ML18275A357
| ML18275A357 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/17/2018 |
| From: | David Silk Operations Branch I |
| To: | Gaydosik T FirstEnergy Nuclear Operating Co |
| Shared Package | |
| ML17312A957 | List: |
| References | |
| 000500 | |
| Download: ML18275A357 (30) | |
Text
If ES-401 PWR Examination Outline Form ES-401-2 Facility: Beaver Valley Unit 1 1LOTI8 Date of Exam:
RO KIA Category 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
- 1.
3 3
3 2
18
- Emergency 2
1 2
2 1
9 Abnormal Plant Tier 4
4 5 4
27 Evolutions Totals 1
2 3
2 3
3 2
3 3
1 3
3 28
- 2.
Plant 2
Systems 2
1 0
1 1
1 1
1 1
1 0
10 Tier 4
4 2 4 4 3 4 4
2 4 3 38 Totals 1
2 3
4 10
- 3. Generic Knowledge and Abilities Category 2
2 3
3 Note:
- 1.
F.nsure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the "Tier Totals" in each KIA category shall not be less than two). (One Tier 3 radiation control KIA is allowed if it is replaced by a K/A 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 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 0.1.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 K/As having an importance rating (IR) of 2.5 or higher shall be selected.
Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and 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 Kl As.
- 8.
On the following pages, enter the KIA numbers, a brief description of each topic, the topics' 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 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 K/A catalog and enter the KIA numbers, descriptions, I Rs, and point totals(#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
,* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
- These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the KIA catalo is used to develo the sam le Ian.
\\IUREG-1021, Revision 11 RO Page 1 of 13 FENOC Facsimile Rev. 0
ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions -Tier 1/Group 1(RO) ad.*
- /APE # / Name I Safety Function K
K K A 'A.. ~
KIA Topic(s)
IR 1
2 3
1
- 2 000007 (EPE 7; BW E02&E10; CE E02)
)(
EK1 Knowledge of the operational implications of 3.3 Reactor Trip, Stabilization, Recovery/ 1 the following concepts as they apply to the reactor
[Question 1]
trip:
EK1.05 Decay power as a function of time (CFR 41.8 / 41.10 / 45.3)
' ).
000008 (APE 8) Pressurizer Vapor Space
)(
\\< AK2. Knowledge of the interrelations between the 2.7 Accident/ 3 Pressurizer Vapor Space Accident and the i! [Q::.ld,tiOn 2]
following:
- 1 AK2.01 Valves (CFR 41.7 / 45.7) 000009 (EPE 9) Small Break LOCA I 3
)(
EA 1 Ability to operate and monitor the following as 3.4
[Question 3]
they apply to a small break LOCA:
EA 1.14 Secondary pressure control
- It-(CFR 41. 7 / 45.5 / 45.6)
- 1F 2.6 000011 (EPE 11) Large Break LOCA I 3
)(
EK2 Knowledge of the interrelations between the
[Question 4]
and the following Large Break LOCA:
EK2.02 Pumps (CFR 41.7 / 45.7) i 000015 (APE 15) Reactor Coolant Pump
)(
AA2. Ability to determine and interpret the following 3.4 Mr"* *nctions / 4 as they apply to the Reactor Coolant Pump I
Malfunctions (Loss of RC Flow):
[C...
.,tion 5]
AA2.09 When to secure RCPs on high stator temperatures (CFR 43.5 / 45.13) 000025 (APE 25) Loss of Residual Heat X
y
. }{
AA 1. Ability to operate and / or monitor the 2.9 r.:emoval System/ 4 following as they apply to the Loss of Residual Heat I [Uuestion 6]
Removal System:
AA 1.08 RHR cooler inlet and outlet temperature indicators (CFR 41.7 / 45.5 / 45.6) 000026 (APE 26) Loss of Component
)(
AK3. Knowledge of the reasons for the following 3.6 Cooling Water/ 8 responses as they apply to the Loss of Component
[Question 7]
Cooling Water:
AK3.02 The automatic actions (alignments) within the CCWS resulting from the actuation of the ESFAS I
(CFR 41.5,41.10 / 45.6 / 45.13) 000027 (APE 27) Pressurizer Pressure
)(
AK1. Knowledge of the operational implications of 3.1 Control System Malfunction / 3 the following concepts as they apply to Pressurizer
[Question 8]
Pressure Control Malfunctions:
AK1.01 Definition of saturation temperature (CFR 41.8 / 41.10 / 45.3)
NUREG-1021, Revision 11 RO Page 2 of 13 FENOC Facsimile Rev. 0
ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions -Tier 1/Group 1(RO) Continued
=
1 E/APE #/Name/Safety Function K
K K A A G KIA Topic(s)
IR 1
2 3
1 2,*
000038 (EPE 38) Steam Generator Tube 1:) 2.4.20 Knowledge of the operational implications of 3.8 Rupture/ 3 EOP warnings, cautions, and notes.
1
[Question 9]
J,.,..
(CFR: 41.10 / 43.5 / 45.13) 000055 (EPE 55) Station Blackout / 6 1;) :,:,:
EA2 Ability to determine or interpret the following as 3.4
[Question 1 OJ
'*i; they apply to a Station Blackout:
EA2.01 Existing valve positioning on a loss of instrument air system (CFR 43.5 / 45.13) 000057 (APE 57) Loss of Vital AC
.) 2.1.27 Knowledge of system purpose and/or 3.9 Instrument Bus / 6 function.
f Q. *?~tiori 11 l (CFR: 41.7) 000058 (APE 58) Loss of DC Power/ 6
}(
AK3. Knowledge of the reasons for the following 3.4
[Question 12]
responses as they apply to the Loss of DC Power:
'.. ;.. : It:'
AK3.01 Use of de control power by D/Gs (CFR 41.5,41.10 / 45.6 / 45.1) 000062 (APE 62) Loss of Nuclear Service
}(
AK3. Knowledge of the reasons for the following 4.0 Water/ 4 responses as they apply to the Loss of Nuclear
[O* *'"'c;tion 13]
Service Water:
1 AK3.03 Guidance actions contained in EOP for I* :
Loss of nuclear service water (CFR 41.4, 41.8 / 45.7) 000065 (APE 65) Loss of Instrument Air/ 8
)
AA2. Ability to determine and interpret the following 3.4
[Question 14]
as they apply to the Loss of Instrument Air:
i AA2.05 When to commence plant shutdown if instrument air pressure is decreasing (CFR: 43.5 / 45.13) 000077 (APE 77) Generator Voltage and
}(
' '.j AK1. Knowledge of the operational implications of 3.3 1,
the following concepts as they apply to Generator Flectric Grid Disturbances I 6 Voltage and Electric Grid Disturbances:
[\\Ji.lcSll011 15]
AA 1.03 Under-excitation
<}.'
(CFR: 41.4, 41.5, 41.7, 41.10/45.8)
(W E04) LOCA Outside Containment I 3
}(
EA 1. Ability to operate and / or monitor the 3.8
[Question 16]
following as they apply to the (LOCA Outside Containment)
EA 1.3 Desired operating results during abnormal and emergency situations.
(CFR: 41. 7 / 45.5 / 45.6)
NUREG-1021, Revision 11 RO Page 3 of 13 FENOC Facsimile Rev. 0
es~,
PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions -Tier 1/Group 1(RO) Continued
,=
t:G
- i E/APE #/Name I Safety Function K K K A '-A KIA Topic(s)
IR 1
2 3
1,s2 (BW E04; W E05) Inadequate Heat
)(
J:
EK2. Knowledge of the interrelations between the 3.7 Transfer-Loss of Secondary Heat Sink / 4 (Loss of Secondary Heat Sink) and the following:
[Question 17]
EK2.1 Components, and functions of control and
[,i}f safety systems, including instrumentation, signals, 1,
interlocks, failure modes, and automatic and 1*.
manual features.
,\\ IX. (CFR: 41.7 / 45.7)
(W E11) Loss of Emergency Coolant l;l r*rf EA2. Ability to determine and interpret the following 3.4 Recirculation / 4 as they apply to the (Loss of Emergency Coolant
[Question 18]
,t Recirculation)
I* !} EA2.1 Facility conditions and selection of appropriate procedures during abnormal and 1*,...
emergency operations.
(CFR: 43.5 I 45.13)
~r-3 3 3 3 4 2 I*
~ KIA Category Point Totals:
Group Point Total:
18 NUREG-1021, Revision 11 RO Page 4 of 13 FENOC Facsimile Rev. 0
ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2(RO)
E/APE # / Name I Safety Function KIA Topic(s)
IR KKK A~~*G 1
2 3
1 ;i *:"'"'
ii=....
,.,,==============l==lo=l===t===l=~:==1=====================1====+====11
' 000001 (APE 1) Continuous Rod
~
i : "'
- 3.5 Withdrawal/ 1
.* * [~i
./
AK1. Knowledge of the operational implications of the following concepts as they apply to Continuous Rod Withdrawal:
[Question 19]
/ i, AK1.07 Effects of power level and control position on flux 000036 (APE 36; BW/A08) Fuel-Handling Incidents / 8
[Question 20]
000037 (APE 37) Steam Generator Tube Leak/ 3
[Question 21]
t 000051 (APE 51) Loss of Condenser
- ar.*,;_::r1- / 4 - -
[C'**-c;tion 22]
I 000068 (APE 68; BW A06) Control Room Evacuation / 8
[Question 23]
000076 (APE 76) High Reactor Coolant Activity/ 9
[Question 24]
, 3W E08; W E03) LOC/\\ Cooidown-De;.;i~ssurizatiun I 4
[Question 25]
NUREG-1021, Revision 11
}(
';\\'. 1:.
- J;....
(CFR 41.8 / 41.10 / 45.3) 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.
. *;: I;:.:.
(CFR: 41.10 / 45.3)
'.k-
}(
l*.,. AK3. Knowledge of the reasons for the following 1,
responses as they apply to the Steam Generator Tube Leak:
I*
AK3.05 Actions contained in procedures for radiation monitoring, RCS water inventory balance, SIG tube failure, and plant shutdown (CFR 41.5,41.10 / 45.6 / 45.13)
AA2. Ability to determine and interpret the following as they apply to the Loss of Condenser Vacuum:
AA2.02 Conditions requiring reactor and/or turbine trip (CFR: 43.5 / 45.13)
AA 1. Ability to operate and / or monitor the following as they apply to the Control Room Evacuation:
AA 1.11 Emergency borate valve controls and indicators (CFR 41.7 / 45.5 / 45.6) 4.2 3.7 3.9 3.9 AK3. Knowledge of the reasons for the following 2.9 responses as they apply to the High Reactor Coolant Activity AK3.05 Corrective actions as a result of high fission-product radioactivity level in the RCS (CFR 41.5,41.10 / 45.6 / 45.13)
EA 1. Ability to operate and I or monitor the following as they apply to the (LOCA Cooldown and Depressurization)
EA 1.3 Desired operating results during abnormal and emergency situations.
(CFR: 41.7 / 45.5 / 45.6) 3.7 RO Page 5 of 13 FENOC Facsimile Rev. 0
ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2(RO) Continued r
E/APE #/Name/Safety Function K K K A *~ G KIA Topic(s)
IR 1 2 3
1
- 2. "'
(CE A 11 **; W E08) RCS Overcooling-
'ij 2.4.35 Knowledge of local auxiliary operator tasks 3.8 Pressurized Thermal Shock / 4 during an emergency and the resultant operational effects.
i.Cbestion 26) i (CFR: 41.10 / 43.5 / 45.13)
,J.,:,
(W E15) Containment Flooding/ 5
}(
I{ EK2. Knowledge of the interrelations between the 2.8 (Containment Flooding) and the following:
(Question 27]
'..../'
EK2.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
(CFR: 41.7 I 45.7) 1 1 2 2 1 2 KIA Category Point Totals:
Group Point Total:
9 NUREG-1021, Revision 11 RO Page 6 of 13 FENOC Facsimile Rev. 0
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems
- Tier 2/Group 1(RO)
I System # I Name K K K *K K K A A A A'.(3 KIA Topic(s)
IR 1
2 3
4 5
6 1 ~2 3
4
/,,,'.;,
003 (SF4P RCP) Reactor
)(
K5 Knowledge of the operational implications 2.8 Coolant Pump of the following concepts as they apply to the RCPS:
(Question 28)
K5.02 Effects of RCP coastdown on RCS parameters
<CFR: 41.5 / 45.7)
- 21J4 (SF1; SF2 CVCS) Ct,emical 1,,
- *
- ~ 2.2.37 Ability to determine operability and/or 3.6 i1 io,iJ voiume Control availability of safety related equipment.
[Question 29)
(CFR: 41.7 / 43.5 / 45.12) 004 (SF1; SF2 CVCS) Chemical
)(
K5 Knowledge of the operational implications 2.6 and Volume Control of the following concepts as they apply to the
', eves:
[Question 30)
K5.29 Reason for sampling for chloride, fluoride, sodium and solids in RCS (CFR: 41.5/45. 7) 005 (SF4P RHR) Residual Heat
)(
K4 Knowledge of RHRS design feature(s) 2.9 Removal and/or interlock(s) which provide or the
[Question 31) following:
K4.03 RHR heat exchanger bypass flow control (CFR: 41.7) t---;
)(
0,
,3F2; SF3 ECCS)
K2 Knowledge of bus power supplies to the 3.6 Emergency Core Cooling following:
[Question 32)
K2.04 ESFAS-operated valves (CFR: 41.7)
- ' 007 (SF5 PRTS) Prnssurizer
)(
K4 Knowledge of PRTS design feature(s) 2.6 ii R~(i5i{Gn,cnch Tank and/or interlock(s) which provide for the following:
[Question 33)
K4.01 Quench tank cooling (CFR: 41.7) 008 (SF8 CCW) Component
)(
A 1 Ability to predict and/or monitor changes in 2.8 Cooling Water parameters (to prevent exceeding design
[Question 34) limits) associated with operating the CCWS controls including:
A 1.01 CCW flow rate (CFR: 41.5 / 45.5) 010 (SF3 PZR PCS) Pressurizer
)(
K2 Knowledge of bus power supplies to the 2.7 Pressure Control following:
I
[Question 35)
K2.04 Indicator for code safety position (CFR:41.7) 012 (SF? RPS) Reactor
)(
K3 Knowledge of the effect that a loss or 3.9 Protection malfunction of the RPS will have on the
[(
lion 36) following:
K3.01 CRDS (CFR: 41. 7 / 45.6)
NUREG-1021, Revision 11 RO Page 7 of 13 FENOC Facsimile Rev. 0
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems
- Tier 2/Group 1 (RO) Continued i
System # I Name K K K K K K A A A A \\3 KIA Topic(s)
IR 1 2 3
4 5
6 1 1;2 3
4 '
013 (SF2 ESFAS) Engineered
}(
'f K6 Knowledge of the effect of a loss or 2.7 Safety Features Actuation malfunction on the following will have on the
[Question 37]
I' ESFAS:
K6.01 Sensors and detectors (CFR: 41. 7 / 45.5 to 45.8) 022 (SF5 CCS) Containment X
A 1 Ability to predict and/or monitor changes in 3.6 Cooling parameters (to prevent exceeding design
[Question 38]
limits) associated with operating the CCS controls including:
f :}
A 1.01 Containment temperature
- <1 (CFR
- 41.5 / 45.5)
II, 022 (SF5 CCS) Containment
)
K2 Knowledge of power supplies to the 3.0 Cooling t',,
following:
[Question 39]
K2.01 Containment cooling fans (CFR: 41.7) 026 (SF5 CSS) Containment X
A 1 Ability to predict and/or monitor changes in 3.6 Spray parameters (to prevent exceeding design (Question 40]
limits) associated with operating the CSS controls including:
A 1.02 Containment temperature (CFR: 41.5 / 45.5) 026 (SF5 CSS) Containment
}( ;
A4 Ability to manually operate and/or monitor 4.5 Spray in the control room:
[Question 41]
A4.01 CSS controls (CFR: 41. 7 / 45.5 to 45.8) 11 '139 (SF4S MSS) Ml'lin ond
)
A3 Ability to monitor automatic operation of the 3.1
-1 r.~;el:Jl Steam MRSS, including:
[Question 42]
A3,02 Isolation of the MRSS (CFR: 41,5 I 45.5) 059 (SF4S MFW) Main
'}(
A2 Ability to (a) predict the impacts of the 3.1 Feedwater following malfunctions or operations on the
[Question 43]
MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
A2.12 Failure offeedwater regulating valves (CFR: 41.5 I 43.5 / 45.3 I 45.13)
O!Jl::I \\SF4S MFW) Main
}( 2.1.30 Ability to locate and operate 4.4 Feedwater components, including local controls.
[Question 44]
(CFR: 41.7 /45.7)
NUREG-1021, Revision 11 RO Page 8 of 13 FENOC Facsimile Rev. 0
~;*--. -
"'......... ---~--
I "ES-401 PWR Examination Outline Form ES-401-2 Plant Systems
- Tier 2/Group 1 (RO) Continued I
A System # / Name K
K K
K K K A A
A G KIA Topic(s)
IR 1
2 3
4 5
6 1 2 3
4 r' '
061 (SF4S AFW)
X K5 Knowledge of the operational implications 3.6 Auxiliary/Emergency F eedwater it of the following concepts as the apply to the AFW:
[Question 45]
K5.01 Relationship between AFW flow and RCS heat transfer (CFR: 41.5 / 45.7) 061 (SF4S AFW)
}(
K6 Knowledge of the effect of a loss or 2.6 Auxiliary/Emergency Feedwater malfunction of the following will have on the AFW components:
[Question 46]
K6.02 Pumps (CFR: 41.7 / 45.7)
~- -~a2 (SF6 ED AC) AC Electrical I
~:f 2.4.20 Knowledge of the operational 3.8 I\\ \\Ji&u 1ot..ifion 1,;*.*
implications of EOP warnings, cautions, and
[Question 47]
I>.
notes.
1*
(CFR: 41.10 / 43.5 / 45.13) 063 (SF6 ED DC) DC Electrical
}( 1*_,
A4 Ability to manually operate and/or monitor 2.8 Oi!':tribution in the control room:
[,
.,tion 48]
A4.01 Major breakers and control power fuses (CFR: 41. 7 / 45.5 to 45.8) 063 (SF6 ED DC) DC Electrical
}(
K1 Knowledge of the physical connections 2.9 Distribution and/or cause-effect relationships between the DC electrical system and the following
[Question 49]
systems:
K1.03 Battery charger and battery (CFR: 41.2 to 41.9 I 45. 7 to 45.8) 064 (SF6 EOG) Emergency
}(
K1 Knowledge of the physical connections 3.1 Diesel Generator and/or cause-effect relationships between the ED/G system and the following systems:
'! :'.1v~~+it:1n 50]
I K1.02 D/G cooling water system I
(CFR: 41.2 to 41.9 I 45.7 to 45.8) 073 (SF? PRM) Process
)
A2 Ability to (a) predict the impacts of the 2.5 Radiation Monitoring following malfunctions or operations on the PRM system; and (b) based on those
[Question 51]
predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
A2.01 Erratic or failed power supply (CFR: 41.5 / 43.5 I 45.3 I 45.13)
NUREG-1021, Revision 11 RO Page 9 of 13 FENOC Facsimile Rev. 0
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems
- Tier 2/Group 1 (RO) Continued
/:., I
. Systen; # I Name K K K K K K A A A A.*. G KIA Topic(s)
IR
_;I 1 2 3
4 5
6 1 2 3
4 076 (SF4S SW) Service Water
)
K3 Knowledge of the effect that a loss or 3.5
[Question 52]
malfunction of the SWS will have on the following:
K3.03 Reactor building closed cooling water
\\
.. (CFR: 41. 7 / 45.6) 076 (SF4S SW) Service Water
}(
i**
K4 Knowledge of SWS design feature(s) 2.5
[Question 53]
f.1-'.,
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) 078 (SF8 IAS) Instrument Air
)
A4 Ability to manually operate and/or monitor 3.1
[Question 54]
in the control room:
A4.01 Pressure gauges
.c' (CFR: 41. 7 / 45.5 to 45.8) 103 (SF5 CNT) Containment
)C A2 Ability to (a) predict the impacts of the 2.9
,I following malfunctions or operations on the
- I
>iesti~n,5.~]..
containment system and (b) based on those Ii
' predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations
[:,
A2.05 Emergency containment entry
.f (CFR: 41.5 / 43.5 / 45.3 I 45.13) 2 3 2 3 3 2 3 3 1 3 3 KIA Category Point Totals:
Group Point Total:
28 NUREG-1021, Revision 11 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
- A K K K K K A A
A :G KIA Topic(s) 1 2
3 4
5 6
1 2 3
4 (l IR 001 (SF1 CRDS) Control Rod
)
i{ K2 Knowledge of bus power supplies to the following:
3.6 Drive K2.02 One-line diagram of power supply to trip
- I f Ouestion 56]
breakers (CFR: 41.7)
II i
011 (SF2 PZR LCS) Pressurizer
}(
K4 Knowledge of PZR LCS design feature(s) and/or 3.7 Level Control interlock(s) which provide for the following:
[Question 57]
K4.05 PZR level inputs to RPS (CFR: 41.7) 015 (SF? NI) Nuclear
)
K6 Knowledge of the effect of a loss or malfunction on 2.6 Instrumentation the following will have on the NIS:
[Question 58]
K6.03 Component interconnections I
I>**
(CFR: 41. 7 / 45. 7) c, 071 (SF9 WGS) Waste Gas
}(
A 1 Ability to predict and/or monitor changes in 2.5 Disposal parameters (to prevent exceeding design limits) associated with Waste Gas Disposal System
[Question 59]
operating the controls including:
A 1.06 Ventilation system (CFR: 41.5 / 45.5)
().
..3F4S SOS) Steam
}(
KS Knowledge of the operational implications of the 2.5
~ Jmp/Turbine Bvpfli:s Control following concepts as the apply to the SOS:
'I 1,.
n [Question 60]
KS.06 Effect of power change on fuel cladding (CFR: 41.5 / 45. 7) 045 (SF 4S MTG) Main Turbine
- x A2 Ability to (a) predict the impacts of the following 2.5 Generator malfunctions or operation on the MT/G system; and (b) based on those predictions, use procedures to
[Question 61]
correct, control, or mitigate the consequences of those malfunctions or operations:
A2.12 Control rod insertion limits exceeded (stabilize secondary)
(CFR: 41.5 / 43.5 I 45.3 I 45.5) 055 (SF4S CARS) Condenser
}(
4':
i::t K1 Knowledge of the physical connections and/or 2.6 Air Removal cause-effect relationships between the CARS and the J
following systems:
[Question 62]
1J,t l;i,
- ./
- r K1.06 PRM system IJ 1tfi*
(CFR: 41.2 to 41.9 / 45. 7 to 45.8) 056 (SF4S CDS) Condensate
}(
K1 Knowledge of the physical connections and/or 2.6 cause-effect relationships between the Condensate I [Ot.iestion 63]
System and the following systems:
I K1.03 MFW (CFR: 41.2 to 41.9 / 45.7 to 45.8)
NUREG-1021, Revision 11 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 2 3 4*
i IR I
072 (SF? ARM) Area Radiation
)(
r A3 Ability to monitor automatic operation of the ARM 2.9 Monitoring 1::
system, including:
[Question 64]
I* '
A3.01 Changes in ventilation alignment (CFR: 41. 7 / 45.5)
- J 075 (SF8 CW) Circulating Water
)(
i' A4 Ability to manually operate and/or monitor in the 3.2 l
control room:
1, [(_/H.<.:Stio.-1 65r II A4.01 Emergency/essential SWS pumps (CFR: 41. 7 / 45.5 to 45.8)
'1"!
KIA Category Point Totals:
2 1 0 1 1 1 1 1 1 1 0 Group Point Total:
10 NUREG-1021, Revision 11 RO Page 12 of 13 FENOC Facsimile Rev. 0
f "
ES 401 Generic Knowledge and Abilities Outline (Tier 3)
~u~ility: Beaver Valley Unit 1 1LOT18 Date of Exam:
Category KIA#
- 1.
G2.1.1 Conduct of Operations
>------+
Topic 2.1.1 Knowledge of conduct of operations requirements.
(CFR: 41.10 / 45.13)
[Question 66]
G2.1.23 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.
(CFR: 41.10 / 43.5 / 45.2 / 45.6)
[Question 67]
Subtotal
'F,"---*-"""'""""'_,,....,. ___________________________ _
' L.
G2.2.13 2.2.13 Knowledge of tagging and clearance procedures.
- Equipment (CFR: 41.1 o / 45.13)
Control
[Question 68]
3 F..
Jtion Control
- 4.
Emergency Procedures/
Plan G2.2.41 2.2.41 Ability to obtain and interpret station electrical and mechanical drawings.
Subtotal (CFR: 41.10 / 45.12 / 45.13)
[Question 69]
G2.3.14 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.
(CFR: 41.12 / 43.4 / 45.10)
[Question 70]
G2.3.15 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
(CFR: 41.12 / 43.4 / 45.9)
[Question 71 J G2.3.7 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions.
, (CFR: 41.12 / 45.10)
[Question 72]
Subtotal G2.4.14 2.4.14 Knowledge of general guidelines for EOP usage.
(CFR: 41.10/45.13)
[Question 73]
G2.4.19 2.4.19 Knowledge of EOP layout, symbols, and icons.
(CFR: 41.10 / 45.13)
[Question 74]
i------+-
G2.4.45 2.4.45 Ability to prioritize and interpret the significance of each Subtotal annunciator or alarm.
(CFR: 41.10 / 43.5 / 45.3 I 45.12)
[Question 75]
Tier 3 Point Total NUREG-1021, Revision 11 RO Page 13 of 13 Form ES-401-3 RO IR 3.8 4.3 3.5 2.9 3.5 3.4 4.1 FENOC Facsimile Rev. 0
ES-401 PWR Examination Outline Form ES-401-2 Facility:
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 1
1 3
Tier 5
3 8
Totals
- 3. Generic Knowledge and 1 2 3 4 7
Abilities Category 2 2 1 2 Note:
- 1.
En.sure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the "Tier Totals" in each KIA category shall not be less than two). (One Tier 3 radiation control KIA is allowed if it 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 NRG revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the 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 K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.
- 7.
The generic (G) Kl As in Tiers 1 and 2 shall be selected from Section 2 of the KIA catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KIAs.
- 8.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics' 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, I Rs, and poliiftotals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 1 O CFR 55.43.
Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
- These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the KIA catalo is used to develo the sam le Ian.
NUREG-1021, Revision 11 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)
=
I E/APE #/Name I Safety Function K
K K
A A :G KIA Topic(s)
IR 1
2 3
1 2 000007 (EPE 7; BW E02&E10; CE E02)
)
EA2 Ability to determine or interpret the following as 4.6 Reactor Trip, Stabilization, Recovery I 1 they apply to a reactor trip:
[Question 76]
1./ \\
EA2.04 If reactor should have tripped but has not done so, manually trip the reactor and carry out
}
(CFR 41. 7 / 45.5 / 45.6)
H Ii
- ,._(
000011 (EPE 11) Large Break LOCA I 3 Ji,,, ;
EA2 Ability to determine or interpret the following as 4.0
[Question 77]
they apply to a Large Break LOCA:
f<.* ****.
EA2.14 Actions to be taken if limits for PTS are IY violated (CFR 43.5 / 45.13) 1)-*.* i*('
011EA2.14 000025 (APE 25) Loss of Residual Heat t;f AA2. Ability to determine and interpret the following 3.4 Removal System / 4 I*,.
as they apply to the Loss of Residual Heat Removal System:
[Question 78]
If I,,:****
AA2.06 Existence of proper RHR overpressure I ;,,
protection IC:
(CFR: 43.5 / 45.13)
I>'
1--
OL~ JL6 (APE 26) Loss of Component
)
2.4.47 Ability to diagnose and recognize trends in an 4.2 Cooling Water / 8 accurate and timely manner utilizing the appropriate
[Question 79]
control room reference material.
(CFR: 41.10 / 43.5 / 45.12)
- .- ~,0056 (APE 56) Loss of Offsite Power I 6
)
2.4.50 Ability to verify system alarm setpoints and 4.0 I [Question 80]
operate controls identified in the alarm response manual.
(CFR: 41.10 / 43.5 / 45.3)
} ',-\\:
000062 (APE 62) Loss of Nuclear Service
. )( 2.1.28 Knowledge of the purpose and function of 4.1 Water I 4 major system components and controls.
[Question 81]
(CFR: 41.7) 3 3 6
KIA Category Point Totals:
Group Point Total:
NUREG-1021, Revision 11 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) c;,'/":i :,:,,
A J'l\\ r*
E/APE # / Name I Safety Function K K K
'.~~
KIA Topic(s)
IR 1
2 3
1 }2
','.,, ~ '.
000003 (APE 3) Dropped Control Rod / 1
)
2.4.46 Ability to verify that the alarms are consistent 4.2 with the plant conditions.
[Question 82]
J (CFR: 41.10 / 43.5 / 45.3 / 45.12) 000028 (APE 28) Pressurizer (PZR) Level
.,,* )
2.4.50 Ability to verify system alarm setpoints and 4.0 Control Malfunction / 2 operate controls identified in the alarm response
_- >{
manual.
[Question 83]
C1:.(
't l; (CFR: 41.10 / 43.5 / 45.3)
- ,.'.~'.~
~ 000068 (APE 68; BVI/ A06) Control Room
'.) Ii: M2. Ability to determine and interpret the following 4.3 II i;:va~i.;t.tior1 i 8 Ii!
as they apply to the Control Room Evacuation:
I
}
[Question 84]
M2.06 RCS pressure
_:,,. i*'.,.
(CFR: 43.5 / 45.13)
I>_;. I'..
1-,*;,
(BW E09; CE A13**; W E09 & E10) Natural jl EA2. Ability to determine and interpret the following 3.9 Circulation/4 as they apply to the (Natural Circulation with Steam
[Question 85]
Void in Vessel with/without RVLIS)
EA2.2 Adherence to appropriate procedures and j!
operation within the limitations in the facility*s license
' t and amendments.
- Y.;,
(CFR: 43.5 / 45.13)
I =
KIA Category Point Totals:
2 2 Group Point Total:
4
ES-401 PWR Examination Outline Form ES-401-2 Plant Systems
- Tier 2/Group 1 (SRO)
- \\;~;
i System # I Name K K K K
K K A *A A A, (,3 KIA Topic(s)
IR 1 2 3
4 5
6 1 \\2 3 4:/
003 (SF4P RCP) Reactor r\\J'
- ~ 2.4.30 Knowledge of events related to system 4.1 Coolant Pump operation/status that must be reported to I
1,} internal organizations or external agencies,
[Question 86]
such as the State, the NRC, or the transmission system operator.
(CFR: 41.10/43.5/45.11) 007 (SF5 PRTS) Pressurizer
)(
A2 Ability to (a) predict the impacts of the 2.9 Relief/Quench Tank
- . following malfunctions or operations on the P S; and (b) based on those predictions, use
[Question 87]
?-,
\\J' procedures to correct, control, or mitigate the
. :' consequences of those malfunctions or operations:
',s.,_
I.,
+:
A2.04 Overpressurization of the waste gas vent header ii II (CFR: 41.5 I 43.5 I 45.3 I 45.13) t*
'C
)(
010 (SF3 PZR PCS) Pressurizer 2.2.38 Knowledge of conditions and limitations 4.5 Pressure Control in the facility license.
[Question 88]
(CFR: 41. 7 / 41.10 / 43.1 / 45.13) o:
,F5 CSS) Containment
)(
. A2 Ability to (a) predict the impacts of the 3.7 S1-j following malfunctions or operations on the
[Question 89]
CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the
. consequences of those malfunctions or I*
operations:
,,,' t
,i A2.08 Safe securing of containment spray when it can be done)
(CFR: 41.5 / 43.5 / 45.3 / 45.13) 1, 073 (SF7 PRM) Process
)(
A2 Ability to (a) predict the impacts of the 3.2 Radiation Monitoring following malfunctions or operations on the PRM system; and (b) based on those l
[C,'uestion 90]
predictions. use procedures to correct, control, or mitigate the consequences of those I
malfunctions or operations:
A2.02 Detector failure (CFR: 41.5 / 43.5 / 45.3 / 45.13) 3 2
5
, KIA Category Point Totals:
Group Point Total:
IJUREG-1021. Revision 11 SRO Page 4 of 6 FENOC Facsimile Rev. 0
ES-401 PWR Examination Outline Form ES-401-2 I
Plant Systems
- Tier 2/Group 2(SRO) a-System # / Name K K K K K K A A A A.G KIA Topic(s)
I 1
2 3
4 5
6 1 2 3
4 ~.
IR I'.* '
034 (SF8 FHS) Fuel-Handling
)(
r:.,,. A 1 Ability to predict and/or monitor changes in 3.2 parameters (to prevent exceeding design limits)
Equipment
(
- ,,c, associated with operating the Fuel Handling
[Question 91)
System controls including:
l.*J A 1.01 Load limits
- ti.
.'\\ (CFR: 41.5 / 45.5) 072 (SF7 ARM) Area Radiation
.*.. ~ 2.2.44 Ability to interpret control room 4.4 Monitoring indications to verify the status and operation of
[Question 92) a system, and understand how operator actions and directives affect plant and system conditions.
(CFR: 41.5 / 43.5 / 45.12) 086 Fire Protection
,?
I{;
A2 Ability to (a) predict the impacts of the 2.9
[Question 93) following malfunctions or operations on the Fire Protection System; and (b) based on those I
I\\>
predictions, use procedures to correct, control, I
or mitigate the consequences of those I
malfunctions or operations:
A2.03 Inadvertent actuation of the FPS due to circuit failure or welding (CFR: 41.5 / 43.5 / 45.3 / 45.13)
KIA Category Point Totals:
1 1 1 Group Point Total:
3 NUREG-1021, Revision 11 SRO Page 5 of 6 FENOC Facsimile Rev. 0
ES 401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Facility: Beaver Valley Unit 1 1LOT18 Date of Exam: 9/17 through 9/28/2018
- I Category KIA#
Topic SRO Only IR
!I
- 1.
G2.1.32 2.1.32 Ability to explain and apply system limits and precautions.
4.0 Conduct of Operations (CFR: 41.10 / 43.2 / 45.12)
[Question 94]
G2.1.4 2.1.4 Knowledge of individual licensed operator responsibilities related 3.8 to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 10CFR55, etc.
(CFR: 41.10 / 43.2)
[Question 95]
Subtotal 2
- 2.
G2.2.17 2.2.17 Knowledge of the process for managing maintenance activities Equipment during power operations, such as risk assessments, work prioritization, Control and coordination with the transmission system operator.
(CFR: 41.10 / 43.5 / 45.13)
[Question 96]
G2.2.35 2.2.35 Ability to determine Technical Specification Mode of Operation.
4.5 (CFR: 41.7 /41.10/43.2/45.13)
[Question 97]
Subtotal 2
- 3.
G2.3.7 2.3.7 Ability to comply with radiation work permit requirements during Radiation normal or abnormal conditions.
Control (CFR: 41.12 / 45.10)
[Question 98]
Subtotal 1
- 4.
G2.4.1 2.4.1 Knowledge of EOP entry conditions and immediate action steps.
Emergency (CFR: 41.10 / 43.5 / 45.13)
Procedures/
Plan
[Question 99]
G2.4.9 2.4.9 Knowledge of low power/shutdown implications in accident (e.g.,
4.2 loss of coolant accident or loss of residual heat removal) mitigation strategies.
(CFR: 41.10 / 43.5 / 45.13)
[Question 100]
Subtotal 2
- :r: er 1 Point Totai 7
NUREG-1021, Revision 11 SRO Page 6 of 6 FENOC Facsimile Rev. 0
ES-401 R ecor d f R. t d KIA 0
eJec e s
F orm ES-401-4 Facility: Beaver Valle~ Unit 1 1LOT18 Date of Exam: 9-17 through 9/28/201u Tier/
Randomly Reason for Rejection Group Selected KIA 1/1 000025 Question 6; Residual Heat Removal (RHRS} is not used as Low Pressure Injection at Beaver Valley, therefore there is no tie between RWST and the RHR M1.22 system. NRC Chief Examiner randomly selected M1.08 as a replacement.
1/2 000036 Question 20; There are no RO tasks performed outside the main control room during an emergency and the resultant operational effects during a Fuel-Handling 2.4.34 Incident. NRC Chief Examiner randomly selected 2.4.31 as a replacement.
1/2 000037 Question 21; There are no automatic actions associated with high radioactivity in S/G sample lines at BV1. NRC Chief Examiner randomly selected AK3.05 as a AK3.10 replacement.
2/1 022 Question 38; BV1 does not operate containment cooling based on containment pressure. NRC Chief Examiner randomly selected A1.01 as a replacement.
A1.02 2/1 073 Question 51; Reselect due to oversampling/duplicate KIA with question 90. NRC Chief Examiner randomly selected A2.01 as a replacement.
A2.02 2/2 028 Question 59; Unable to write a discriminatory question on Hydrogen purge control, and the Hydrogen Recombiners are Retired In Place. NRC Chief A1.02 Examiner randomly selected system 071 and A1.06 as a replacement.
1/1 000025 Question 78; Residual Heat Removal (RHRS} is not used as Low Pressure Injection at Beaver Valley, therefore there are no limitations on LPI flow and AA2.05 temperature rates of change. NRC Chief Examiner randomly selected AA2.06 as a replacement.
1/1 000026 Question 79; Loss of Component Cooling Water has no immediate actions required and the KIA is RO required knowledge. NRC Chief Examiner randomly 2.4.49 selected 2.4.47 as a replacement.
1/2 000003 Question 82; Dropped Control Rod has no ties to the generic KIA, status of the safety functions, therefore we cannot write a discriminatory question. NRC Chief 2.4.21 Examiner randomly selected 2.4.46 as a replacement.
2/1 026 Question 89; We cannot write a discriminatory question regarding Containment Spray and the response to Reflux boiling pressure spike when first going on A2.01 recirculation. NRC Chief Examiner randomly selected A2.08 as a replacement.
1/1 062 Question 81; Unable to write a discriminatory question regarding Loss of Nuclear Service Water - Knowledge of the purpose and function of major 2.1.28 system components and controls. NRC Chief Examiner randomly selected 2.1.23 as a replacement.
NUREG-1021, Revision 11 FENOC Facsimile Rev. 1
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Beaver Valley Unit 1 Examination Level RO IXI SROD Date of Examination: 9/17 thru 9/28, 2018 Operating Test Number: BV1LOT18 NRC Administrative Topic (See Note)
Conduct of Operations (RO A 1.1)
Conduct of Operations (RO A 1.2)
Equipment Control (RO A2)
Radiation Control (RO A 3)
Emergency Plan (RO A4)
Type Code*
D,R M,R D,R D,R Describe activity to be performed 2.1.7 (4.4)
Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
Calculate The RCS Initial Void Volume And Final Void Volume (IAW 1 OM-6.4.Q, "Response To Void In Reactor Vessel") (RO)
Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 1 OCFR55, etc.
Determine if License Status is Maintained Active (RO)
Ability to determine operability and/or availability of safety related equipment.
JPM 1AD-027 Complete Surveillance of RHR Pump (RO) 2.3.11 (3.8)
Ability to control radiation releases.
JPM 1AD-010 NOT EVALUATED Determine GW Storage Tank Discharge Bleed Flow Rate (RO)
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking on topics (which would require all 5 items).
- Type Codes & Criteria NUREG-1021, Revision 11 (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 0
Administrative Topics Outline
,,-~'-'-'-"'~*=========================================
ES-301 Form ES-301-1 Facility: Beaver Valley Unit 1
- amination Level RO D SRO [BJ Date of Examination: 9/17 thru 9/28, 2018 Operating Test Number: BV1LOT18 NRC Administrative Topic (See Note)
Conduct of Operations (SRO A 1.1)
Conduct of Operations (SRO A 1.2)
-1Uipment Control (SRO A 2)
Radiation Control (SRO A 3)
~mergency Pian 11 (SRO A 4)
Type Code*
N,R M,R D,R D,R D,R Describe activity to be performed 2.1.7 (4.7)
Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
Review The RCS Initial Void Volume And Final Void Volume (IAW 10M-6.4.Q, "Response To Void In Reactor Vessel") and then Determine the Vent Time. (SRO)
Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 1 OCFR55, etc.
Evaluate Operators Work History to Determine if License Status is Active (SRO)
Ability to determine operability and/or availability of safety related equipment.
Review/Approve Completed Surveillance of RHR Pump (SRO)
Ability to control radiation releases.
JPM 1AD-023 Review/Approve LW Discharge (SRO) 2.4.30 (4.1)
Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.
JPM 1AD-020 Determine Event Notification Times (SRO)
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 NUREG-1021, Revision 11 (C)ontrol Room, (S)imulator, or Class{R)oom (D)irect from bank (:: 3 for ROs; _s 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (.:: 1)
(P)revious 2 exams (:: 1 ; randomly selected)
FENOC Facsimile Rev 0
ES 301 C
t IR on ro
/I Pl t S t
oom n-an
,ys ems 0 tr u me F orm ES 301 2 Facility: Beaver Valle~ Unit 1 Date of Examination: 9/17 thru 9/28, 2018 Exam Level: RO 00 SRO(I) D SRO(U) D Operating Test No.: BV1LOT18 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U)
System/ JPM Title Type Code*
Safety Function S1 - Raise Reactor Power To 10-a Amps (1CR-640)
A,D,L,S 1
S2 - Align SI Pumps for Hot/Cold Leg Recirculation (1CR-570)
A, D, EN, L, 2
s S3 - Depressurize the RCS During Natural Circulation Cooldown A,D,L,S 3
(1CR-554)
S4 - Isolate a Faulted Steam Generator via Att 1.2-Y AFW Valve Failure A, N,L,S 4S (1CR-715 NEW)
S5 - Verify CIB Valve Alignment (1CR-045)
D, EN, L, S 5
S6 - Synchronize and Load EOG No. 2 (1CR-024)
D, L, S 6
S7 - Respond to an Intermediate Range Malfunction (1CR-106)
D, L,S 7
SB - Respond to a Loss of Secondary Component Cooling Water (1CR-A,N,S 8
557 NEW)
In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
P1 - Startup the Wide Range Hydrogen Analyzer (1 PL-047)
D,E 5
P2 - AMSAC System Trouble - FT Failure (1 PL-148)
D,E 7
P3 - Verification of Cold Leg Recirculation Components (1 PL-028)
D,E,R 2
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 / SRO-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
- 11;
- ::11;:::1 (EN)gineered safety feature
- ,
- 1 / ~ 1 / ~ 1 { control room system)
(L)ow-power / Shutdown
~1,~11;:::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
- 11~11;
- ::1 (S)imulator NU RE G-1021, Revision 11 FENOC Facsimile
ES 301 C
t IR on ro
/I Pl t S t oom n-an
,ys ems Otl' u me F orm ES 301 2 Facility: Beaver Valle~ Unit 1 Date of Examination: 9/17 thru 9/28, 2018 JI Exam Level: RO D SRO(I) 000 SRO(U) D Operating Test No.: BV1LOT18 NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U)
System/ JPM Title Type Code*
Safety Function S1 - Raise Reactor Power To 10-s Amps (1CR-640)
A, D, L, S 1
S2 -Align SI Pumps for HoUCold Leg Recirculation (1CR-570)
A, D, EN, L, 2
s S3 - Depressurize the RCS During Natural Circulation Cooldown A,D,L,S 3
(1CR-554)
S4 - Isolate a Faulted Steam Generator via Att 1.2-Y AFW Valve Failure A,N,L,S 4S (1CR-NEW)
S6 - Synchronize and Load EOG No. 2 ( 1 CR-024)
D, L, S 6
S7 - Respond to an Intermediate Range Malfunction (1CR-106)
D, L, S 7
SB - Respond to a Loss of Secondary Component Cooling Water A, N,S 8
(1CR-NEW)
In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
P1 - Startup the Wide Range Hydrogen Analyzer (1 PL-047)
D,E 5
P2 - AMSAC System Trouble - FT Failure (1 PL-148)
D,E 7
P3 - Verification of Cold Leg Recirculation Components (1 PL-028)
D,E,R 2
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 / SRO-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/.::1/.::1 (EN)gineered safety feature
.:: 1 / ~ 1 / ~ 1 ( control room system)
(L)ow-power / 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 (S)imulator NUREG-1021, Revision 11 FENOC Facsimile
C Scenario Outline II Facility:
BVPS Unit 1 Scenario No. 1 Op Test No.:
1LOT18 NRC Examiners:
Candidates:
(10): 100% power, BOL, Equ. XE Conditions, CB "D" @230 steps, RCS boron -
Conditions:
__ ppm.
Turnover:
Maintain 100% power.
PCV-1 RC-456 isolated due to seat leakage, block valve closed. TS 3.4.11, Condition A FW-P-3A Out of service, TS 3.7.5, Condition B Critical Tasks:
- 1. CT-11 (E-0.0) Close CNMT isolation valves
- 2. CT-16 (E-1.C) Stop RCP's
- 3. CT-51 (FR-S.1.B) Start Auxiliary Feedwater pumps
- 4. CT-52 (FR-S.1.C) Initiate negative reactivity Malf. No.
Event Type Event Description Event j
,i l\\T-.
... ~-......
Ii 1
AUXl3G (0 0)
(C) BOP, SRO VS-F-4A spurious trip, requires BOP to start VS-F-4B.
(TS) SRO 2
PRS06A (0 0) I 00 60 (I) ATC, SRO LT-lRC-459 drifts high, requires ATC to place alternate (TS) SRO channel in service..
3 TURl5 (0 0) 78 10 (C) ATC, SRO Turbine valve position limiter fails low, causes - 100 mw (TS) SRO load reduction. ATC required to Borate RCS.
4 FWM09C (7 0) 25 0 (C) BOP, SRO "C" SG Feedwater valve, FCV-1 FW-498, begins oscillating, requiring BOP to manually control level.
5 TUR03E (0 0) 15 8 (C) BOP, SRO Turbine high bearing vibration requires crew to manually trip the unit.
6 CRF12A (M) ALL Failure of Automatic and manual Reactor trip from the CRF12B control room requires entry into FR-S.1.
7 CRF02A (5 0)
(C) ATC, SRO Control Rod automatic insertion failure, A TC must manually insert control rods.
INH20 8
INH21 (C) BOP, SRO All Aux Feedwater pumps fail to automatically start, requires INH35 BOP to start AFW pumps.
INH36 9
SIS08 (M) ALL 1200 gpm LOCA occurs on "B" Loop 10 INH49 (C) BOP, SRO Train "A" CIA fails to automatically actuate.
11 VLV-SEAlO (0 0) 100 (C) ATC, SRO MOV-lCH-381 fails to automatically close on CIA signal (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal
~ FR-S.l -
E E ES-1.2 NUREG-1021, Revision 11 FENOC Facsimile Rev. 0
Appendix D 1 L 18N1 Scenario Outline
'ter taking the shift at 100% power, BOL, Leak Collection Exhaust fan, 1 VS-F-4A, will trip, the crew will
' appficable TS and License Requirements Manual.
LT-lRC-459 will then drift high. The ATC will recognize the failure and respond IAW AOP 1.4.1 IOA's to remove the failed channel from service. The SRO will transition to the Instrument Failure procedure, 1 OM-6.4.IF and direct the A TC complete the removal of the 459 channel from service. and then review applicable Technical Specifications in effect for the failed level transmitter.
A malfunction will then occur with the turbine valve position limiter causing a load rejection, in response to the load rejection, control rods will auto insert bringing in the Bank D low alarm, requiring the ATC to Borate the RCS and withdraw control rods to clear the alarm.
At the same time, a malfunction will occur with the "C" main feed regulating valve, FCV-lFW-498, causing oscillations in the "C" SG level requiring the BOP to manually stabilize and control level.
The SRO will enter AOP 1.35.2, "Load Rejection," to stabilize the plant and address DNB technical specifications A main turbine bearing #5 will exhibit high vibrations, at 15 mils the ARP for A 7-104, probable cause 5 will require a unit trip.
The ATC will unsuccessfully attempt to trip the reactor from BB-Band BB-A.
The SRO will enter FR-S.1 with the ATC and BOP performing the IOA's.
i ~--..11i1ule after Emergency boration flow is established in FR-S.l, if the crew previously dispatched an operator locally trip the reactor, the reactor will be locally tripped. The A TC will verify reactor power is <5% after
.nich the SRO will return to E-0.
1 min. after the local Rx trip, a 1200 gpm LOCA will occur on the B loop resulting in an automatic SI actuation.
Additional malfunctions that occur during the ATWS condition are that all available Aux feed water pumps fail to automatically start, all can be manually started. (1FW-P3A was OOS on turnover)
The safety injection that occurred will fail to actuate the train "A" CIA signal, and MOV-lCH-381 (a train "B" CIA valve) will fail to automatically close. The crew will be required to isolate the containment penetration via either manually actuating Train "A" CIA or manually closing MOV-1 CH-381.
After returning to E-0, the SRO will determine that the RCS is not intact and transition to E-1. The scenario will be terminated at the lead evaluators discretion after the crew transitions to ES-1.2 and initiates a RCS Cooldown to Cold Shutdown.
Expected procedure flow path is E FR-S.1 -
E E ES-1.2.
NUREG-1021, Revision 11 FENOC Facsimile Rev. 0
A d" D
,ooen 1x s
- o r cenario utme
~ Facility:
BVPS Unit 1 Scenario No. 2 Op Test No.:
1LOT18 NRC Examiners:
Candidates:
SRO i
1, Ii i;
ATC BOP Initial IC __ (29): 100% power, EOL, Equ. XE Conditions, CB "D" @ 230 steps, RCS boron -
Conditions:
__ ppm.
Turnover:
Maintain 100% power.
PCV-1 RC-456 isolated due to seat leakage, block valve closed. TS 3.4.11, Condition A 1FW-P-3A Out of service, TS 3.7.5, Condition B Control Rods are in manual due to a circuit malfunction, I&C is investigating the problem using work order instructions, automatic rod control is not available.
Critical Tasks:
- 1. E-0.D Manually actuate 1 train of Safety Injection
- 2. FR-H.1.A Establish feed flow to SG before Feed and Bleed is required
- 3. E-2.A Isolate faulted SG Event Malf. No.
Event Type Event Description No.
1 PRS12 (0 0) 85 45 (I) ATC, SRO Master Pressure controller drifts to 85%, requires A TC to ASlS (TS) SRO manually control RCS pressure.
2 IOR X12l027L (12 0)
(TS) SRO Auto Stop Oil Pressure Switch, PS-1 TB-417, fails ON "A" Station air compressor trips, "B" fails to auto start - BOP 3
AUX02A (C) BOP, SRO manually starts "B" air compressor, Diesel Air compressor fails to auto start but will start locally, 4
MSS18C (0 0) 2.5E5 (R) ATC "C" SG Atmospheric valve fails open causes Rx overpower, 300 0 (N) BOP, SRO requires power reduction.
CRF04BV (2 2) I 2 Rods drop during power reduction - requires manual Rx 5
CRF04BT (2 4) 1 (C) ATC, SRO trip.
6 TRG 3 'IMF (M) ALL "C" SG Steam Break in turbine building occurs on Rx trip -
MSS l 8C (0 0) 6E5' requires Safety Inj.
7 SISIOA (C) ATC, SRO Automatic SI fails to actuate - requires manual actuation.
SIS I OB 8
VLV-MSS18 100 (C) BOP, SRO "C" SG Mainsteam line isolation valve fails to auto close.
VLV-SGBI7 100 9
VLV-SGB18 100 (C) BOP, SRO SG Sample valves fail to auto close, TV-I SS-117 A,B,C VLV-SGB19 100 10 FWM11C(J30)
(M) ALL Loss of all Aux Feedwater flow, requires entry into FR-H.1 FWMI IA (0 0)
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal 0 --+ FR-H.1 --+ E-0 --+ E-2.
NUREG-1021, Revision 11 FENOC Facsimile Rev. 0
Appendix D 1L 18N2 Scenario Outline
~er taking the shift at 100% power, the Master Pressure controller will fail to 85% causing the PRZR spray
.ilves to open. IA W AOP 1.4.1, the A TC will close the spray valves to stabilize RCS pressure. The A TC will manually control PRZR pressure for the remainder of the scenario. The SRO will address Technical Specifications for DNB.
An Auto Stop Oil Pressure Switch, PS-I TB-417, will then fail, the crew will identify the failure and the SRO will evaluate applicable Technical Specifications for the failure.
- T!-ic "A'. station air compressor will trip with an auto start failure of the "B" station air compressor. The SRO will direct activities in accordance with AOP 1.34.1, "Loss of Station Instrument Air", the BOP will manually start the "B" station air compressor.
The "C" steam generator Atmospheric steam dump valve, PCV-1MS-101C, will fail open causing the RCS temperature to decrease and an increase in Rx power. The crew will recognize the increase in power and reduce power. If necessary, Operations management will direct the crew to lower Rx power to 90% while maintenance can attempt to isolate the flow path.
The crew will identify the failed valve and attempt to close the valve from the control room. After the failed closure attempt, the crew will dispatch an operator to locally isolate or take actions to locally close the valve, all attempts to isolate the leak will not be successful.
The A TC will insert the control rods in response to the turbine load reduction, when bank D lowers to less than 215 steps, 2 rods will drop during the rod insertion, the immediate actions of AOP 1.1.8, "Rod Inoperability",
- ill be taken and the reactor will be manually tripped due to more than 1 rod being dropped. When the reactor
, tripped, the "A", "B" and "AE" 4kv buses will become de-energized on the transfer to offsite power.
Upon the Rx trip, a steam header break will occur in the turbine building. An automatic steamline isolation will occur, however the "C" Mainsteam line isolation valve will fail to automatically close requiring the BOP to mm;n1ally close it. -lhe fault will also result in an SI being required, however, Safety Injection will not automatically actuate, requiring manual actuation to initiate SI flow.
Aux Feedwater malfunctions will occur such that the turbine driven pump, FW-P-2 trips on start, FW-P-3B will not start and 1FW-P-3A was OOS on turnover.
The crew will enter E-0 on the reactor trip, and then enter FR-H-1 due to no auxiliary feed water being available. After Feedwater has been established using either the dedicated Feedwater pump, FW-P-4 or either Main feed pump, the crew will return to E-0 and progress to diagnose the "C" SG as being faulted and enter E-2 to isolate the "C" SG.
The scenario will be terminated when the crew determines transition to ES-1.1 is appropriate.
Expected procedure flow path is E FR-H.1 -
E E-2.
NUREG-1021, Revision 11 FENOC Facsimile Rev. O
A d" D
,ppen 1x s
- o r cenano utme 11 Facility:
BVPS Unit 1 Scenario No. 3 Op Test No.:
1LOT18 NRC Examiners:
Candidates:
SRO ATC BOP Initial IC_: -4% power, EOL, Xe increasing, CB "D" @_ steps, RCS boron -__ ppm.
Conditions:
!: Jumover:
Continue power increase IA W reactivity plan and commence turbine roll.
PCV-1 RC-456 isolated due to seat leakage, block valve closed. TS 3.4.11, Condition A Critical Tasks:
- 1. E-3.A Isolate Ruptured SG
- 2. E-3.B Establish/maintain RCS temperature
- 3. ECA-3.3.A Terminate Safety Injection Event Malf. No.
Event Type Event Description No.
1 NIA (R) ATC Raise power (N) SRO 2
NIA (N) BOP, SRO Startup "B" EHC pump, 1LO-M-9B, shutdown "A" EHC pump, 1LO-M-9A.
3 BST-CCW006 1 (C) BOP, SRO "A" CCR pump trips, Auto start failure of "B" CCR pump.
CCW3A (TS) SRO 4
CND01 (0 0) 100 (C) BOP, SRO MOV-1 CN-105 spuriously opens 5
PRS08E (8 0) 2500 15 (I) ATC, SRO PT-lRC-445 fails high, PORV's 455D & 456 open, ATC Ii (TS) SRO required to manually close PORV, PCV-1RC-455D 6
RCSlOB (0 0) 30 (C) ATC, SRO "B" RCP high vibration and trip - will require manual Rx trip
.. RCS08R (7 3)
- I 7
RCS03B (1 0) 450 (M)-ALL "B" SG - 425 gpm tube rupture 8
VLV-SGBOI,02,03 (C) BOP, SRO SG BD isolation failure, requires manual valve closure.
Steam dump, PCV-1MS-106A fails open following cooldown 9
MSS08A (0 0) 100 (C) BOP, SRO during E-3. Crew required to isolate steam lines and control RCS temperature via atmospheric steam dumps PRS09A (2 120) 0 PRZR spray valves and remaining PORV fail to open during 10 PRS09B (2 120) 0 (C) ATC, SRO depressurization in E-3, will require transition to ECA-3.3 IOR X07I0970 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal E E ECA-3.3 NUREG-1021, Revision 11 FENOC Facsimile Rev. 0
Appendix D 1 L 18N3 Scenario Outline
-*1e crew will assume the shift at -4% power with instructions to raise power in accordance with the reactivity 1_,an and lOM-52.4.A., Additionally, the turnover includes direction to place the "B" EHC pump, 1LO-M-9B, in service and place the "A" EHC pump, 1LO-M-9A in standby.
After the power increase is commenced and Mode 1 is declared. The "A" CCR pump will then trip due to a faulty breaker with a failure of the "B" to auto start. The crew will enter AOP 1.15.1. The BOP will manually start the "B"; the crew will dispatch an operator to place the "C" pump in service on the "AE" 4kv bus. The SRO will then address Technical Specifications.
The Condenser Hotwell Hi Level Bypass valve, MOV-lCN-105, will then spuriously open causing condenser hotwell level to drop. The crew will respond to the low level alarm using the Alarm Response procedure and, diagnose, identify and close MOV-lCN-105.
PT-lRC-445 will fail high causing PORV 455D and 456 to open, (per turnover, 456 previously isolated with block valve closed.) The ATC will be required to manually close PORV, PCV-1RC-455D IAW AOP 1.4.1 immediate operator actions. The US will enter AOP 1.4.1 and then transition to 1 OM-6.4.IF, Attachment 2 and determine applicable Tech Spec actions.
The "B" RCP will then show signs of high vibration, the crew will respond using AOP 1.6.8, "Abnormal RCP Operation". After diagnosing and monitoring, the vibration will increase in severity to the point where the RCP will trip. Since the plant is less than 30% power, the RCP trip will not cause an automatic Rx trip. The crew will identify the loss of the RCP and manually trip the reactor.
_ _, a result of the reactor trip a 425 gpm SGTR occurs on the "B" SG. The crew will progress through E-0 and diagnose the "B" SG as ruptured and transition to E-3. While the crew is isolating the "B" SO, the BOP will identify that the Blowdown valve will not close and procedurally close the backup cnmt isolation valve.
Following the cooldown to target temperature, Condenser Steam Dump valve, PCV-1MS-106A will fail open, the crew will identify the failure and isolate the mainsteam lines and stabilize temperature using the "A" and "C" SG atmospheric steam dump valves.
When the crew attf"mpts to depressurize the RCS, the spray valves will not function, nor will the PRZR PORV's, 456 was previously isolated on turnover-block valve will not open, 455D CS was placed in "CLOSE" per event 5, valve will not open. 455C will fail to open via control switch, crew will then transition to ECA-3.3.
The scenario is terminated when the crew establishes a normal charging flow path in ECA-3.3.
Expected procedure flow path is E-0-. E-3-. ECA-3.3.
NUREG-1021, Revision 11 FENOC Facsimile Rev. 0