.," I;'*

I* (CFR: 41.5 / 43.5 / 45.3 / 45.13)

, .... ,._;,.. K4 Knowledge of CSS design feature(s) and/or 026 Containment Spray )

I

. .* interlock(s) which provide for the following:

3.7* 1

[Question 41]

• • • • • K4.09 - Prevention of path for escape of 1* **; radioactivity from containment to the outside

.'  ;

(interlock on RWST isolation after swapover)

. (CFR: 41.7) 039 Main and Reheat Steam )( 2.2.38 - Knowledge of conditions and 3.6 1 limitations in the facility license.

[Question 42)

(CFR: 41.7 / 41.10 / 43.1/45.13) 039 Main and Reheat Steam }( K3 Knowledge of the effect that a loss or 2.8* 1 malfunction of the MRSS will have on the

[Question 43) following:

K3.06-SDS I (CFR: 41. 7 I 45.6) 059 Main Feedwater }( ..

A 1 Ability to predict and/or monitor changes in 2.7* 1

[Question 44)

• " parameters (to prevent exceeding design limits) associated with operating the MFW controls including:

A 1.03 - Power level restrictions for operation of MFW pumps and valves.

(CFR: 41.5 I 45.5)

NUREG-1021, Revision 10 RO Page 8of13 FENOC Facsimile Rev. 0

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1(RO) Continued System # I Name K K K K K K A A A A G KIA Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 061 Auxiliar-YtEmergency )( K6 Knowledge of the effect of a loss or 2.5 1 Feedwater malfunction of the following will have on the AFW components:

[Question 45]

K6.01 - Controllers and positioners (CFR: 41.7 / 45.7) 062 AC Electrical Distribution ) I: K4 Knowledge of ac distribution system design 2.8* 1 feature(s) and/or interlock(s) which provide for

[Question 46] the following:

I:*.

I K4.03 - Interlocks between automatic bus transfer and breakers (CFR: 41.7) 063 DC Electrical Distribution x A4 Ability to manually operate and/or monitor in the control room:

2.8* 1

[Question 47]

A4.01 - Major breakers and control power fuses

'*  :  ;< (CFR: 41.7 / 45.5 to 45.8) l'j~ ~;'.~

063 DC Electrical Distribution A2 Ability to (a) predict the impacts of the 2.5 1 following malfunctions or operations on the DC

[Question 48] electrical systems; and (b) based on those

y

predictions, use procedures to correct, control,

' 1: ...

.. or mitigate the consequences of those malfunctions or operations:

A2.01 - Grounds (CFR: 41.5 / 43.5 / 45.3 / 45.13) 064 Emergency Diesel Generator x .. '*'

1*

K2 Knowledge of bus power supplies to the following:

2.7* 1

[Question 49] '  :. 1:*: K2.01 - Air compressor Ir.;. (CFR: 41.7) 073 Process Radiation x *.

. In K5 Knowledge of the operational implications 2.5 1 Monitoring as they apply to concepts as they apply to the PRM system:

[Question 50]

K5.01 - Radiation theory, including sources, types, units, and effects (CFR: 41.5 / 45. 7) 076 Service Water )( A 1 Ability to predict and/or monitor changes in 2.6* 1 parameters (to prevent exceeding design

[Question 51] limits) associated with operating the SWS controls including:

I*

A 1.02 - Reactor and turbine building closed

... cooling water temperatures (CFR: 41.5 / 45.5)

NUREG-1021, Revision 10 RO Page 9of13 FENOC Facsimile Rev. 0

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems -Tier 2/Group 1(RO) Continued System #I Name K K K K K K A A A A G KIA Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 ,,

076 Service Water )( 2.1.27 - Knowledge of system purpose and/or 3.9 1 function.

[Question 52]

1(

(CFR: 41.7) 078 Instrument Air )( A4 Ability to manually operate and/or monitor 3.1 1 in the control room:

[Question 53]

!,,, A4.01 - Pressure gauges (CFR: 41.7 / 45.5 to 45.8) x 078 Instrument Air K3 Knowledge of the effect that a loss or 3.4 1 malfunction of the IAS will have on the

[Question 54] following:

K3.02 - Systems having pneumatic valves and controls

(CFR: 41. 7 I 45.6) 103 Containment )( *',

K1 Knowledge of the physical connections 3.6 1

and/or cause-effect relationships between the

[Question 55] f,,, containment system and the following

X< t\ ", systems:

I ~1~'. K1.01 -CCS IL' *<**

/;i(;~ (CFR: 41.2 to 41.9 / 45.7 to 45.8) 3 2 3 3 2 2 2 3 2 3 3 KIA Category Point Totals: Group Point Total: 28 NUREG-1021, Revision 10 RO Page 10 of 13 FENOC Facsimile Rev. 0

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

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

IR #

1 2 3 4 5 6 1 2 3 4 f

011 Pressurizer Level Control )( A2 Ability to (a) predict the impacts of the 3.4 1 following malfunctions or operations on the PZR

[Question 56]

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

A2.10 - Failure of PZR level instrument - high (CFR: 41.5 / 43.5 / 45.3 / 45.13) 014 Rod Position Indication )( A4 Ability to manually operate and/or monitor in 3.3 1 the control room:

[Question 57]

A4.01 - Rod selection control (CFR: 41.7 / 45.5 to 45.8) 016 Non-nuclear )( I*

A3 Ability to monitor automatic operation of the 2.9* 1 Instrumentation NNIS, including:

A3.02 - Relationship between meter readings

[Question 58]

.\

.. and actual parameter value (CFR: 41. 7 I 45.5)

' ~*

017 In-core Temperature )( *: ... \I:;: K6 Knowledge of the effect of a loss or 2.7 1 Monitor .,... *** malfunction of the following ITM system

"'*', components:

[Question 59]

• .

K6.01 - Sensors and Detectors (CFR: 41.7 / 45.7) '

+ ******

--.7 Containment Iodine )( *, " K1 Knowledge of the physical connections 3.4* 1

' :ioval and/or cause effect relationships between the CIRS and the following systems:

[Question 60]

. K1.01 - CSS

" '"':;,:. (CFR: 41.2 to 41.9 I 45. 7 to 45.8) 029 Containment Purge )(  ; I**;. K3 Knowledge of the effect that a loss or 2.9* 1

'>,. *,, I*:'* malfunction of the Containment Purge System

[Question 61]

.. I*

will have on the following:

1 **

,\ K3.02 - Containment entry

. (CFR: 41. 7 / 45.6)

':;,;,.,.

055 Condenser Air Removal 2.4.34 - Knowledge of RO tasks performed 4.2 1

~*

outside the main control room during an II

[Question 62]

emergency and the resultant operational effects.

1)1,~t.* (\ (CFR: 41.10 I 43.5 I 45.13) 071 Waste Gas Disposal k .*

)( K4 Knowledge of design feature(s) and/or 2.5* 1 I interlock(s) which provide for the following:

[Question 63]

K4.03 - Tank loop seals (CFR: 41.7)

NUREG-1021, Revision 10 RO Page 11 of 13 FENOC Facsimile Rev. 0

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2(RO) Continued System #I Name K K K K K K A A A A G KIA Topic(s) 1 2 3 4 5 6 1 t 3 4 IR 072 Area Radiation Monitoring x .

.*

K5 Knowledge of the operational implications of the following concepts as they apply to the 2.5 1

[Question 64] ARM system:

K5.02 - Radiation intensity changes with

• .... source distance I

.* (CFR: 41.5 I 45. 7) 086 Fire Protection )( A 1 Ability to predict and/or monitor changes in 2.9 1 parameters (to prevent exceeding design

[Question 65]

limits) associated with Fire Protection System operating the controls including:

A 1.01 - Fire header pressure (CFR: 41.5 I 45.5)

KIA Category Point Totals: 1 0 1 1 1 1 1 1 1 1 1 Group Point Total: 10 NUREG-1021, Revision 10 RO Page 12of13 FENOC Facsimile Rev. 0

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

.:ility: BVPS UNIT 1 RO Date of Exam 10/31 thru 11/18/2016 RO Category KJA# Topic IR #

1. 2.1.19 Ability to use plant computers to evaluate system or component status. 3.9 Conduct (CFR: 41.10 / 45.12) of Operations Question 66 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, 3.9 etc.

(CFR: 41.10 / 45.12)

Question 67 Subtotal

2. 2.2.25 Knowledge of the bases in Technical Specifications for limiting Equipment conditions for operations and safety limits.

Control (CFR: 41.5 / 41.7 / 43.2)

Question 68 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded 3.1 power sources, on the status of limiting conditions for operations.

(CFR: 41.10 / 43.2 / 45.13)

Question 69 2.2.43 Knowledge of the process used to track inoperable alarms. 3.0 (CFR: 41.10 / 43.5 / 45.13)

Question 70 Subtotal

3. 2.3.7 Ability to comply with radiation work permit requirements during normal*

Radiation or abnormal conditions.

Control (CFR: 41.12 / 45.10)

Question 71 2.3.11 Ability to control radiation releases. 3.8 (CFR: 41.11 I 43.4 I 45.10)

Question 72 2.3.12 Knowledge of radiological safety principles pertaining to licensed 3.2 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 73 Subtotal

4. 2.4.6 Knowledge of EOP mitigation strategies.

Emergency (CFR: 41.10 / 43.5 / 45.13)

Procedures/

Question 74 Plan 2.4.50 Ability to verify system alarm setpoints and operate controls identified in 4.2 the alarm response manual.

(CFR: 41.10 / 43.5 / 45.3)

Question 75 Subtotal Tier 3 Point Total NUREG-1021, Revision 10 RO Page 13of13 FENOC Facsimile Rev. O

ES-401 PWR Examination Outline Form ES-401-2 Facility: BVPS UNIT 1 SRO Date of Exam 10/31 thru 11118/2016 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

2.

3 2 5 Plant Systems 2 0 2 1 3 Tier 5 3 8 Totals 1 2 3 4 7

3. Generic Knowledge and Abilities Category 2 2 2 1 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).

(One Tier 3 Radiation Control KIA is allowed if the KIA is replaced by a KIA from another Tier 3 Category).

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points. *'

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KJA 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) KJAs 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 K/As.

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 a category 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.

G* Generic KIAs NUREG-1021, Revision 10 SRO Page 1 of 6 FENOC Facsimile Rev. 0

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

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

1 2 3 1 2 000026 Loss of Component Cooling Water I )' AA2. Ability to determine and interpret the following 2.9* 1 8 as they apply to the Loss of Component Cooling

[Question 76] Water:

. J:

I* ..

• AA2.04 - The normal values and upper limits for the temperatures of the components cooled by CCW

• ....*1.*;_; ~' (CFR: 43.5 I 45.13) 000040 Steam Line Rupture - Excessive I* ) 2.4.41 - Knowledge of the emergency action level 4.6 1 Heat Transfer I 4 I*

.. thresholds and classifications.

1**

[Question 77]

I*

(CFR: 41.10/43.5/45.11) 000054 Loss of Main Feedwater I 4

• AA2. Ability to determine and interpret the following 3.3* 1

~

Ii ..* i.;/* as they apply to the Loss of Main Feedwater (MFW):

l~l f~'

[Question 78]

AA2.08 - Steam flow-feed trend recorder (CFR: 43.5 I 45.13)

• • ~

000055 Station Blackout I 6 -,_ ',:: ~ 2.2.44 - Ability to interpret control room indications to 4.4 1

.. verify the status and operation of a system, and '

[Question 79]

..... '>** understand how operator actions and directives affect plant and system conditions.

.' (CFR: 41.5 I 43.5 I 45.12)

~*~ l*f~.

000057 Loss of Vital AC Inst. Bus I 6 AA2. Ability to determine and interpret the following 3.8* 1 as they apply to the Loss of Vital AC Instrument Bus:

jii~~

[Question 80]

AA2.02 - Core flood tank pressure and level indicators ki . (CFR: 43.5 I 45.13) 000077 Generator Voltage and Electric Grid  ;* 2.4.31 - Knowledge of annunciator alarms, 4.1 1 Disturbances I 6 I) indications, or response procedures.

[Question 81] I L*

(CFR: 41.10 I 45.3) 3 3 6 KIA Category Point Totals: Group Point Total:

NUREG-1021, Revision 10 SRO Page 2 of 6 FENOC Facsimile Rev. 0

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

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

b.._ 1 2 3 1 2 000060 Accidental Gaseous Radwaste Rel. *,*~ 2.4.50 - Ability to verify system alarm setpoints and 4.0 1 19 1,*"

• "j

t operate controls identified in the alarm response manual.

[Question 82]

\ .. (CFR: 41.10 / 43.5 / 45.3) 000074 lnad. Core Cooling/ 4 I*~? '/k *. EA2 Ability to determine or interpret the following as they apply to Inadequate Core Cooling:

4.6* 1

[Question 83]

EA2.08 - The effect of turbine bypass valve operation on RCS temperature and pressure (CFR 43.5 / 45.13)

W/E01 Rediagnosis / 3 :x EA2. Ability to determine and interpret the following as they apply to the (Reactor Trip or Safety Injection 4.0 1

[Question 84]

• .

• I* Rediagnosis) ij~II!

EA2.1 - Facility conditions and selection of

[~;.*; appropriate procedures during abnormal and emergency operations (CFR: 43.5145.13)

I~:~

l ',~ ..

W/E10 Natural Circ. / 4 2.1.20 - Ability to interpret and execute procedure 4.6 1 J ,: steps.

[Question 85]

l'>

1*** (CFR: 41.10 / 43.5 / 45.12) t\/A Category Point Totals: 2 2 Group Point Total: 4 NUREG-1021, Revision 10 SRO Page 3 of 6 FENOC Facsimile Rev. 0

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

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

1 2 3 4 5 6 1 2 3 4

=-

004 Chemical and Volume }( 2.4.8 - Knowledge of how abnormal operating 4.5 1 Control procedures are used in conjunction with EOPs.

I

[Question 86] (CFR: 41.10 I 43.5 I 45.13) 008 Component Cooling Water }( A2 Ability to (a) predict the impacts of the 3.5 1 following malfunctions or operations on the

[Question 87] I 11 CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

I A2.02 - High/low surge tank level I

(CFR: 41.5 I 43.5 I 45.3 I 45.13) 2.4.6 - Knowledge of EOP mitigation 059 Main Feedwater

..'). strategies .

4.7 1

[Question 88] ..,

(CFR: 41.10 I 43.5 I 45.13)

I *..

061 Auxiliary/Emergency }( A2 Ability to (a) predict the impacts of the 3.6* 1 Feedwater following malfunctions or operations on the

.. AFW; and (b) based on those predictions, use

[Question 89]

procedures to correct, control, or mitigate the consequences of those malfunctipns or operations:

.. A2.02 - Loss of air to steam supply valve

'i' ,!

1>

."*.' (CFR: 41.5 I 43.5 I 45.3 I 45.13) 103 Containment ')~ ,,~ *...* A2 Ability to (a) predict the impacts of the 2.6* 1 I"* I following malfunctions or operations on the

[Question 90] I containment system and (b) based on those

• . predictions, use procedures to correct, control, or mitigate the consequences of those I

malfunctions or operations A2.01 - Integrated leak rate test (CFR: 41.5 I 43.5 I 45.3 I 45.13) 3 2 KIA Category Point Totals: Group Point Total: 5 NUREG-1021, Revision 10 SRO Page 4 of 6 FENOC Facsimile Rev. 0

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

System #I Name K K K K K K A A A A G KIA Topic(s) 1 2 3 4 5 6 1 2 3 4 IR #

015 Nuclear Instrumentation }( 2.2.25 - Knowledge of the bases in Technical 4.2 1 Specifications for limiting conditions for

[Question 91] operations and safety limits .

• . (CFR: 41.5 I 41.7 I 43.2) 035 Steam Generator )( . A2 Ability to (a) predict the impacts of the 4.6 1 following malfunctions or operations on the GS;

[Question 92]

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

A2.06 - Small break LOCA (CFR: 41.5 I 43.5 I 45.3 I 45.5) 041 Steam Dump/Turbine }( A2 Ability to (a) predict the impacts of the 3.9 1 Bypass Control following malfunctions or operations on the

[Question 93]

"'* SOS; and (b) based on those predictions or I . .' mitigate the consequences of those malfunctions or operations:

1: ** A2.02 - Steam valve stuck open (CFR: 41.5 I 43.5 / 45.3 I 45.13)

KIA Category Point Totals: 2 1 Group Point Total: 3 NUREG-1021, Revision 10 SRO Page 5 of 6 FENOC Facsimile Rev. 0

ES 401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: BVPS UNIT 1 SRO Date of Exam 10/31thru11/18/2016 SRO Only Category KIA# Topic IR 1 -#-

1. 2.1.4 Knowledge of individual licensed operator responsibilities related to shift 3.8 Conduct staffing, such as medical requirements, "no-solo" operation, maintenance of Operations of active license status, 10CFR55, etc.

(CFR: 41.10 I 43.2)

[Question 94]

2.1.23 Ability to perform specific system and integrated plant procedures during 4.4 all modes of plant operation.

(CFR: 41.10 I 43.5 / 45.2 / 45.6)

[Question 95]

Subtotal 2

2. 2.2.35 Ability to determine Technical Specification Mode of Operation.

Equipment (CFR: 41.7 / 41.10 / 43.2 / 45.13)

Control

[Question 96]

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

(CFR: 41.7 / 43.5 / 45.12)

[Question 97]

Subtotal 2

3. 2.3.6 Ability to approve release permits.

Radiation (CFR: 41.13 I 43.4 / 45.10)

Control

[Question 98]

2.3.13 Knowledge of radiological safety procedures pertaining to licensed 3.8 operator duties, such as response to radiation monitor alarms..

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

(CFR: 41.12 / 43.4 / 45.9/ 45.10)

[Question 99]

Subtotal 2

4. 2.4.5 Knowledge of the organization of the operating procedures network for Emergency normal, abnormal, and emergency evolutions.

Procedures/

Plan (CFR: 41.10 / 43.5 / 45.13)

[Question 100]

Subtotal 1 Tier 3 Point Total 7 NUREG-1021, Revision 10 SRO Page 6 of 6 FENOC Facsimile Rev. 0

ES-401 Record of Rejected K/As Form ES-401-4 I Facility: BVPS Unit 1 Date of Exam 10/31thru11/18/2018 Operating Test No.: BV1LOT16 NRC I

Tier I Randomly Reason for Rejection Group Selected KIA 1/1 000011 Question #4; There is no interface between Safety Injection and EA2.02 Residual Heat Removal (RHRS) at Beaver Valley. Randomly selected 000011 EA2.07 as a replacement.

1/1 W/E04 Question #15; Unable to write a discriminatory question due to the limited scope of the LOCA Outside Containment procedure.

EK3.1 Randomly selected W/E EK3.2 as a replacement.

2/1 007 K5.02 Question #33; The PRT is not used to form a steam bubble in the PZR. Randomly selected 007 K1 .01 as a replacement. Outside K5 due to importance ratings were <2.5.

2/1 022 A2.06 Question #40; Beaver Valley does not have Containment Cooling System (CCS) pumps. Randomly selected 022 A2.01 as a replacement.

2/1 059 A1 .07 Question #44; Beaver Valley Main Feedwater pumps are motor driven pumps and do not have variable speed control. Randomly selected 059A1.03 as a replacement.

2/1 073 K5.02 Question #50; K5.02 - Radiation intensity changes with source distance. Reselected due to oversampling. KIA was similar to Q64.

Randomly selected 073 K5.01 as a replacement.

2/2 011 A2.08 Question #56; Pressurizer level is not compensated at Beaver Valley.

Randomly selected 011 A2.10 as a replacement.

2/2 027 K2.01 Question #60; Containment Iodine Removal fans are no longer used at Beaver Valley. Randomly selected 027 K1 .01 as a replacement.

2/2 086 A1.02 Question #65; Beaver Valley does not have a fire water storage tank which has design limits which are monitored by the control room staff. The tank is used for fire protection outside the protected area. Randomly selected 086A1.01 as a replacement.

1/1 000077 Question #81; Unable to write a discriminatory SRO level question for Knowledge of EOP entry conditions and immediate action steps.

SRO 2.4.1 Randomly selected 000077 G2.4.31 as a replacement.

1/2 W/E10 Question #85; Reselected due to overlap with Audit Exam.

Randomly selected W/E10 G2.1.20 as a replacement.

SRO 2.1.32 NUREG-1021, Revision 10 FENOC Facsimile Rev. 0

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Beaver Valley Unit 1 Date of Examination: 10/31 thru 11/18 2016

-xamination Level RO !RI SROD Operating Test Number BV1LOT16 NRC Administrative Type Describe activity to be performed Topic Code*

{See Note)

Conduct of D,R 2.1.25 (3.9)

Operations Ability to interpret reference materials, such as graphs, curves, tables, etc.

(RO A 1.1)

JPM 1AD-001 Calculating a Shutdown Margin Following a Stuck Rod Conduct of D,R 2.1.7 (4.4)

Operations Ability to evaluate plant performance and make operational judgments based (RO A 1.2) on operating characteristics, reactor behavior, and instrument interpretation.

JPM 1AD-003 Perform a Quadrant Power Tilt Ratio Calculation Equipment N, R 2.2.41 (3.5)

Control Ability to obtain and interpret station electrical and mechanical drawings.

~OA2)

JPM 1AD-040 Identify Isolation Boundary Points on Plant VOND Radiation Control D,R 2.3.7 (3.5)

(RO A 3) Ability to comply with radiation work permit requirements during normal or abnormal conditions.

JPM 1AD-012 Select RWP and Determine Maximum Allowable Stay Time Emergency Plan NOT EVALUATED (RO A 4)

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

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

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

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

NUREG-1021, Revision 10 FENOC Facsimile Rev. 0

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Beaver Valley Unit 1 Date of Examination: 10/31thru11/18 2016

• xamination Level RO D SRO IBJ Operating Test Number BV1LOT16 NRC Administrative Type Describe activity to be performed Topic Code*

(See Note)

Conduct of D, R 2.1.20 (4.6)

Operations Ability to interpret and execute procedure steps.

(SROA1.1)

JPM 1AD-029 Prepare Partial OST [1 OST-1.1] for Performance Conduct of D,R 2.1.7 (4.7)

Operations Ability to evaluate plant performance and make operational judgments based (SRO A 1.2) on operating characteristics, reactor behavior, and instrument interpretation.

JPM 1AD-009 Review a Quadrant Power Tilt Ratio Calculation Equipment N, R 2.2.41 (3.9)

Control '

Ability to obtain and interpret station electrical and mechanical drawings.

-~RO A 2)

JPM 1AD-039 Identify Isolation Boundary Points on Plant VOND, then determine diesel operability Radiation Control D,R 2.3.4 (3.7)

(SRO A 3) Knowledge of radiation exposure limits under normal or emergency conditions.

JPM 1AD-038 Determine Emergency Exposure Authorization Limits Emergency Plan D,R 2.4.44 (4.4)

(SRO A 4) Knowledge of emergency plan protective action recommendations.

JPM 1AD-037 Determine Protective Action Recommendations (Part 1)

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

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

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

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

NUREG-1021, Revision 10 FENOC Facsimile Rev. 0

ES 301 C on t ro IR oom /I n-Pl an t S,ys tems 0 utrme Form ES -301 -2 Facility: Beaver Valle~ Unit 1 Date of Examination: 10/31/ thru 11/18 2016 Exam Level: RO l:8:I SRO(I) D SRO(U) D Operating Test No.: BV1LOT16 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U)

System I JPM Title Type Code* Safety Function S 1 - Withdraw Shutdown Bank "A" ( 1CR-085) S, M, L 1 S2 - Perform Manual Makeup to the Charging Pump Suction (1CR-581) S,D,A 2 S3 - Depressurize RCS During SGTR (1 CR-638) S,D,A 3 S4 - Respond to a Loss of the RHR System (1 CR-694) S, M, L, A 4P S5 - Transfer from Bypass to Main Feed Regulating Valve (1 CR-520) S,D,A 4S S6 - Manually Actuate CIB (1 CR-578) S,D,A,EN 5 S7 - Transfer Bus 1AE From Emergency To Normal Feed (1CR-097) S,D 6 S8 - Verify CREVs Isolation (1 CR-662) S,N,A 7 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1 - Vent the Charging Pump Suction Header'(1 PL-057) R,D,EN,P 2 P2 - Startup the Dedicated Auxiliary Feedwater Pump [FW-P-4] (1 PL-007) D,E 4S P3 - Locally Start the No. 1 Emergency Diesel Generator (1 PL-606) D,E, EN 6

@ 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 SRO-I I SRO-U

-~------- -- -------

(A)lternate Path 4-6 /4-6 /2-3 (C)ontrol room (D)irect from bank S9/:S8/:S4 (E)mergency or abnormal in-plant ~1/;::11;::1 (EN)gineered safety feature  ;:: 1 I <! 1 I .'."'. 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  ::;; 3/ ::; 3 I ::; 2 (randomly selected)

{R)CA ~1/;::1/<!1 (S)imulator NUREG-1021, Revision 10 FENOC Facsimile Rev. 0

ES 301 C on tro IR oom /I n-Pl an t S;ys tems 0 utrme Form ES -301 -2 Facility: Beaver Valle~ Unit 1 Date of Examination: 10/31/thru 11/182016 Exam Level: RO D SRO(I) IZl SRO(U) D Operating Test No.: BV1LOT16 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U)

System I JPM Title Type Code* Safety Function S1 - Withdraw Shutdown Bank "A" (1 CR-085) S, M, L 1 S2 - Perform Manual Makeup to the Charging Pump Suction (1CR-581) S,D,A 2 S3 - Depressurize RCS During SGTR (1 CR-638) S,D,A 3 S5 - Transfer from Bypass to Main Feed Regulating Valve (1 CR-520) S,D,A 4S S6 - Manually Actuate CIB (1CR-578) S,D,A,EN 5 S7 - Transfer Bus 1ae From Emergency To Normal Feed (1 CR-097) S,D 6 S8 - Verify CREVs Isolation (1CR-662) S,N,A 7 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1 - Vent the Charging Pump Suction Header (1 PL-057) R,D,EN,P 2 P2 - Startup the Dedicated Auxiliary Feedwater Pump [FW-P-4] (1 PL-007) D,E 4S P3 - Locally Start the No. 1 Emergency Diesel Generator (1 PL-606) D,E,EN 6

@ 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 SRO-I I SRO-U (A)lternate Path 4-6 /4-6 /2-3 (C)ontrol room (D)irect from bank :59/:58/:54 (E)mergency or abnormal in-plant 2:1/2:1/2:1 (EN)gineered safety feature  ;:: 1 I;:: 1 I~ 1 (control room system)

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

(R)CA 2:1/2:1/2:1 (S)imulator NUREG-1021, Revision 10 FENOC Facsimile Rev. 0

A.ppen d"1x D . 0 utrme s cenano 1L16N1 Facility: BVPS Unit 1 Scenario No. 1 Op Test No.: BV1LOT16 NRC Examiners: Candidates: SRO ATC BOP Initial IC-62 (17): 67% power, MOL, Equ. XE Conditions, CB "D" @ 177 steps, Conditions: RCS boron - 985 ppm, 1FW-P-3A OOS Turnover: Maintain 67% power.

Critical Tasks: 1. CT-2 (E-0.D) Crew manually actuates at least 1 train of SIS

2. CT-51 (FR-S.1.B) Crew starts AFW pumps

3. CT-52 (FR-S.l.C) Crew inserts negative reactivity Event Malf. No. Event Type Event Description No.

(I,A) ATC, SRO 1 PRS06A Pressurizer level transmitter, 1RC-LT-459 drifts low.

(TS) SRO (I,A) BOP, SRO 2 XMT-MSS021A PT-1 MS-446 fails low.

(TS) SRO 3 CHS03 (C,A) ATC, SRO Isolable 25 gpm RCS leak on letdown line. (AOP 1.6. 7)

(R) ATC Main feedwater pump trip, requires turbine runback and 4 FWMOlA (C,A) BOP, SRO manual rod insertion.

GENO I, Spurious Gen Trip with auto & manual Rx trip failures 5 (M) ALL CRF12A, 12B (ATWS) 6 IOR X06i068C (C) ATC, SRO MOV-lCH-350 failed closed 7 INH20,21,36 (C) BOP, SRO All AFW pumps fail to auto start 8 RCS02A (M) ALL 950 gpm LOCA 9 VLV-MSS03,04 (C) BOP, SRO Reheat steam failure to auto isolate.

10 SIS 1OA, SIS 1OB (C) ATC, SRO Automatic SI actuation failure (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal E FR-S. l - E E Terminate after evaluating SI termination criteria.

NUREG-1021, Revision 10 FENOC Facsimile Rev O

Appendix D Scenario Outline 1L16N1 After taking the shift at 67% power with AFW pump, 1FW-P-3A OOS, Pressurizer level transmitter, lRC-LT-459 will drift low. The crew will diagnose the indications and IAW AOP 1.4.1, Process Control Failure, remove the failed channel from service and ensure the plant is stable, the SRO will transition to the instrument failure procedure for further channel actions and will address Tech Specs for the failed channel.

The non-selected Turbine First Stage pressure transmitter, lMS-PT-446 will fail low. IA W the instrument failure procedure, the crew will take action to transfer the condenser steam dump control to "Steam Pressure" mode. The SRO will address Tech Specs for the failed channel.

An isolable 25 gpm leak will then occur on the letdown line, the crew will enter AOP 1.6. 7, Excessive Primary Plant Leakage, to evaluate leak rate and leak location. The crew will determine the leak rate is

> 10 gpm and is isolable.

The "A" Main feed pump will then trip, the crew will enter AOP 1.24.1, Loss of Main Feedwater, and determine that a power reduction to <52% is required. The crew will reduce power IA W AOP 1.24.1.

When reactor power lowers to< 61 %, the Main Unit Generator will spuriously trip. The reactor will fail to automatically trip as expected due to the MUG trip. The crew will identify the automatic Rx trip failure.

The SRO will direct the crew to manually trip the Rx and perform IOA's of E-0. The A TC will attempt to manually trip the Rx which will also fail. The SRO will direct the crew to perform IOA's for FR-S.1, Response to Nuclear Power Generation - ATWS. Th~ control rods will fail to automatically insert, the A TC will place the Rod Control system in manual and begin inserting rods. When control bank "D" inserts to <150 steps, an "Urgent Failure" will occur in the Rod control system, stopping all rod motion. When the crew attempts to align the Emergency Boration flowpath, the Emergency Boration Valve, MOV-1 CH-350, will fail to open. The crew will align an alternate boration flow path by aligning the Charging pump suction to the RWST. At the lead evaluator's discretion, when an emergency boration flowpath is aligned, the reactor will be locally tripped via a field operator if dispatched.

Additionally, all available AFW pumps will fail to automatically start, the BOP will start the Turbine Driven AFW pump and the "B" Motor driven AFW pump. The BOP will recognize that Reheat steam failed to automatically isolate on the Turbine Trip and manually close, MOV-lMS-lOOA and lOOB.

When the Rx is locally tripped and verified, the crew will transition back to E-0, Reactor Trip Response, coincident with the local Rx trip, a 950 gpm LOCA will occur on the "A" Loop cold leg, While performing the IOA's of E-0, the crew will recognize that RCS pressure and level are reducing and that conditions require a Safety Injection which failed to automatically actuate. The crew will actuate SI and continue in E-0. The crew will progress thru E-0 and transition to E-1 after diagnosing that containment pressure and sump level are not consistent with pre-event values.

The scenario will be terminated after the crew has evaluated SI termination criteria in E-1.

Expected procedure flow path is E-0 ~ FR-S .1 ~ E-0 ~ E-1.

NUREG-1021, Revision 10 FENOC Facsimile Rev 0

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. 0 utrme 1L 16N2 Facility: BVPS Unit 1 Scenario No. 2 Op Test No.: BV1LOT16 NRC Examiners: Candidates: SRO ATC BOP Initial IC-64 (18): 100% power, MOL, Equ. XE Conditions, CB "D" @ 228 steps, Conditions: RCS boron - 870 ppm. 1FW-P-3A OOS Turnover: Maintain 100% power.

Critical Tasks: 1. CT-18 (E-3.A) Crew isolates ruptured SG

2. CT-19 (E-3.B) Crew establishes/maintains temperature

3. CT-20 (E-3.C) Crew depressurizes RCS to meet SI termination criteria Event Malf. No. Event Type Event Description No.

"C" SG, selected Main steam flow transmitter, lMS-FT-494 1 XMT-MSS039A (I,A) BOP, SRO fails low, requires manual control of feedwater and placing alternate channel in service.

(I,A) ATC, SRO PRZR pressure control transmitter, PT-I RC- 444 fails high, 2 PRS08D (TS) SRO requires closing PORV and manual PRZR pressure control.

(C,A) ATC, SRO 3 RCS03A 22 gpm SG Tube leak on "A" SG. (AOP 1.6.4)

(TS) SRO (R) ATC 4 SG tube leak requires plant S/D IA W AOP 1.51.1.

(N) BOP, SRO 5 RCS03A (M) ALL 650 gpm SGTR occurs on "A" SG during S/D.

6 INH40 (C) ATC, SRO "B" HHSI pump auto start failure on SI.

7 VL V-SGBO 1,02,03 (C) BOP, SRO SG BD isolation failure, requires manual valve closure.

Condenser steam dump fails open following cooldown, 8 MSS08C (C) BOP, SRO requires Main steam line isolation.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal E-0 __,. E-3 NUREG-1021, Revision 10 FENOC Facsimile Rev 0

Appendix D Scenario Outline 1L 16N2 After taking the shift at 100% power with AFW pump, 1FW-P-3A OOS., the "C" SG selected Steam flow transmitter will fail low requiring the BOP to take manual feedwater control IA W AOP 1.4.1, Process Control Failure. After the plant has been stabilized, the SRO will transition to the instrument failure procedure and direct placing the alternate steam flow channel in service, Feedwater control will then be returned to automatic.

The controlling PRZR pressure channel, PT-lRC-444 will drift high causing RCS pressure to lower due to the pressurizer spray valves and a PORV opening. The crew will initially respond IA W AOP 1.4.1, identify the failure and close the spray valves and PORV, the SRO will then transition to 1OM-6.4.IF, attachment 2 to address the failed channel. The ATC controls PRZR pressure by manually operating the PRZR heaters and spray valves, or manual control of the PRZR pressure master controller. The SRO will address applicable TS entered due to the instrument failure.

Subsequently, a 22 gpm SG tube leak will develop on the "A" SG. AOP 1.6.4 will be entered and the leak will be quantified. Due to the leak rate, AOP 1.6.4 will provide direction to enter Mode 3 IA W AOP 1.51.1. The SRO will address Technical Specifications which also will require Mode 3 entry.

The crew will initiate an emergency shutdown IA W AOP 1.51.1, when Rx power is reduced to <94%, the tube leak will become a 650 gpm tube rupture. The crew will identify degrading plant parameters and the SRO will direct a pre-emptive reactor trip and enter E-0.

An automatic Safety Injection will occur upon the Rx trip, the "B" HHSI pump will fail to automatically start on the SI signal, the A TC will identify the failure and manually start the pump. Additionally, the steam generator blowdown system will fail to automatically isolate requiring the BOP to identify and isolate the SG Blowdown system.

The crew will proceed thru E-0, perform diagnostics and determine that indications of a SGTR exist, the SRO will transition to E-3 to take actions to address the tube rupture.

After the crew identifies the "A" SG as the ruptured SG and isolates it, a target temperature will be determined and a cooldown commenced. A condenser steam dump valve will fail open during the cooldown, when the cooldown to target temperature is reached, the BOP will identify the failed open steam dump and report it to the crew. The SRO will direct the BOP to manually close the Main steam line isolation valves, requiring the BOP to stabilize RCS temperature using the "B" and "C" SG atmospheric steam dump valves.

The scenario will be terminated when the crew terminates SI and establishes a normal charging flow alignment IA W E-3.

Expected procedure flow path is E-0---+ E-3 NUREG-1021, Revision 10 FENOC Facsimile Rev 0

Appendix D Scenario Outline 1L 16N3 Facility: BVPS Unit 1 Scenario No. 3 Op Test No.: BV1LOT16 NRC Examiners: Candidates: SRO ATC BOP Initial IC-66(5): ~5% power, BOL, CB "D"@ 109 steps, RCS boron - 1750 ppm.

Conditions:

Turnover: Raise Rx power and place turbine online.

Critical Tasks: 1. CT-1 (E-0.A) Crew manually trips the reactor.

2. CT-24 (E-0.C) Energize 1 AC emer bus

3. CT-9 (E-0.L) Establish flow from RPRW pump Event Malf. No. Event Type Event Description No.

(R) ATC 1 Power increase to> P-10.

(N) SRO "B" Bypass feed regulating valve fails asis in Auto. Requires 2 FWM08B (C,A) BOP, SRO manual control CHS22 Failure ofFCV-lCH-122 controller, requires manual control 3 (C,A) ATC, SRO X06D088M of PRZR lvl.

4 (N) BOP, SRO Startup standby Turbine plant River water pump (C,A) BOP, SRO 5 NIS08B N-42 Instrument power fuse blown. (>P6 and< PIO)

(TS) SRO 6 XMT-CNM004A (TS) SRO CH 2, CNMT Pressure transmitter fails High, PT-1 LM-1 OOB 7 SIS I OB (M) All Inadvertent Train "B" SI with Rx trip failure.

8 (C,A) ATC, SRO Manual Rx trip EPS04E, 04F Loss of lAE and lDF 4kv Busses on Rx trip w/ EDG auto 9 (C) BOP, SRO INH53, 54 start failures.

Reactor plant River water pump auto start failures on 10 INH32, 33 (C) BOP, SRO Sequencer, requires manually starting WR-P-lA and lB.

Letdown isolation on SI unable to be recovered due to failure 11 CHS21A (C) ATC, SRO of LCV -1 CH-460A, requires Excess letdown.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal E-0 ~ ES-1.1 NUREG-1021, Revision 10 FENOC Facsimile Rev 0

Appendix D Scenario Outline 1L16N3 The crew will assume the shift at approximately 5% (4.8%) power with instructions to raise power to place the turbine online IAW the reactivity plan and lOM-52.4.A. The ATC will initiate a dilution and withdraw control rods.

After the power has raised to >5.5%, "B" SG Bypass Feed regulating valve will fail asis in Auto, failure will become evident as Rx power continues to be raised, the BOP will be required to identify the malfunction and take action to control FCV-lFW-489 in manual IAW AOP 1.4.1 Process Control Failure.

Additionally, at >5.5% Rx power, FCV-lCH-122 will fail closed in Auto, the ATC will be required to identify the failure and manually control FCV-1 CH-122 IA W AOP 1.4.1, to maintain PRZR level A field operator will then report that the "A" Turbine Plant River Water pump, has a significant oil leak and needs to be shutdown. IA W 1OM-30.4.N, Standby Turbine Plant River Water Pump Startup, the BOP will startup 1WR-P-6B and shutdown 1WR-P-6A.

An instrument power fuse will then blow for Power Range Nuclear instrument, N-42. The crew will identify the N-42 blown fuse failure and the SRO will enter AOP 1.2.1 C, Power Range Channel Malfunction, and direct the BOP to remove the failed channel from service. The SRO will address Tech Specs for the failed instrument.

After the crew has removed N42 from service, CH 2 containment pressure transmitter, PT-lLM-lOOB fails high. The SRO will enter 1OM-1.4.IF and review the Technical Specifications. The SRO will then contact I&C to trip,the applicable bistables.

After the SRO has determined the appropriate Technical Specifications for the CNMT pressure channel, a spurious Train "B" Safety Injection signal will occur with an automatic Rx trip failure. The crew will recognize the automatic Rx trip failure and the SRO will direct the ATC to manually trip the Rx and perform the IOA's of E-0.

Upon the Rx trip, both Emergency 4Kv buses will deenergize with auto start failures of both Emergency Diesel Generators. The BOP will start an EDG IA W E-0 IOA's (RNO actions).

Upon EDG start, each respective River water pump will fail to auto start via sequencer, the crew will identify the auto start failure and start each R W pump.

The crew will continue progressing thru E-0 and perform diagnostic steps and determine that no accident has occurred and plant conditions support Termination of Safety Injection and transition to ES-1.1. LCV-1CH-460A fails closed upon the SI signal and will not be able to be reopened requiring the crew to place Excess letdown in service.

The scenario will be terminated when the crew establishes Excess letdown flow.

Expected procedure flow path is E-0 ~ ES-1.1 NUREG-1021, Revision 10 FENOC Facsimile Rev 0

Appendix D Scenario Outline 1L 16N5 Facility: BVPS Unit 1 Scenario No. 5 Op Test No.: BV1LOT16 NRC Examiners: Candidates: SRO ATC BOP Initial IC- 68 (10): 100% power, BOL, Equ. XE Conditions, CB "D"@ 228 steps, Conditions: RCS boron - 1210 ppm, 1FW-P-3A OOS Turnover: Maintain 100% power.

Critical Tasks: 1. CT-10 (E-0.M) Crew closes upstream PORV Block valve.

2. CT-11 (E-0.0) Crew closes CNMT isolation valves.

3. CT-43 (FR-H.1.A) Crew establishes feedwater flow before feed and bleed required.

Event Malf. No. Event Type Event Description No.

VCT Level Transmitter, lCH-LT-112 fails low causing auto 1 CHS20B (I,A) A TC, SRO makeup to occur.

"A" Feedwater flow transmitter fails high, requires manual 2 FWM14B (C,A) BOP, SRO control of Feedwater control valve and placing alternate channel in service and return to auto control.

(C,A) ATC, SRO 3 NIS03D N44 failed high, control rods automatically insert. (AOP 1.1.3)

(TS) SRO (N) BOP, SRO 4 N44 removal from service. (AOP 1.2.1 C)

(TS) SRO (C,A) BOP, SRO 5 CRF04 Dropped Rod, requires turbine load reduction (AOP 1.1.8)

(TS) SRO 6 CRF04BP (C,A) ATC, SRO 2nd Dropped Rod, requires manual Rx trip.

7 (M) ALL Reactor Trip lRC-PT-445 fails high on Rx trip, PORVs open, requires 8 PRS08E C)ATC, SRO closing Block valves.

INH49 Train "B" CIA Actuation failure with MOV-lCH-378 (Tm A) 9 (C) BOP, SRO VLV-SEA09 auto close failure.

Loss of ALL Feedwater - FR-H.1 with main feed pump 10 (M) ALL recovery.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal E-0---+ FR-H.1 ---+ E-0 NUREG-1021, Revision 10 FENOC Facsimile Rev 0

Appendix D Scenario Outline 1L 16N5 After taking the shift at 100% power with AFW pump 1FW-P-3A OOS, VCT level transmitter, lCH-LT-112 will fail low causing an automatic Makeup to occur. The ATC will diagnose the indications and IAW AOP 1.4.1, Process Control Failure, turn the blender off to stop the makeup and ensure the plant is stable, the SRO will transition to the instrument failure procedure for further channel actions.

The channel 3, "A" SG feed flow transmitter, FT-lFW-477, will then drift high, the crew will recognize the SG level perturbation and IAW AOP 1.4.1, the BOP will place the controller for lFW-FCV-478 in manual and restore SG level, the SRO will transition to the instrument failure procedure for additional channel removal actions and place the alternate channel in service, the BOP will then return lFW-FCV-4 78 to automatic control.

Power Range Nuclear instrument, N-44 will then fail high causing the control rods to automatically insert.

The crew will perform the Immediate Operator Actions for AOP 1.1.3, Unexpected Control Rod Movement. The ATC will identify the N-44 failure and place the rods in manual. The SRO will then transition to AOP 1.2.1 C, Power Range Channel Malfunction, and direct the BOP to remove the failed channel from service. The SRO will address Tech Specs for the failed instrument.

A control rod will then drop, the crew will enter AOP 1.1.8 for an Inoperable Rod. Due to the magnitude of the RCS temperature drop, the crew will be required to lower power to restore RCS temperature. The SRO will address Tech Specs for the dropped rod.

After the crew has completed a power reduction and stabilized the plant, a 2nd control rod will d,rop. The ATC will recognize that 2 control rods are now dropped. Due to 2 dropped rods, IA W AOP 1.1.8 IOA's, the SRO will direct the A TC to manually trip the Rx and enter E-0.

When the Rx is manually tripped, lRC-PT-445 will fail high causing 2 PORV's to open resulting in a Safety Injection signal, the ATC will recognize the open PORV's with lowering RCS and manually close the valves.

The safety injection that occurred as a result of the PORV's opening, will fail to actuate the train "B" CIA signal, and MOV-lCH-378 (a train "A" CIA valve) will fail to automatically close. The crew will be required to isolate the containment penetration via either manually actuating Train "B" CIA or manually closing MOV-lCH-378.

On the trip, the turbine driven AFW pump, lFW-P-2 will start but not produce any flow, the remaining available motor driven AFW pump, 1FW-P-3B will start but will trip when the SI Manual actuation PB's are depressed. When "Verifying AFW Status" in E-0, the crew will identify that all auxiliary feed water pumps have failed, the SRO will transition to FR-H. l.

IA W FR-H.1 direction the crew will restore feedwater flow by starting a main feedwater pump. After feed flow is verified, the SRO will return to E-0 at which point the scenario will be terminated.

Expected procedure flow path is E FR-H.1 -t E-0 NUREG-1021, Revision 10 FENOC Facsimile Rev 0