ML081260324
Jump to navigation
Jump to search
| ML081260324 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 04/02/2008 |
| From: | NRC/RGN-II |
| To: | Virginia Electric & Power Co (VEPCO) |
| References | |
| 50-280/08-301, 50-281/08-301 | |
| Download: ML081260324 (123) | |
See also: IR 05000280/2008301
Text
{{#Wiki_filter:Draft Submittal
(Pink Paper)
DRAFT Written Exam r6uality GRcklist (E£-401-6)
& Written Exam Sample Plan
SLt,eey 02 tJ08-M/
:DRA-F7
ES-401 , Rev. 9 PWR Examination Outline Form ES-401-2
Facility: >Ar/V Date of Exam: ~1f ~
t. //
I
RO KIA Category Points SRO-Only Points
Tier Group
K K K K K K A A A A G A2 G* Total
1 2 3 4 5 6 1 234 * Total
1. 1 3 3 3 4 3 2 18 2 4 6
Emergency & 2
2 1 1 2 0 3 9 2 2 4
Abnormal Plant N/A N/A
Evolutions Tier Totals 4 4 5 4 6 4 27 4 6 10
1 4 3 3 2 2 3 1 2 2 4 2 28 2 3 5
2.
2 1 0 0 2 1 1 1 1 1 1 1 10 2 1 3
Plant
Systems Tier Totals 5 3 3 4 3 4 2 3 3 5 3 38 4 4 8
3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7
Categories
3 3 2 2 2 2 1 2
1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO
and SRO-only outlines (l.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals"
in each KIA category shall not be less than two).
2. The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by +/-1 from that specified in the table
based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do
not apply at the facility should be deleted and justified; operationally important, site-specific systems that are
not included on the outline should be added. Refer to ES-401 , Attachment 2, 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.
8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance
ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter
the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other
than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note
# 1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs,
and point totals (#) on Form ES-401-3. Limit SRO selections to KIAs that are linked to 10 CFR 55.43.
ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
007EK2.02 Reactor Trip - Stabilization - Recovery 2.6 2.8 D~DDDDDDDDD Breakers, relays and disconnects
/1
Knowledge of electrical power supplies to
the following:(CFR: 41.7)
008AK2.01 Pressurizer Vapor Space Accident / 3 2.7 2.7 D~DDDDDDDDD Valves
Knowledge of electrical power supplies to
the following:(CFR: 41.7)
011EK1.01 Large Break LOCA / 3 4.1 4.4 ~ DDDDDDDDDD Natural circulation and cooling, including reflux boiling.
Knowledge of the physical connections
and/or cause-effect relationships between
(SYSTEM) and the following:(CFR: 41.2 to
41.9/45.7 to 45.8)
022AA2.03 Loss of Rx Coolant Makeup / 2 3.1 3.6 DDDDDDD ~ DDD Failures of flow control valve or controller
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 / 43.5 / 45.3 / 45.13)
026AA1.07 Loss of Component Cooling Water / 8 2.9 3 DDDDDD~DDDD Flow rates to the components and systems that are
serviced by the CCWS; interactions among the
Ability to predict and/or monitor changes in components
parameters associated with operating the
(SYSTEM) controls including:(CFR: 41.5/
45.5)
027AA1.01 Pressurizer Pressure Control System 4 3.9 0 DDDDD ~ DDDD PZR heaters, sprays, and PORVs
Malfunction / 3
Ability to predict and/or monitor changes in
parameters associated with operating the
(SYSTEM) controls including:(CFR: 41.5 /
45.5)
Page 1 of 3 4/2/2008 2:55 PM
ES*401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES*401*2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
029EK3.10 ATWS 11 4.1 4.1 0 0 ~ 0 0 0 0 0 0 0 0 Manual rod insertion
Knowledge of the effect that a loss or
malfunction of the (SYSTEM) will have on
the following:(CFR: 41.7 1 45.6)
038EK1.01 Steam Gen. Tube Rupture 13 3.1 3.4 ~ 0 0 0 0 0 0 0 0 0 0 Use of steam tables
Knowledge of the physical connections
and/or cause-effect relationships between
(SYSTEM) and the following:(CFR: 41.2 to
41.9/45.7 to 45.8)
040AA1.24 Steam Line Rupture - Excessive Heat 3.8 3.8 DOD D D D ~ D D D D Main steam header pressure gauges
Transfer 14
Ability to predict and/or monitor changes in
parameters associated with operating the
(SYSTEM) controls including:(CFR: 41.5 1
45.5)
054AK1.02 Loss of Main Feedwater 14 3.6 4.2 ~ 0 0 0 0 0 0 00 0 0 Effects of feedwater introduction on dry S/G
Knowledge of the physical connections
and/or cause-effect relationships between
(SYSTEM) and the following:(CFR: 41.2 to
41.9/45.7 to 45.8)
055EA2.05 Station Blackout 16 3.4 3.7 0 000 0 0 D ~ D D 0 When battery is approaching fully discharged
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 1 43.5 1 45.3 1 45.13)
056AA1.04 Loss of Off-site Power 16 3.2 3.1 D 0 0 D D D ~ 0 D D D Adjustment of speed of ED/G to maintain frequency and
voltage levels
Ability to predict and/or monitor changes in
parameters associated with operating the
(SYSTEM) controls including:(CFR: 41.5 1
45.5)
Page 2 of 3 4/2/2008 2:55 PM
ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
057AA2.07 Loss of Vital AC Inst. Bus / 6 3.3 3.5 D D D D D D D ~ D D D Valve indicator of charging pump suction valve from
RWST
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 / 43.5 / 45.3 / 45.13)
062AG2.1.32 Loss of Nuclear Svc Water / 4 3.8 4.0 D DOD D D D D D D ~ Ability to explain and apply all system limits and
precautions.
This is a Generic, no stem statement is
associated.
065AK3.03 Loss of Instrument Air I 8 2.9 3.4 D D ~ D D D D D D D D Knowing effects on plant operation of isolating certain
equipment from instrument air
Knowledge of the effect that a loss or
malfunction of the (SYSTEM) will have on
the following:(CFR: 41.7 / 45.6)
we04EG2.4.3 LOCA Outside Containment / 3 4.2 4.1 D D D D D D D D D D ~ Knowledge of annunciators alarms, indications or
response procedures
This is a Generic, no stem statement is
associated.
WE05EK3.2 Inadequate Heat Transfer - Loss of 3.7 4.1 D D ~ D D D D D D D D Normal, abnormal and emergency operating procedures
Secondary Heat Sink I 4 associated with (Loss of Secondary Heat Sink).
Knowledge of the effect that a loss or
malfunction of the (SYSTEM) will have on
the following:(CFR: 41.7 / 45.6)
WE 12EK2.2 Steam Line Rupture - Excessive Heat 3.6 3.9 D ~ D D D D D D D D D Facility's heat removal systems, including primary
Transfer /4 coolant, emergency coolant, the decay heat removal
Knowledge of electrical power supplies to systems and relations between the proper operation of
the following:(CFR: 41.7) these systems to the operation of the facility.
Page 3 of 3 4/2/2008 2:55 PM
ES-401, REV 9 T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
001AK2.06 Continuous Rod Withdrawal / 1 3 3.1 D ~ D D D D D D D D D T-ave.lref. deviation meter
Knowledge of electrical power supplies to
the following:(CFR: 41.7)
003AG2.4.1 Dropped Rod /1 4.6 4.8 D D D D D D D D D D ~ Knowledge of EOP entry conditions and immediate action
steps.
This is a Generic, no stem statement is
associated.
005AA2.01 Inoperable/Stuck Control Rod / 1 3.3 4.1 D D D D D D D ~ D D D Stuck or inoperable rod from in-core and ex-core NIS, in-
core or loop temperature measurements
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 / 43.5 I 45.3 I 45.13)
061AA2.05 ARM System Alarms / 7 3.5 4.2 DOD D D D D ~ D D D Need for area evacuation; check against existing limits
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 / 43.5 / 45.3 / 45.13)
068AK3.09 Control Room Evac. /8 3.9 4.4 D 0 ~ D D D D D D D D Transfer of the following to local control: charging pumps,
charging header flow control valve, PZR heaters and
Knowledge of the effect that a loss or boric acid transfer pumps
malfunction of the (SYSTEM) will have on
the following:(CFR: 41.7 / 45.6)
076AK3.05 High Reactor Coolant Activity I 9 2.9 3.6 D D ~ D D D D D D D D Corrective actions as a result of high fission-product
radioactivity level in the RCS
Knowledge of the effect that a loss or
malfunction of the (SYSTEM) will have on
the following:(CFR: 41.7 I 45.6)
Page 1 of 2 4/2/2008 2:55 PM
ES-401, REV 9 T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
we02EG2.4.3 SI Termination 1 3 4.2 4.1 D D D D D D D D D D ~ Knowledge of annunciators alarms, indications or
response procedures
This is a Generic, no stem statement is
associated.
WE13EK1.2 Steam Generator Over-pressure 14 3.0 3.3 ~ D D D D D D D D D D Normal, abnormal and emergency operating procedures
associated with (Steam Generator Overpressure).
Knowledge of the physical connections
and/or cause-effect relationships between
(SYSTEM) and the following:(CFR: 41.2 to
41.9 1 45.7 to 45.8)
WE16EA2.2 High Containment Radiation 19 3.0 3.3 D D D D D D D ~ D D D Adherence to appropriate procedures and operation
within the limitations in the facility's license and
Ability to (a) predict the impacts of the amendments.
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 1 43.5 1 45.3 1 45.13)
Page 2 of 2 4/2/2008 2 :55 PM
ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
003K3.01 Reactor Coolant Pump 3.7 4.0 0 0 ~ 0 0 0 0 0 0 D 0 RCS
Knowledge of the effect that a loss or
malfunction of the (SYSTEM) will have on
the following:(CFR: 41.7 1 45.6)
003K6.14 Reactor Coolant Pump 2.6 2.9 o~ooo Starting requirements
Knowledge of the effect that a loss or
malfunction of the following will have on
the (SYSTEM):(CFR: 41.7/45.7)
004A2.13 Chemical and Volume Control 3.6 3.9 0 0 0 0 0 0 0 ~ 0 0 0 Low RWST
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.51 43.5 1 45.3 1 45.13)
004K5.15 Chemical and Volume Control 3.3 3.5 0 0 0 0 ~ 0 0 0 0 0 0 Boron and control rod reactivity effects as they relate to
MTC
Knowledge of the operational implications
of the following concepts as they apply to
the (SYSTEM):(CFR: 41.5/45.7)
004K6.13 Chemical and Volume Control 3.1 3.3 ooooo~oo 00 Purpose and function of the boration/dilution batch
controller
Knowledge of the effect that a loss or
malfunction of the following will have on
the (SYSTEM):(CFR: 41.7 1 45.7)
005K2.03 Residual Heat Removal 2.7 2.8 0 ~ 0 0 0 0 0 0 D D RCS pressure boundary motor-operated valves
Knowledge of electrical power supplies to
the following:(CFR: 41.7)
006A3.05 Emergency Core Cooling 4.2 4.3 0 0 0 0 0 0 ~ 0 0 Safety Injection Pumps
Ability to monitor automatic operations of
the (SYSTEM) including:(CFR: 41.7 1 45.5)
Page 1 of 4 4/2/2008 2:55 PM
ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
007K4.01 Pressurizer Relief/Quench Tank 2.6 2.9 D D D ~ D D D D D D D Quench tank cooling
Knowledge of (SYSTEM) design feature(s)
and or interlock(s) which provide for the
following:(CFR: 41.7)
008A1.04 Component Cooling Water 3.1 3.2 D D D D D D ~ D D D D Surge tank level
Ability to predict and/or monitor changes in
parameters associated with operating the
(SYSTEM) controls including:(CFR: 41.5/
45.5)
008K1.05 Component Cooling Water 3.0 3.1 ~ D D D D D D D D D D Sources of makeup water
Knowledge of the physical connections
and/or cause-effect relationships between
(SYSTEM) and the following:(CFR: 41.2 to
41.9/45.7 to 45.8)
010K6.03 Pressurizer Pressure Control 3.2 3.6 D D D D D ~ D D D D D PZR sprays and heaters
Knowledge of the effect that a loss or
malfunction of the following will have on
the (SYSTEM):(CFR: 41.7 / 45.7)
012G2.1.27 Reactor Protection 3.9 4 DDDDDDDDDD~ Knowledge of system purpose and or function.
This is a Generic, no stem statement is
associated.
013K2.01 Engineered Safety Features Actuation 3.6 3.8 D ~ D D D D D D D D ESFAS/safeguards equipment control
Knowledge of electrical power supplies to
the following:(CFR: 41.7)
022K1.01 Containment Cooling 3.5 3.7 ~ D D D D D D D D D D SWS/cooling system
Knowledge of the physical connections
and/or cause-effect relationships between
(SYSTEM) and the following:(CFR: 41.2 to
41.9/45.7 to 45.8)
Page 2 of 4 4/2/2008 2:55 PM
ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
026A4.01 Containment Spray 4.5 4.3 0 0 0 0 0 0 0 0 0 ~ 0 CSS controls
Ability to manually operate and/or monitor
in the control room:(CFR: 41.7 / 45.5 to
45.8)
039K5.01 Main and Reheat Steam 2.9 3.1 0 0 0 0 ~ 0 0 0 0 0 0 Definition and causes of steam/water hammer
Knowledge of the operational implications
of the following concepts as they apply to
the (SYSTEM):(CFR: 41.5/45.7)
059A4.03 Main Feedwater 2.9 2.9 0 0 0 D 0 0 0 D D ~ D Feedwater control during power increase and decrease
Ability to manually operate and/or monitor
in the control room:(CFR: 41.7 / 45.5 to
45.8)
059A4.10 Main Feedwater 3.9 3.8 0 0 0 D D D DOD ~ 0 ICS
Ability to manually operate and/or monitor
in the control room:(CFR: 41.7 / 45.5 to
45.8)
061K4.01 Auxiliary/Emergency Feedwater 3.1 3.3 0 0 0 ~ 0 0 0 000 Turbine trip, including overspeed
Knowledge of (SYSTEM) design feature(s)
and or interlock(s) which provide for the
following:(CFR: 41.7)
062A3.05 AC Electrical Distribution 3.5 3.6 oooooo~oo Safety-related indicators and controls
Ability to monitor automatic operations of
the (SYSTEM) including:(CFR: 41.7 / 45.5)
063A2.01 DC Electrical Distribution 2.5 3.2 0 0 0 0 0 0 0 ~ 0
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5/ 43.5 / 45.3 / 45.13)
Page 3 of 4 4/2/2008 2:55 PM
ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
064A4.06 Emergency Diesel Generator 3.9 3.9 0 0 0 0 0 0 0 0 0 ~ 0 Manual start, loading and stopping of the ED/G
Ability to manually operate and/or monitor
in the control room:(CFR: 41.7 1 45.5 to
45.8)
.064K3.03 Emergency Diesel Generator 3.6 3.9 o~ooooo 00 ED/G (manual loads)
Knowledge of the effect that a loss or
malfunction of the (SYSTEM) will have on
the following:(CFR: 41.7 1 45.6)
073K1.01 Process Radiation Monitoring 3.6 3.9 ~ 0 0 0 0 0 0 0 00 0 Those systems served by PRMs
Knowledge of the physical connections
and/or cause-effect relationships between
(SYSTEM) and the following:(CFR: 41.2 to
41.9/45.7 to 45.8)
076K1.12 Service Water 1.9 2.1 ~ 0 0 0 0 0 0 0 0 0 0 Intake screen system
Knowledge of the physical connections
and/or cause-effect relationships between
(SYSTEM) and the following:(CFR: 41.2 to
41.9/45.7 to 45.8)
078K2.01 Instrument Air 2.7 2.9 D ~ D D D D D D DOD Instrument air compressor
Knowledge of electrical power supplies to
the following:(CFR: 41.7)
103G2.2.12 Containment 2.9 4.0 0 0 0 0 0 0 0 0 0 ~ Ability to apply Technical Specifications for a system
This is a Generic, no stem statement is
associated.
103K3.02 Containm ent 3.8 4.2 0 0 ~ 0 0 0 0 0 0 0 0 Loss of containment integrity under normal operations
Knowledge of the effect that a loss or
malfunction of the (SYSTEM) will have on
the following:(CFR: 41.7 1 45.6)
Page 4 of 4 4/2/2008 2:55 PM
~
ES-401, REV 9 T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 Ai A2 A3 A4 G TOPIC:
RO SRO
001G2.1.27 Control Rod Drive 3.9 4 D D D D D D D- D D D ~ Knowledge of system purpose and or function.
This is a Generic, no stem statement is
associated.
002K6.02 Reactor Coolant 3.6 3.8 DDDDD~DDDDD RCP
Knowledge of the effect that a loss or
malfunction of the following will have on
the (SYSTEM):(CFR: 41.7 / 45.7)
011A2.10 Pressurizer Level Control 3.4 3.6 DDDDDDD~DDD Failure of PZR level instrument - high
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 / 43.5 / 45.3 / 45.13)
015K5.10 Nuclear Instrumentation 2.8 3.0 DDDD~DDDDDD Ex-core detector operation
Knowledge of the operational implications
of the following concepts as they apply to
the (SYSTEM):(CFR: 41.5 / 45.7)
029A3.01 Containment Purge 3.8 4.0 DDDDDDDD~DD CPS isolation
Ability to monitor automatic operations of
the (SYSTEM) including:(CFR: 41.7 / 45.5)
035K1.01 Steam Generator 4.2 4.5 ~DDDDDDDDDD MFW/AFW systems
Knowledge of the physical connections
and/or cause-effect relationships between
(SYSTEM) and the following:(CFR: 41.2 to
41.9 / 45.7 to 45.8)
Page 1 of 2 4/2/2008 2:55 PM
ES-401, REV 9 T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
045A1.05 Main Turbine Generator 3.8 4.1 D D D D D D ~ D D D D Expected response of primary plant parameters
(temperature and pressure) following T/G trip
Ability to predict and/or monitor changes in
parameters associated with operating the
(SYSTEM) controls including:(CFR: 41.5 I
45.5)
071K4.04 Waste Gas Disposal 2.9 3.4 D D D ~ D D D D D D D Isolation of waste gas release tanks
Knowledge of (SYSTEM) design feature(s)
and or interlock(s) which provide for the
following:(CFR: 41.7)
079K4.01 Station Air 2.9 3.2 D D D ~ D D D D D D D Cross-connect with lAS
Knowledge of (SYSTEM) design feature(s)
and or interlock(s) which provide for the
following:(CFR: 41.7)
086A4.05 Fire Protection 3.0 3.5 D D D D D D D D D ~ D Deluge valves
Ability to manually operate and/or monitor
in the control room:(CFR: 41.7 / 45.5 to
45.8)
Page 2 of 2 4/2/2008 2:55 PM
ES-401, REV 9 T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
G2.1.1 Conduct of operations 3.8 4.2 DDDDDDDDDD ~ Knowledge of conduct of operations requirements.
This is a Generic, no stem statement is
associated.
G2.1.14 Conduct of operations 3.1 3.1 DDDDDDDDDD ~ Knowledge of criteria or conditions that require plant-wide
announcements, such as pump starts, reactor trip, mode
This is a Generic, no stem statement is changes, etc.
associated. .
G2.1.7 Conduct of operations 4.4 4.7 DDDDDDDDDD ~ Ability to evaluate plant performance and make
operational judgments based on operating
This is a Generic, no stem statement is characteristics, reactor behavior and instrument
associated. interpretation.
G2.2.13 Equipment Control 4.1 4.3 DDDDDDDDDD~ Knowledge of tagging and clearance procedures.
This is a Generic, no stem statement is
associated.
G2.2.22 Equipment Control 4.0 4.7 DDDDDDDDDD~ Knowledge of limiting conditions for operations and safety
limits.
This is a Generic, no stem statement is
associated.
G2.2.28 Equipment Control 2.6 3.5 DDDDDDDDDD~ Knowledge of new and spent fuel movement procedures.
This is a Generic, no stem statement is
associated.
G2.3.2 Radiation Control 2.5 2.9 DDDDDDDDDD~ Knowledge of the facility ALARA program
This is a Generic, no stem statement is
associated.
Page 1 of 2 4/2/2008 2:55 PM
ES-401, REV 9 T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
G2.3.4 Radiation Control 3.2 3.7 D D D D D D D D D D ~ Knowledge of radiation exposure limits under normal and
emergency conditions
This is a Generic, no stem statement is
associated.
G2.4.11 Emergency Procedures/Plans 4.0 4.2 D D D D D D D D D D ~ Knowledge of abnormal condition procedures.
This is a Generic, no stem statement is
associated.
G2.4.49 Emergency Procedures/Plans 4.6 4.4 D D D D D D D D D D ~ Ability to perform without reference to procedures those
actions that require immediate operation of system
This is a Generic, no stem statement is components and controls.
associated.
Page 2 of 2 4/2/2008 2:55 PM
ES-401, REV 9 SRO T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
007EG2.4.4 Reactor Trip - Stabilization - Recovery 4.5 4.7 DDDDDDDDDD~ Ability to recognize abnormal indications for system
/1 operating parameters which are entry-level conditions for
This is a Generic, no stem statement is emergency and abnormal operating procedures.
associated.
008AA2.26 Pressurizer Vapor Space Accident / 3 3.1 3.4 D D D D D D D ~ D D D Probable PZR steam space leakage paths other than
PORV or code safety
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 / 43.5 / 45.3 / 45.13)
038EG2.4.6 Steam Gen. Tube Rupture / 3 3.7 4.7 D D D D D D D D D D ~ Knowledge symptom based EOP mitigation strategies.
This is a Generic, no stem statement is
associated.
054AA2.04 Loss of Main Feedwater / 4 4.2 4.3 D D D D D D D ~ D D D Proper operation of AFW pumps and regulating valves
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 / 43.5 / 45.3 / 45.13)
055EG2.4.30 Station Blackout / 6 2.7 4.1 D D D D D D D D D D ~ Knowledge of events related to system operations/status
that must be reported to internal orginizations or outside
This is a Generic, no stem statement is agencies.
associated.
we05EG2.4.6 Inadequate Heat Transfer - Loss of 3.7 4.7 D D D D D D D D D D ~ Knowledge symptom based EOP mitigation strategies.
Secondary Heat Sink / 4
This is a Generic, no stem statement is
associated.
Page 1 of 1 4/2/2008 2:56 PM
ES-401, REV 9 SRO T1'G2 PWR, EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
060AG2.4.6 Accidental Gaseous Radwaste ReI. /9 3.7 4.7 D D D D D D D D D D ~ Knowledge symptom based EOP mitigation strategies.
This is a Generic, no stem statement is
associated.
061AG2.2.25 ARM System Alarms / 7 3.2 4.2 D D D D D D D D D D ~ Knowledge of the bases in Technical Specifications for
limiting conditions for operations and safety limits.
This is a Generic, no stem statement is
associated.
069AA2.02 Loss of CTMT Integrity 15 3.9 4.4 D D D D D D D ~ D D D Verification of automatic and manual means of restoring
integrity
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 / 43.5 / 45.3 I 45.13)
WE03EA2.1 LOCA Cooldown - Depress. I 4 3.4 4.2 D D D D D D D ~ D D D Facility conditions and selection of appropriate
procedures during abnormal and emergency operations.
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.51 43.51 45.3 I 45.13)
Page 1 of 1 4/2/2008 2:56 PM
ES-401, REV 9 SRO T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
013G2.1.23 Engineered Safety Features Actuation 4.3 4.4 DDDDDDDDDD~ Ability to perform specific system and integrated plant
procedures during all modes of plant operation.
This is a Generic, no stem statement is
associated.
039A2.03 Main and Reheat Steam 3.4 3.7 DDDDDDD~DDD Indications and alarms for main steam and area radiation
monitors (during SGTR)
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 / 43.5 / 45.3 / 45.13)
062A2.03 AC Electrical Distribution 2.9 3.4 DDDDDDD~DDD Consequences of improper sequencing when transferring
to or from an inverter
Ability to (a) predict the impacts of the
following on the'(SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 / 43.5 / 45.3 / 45.13)
073G2.1.10 Process Radiation Monitoring 2.7 3.9 DDDDDDDDDD~ Knowledge of conditions and limitations in the facility
license
This is a Generic, no stem statement is
associated.
078G2.1.20 Instrument Air 4.6 4.6 DDDDDDDDDD~ Ability to execute procedure steps.
This is a Generic, no stem statement is
associated.
Page 1 of 1 4/2/2008 2:56 PM
ES-401, REV 9 SRO T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
034A2.01 Fuel Handling Equipment 3.6 4.4 D D D D D D D ~ D D D Dropped fuel element
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 / 43.5 / 45.3 / 45.13)
071G2.4.4 Waste Gas Disposal 4.5 4.7 D D D D D D D D D D ~ Ability to recognize abnormal indications for system
operating parameters which are entry-level conditions for
This is a Generic, no stem statement is emergency and abnormal operating procedures.
associated.
086A2.04 Fire Protection 3.3 3.9 D D D D D D D ~ D D D Failure to actuate the FPS when required, resulting in fire
damage
Ability to (a) predict the impacts of the
following on the (SYSTEM) and (b) based
on those predictions, use procedures to
correct, control, or mitigate the
consequences of those abnormal
operation:(CFR: 41.5 / 43.5 / 45.3 / 45.13)
Page 1 of 1 4/2/2008 2:56 PM
ES-401, REV 9 SRO T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
RO SRO
G2.1.11 Conduct of operations 3.0 3.8 D D D D D D D D D D ~ Knowledge of less than one hour technical specification
action statements
This is a Generic, no stem statement is
associated.
G2.1.20 Conduct of operations 4.6 4.6 D D D D D D D D D D ~ Ability to execute procedure steps.
This is a Generic, no stem statement is
associated.
G2.2.22 Equipment Control 4.0 4.7 D D D D D D D D D D ~ Knowledge of limiting conditions for operations and safety
limits.
This is a Generic, no stem statement is
associated.
G2.2.6 Equipment Control 3.0 3.6 D D D D D D D D D D ~ Knowledge of the process for making changes to
procedures
This is a Generic, no stem statement is
associated.
G2.3.8 Radiation Control 2.3 3.2 D D D D D D D D D D ~ Knoledge of the process for performing a planned
gaseous radioactive release
This is a Generic, no stem statement is
associated.
G2.4.1 Emergency Procedures/Plans 4.6 4.8 D D D D D D D D D D ~ Knowledge of EOP entry conditions and immediate action
steps.
This is a Generic, no stem statement is
associated.
G2.4.28 Emergency Procedures/Plans 3.2 4.1 D D D D D D D D D D ~ Knowledge of procedures relating to emergency
response to sabotage.
This is a Generic, no stem statement is
associated.
Page 1 of 1 4/2/2008 2:56 PM
Draft Submittal
(Pink Paper)
Reactor Operator Written Exam
~ u£l?y ~ o08-c:J-oj
J) e Arr
Question #: KA 008.AK2.01 . Unit 1 conditions are as follows:
RX Power is 50% and stable PZR Pressure is 2100 and decreasing B/U Heaters are in auto and energized. Which ONE of the following correctly describes the instrument failure and associated plant response that would cause the plant conditions above?
References
ND-93-3
Choice_a: COMMENT
PT-445 has failed high causing PORV-1456 is open A - Correct -
B - Wrong - PT-445 does not control PORV-1455C
C. Wrong - If PT-444 failed high B/U heaters would be
off.
Choice_b: D. Wrong - If PT-444 failed high B/U heaters would be off.
PT-445 has failed high causing PORV-1455C is open
Choice_c:
PT-444 has failed high causing PORV-1456 is open
Choice_d:
PT-444 has failed high causing PORV-1455C is open
Question #: 2 KA 011.EK3.12
Given the following plant conditions:
-A large break LOCA has occurred on Unit 1.
-The crew has transitioned to EOP-ES-1.3, TRANSFER TO COLD LEG RECIRCULATION.
-The operators have progressed in EOP-ES-1.3 to the point where SI suction has been aligned to the Containment Recirc.
Sump.
-One SI pump has been started when the following indications are noted:
The running SI pump amps are oscillating
SI flow is abnormally low and oscillating
Which ONE of the following correctly describes the reason for the abnormal SI pump indications and the initial actions the crew
should take?
Choice_a:
The Recirculation Sumps are blocked. Place both LHSI pumps in
Pull-to-Lock and continue performing the remaining steps of ES-1.3
Choice_b:
The running LHSI pump suction is blocked. Start the other LHSI
pump and continue performing the remaining steps of ES-1.3.
Choice_c:
The Recirculation Sumps are blocked. Transition to 1-EOP-ECA-
1.1, "Loss of Emergency Coolant Recirculation, and add makeup
II
to the RWST.
Choice_d:
The running LHSI pump suction is blocked. Transition to 1-EOP-
ECA-1.1, "Loss of Emergency Coolant Hecirculatlon," and start the
other LHSI pump.
COMMENT
A. Incorrect. Both recirc sumps are not known to be blocked. The running SI pump should be stopped. The operators should
remain in ES-1.3 and start the other LHSI pump.
B. Correct. Oscillating amps and flow indicate sump blockage on that pump. The operators should remain in ES-1.3 and attempt to
start the other LHSI pump.
C. Incorrect. Both recirc sumps are not known to be blocked. This action is correct if both pumps were blocked.
D. Incorrect. It is correct to transition to start the other LHSI pump but a transition to ECA-1.1 is not required to perform this.
Knowledge of the reasons for the following responses as the apply to the Large Break LOCA: Actions contained in EOP for
emergency LOCA (large break)
4.4 /4.6
References
EOP-ES-1.3, TRANSFER TO COLD LEG RECIRCULATION
EOP-ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION
Question #: 3 KA 007.EK2.02
Given the following plant conditions:
Unit 1 has just tripped.
Breakers 15A 1, B1, and C1, Transfer Bus to SS Bus Supply Breakers, are closed.
Unit 2 is in Intermediate Shutdown and due to electrical problems, 25A 1 and 25B1, Transfer Bus to SS Bus Supply Breakers,
are open.
Breaker 25C1, Transfer Bus to SS Bus Supply Breaker, is closed.
Which ONE of the following describes the effect this electrical lineup has on Station equipment?
References
ND-90.3-LP-7, SS and Emergency Dist Prot and Control
Choice_a: COMMENT
All SS buses on both Units will experience a load shed signal. A. Incorrect. Only 25C1 on Unit 2 is aligned to the SS
Bus and is therefore the only bus on Unit 2 that will load
shed.
B. Correct. Only Unit 1 and Unit 2 "C" SS busses are
properly aligned for load shed to occur.
Choice_b: C. Incorrect. An EDG start and load signal sequence has
Only "C" SS buses on both Units will experience a load shed not been met.
signal.
D. Incorrect. Auto Start Inhibit feature is activated via SI
or High High CLS signal.
References
Choice_c: ND-90.3-LP-7, SS and Emergency Dist Prot.and Control
#2 and #3 EDGs will start and load on the respective 2H and 2J Com prehension / Anal
buses.
Choice_d:
Component Cooling Water Pump, 1-CC-P-1A will receive an auto-
start inhibit signal.
Question #: 4 KA 022.AA2.03
Given the following plant conditions: -Unit 1 is operating at 100% -Letdown flow, CH-FI-1150, is oscillating -"CHG PP TO REGEN HX HI-La FLOW" is in alarm Which ONE of the following describes the malfunction in progress? Choice_a: The Regenerative Heat Exchanger has developed a tube leak. Choice_b: 1-CC-TCV-103, NRHX Temperature Control Valve, has failed open. Choice_c: 1~CH-PCV-1145, Letdown Pressure Control Valve, has failed closed. Choice_d: 1-CH-FCV-1122, Charging Line Flow Control Valve, has failed closed. COMMENT A. Incorrect. A leak from the tubes would not decrease the amount of cooling provided to letdown. B. Incorrect. This would increase cooling. C. Incorrect. This would increase the amount of backpressure on the system and reduce the likelihood of flashing. D. Correct. A minimum of 25 gpm is needed to prevent flashing downstream of the letdown orifice. References ND-88.3-LP-2, Charging & Letdown Com prehension / Anal References ND-88.3-LP-2, Charging & Letdown
Question #: 5 KA 026.AA1.07
Given the following plant conditions: A loss of all AC power occurred offsite power was restored after 10 minutes, The actions of ECA-O.1, LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED, are being performed to start a CC Pump . Why are 1-CC-TV-140A and 8, RCP CC Return Manual Isolation Valves verified closed prior to restarting the CC Pump?
References
ND-95.3-LP-18-DRR
1-ECA-O.1
, LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED
Choice_a: COMMENT
Protect CC availability by precluding steam formation in the CC A. Correct. Prevents the introduction of steam into the CC
piping. system when CC pump is started.
B. Incorrect. Plausible, the loss of SW due to the loss of
power is addressed at step 12 of the procedure. This
would not place CC loads at minimum, however.
Choice_b: C. Incorrect. Plausible, step 1 also addresses closing
Reduce CC heat loads to the minimum based on SW loads. RCP Seal injection valves to prevent possible seal
damage.
D. Incorrect. Plausible, isolating CC to the RCPs would
increase flow to other components but would not
maximize flow to the CVCS.
Ability to operate and/or monitor the following as they
Prevent damage to the RCP bearings due to excessive cooldown apply to the Loss of Component Cooling Water: Flow
rate. rates to the components and systems that are serviced by
the CCWS; interactions among the components
Choice_d:
Maximize flow to the CVCS components for reestablishing
charging, letdown and seal return.
Question #: 6 KA 027.AA1.01
Unit 1 conditions are as follows: Reactor Power is 100 % Pressurizer' Pressure control is in automatic Which one of the following correctly describes the immediate response of the pressure control system if the master pressure controller refrence fails high?
References
ND-93-3-LP-5
Cholcea: COMMENT
Both spray valves close and proportional heaters go to maxim urn
output
1- Correct - Changing setting reduces the output from the
controller and raises the demanded pressure setpoint.
The reduction results in spray valve closure and heaters
turning fully on.
Choice_b: 2- Wrong - Proportional heaters are normally energized.
Both spray valves close and only back-up heaters energize An increase in pressure setpoint would ensure they are
on at full .
3- Wrong This is the opposite reaction to actual pressure
being lower than setpoint.
Both spray valves open and proportional heaters go to minimum
output
4- Wrong - This is the opposite reaction to actual pressure
being lower than setpoint.
Choice_d:
Both spray valves open and both proportional and backup heaters
deenergize
Question #: 7 KA 029.EK3.10
Unit 1 conditions are as follows: An ATWS is in progress The reactor is NOT tripped The Turbine Generator is tripped RCS Pressure is 2335 PSI and rising Rods are automatically stepping in at 45 steps per/m in All S/G levels are 10% NR and decreasing IIAII MDAFW Pump is providing a total AFW Flow of 300GPM The crew has just transitioned from E-O.O to FR-S.1 Which of the following correctly describes the immediate action required per 1-FR-S.1 based on plant conditions and the reason for performing the actions?
References
FR-S.1
Choice_a: COMMENT
Manually insert control rods to maximize reactivity insertion. A. Correct - per FR-S.1
B. Incorrect - Operation of the AFW system is not an
immediate action of FR-S.1 Plausable - AFW flow is
lower than required by step 10 RNO of FR-S.1
C. Incorrect - Emergency Boration is not an immediate
Choice_b: action of FR-S.1 Plausable - Method of adding reactivity
Start all available AFW Pumps to increase AFW Flow to at least
700GPM to ensure secondary heat sink.
D. Incorrect - RCS pressure control is not an immediate
action of FR-S.1. Plausable - RCS pressure is high
enough to warrant action in FR-S.1 step 5
Choice_c:
Initiate Emergency Boration to insert negative reactivity.
Choice_d:
Open at least one PORV and Block Valve to maintain RCS
Pressure to less than 2335PSI.
Question #: 8 KA 038.EK1.01
A SGTR has occurred on Unit 1. Current conditions are:
- RCS pressure 1350 psig
- RCS temperature (CETCs) 542 of
- SG pressures 1000 psig (A) 1050 psig (B) 960 psig (C)
- SG 1B has been confirmed as the ruptured SG
While performing the steps of 1-E-3, "Steam Generator Tube Rupture", the Unit Supervisor found ALL available copies of the
procedure had an illegible page. This page contained the required temperatures for determining ReS cooldown temperatures.
The US directs you to use the steam tables to determine the required RCS (core exit) temperature with an allowance of 50 of
for subcooling.
The required core exit temperature after the RCS cooldown is...
(Steam Tables' provided)
References
1-E-3
,STEAM GENERATOR TUBE RUPTURE
Choice_a: COMMENT
492 of C. Correct - 1050psig for lowest ruptured SG + 14.7 =
1065 psia = -553(Saturation temp) - 50 (subcooling)=
503
A. Incorrect - 960 psig + 14.7 = 975 psia = 542 - 50= 492
Plausible if candidate used the lowest of all SG
pressures.
Choice_b:
500 of B. Incorrect - 1050 psig = 550 - 50= 500 Plausible if
candidate picks the lowest ruptured SG pressure but does
not account for psia vs psig in steam tables.
D. Incorrect - (Sat temp. for 1065 psia wi No subcooling
added in) Plausible if candidate forgets to subtract
another 50 of for subcooling.
Knowledge of the operational implications of the
following concepts as they apply to the SGTR: Use of
steam tables
Choice_d:
553 of
Question #: 9 KA 040.AA1.24
Given the following plant conditions exist on Unit 1: -Unit is at 20% power. -A faulted steam generator has occurred. -RCS hot leg temperatures: 547F (A), 544F (B), 545F (C) -RCS cold leg temperatures: 545F (A), 530F (B), 543F (C) -S/G pressures: 525 psig (A), 515 psig (B), 530 psig (C) -Steam line flow is 50% of rated on A & B S/Gs. -Steam line flow is 35% of rated on C S/G. -Containrnent pressure (Channel): 8 psig (1), 7.5 psig (2), 7.5 psig (3), 8 psig (4)
For the given plant conditions, which ONE of the following correctly describes whether or not a Main Steam Line Isolation should or should not have occurred and the reason?
(Have licensee provide actual steam line flow numbers)
References
T.S. TABLE 3.7-4
ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS
INSTRUMENT SETTING
Choice_a: COMMENT
A MSIL should have occurred because of high steam line flow Answer - A
coincident with low steam line pressures. Steam line isolation signal generated by:
(1) 1/2 channels High steam flow per steam line in 2/3 of
the steam lines
IN COINCIDENCE WITH EITHER
Choice_b:
A MSIL should have occurred because of high steam line flow (2) 2/3 steam lines low pressure (1 detectorl line)
coincident with low loop average temperatures. OR
(3) 2/3 RCS loops low Tavg signal (1 detectorl loop)
1-E-2, Faulted SG Isolation
Choice_c: Com prehension 1 Anal
A MSIL should NOT have occurred because Containment pressure Ability to operate and 1 or monitor the following as they
is below the required setpoint to receive a MSIL signal. apply to
the Steam Line Rupture: Main steam header pressure
gauges
3.8/3.8
Choice_d:
A MSI L should NOT have occurred because two S/Gs have
pressures above the isolation setpoint and only one indicates high
steam flow.
Question #: 10 KA 054.AK1.02
Given the following plant conditions: -Unit 1 has experienced a loss of all feed condition -The SRO has entered into 1-FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK and directed the OATC to trip all RCPs -Several minutes later the auxiliary operator reports that the AFW booster pump is available -RCS hot leg tem peratures are 570 OF and slowly decreasing -All 3 S/Gs are reading approximately 5% Wide Range level When recovering feed for these conditions, which ONE of the following statements correctly describes how feed must be aligned per 1-FR-H.1 and the reason? Choice,.a: Feed ONE S/G at the maximum flow rate possible in order to reestablish a heat sink as soon as possible. Choice_b: Feed ONE S/G at the minimum verifiable flow rate possible in order to minimize thermal stresses applied to SG U-tubes. Choice_c: Feed ALL S/Gs at the maxim um flow rate possible in order to reestablish a heat sink as soon as possible. Choice_d: Feed ALL S/Gs at the minimum verifiable flow rate possible in order to minimize thermal stresses applied to SG U-tubes. COMMENT A. Incorrect. This is correct if the indications given supported a complete loss of heat sink, l.e. RCS hot leg temp increasing. B. Incorrect. All available S/Gs should be fed to keep the U-tubes wetted until level is restored on the NR. C. Incorrect. All available S/Gs should be fed at the minimum flow to avoid stresses to the U-tubes. D. Correct. This is an acceptable mitigating strategy to recover the secondary heat sink if RCS Hot leg temp is stable or lowering. Memory Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): Effects of feedwater introduction on dry S/G 3.6/4.2 References 1-FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK ND-95.3-LP-41, FR-H.1
Question #: 11 KA 055.EA2.05
Given the following plant conditions: -Unlt 1 experienced a loss of all AC power at 09:38 -The total amount of DC loads on the battery at this time is 225 amps If the DC loading does not change, which ONE of the following is the LATEST time by which AC power must be restored to ensure that the Station Batteries have NOT reached the end of their design discharge rating? Choice_a: 11:38 Choice_b: 13:38 Choice_c: 15:38 Choice_d: 17:38 COMMENT A. Incorrect. Plausible, this is the two hour minimum acceptable time for the batteries to provide power during a blackout condition. B. Incorrect. Plausible if the operator performs an incorrect math calculation or does not know the correct battery rating of 1800 amp-hrs. C. Incorrect. Plausible if the operator performs an incorrect math calculation or does not know the correct battery rating of 1800 amp- hrs. D. Correct. 1800 amp-hrs/225 amps = 8 hrs References ND-80-LP-13, Batteries ND-90-LP-6, 125 VDC Distribution Memory Ability to determine or interpret the following as they apply to a Station Blackout: When battery is approaching fully discharged 3.4/3.7 References
Question #: 12 KA 056.AA1.04
Plant conditions are as follows; - IIB II diesel generator is running and independently supplying the emergency bus @ 60hz - Governor Speed Droop is set to 30 Which ONE of the following correctly describes the response of the EDG frequency to changes in load?
References
ND-90.3-LP-1
Choice_a: COMMENT
As additional load is placed on the bus, frequency will lower and
stabilize at a value lower than 60hz.
A. Correct
B. Wrong - Improper assumption on how the speed droop
mechanism works. Plausible - as load increases, power
output is needed to increase.
Choice_b:
As additional load is placed on the bus, frequency will rise and C. Wrong - With the speed droop setting at 30, the speed
stabilize at a value higher than 60hz. of the machine will not increase to raise frequency but
lower in'an attempt to share load. Plausible - that is how
the system would respond with the speed drop at zero.
D. Wrong - The speed droop works independent of
another source being paralleled with. Plausible - speed
droop is intended to allow paralleling sources without
As additional load is placed on the bus, frequency will lower overloading the generator.
slightly but will be restored to 60 hertz by the automatic governor
control system.
Choice_d:
There will be no effect since this setting only has an effect on the
diesel operating characteristics when it is operating in parallel
with another source.
Question #: 13 KA 057.AA2.19
Given the following plant conditions: -Unlt 1 shutdown 72 hrs earlier and is in an Intermediate Shutdown condition. -RCS Temperature is 310°F. -Both trains of RHR are in service. -A loss of UPS 1A2 occurs. Which ONE of the following correctly describes the effect of the loss of UPS 1A20n RHR? Choice_a: RHR outlet temp increases. Component Cooling flow to RHR train A HX is lost. Choice_b: RHR discharge flow decreases to zero. Component Cooling flow to RHR train A HX is lost. Choice_c: RHR outlet temp remains the same. Component Cooling flow to RHR train A HX remains the same. Choice_d: RHR discharge flow increases. Component Cooling flow to RHR train A HX remains the same. COMMENT A. Incorrect. Plausible, a loss of vital bus I from UPS 1A1 causes a loss of CC flow to A HX. B. Incorrect. Plausible, if the candidate believes that RH-FCV-1758 fails closed on loss of power from Vital Bus III via UPS 1A2. C. Incorrect. Plausible, if candidate believes that 1758 fails as is with no affect on CC flow. D. Correct. Vital Bus III causes the RH-FCV-1758, RHR HX Flow Control Valve, to fail open. CC flow to the HX is not affected. Comprehensive Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus. 4.0/4.3 References 1-AP-10.02, LOSS OF VITAL BUS II 1-AP-27.00,LOSS OF DECAY HEAT REMOVAL CAPABILITY
Question #: 14 KA 062.AG2.1.32
Which ONE of the following correctly states (1) the minimum required level at the high level intake structure
lAW O-AP-12.01, Loss of Intake Level, and (2) the reason for maintaining the minimum level?
Choice_a: 17.2 feet to ensure cooldown on the non-accident unit and one In- Service RSHXs in service on the Unit with a DBA LOCA. Choice_b: 17.2 feet to ensure cooldown on the non-accident unit and two In- Service RSHXs in service on the Unit with a DBA LOCA. Choice_c: 23.5 feet to ensure cooldown on the non-accident unit and one In-Service RSHXs in service on the Unit with a DBA LOCA. Choice_d: 23.5 feet to ensure cooldown on the non-accident unit and two In- Service RSHXs in service on the Unit with a DBA LOCA. COMMENT A. Incorrect. Plausible, if candidate believes that only one RSHS is required. B. Correct. C. Incorrect. Plausible, if candidate mistakes the AP caution concerning 23.5 ft for no restrictions on CC HX SW flow and only 1 HX required for DBA LOCA. D. Incorrect. Plausible, if candidate mistakes improper level with correct # of HX for DBA LOCA criteria. Memory Loss of Nuclear Service Water: Abitity to explain and apply all system limits and precautions. References 1-AP-12.01, LOSS OF INTAKE CANAL LEVEL
Question #: 15 KA 065.AK3.03
Unit 1 is operating at 50% after a shutdown for maintenance. Containment Instrument Air header is 70psi and decreasing. Which one of the following correctly describes the air supply to the PZR PORVls and air pressure to the Letdown isolation valves?
References
ND92.1-LP-1
Choice_a: COMMENT
PZR PORVls are supplied from com pressed air back-up A. Correct - PORVls are supplied from a 80# regulator
Letdown isolation valves air pressure is above minimum to keep supplied from compressed air bottles. Letdown isolation
valves fully open valves require 65# to remain fully open.
B. Wrong-
C. Wrong -PORVls are supplied from a 80# regulator
Choice_b: supplied from compressed air bottles. L
PZR PORVls are supplied from compressed air back-up
Letdown isolation valves air pressure is below minimum to keep D. Wrong-
valves fully open
Choice_c:
PZR PORVls are supplied from instrument air header
Letdown isolation valves air pressure is above minimum to keep
valves fully open
Choice_d:
PZR PORVls are supplied from instrument air header
Letdown isolation valves air pressure is below minimum to keep
valves fully open
Question #: 16 KA WE04. EG2.4.31
Given the following plant conditions: -Unit 1 has tripped due to lowering Pressurizer level. -The SRO has entered 1-E-O, REACTOR TRIP OR SAFETY INJECTION. Which ONE of the following valid alarms REQUIRES transition to 1-ECA-1.2, LOCA OUTSIDE CONTAINMENT? Choice_a: O-RMA-D5, VENT STACK #2 RAD MON TRBL. Choice_b: O-RMA-C5, PROCESS VENT RAD MON TRBL. Choice_c: 1-1A-E5, SW RS HX VV PIT A HI LVL. Choice_d: 1-ACB-C4, LOW COND SUMP A LEVEL HI-HIILO-LO COMMENT A. Correct. Annunciated by 1-VG-RM-131A/B/C in alarm B. Incorrect. Annunciated by 1-GW-RM-130A, Rad Monitor Process Vent Particulate Detector, 1-GW-RM-130B, Rad Monitor Process Vent Noble Gas Detector, or 1-GW-RM-130C, Rad Monitor Process Ionization Chamber. Plausible, doesn't detect gas stream from aux bldg. C. Incorrect. Plausible, indication of leak outside containment. Not located as part of 1-E-O transition guidance. D. Incorrect. Plausible, indication of leak outside containment. Not located as part of 1-E-O transition guidance. Memory Inadequate Heat Transfer: Knowledge of annunciators alarms and indications, and use of the response instructions. References 1-E-O, LOSS REACTOR TRIP OR SAFETY INJECTION 1-ECA-1.2, LOCA OUTSIDE CONTAINMENT ND-93.5-LP-3
Question #: 17 KA WE12.EK2.2
Unit 1 was at 100 % power when a steam rupture occurred inside containment.
The crew is performing 1-ECA-2.1, Uncontrolled Depressurization of all SGIS .
RCS Tern perature is 450 degrees F and increasing slowly
RCS pressure is 2250 psi and increasing slowly
One charging pump is running and aligned to the RCS
All main steam trip valves and main steam trip bypass valves are closed
All SG pressures are 215 psig and decreasing
11
All SG levels are 20 NR and decreasing
Total auxiliary feedwater flow is 700 gpm
Which ONE of the following correctly describes how heat removal from the RCS must be accomplished per 1-ECA-2.1 for the
given plant conditions?
References
ECA-2.1
Choice_a: COMMENT
RCS forced circulation and dumping steam from S/Gl s A. Correct lAW 1-ECA-2.1
B. Wrong - Natural circulation is only required if forced
circulation is required to be secured or RCP cannot be
started.
C/o. Wrong - Plausable - methods to remove heat from
Choice_b: primary
RCS natural circulation and dumping steam from S/G1s
Choice_c:
RCS forced circulation and RCS "Feed and Bleed"
Choice_d:
RCS natural circulation and RCSIIFeed and Bleed"
Question #: 18 KA WE05.EK3.2
Given the following plant conditions:
-Unlt 1 experienced a Reactor Trip.
-E-O, Reactor Trip or Safety Injection," was completed and the team transitioned to ES-O.1, "Reactor Trip Response."
II
-Subsequently all AFW pumps were lost and the team entered FR-H.1, "Response to a Loss of Secondary Heat Sink."
-The operators have started a MFW pump and the SRO directs the RO to use the FRV bypasses to control flow to the SG.
-Annunciator A-F-3, SIINIATIATED TRAIN A is NOT LIT
-Annunciator A-F-4, SIINIATIATED TRAIN B is LIT
Which ONE of the following correctly describes the minimum actions necessary to open the FRV bypass valves?
Choice_a:
Reset the B train SI Signal from the MCR only.
Choice_b:
Reset the B train SI signal from the MCR .first and then depress the
S/G level reset pushbuttons.
Choice_c:
Depress the S/G level reset pushbuttons only.
Choice_d:
Locally block or clear the A train signals and then reset the A & B
train SI signals from the MCR.
COMMENT
A. Incorrect. Plausible, if the candidate believes that a local reset is required for a partial SI initiation signal.
B. Incorrect. Plausible, if candidate believes conditions for block to be in effect based on one SI annunciator not lit.
C. Correct.
D. Incorrect. Plausible, if candidate believes a full SI signal is required to close the FW valves.
Memory
Loss of Secondary Heat Sink: Knowledge of the reasons for the following responses as they apply to the (Loss of Secondary Heat
Sink) Normal, abnormal and emergency operating procedures associated with (Loss of Secondary Heat Sink).
References
1-E-O, LOSS REACTOR TRIP OR SAFETY INJECTION
Question #: 19 KA 001.AK2.06
Unit one reactor power is at 85% and stable
Control rods are in automatic.
The following alarms just actuated:
H-A-4 TavelTref Deviation
H-C-4 Low Tave to FW Cont
Tave is 569F
Which one of the following correctly describes how control rods will initially respond wI no operator action?
References
ND-93.3-LP-3/LP-2
Choice_a: COMMENT
Rods step out at 72 SPM A. Correct - The alarms indicate that median Tave failed
low. With TavelTref deviation alarm in, deviation is>
than 5F. This creates a demand for outward rod motion
of 72 spm.
B. Wrong - With TavelTref deviation alarm in, deviation
is > than 5F. A demand for outward rod motion of 72 spm
Choice_b: is generated. Plausible - when TavelTref deviation is
Rods step out at 40 SPM 4F a demand for rod motion of 40 spm is generated. If
applicant references HFPO program Tave
C. Wrong -With TavelTref deviation alarm in, deviation is
> than 5F. A demand for rod motion of 8 requires a
temperature difference of betwee. 1.5 and 3
D. Wrong - The combination of the 2 alarms indicates
Rods step in at 8 SPM median Tave failed low generating an outward rod
demand. Plausible - if applicant calculated reference
delta T and finds out that there is less than 1 F
difference..
Referenced TavslTref deviation alarm due to no
Choice_d: deviation meter.
Rods remain at the current rod height
Question #: 20 KA 003.AG2.4.1
Given the following plant conditions:
-Unit 1 is operating at 100% power.
II
-One Bank "A Rod Bottom Light has dropped into the core.
-At the same time as the initial rod drop occurred, another Bank "A" Rod is noticed to be moving erratically into the core
(Moving several steps then stopping then moving several more steps)
-1E-E3, DELTA FLUX DEVIATION in alarm.
Which ONE of the following actions is required to be taken IMMEDIATELY?
Choice_a:
Trip the reactor.
Choice_b:
Reduce Reactor Power to
between 70% - 74%.
Choice_c:
Initiate boration to reduce power to < 90% rated with no rod
movement.
Choice_d:
Place the ROD CNTRl MODE SEl switch to the MAN position.
Power reduction is not required.
COMMENT
A. Correct. Required if more than one rod is affected.
B. Incorrect. Required by Control Rod misalignment procedure.
C. Incorrect. Required by misalignment procedure to return delta flux to target band within 15 minutes by ARP and T.S.
D. Incorrect Correct if only one rod affected.
Memory
Dropped Control Rod: Knowledge of EOP entry conditions and immediate action steps.
4.3/4.6
References
O-AP-1, ROD CONTROL SYSTEM MALFUNCTION
O-AP-1.01, CONTROL ROD MISALIGNMENT
Question #: 21 KA 005.AA2.01
Unit 1 started a plant shutdown from 100% power with all control rods fully withdrawn. The following conditions currently exist: The plant is at 80% power ramping down at 1/4 %/min Bank D rod position indication is 190 steps withdrawn Upper Ion Chamber Deviation is alarming Lower Ion Chamber Deviation is normal NIS PR Channel Average Flux Deviation is normal Which ONE (1) of the following correctly describes the cause of the given indications?
References
ND-93.2-LP-4
Choice_a: COMMENT
On Bank D rod is stuck at the fully withdrawn position A. Correct - The stuck rod will cause the Upper Ion
Chamber Deviation to exceed lt's 20/0 threshhold.
B. Wrong - If a rod fell to the bottom of the core it would
also cause the Lower Ion Chamber Deviation to alarm.
C. Wrong - A loss of High Voltage power supply to a
Choice_b: channel would cause Both Upper and Lower Alarms as
One Bank D rod has fallen to the bottom of the core well as NIS PR Average Flux Deviation.
D. Wrong - Loss of a summing and level amp would only
cause the NIS PR Channel Average Flux Deviation
Alarm.
Choice_c:
Loss of High Voltage power supply to a PR Channel
Choice_d:
Summing and level amplifier failure for one PR Channel
Question #: 22 KA 061.AA2.05
Given the following plant conditions:
-Unlt 1 is in a Refueling Shutdown condition
-There are 35 individuals in containment performing various maintenance activities
-A valid HIGH level alarm is received on 1-RM-RMS-162, Manipulator Crane Area
-Containment Purge is in operation
-No fuel movements are currently in progress
II
Which ONE of the following actions is required lAW 1-RM-K8, "1-RM-RI-162 HIGH alarm response procedure?
Choice_a:
Immediately evacuate containment.
Choice_b:
Verify containment purge is isolated.
Choice_c:
Verify containment air recirc is isolated.
Choice_d:
Initiate O-AP-22.00, FUEL HANDLING
ABNORMAL CONDITIONS.
COMMENT
A. Incorrect. ALERT & HIGH level requires notification and discussion of appropriate action with HP prior to directing an evacuation
of containment.
B. Correct. Containment isolation should occur at the HIGH setpoint.
C. Incorrect. This is not directed by the Alarm Response.
D. Incorrect. With no abnormal conditions in containment concerning fuel handling, this step is NA.
Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms: Need for area
evacuation; check against existing limits
3.5/4.2
References
1-RM-K8, 1-RM-RI-162 HIGH
Question #: 23 KA 068.AK3.09
Given the following plant conditions; Unit 1 was at 100% power with IB' Charging Pump out of service for maintenance. Noxious fumes forced the evacuation of the control room. 0-AP-20.00, Main Control Room Inaccessibility was entered. Which ONE of the following identifies the MINIMUM required manipulation(s) to transfer IIAII CCP control to the ASDP? Choice_a: Place IIHII Group Transfer Switch to IIl0CAlII only Choice_b: Place IIJII Group Transfer Switch to IIl0CAlII only Choice_c: Place IIJII Group Transfer Switches to IIl0CAlII and the switches for half station & Manual/Auto controller 1-CH-FCV-1122 to "lOCAL." Choice_d: Place IIHII Group Transfer Switch to IIl0CAlII and the switch for half station controller 1-CH-FCV-1122 to IIl0CAllI * COMMENT A. Incorrect. . Plausible if candidate reads the question to mean only the Charging Pump and not the flow controller also. B. Incorrect. This alone will not provide a flow path with flow control for charging. The IIJII Group Transfer Switch would provide power to IICII CCP if on alternate power supply. C. Incorrect. Correct if both IIAII & 'B II CCPs were available, or if IICII was on it's alternate power supply. Only IIJII power supply is required for IIBII CCP. D. Correct. This will transfer power for the IIAII CCP and the flow control valve. Memory Knowledge of the reasons for the following responses as they apply to the Control Room Evacuation: Transfer of the following to local control: charging pumps, charging header flow control valve, PZR heaters, and boric acid transfer pumps 3.9/4.4 References 0-AP-20.00
Question #: 24 KA 076.AK3.05
Given the following plant conditions: -Unit 1 is @ 1000/0 power -A ramp to 750/0 is in progress to perform turbine valve testing -Mlxed Bed Demin 1A is in service -Mlxed Bed Demin 1B is in standby -Annunciator 1-RM-E7, RC LON HIGH ALERT/FAILURE, is in alarm Which ONE of the following correctly describes the actions the RO must perform lAW 1-0P-CH-011, CVCS MIXED BED DEMIN OPERATIONS, to minimize the consequences of this event? Choice_a: Align BOTH 1A and 1B mixed bed dem in in service with the Cation IX. Choice_b: Place the 1A demin in standby. Align 1B mixed bed demin in service with the Cation IX. Choice_c: Align BOTH 1A and 1B mixed bed demin in service without the Cation IX.
. Choice_d:
Place the 1A demin in standby. Align 1B mixed bed demin in service without the Cation IX. COMMENT A. Incorrect. The Cation IX is not addressed by the ARP or the OP and would not be appropriate for this condition. Used to increase pH by removing excess free lithium. B. Incorrect. The Cation IX would not be placed in service for this condition. C. Incorrect. Both IXs are not placed in service simultaneously. D. Correct. Only one mixed bed demin is normally placed into service. The Cation IX would not be placed in service for this condition. Memory Knowledge of the reasons for the following responses as they apply to the High Reactor Coolant Activity: Corrective actions as a result of high fission-product radioactivity level in the RCS 2.9/3.6 References 1-RM-E7, RC LON HIGH ALERT/FAILURE 1-0P-CH-011, CVCS MIXED BED DEMIN OPERATIONS ND-88.3-LP-2-DRR, CHARGING AND LETDOWN
Question #: 25 KA WE13.EK1.2
Plant conditions are as follows: Unit 1 operators are performing FR-H.2 "Response to S/G Overpressure. II - A S/G pressure is 1145 psig and stable - A S/G lev~1 is 50% and rising. - RCS Thot temperatures are 545 degrees F and stable - All three RCPs are running. Attempts to dump steam from "A" SG were unsuccessful Which ONE of the following correctly describes the required actions based on plant conditions?
References
FR-H.2
Choice_a: COMMENT
Isolate the steam supply to the TOAFW pump from "A" SG and dump A. Correct - lAW 1-FR-H.2
steam from the unaffected SGls to reduce RCS temperature
S. Wrong - AFW is isolated - Plausable is operator
believes that increasing AFW will help C/O the S/G
C/D. Wrong - Plausable if operator bel;ieves that SG
Slowdown is an approved method of reducing S/G
Choice_b: pressure
Increase AFW flow to the "A" SG and dump steam from the unaffected
SGls to reduce RCS temperature
Choice_c:
Isolate the steam supply to the TOAFW pump from "A" SG and reduce "A"
SG pressure using the S/G blowdown system
Choice_d:
Increase AFW flow to the "A" SG and reduce "A" SG pressure using
the S/G blowdown system
Question #: 26 KA WE16.EA2.2
Given the following plant conditions: @ 10:00 AM Unit 1 experienced a LOCA that resulted in significant core damage. @ 11:20 AM Peak Containment Pressure is observed to be 6.2 psig. @ 11:30 AM Peak Containment Radiation level is observed to be 110,000 R/hr. @ 11:35 AM (Current time) Containment radiation level is 90,000 R/hr, Containment pressure is 5.0 psig. Both are slowly lowering. Which ONE of the following correctly describes whether or not the crew must use Adverse Values when implementing the EOPs and the reason? Choice_a: Adverse values MUST still be used. Adverse containment conditions now exist due to the current containment pressure. Choice_b: Adverse values MUST still be used. Adverse containment radiation values previously existed. Choice_c: Adverse values MUST still be used. Adverse containment conditions now exist due to the current containment radiation dose rate. Choice_d: Adverse values are NO longer required to be used. Adverse containment conditions no longer exist. COMMENT A. Incorrect. Containment press for adverse conditions is 20 psia. 5 psig + 14.7 = 19.7 psia B. Correct. Rad levels exceeded setpoint and therefore instrument operability must be evaluated. C. Incorrect. Current rad levels are below the adverse value of > 1 E 5 R/hr. D. Incorrect. Adverse values are required to be used. Adverse containment conditions no longer exist. Ability to determine and interpret the following as they apply to the (High Containment Radiation): Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments. 3.0/3.3 References EOP Continuous Action Page
Question #: 27 KA WE02. EG2.4.31
Given the following plant conditions: * A Small Break LOCA has occurred on Unit 1 * RCS pressure is 1100 psig and slowly lowering * The crew is perform ing the actions of ES-1 .2, Post LOCA Cooldown and Depressurization * The crew is depressurizing the RCS per step 14 of ES-1.2 * Pressurizer level is 36% and slowly increasing * Both LHSI pumps have been stopped * One HHSI pump has been stopped * Normal charging is aligned When the depressurization is stopped, the SRO directs the RO to verify that SI is not required. The following conditions are noted: * RCS subcooling is 35°F and trending DOWN * 1C-D8, PRZR LO LVL annunciator is lit Based on these indications which ONE of the following actions is required lAW ES-1.2? Choice_a: Manually start the charging pumps and align HHSI to the cold legs. Choice_b: Start the last charging pump secured and leave normal charging aligned. Choice_c: Re-initiate SI by using the manual SI pushbuttons. Choice_d: Manually start the LHSI pumps and align LHSI to the cold legs. COMMENT A. Correct per ES-1.2. Step 20 of ES-1.2 states to control PZR level by using LHSI or charging flow. LHSI has been secured. Step 25 and the C.A. steps address restarting SI pum ps if subcooling drops below 30 degrees F or PZR level <22%. PZR low level alarm comes in at 5% off program (22% for 0% Rx power). Press is not given, but SI accumulators are not stated as isolated yet, therefore Rx press is still> LHSI shutoff head. B. Wrong - Plausible due to starting last pump would stop pressure decrease. C. Wrong - Plausible due to it would reinitiate inflow into primary and stop pressure decrease D. Wrong - Plausible. Cooldown and initial depressurization have occurred to increase PZR level. A decreasing pressure indicates a leak/shrink rate greater than SI flow and ReS CD. Depressurization is appropriate in combination with increasing SI flow. HHSI flow is directed by the EOP and charging flow is directed by the ARP. Analysis / Comprehension SI Termination: Knowledge of annunciators alarms and indications, and use of the response instructions. References ES-1.2, POST LOCA COOLDOWN AND DEPRESSURIZATION ND-95.3-LP-9, ES-1.2, POST-LOCA COOLDOWN AND DEPRESSURIZATION 1C-D8, PRZR LO LVL
Question #: 28 KA 003.K6.14
A RCP breaker failed to close while staring a RCP. The following conditions were noted during the investigation into the Problem: Hot Leg Loop Isolation Valve Open Cold Leg Loop Isolation Valve Open Bearing Lift Pump Running with a discharge pressure of 400# 4160V busses energized Which one of the following is a possible cause of the RCP motor breaker failing to close?
References
ND-88.1-LP-6
Choice_a: COMMENT
A failure of the RCP speed sensing relay A. Correct - An input from the speed sensing unit is
required for RCP start.
B. Wrong - It takes the Loop Bypass Line Isolation Valve
position and Cold Leg Loop Isolation Valve position to
prevent RCP start
Choice_b: C. Wrong - RCP Motor stator temperature has no input
Loop Bypass Line Isolation Valve Limit Switch indicating open into start logic. Plausable becouse a temperature from
the speed sensing unit is required.
D. Wrong - Oil reservoir level has no input into RCP start
logic. Plausable - Not having oil could damage the
pump.
Choice_c:
RCP motor stator temperature element failed high
Choice_d:
RCP oil reservoir low level
Question #: 29 KA 004.A2.13
Plant conditions are as follows: Unit 1 is in a refueling outage with core offload in progress Unit 2 is at 100 % Unit 1 RWST level is 50/0 in preparation for tank inspection. Which one of the following correctly identifies the source of makeup water to the Unit 1 refueling cavity lAW 1-AP-22.01 'Loss Of Refueling Cavity Level"? Choice_a: Unit 1 HHSI from Unit 2 RWST Choice_b: Unit 1 HHSI from Unit 1 VCT Choice_c: Unit 1 LHSI from Unit 2 RWST Choice_d: Unit 1 LHSI from Unit 1 VCT COMMENT A. Correct - With a low RWS <<6%) only HHSI is available to be used with RWST crosstie. B. Incorrect - VCT is not used to replenish refueling cavity C. Incorrect -LHSI is not used with RWST crosstie. o - Incorrect - LHSI is not used with RWST crosstie. References 1-AP-22.01
Question #: 30 KA 005.K2.03
Which one of the following correctly describes the effect of deenergizing Bus 1J1 on the RHR system?
References
ND-88.2 / 90.3
Choice_a: COMMENT
Only 1-MOV-1701, RHR Suction Isolation, would be deenergized A. Correct - MOV-1701 is deenergized - suction valves
Only the ability to remotely establish an RHR suction path is lost are in series= suction path lost
B. Incorrect - Discharge valves are in parallel
C. Incorrect - Discharge valves are in parallel
Choice_b: D. Incorrect - suction valves are in series=> suction path
Only 1-MOV-1720B, RHR Discharge Isolation, would be lost
deenergized
Only the ability to remotely establish an RHR discharge path is lost
Choice_c:
1-MOV-1701, RHR Suction Isolation, and 1-MOV-1720B, RHR
Discharge Isolation, would be deenergized
The ability to remotely establish both an RHR suction and discharge path is lost
Choice_d:
1-MOV-1701, RHR Suction Isolation, and 1-MOV-1720B, RHR
Discharge Isolation, would be deenergized
The ability to remotely establish both an RHR suction and discharge path is
available
Question #: 31 KA 006.A3.05
Which ONE of the following identifies the time requirement and the basis for manually securing one of two running Low Head Safety Injection pumps if ReS pressure is greater than 185 psig?
References
ND91-LP-213
Choice_a: COMMENT
30 minutes, prevent overheating of a LHSI pump running at shutoff A - Correct
head .
B - Incorrect - Time requirement is 30 min
C - Incorrect - Motor heatload would be reduced with
pump flow decreased.
Choice_b: D - Incorrect - Time requirement is 30 min. Motor
60 minutes, prevent overheating of a LHSI pump running at heatload would be reduced with pump flow decreased.
shutoff head
Choice_c:
30 minutes, prevent overheating of a LHSI pmup motor running at
shutoff head
Choice_d:
60 minutes,prevent overheating of a LHSI pmup motor running at
shutoff head
Question #: 32 KA 004.K6.13
Which ONE of the following correctly describes the condition that will directly cause an automatic closure of the letdown orifice isolation valves (HCV-1200 A, B, C)? Choice_a: No charging pumps running Choice_b: Letdown high flow signal Choice_c: High VCT level Choice_d: Low VCT level COMMENT A - Correct- B - Incorrect C. Incorrect D. Incorrect References ND-88.3-LP-2C
Question #: 33 KA 008.K1.02
Unit one is operating at 100% power and all systems are in their normal configuration. CC discharge header radiation level and CC surge tank level are steadily rising. Which one of the following components, if leaking, can cause an automatic valve closure due to increased CCW flow?
References
ND-88.5-LP-1
Choice_a: COMMENT
RCP Thermal Barrier - CC-TV-120 A - Correct - High flow can cause CC-TV-120 or CC-TV-
140
B. & 0 - Plausable - can cause inleakage into CCW
Syatem - but now flow associations
C. Wrong - is isolated during normal operation.
Choice_b: Plausable due to can cause CCW inleakage - and has a
Primary Sample Cooler flow isolation of CC-HCV-108
Choice_c:
Excess Letdown Heat Exchanger
Choice_d:
Non-regenerative Heat Exchanger
Question #: 34 KA 010.K6.03
Given the following plant conditions for Unit 1: Reactor Power - 100% A 20% power reduction is required for emergent maintenance Boration is initiated to allow for the power reduction with NO rod movement As the down power is initiated the PZR Pressure Master Controller sticks at 30%
Which one of the following correctly describes the difference in the final conditions for this power reduction (PZR Pressure Master Controller sticks) when compared to the power reduction under normal conditions (PZR Pressure Master Controller working normally)?
Choice_a: . PZR pressure would be HIGHER during the down power with the PZR controller malfunction Choice_b: PZR pressure will be LOWER during the power reduction with the PZR controller malfunction Choice_c: PZR level will be HIGHER during the power reduction with the PZR controller malfunction Choice_d: PZR level will be LOWER during power reduction with the PZR controller malfunction COMMENT A - Correct - With the master pressure controller in manual, spray will not initiate as pressure increases in the PZR due to the increased level. B. Incorrect C. PZR Level will be the same as during a down power with the Master pressure controller in auto D. PZR Level will be the same as during a down power with the Master pressure controller in auto References ND-93.3-LP-5
Question #: 35 KA 012.G2.1.27
Unit 1 is at 100 % power Reactor Trip Breaker IIAII (RTA) is racked out for maintenance and closed Reactor Trip Bypass Breaker IIAII (BYA) is racked in and closed A Reactor trip signal for channel IIAII RPS is received Which one of the following correctly describes the operation of the Reactor trip and bypass breakers with the plant conditions given? ** Need licensee to verify what alarms would be in for this condition** Choice_a: Only Reactor Trip Breaker IIAII trips open Choice_b: Reactor Trip Breaker IIAII trips open and Reactor Trip Bypass Breaker IIAII trips open Choice_c: Only Reactor Trip Bypass Breaker IIAII trips open Choice_d: Reactor Trip Breaker IIBII trips open and Reactor Trip Bypass Breaker IIA IItrips open COMMENT A. Correct- B. Incorrect - IIAII Bypass breaker receives a trip signal from B RPS C. Incorrect - IIAII Bypass breaker receives a trip signal from B RPS D. Incorrect - A B RPS trip signal would cause this References ND-93.3-LP-10
Question #: 36 KA 013.K2.01
Unit 1 is at 90% power and ramping down due to excessive temperatures on the IIAII DC bus. During the shutdown a short caused the IIAII DC bus to deenergized and IIFII Transfer Bus feeder to open. Prior to any operator action, which one of the following correctly describes the status of the power supplies for the IIAII train SI components?
References
ND-90.3-PP-213
Choice_a: COMMENT
#1 EDG is running with its output breaker open. Local manual A. Correct - With a loss of power to the 1H bus #1 EDG
breaker operation is required to energize IIAII train SI components will start, but with a loss of the "A II DC bus the output
breaker will loose control power and is required to be
closed manually
B. Wrong - A loss of power to the IIFII transfer bus
deenergizes the 1H bus which powers the IIAII train SI
Choice_b: components. Plausable - E trans bus powers 2H Bus, and
#1 EDG is running with its output breaker open. IIAII train SI if normal power is available then EDG bkr would not
components have power available from Reserve Station Service shut.
Transformers
C. Wrong - EDG does not require DC bus to start.
D. Wrong - EDG does not require DC bus to start. A loss of
power to the II FII transfer bus deenergizes the 1H bus
which powers the IIAII train SI components.
#1 EDG must be started manually. Local manual breaker operation
is required to provide power to IIAII train SI components
Choice_d:
#1 EDG must be started manually. IIAII train SI components have
power available from Reserve Station Service Transformers
Question #: 37 KA 003.K3.01
Given the following plant conditions: * Reactor Startup is in progress below the ECP * Control Bank "C" is at 35 steps withdrawn * Reactor Coolant System pressure is 2235 psig and stable * Tavg is 547 of and stable * "A" Reactor Coolant Pump trips Which ONE of the following correctly describes the effect on the plant? Choice_a: Tavg will decrease Choice_b: The reactor will automatically trip Choice_c: Source Range count rate will decrease (with no reactor trip) Choice_d: RCS pressure will increase COMMENT A. - Correct B. - Incorrect. RCP trip not in effect C. - Incorrect. No effect on count rate until critical D. - Incorrect. No effect on RCS pressure. Not at power. Knowledge of the effect that a loss or malfunction of the RCPS will have on the following: RCS 3.7/4.0 References
ND-83-LP-3-DRR, Thermodynamic Terminology and Behavior of Water
Question #: 38 KA 076.K1.12
Which one of the following correctly describes the components isolated by placing a stop log in the 10 high level screenwall? Choice_a: One flow path to Unit 1 component cooling heat exchangers One flow path to both units' charging pump service water Choice_b: One flow path to Unit 1 component cooling heat exchangers One flow path to Unit 1 recirc spray heat exchangers Choice_c: One flow path to Unit 1 bearing cooling heat exchangers One flow path to both units' charging pum p service water Choice_d: One flow path to Unit 1 bearing cooling heat exchangers One flow path to Unit 1recirc spray heat exchangers COMMENT A. Correct B. Incorrect - RS Spray not on 10 on 1CIA C. Incorrect - Bearing cooling not on 10 - (1C) D. Incorrect - Bearing cooling not on 10 - (1C) References ND-89.5-LP-2
Question #: 39 KA 078.K2.01
Unit 2 is at 100 % power Unit 2 TB instrument air compressor (2-IA-C-1) is running in auto and maintaining normal air pressure. A Loss of off-site power and a Unit 1 81 have occurred. Which one of the following correctly describes the power source and status of the unit 2 TB instrument air compressor (2-IA-C-1) for the plant conditions given above?
References
ND-90.3-LP-5
Choice_a: COMMENT
2J Bus is deenergized A - Correct - 25J3 trip on unit 1 81 requires the 2J bus to
Compressor is not running be reenergized.
B - Incorrect - Compressor is powered from 2J
C - Incorrect - 2H is powered from 2EDG / Compressor is
powered from 2J
Choice_b:
2H Bus is energized from #2 EDG D. - Incorrect - 25J3 trip on unit 1 81 requires the 2J buss
Compressor is running in auto to be reenergized.
Choice_c:
2H Bus is deenergized
Com pressor is not running
Choice_d:
2J Bus is energized from #3 EDG
Compressor is running in auto
Question #: 40 KA 103.G2.1.12
Given the following plant conditions: * Unit 1 is currently in an INTERMEDIATE SHUTDOWN condition * Unit restart is scheduled to commence this shift * Tavg is 510 of * RWST is 50 of * Service Water temperature is 60 of * Containment temperature is 100 of * Containment pressure is 15 psig Which ONE of the following is the required action lAW Tech Specs to continue with the restart (if any)? Choice_a: No action with regard to containment pressure is required to continue to Power Operation. Choice_b: No action with regard to containment pressure is required to enter Hot Shutdown, however, the crew must restore Containment Pressure to within acceptable limits before entering Power Operatlon. Choice_c: The crew must restore containment pressure to within acceptable limits before entering Hot Shutdown, but can remain in Intermediate Shutdown indefinitely at the current containment conditions. Choice_d: The crew cannot enter Hot Shutdown without restoring containment pressure to within acceptable limits and must restore containment pressure within 1 hour while in Intermediate Shutdown COMMENT A - Incorrect. A mode change to power operation would not be allowed. B - Correct. With the containment air partial pressure outside the acceptable operation range, restore the air partial pressure to within acceptable limits within 1 hour or be in at least HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 30 hours C - Incorrect. The Unit is already below the T.S. required Hot Shutdown condition and is only req'd to be in HSD .. D - Incorrect. ISD is a lower mode than HSD. The crew can enter HSD mode and then have up to one hr to correct.. Memory Ability to apply technical specifications for a system. RO 2.9 SRO 4.0 References Tech Spec 3.8.D & 3.4.A.3
Question #: 41 KA 073.K1.01
Which ONE of the following correctly describes the automatic actions that occur when a HIGH alarm is received on Condenser Air Ejector Radiation Monitor, RM-SV-111?
References
Choice_a: COMMENT
OPENS SV-TV-102 (Air Ejector Containment Isolation TV) and A - Correct
CLOSES SV-TV-1 03 (Air Ejector Atmospheric Vent Isolation Valve).
8 - One correct one incorrect
C - Both incorrect
o - One correct one incorrect
Choice_b:
OPENS SV-TV-102 (Air Ejector Containment Isolation TV) and
OPENS SV-TV-1 03 (Air Ejector Atmospheric Vent Isolation
Valve).
Choice_c:
CLOSES SV-TV-1 02 (Air Ejector Containment Isolation TV) and
OPENS SV-TV-103 (Air Ejector Atmospheric Vent Isolation
Valve).
Choice_d:
CLOSES SV-TV-102 (Air Ejector Containment Isolation TV) and
CLOSES SV-TV-103 (Air Ejector Atmospheric Vent Isolation
Valve).
Question #: 42 KA 007.K4.01
Unit 1 is operating at 100% power and the following conditions exist in the Unit 1 PRT: Temperature - 130F Pressure - 8# Which one of the following correctly describes the action required lAW 1-0P-RC-11 for the given plant conditions?
References
ND-88.1-LP-3
Choice_a: COMMENT
The PRT must besprayed/drained to decrease A. Correct
temperature
B. Alarm for Pressure is 10#
C. venting/purging is only for pressure control
D. Alarm for Pressure is 10#
Choice_b:
The PRT must be sprayed/drained to decrease
pressure
Choice_c:
The PRT must be vented/purged to decrease
tem perature
Choice_d:
The PRT must be vented/purged to decrease
pressure
Question #: 43 KA 064A.4.06
Plant conditions are as follows:
- A Loss of Off-Site Power has occurred
- #1 Emergency Diesel Generator has started but failed to auto load
- It has been determ ined that the auto-closure circuit for 15H3, #1 EDG Output
Breaker, is inoperable and that 15H3 can be manually closed
II
- When the operator places the sync switch for 15H3 to 1I0N he observes 120 volts on
the "lncorntnq" meter, 0 volts on the "runnlnq" meter, and the synchroscope is
ll
stationary at 113-0'clock
Which ONE of the following correctly describes the actions (if any) required to close 15H3?
References
ND-90.3-LP-7 pg. 18
Choice_a: COMMENT
No additional action is necessary. Close 15H3. A. Correct because the synchroscope has been turned
on, there is no over-current ordifterential and the aux trip
relay does not need to be reset (ND-90.3-LP-7 pg. 18).
Therefore, all criteria for manually closing the breaker
are met.
B. Incorrect because the bus is dead. Raising EDG
speed will not synchronize the phases.
Choice_b: C. Incorrect, field flash PB does not need to be pushed.
Raise EDG speed until the synchroscope is turning slowly in the
fast direction, then close 15H3 at 1111 o'clock", D. Incorrect because raising the EDG voltage will not
raise running voltage. Incoming voltage is the EDG
voltage (not running voltage).
Choice_c:
Momentarily press the "field flash" pushbutton, close 15H3.
Choice_d:
Raise EDG voltage until the running meter indicates 120 volts,
close 15H3.
Question #: 44 KA 008.A1.04
Unit 1 conditions are as follows: A normal"plant cooldown is in progress Residual heat removal cooling is not yet in service The component cooling surge tank high-low level annunciator (VSP-D7) is alarming CCW surge tank level is decreasing slowly Which one of the following conditions would cause the plant indications given above?
References
O-VSP-D7
Choice_a: COMMENT
A tube leak in one of the com ponent cooling heat exchangers A. Correct - CCW is at a higher pressure than SW - a leak
would cause flow from the CCW system to the SW system.
B. Wrong - The RCP thermal barrier leak would cause
inleakage into the CCW system.
Choice_b: C. Wrong - Would only cause primary plant contraction
A leak in an reactor coolant pump thermal barrier heat exchanger
D. Wrong - The relief is located on the top of the tank
Choice_c:
An excessive primary plant cooldown rate
Choice_d:
A leaking relief valve on the com ponent cooling water surge tank
Question #: 45 KA 059.A4.03
During a unit 1 shut down the following Feedwater System conditions exist; IIAII MFP is in service Feed control is in the process of being transferred from the Feed Reg Valves to the Feed Reg Bypass Valves with the FRV*s and bypass valves sharing flow. Total feed flow is 3500gpm IIAII Main Feed Recirculation valve is in Manual for testing Indicated main feed flow to all S/Gls begins increasing without any operator action. Which one of the following correctly describes the cause of the increase in flow for the given plant conditions? Choice_a: Loss of all air to all S/G Feedwater Regulating Bypass Valves Choice_b: Loss of all air to all S/G Feedwater Regulating Valves Choice_c: "A" MFP discharge flowmeter failing low Choice_d: "A" MFP discharge flowmeter failing high COMMENT A. Correct - A loss of air will cause the valves to close and cause pressure to increase therefore increasing flow through FRV. During SID the bypass valves also bypass the flowmeters for the S/Gls B. Wrong - FRV*s fail closed = decreasing flow to S/Gls C. Wrong - If the Discharge flowmeter fails low the MFP recirc valves will open causing flow to decrease D. Wrong - In manual the valve will not shut if flow goes over 5000GPM NEW References ND-89.3-LP-3
Question #: 46 KA 022.K1.01
Unit 1*is operating at 100 % power The containment air cooling system is in Normal mode A complete loss of air to hand control valves 1- CC-HCV 101AlB &102A1B has occurred. Which one of the following correctly describes the status of the containment air cooling system if no operator actions taken? Choice_a: Component Cooling flow is isolated from the containment air coolers Chilled Component Cooling flow is isolated from the containment
- Choice_b:
Component Cooling flow is directed to the containment air coolers Chilled Component Cooling flow is isolated from the containment Choice_c: Component Cooling flow is isolated from the containment air coolers Chilled Component Cooling flow is directed to the containment Choice_d: Component Cooling flow is directed to the containment air coolers Chilled Component Cooling flow is directed to the containment COMMENT New A. Correct - Upon a loss of air all valves fail shut isolating cooling water flow. B. Incorrect - Upon a loss of air all valves fail shut isolating cooling water flow. C. Incorrect - Upon a loss of air all valves fail shut isolating cooling water flow. D. Incorrect - Upon a loss of air all valves fail shut isolating cooling water flow. All answers plausible if the failing positions of the supply valves are misunderstood. References ND-88.5-LP-1
Question #: 47 KA 026.A4.01
A Consequence Limiting Safeguards actuation (Hi-Hi CLS) is actuated. Which ONE of the following Containment Spray System manipulations can be accomplished from benchboard 1-1 in the Control Room without CLS signal being reset? Choice_a: Close MOV-RS-155A, Containment Recirculation Sump Isolation Valve. Choice_b: Close MOV-CS-102A, Caustic Additive Tank (CAT) Isolation Valve. Choice_c: Stop Containment Spray Pump 1A. Choice_d: Close MOV-CS-100A, Containment Spray Suction Valve. COMMENT A - Correct - Can be operated from benchboard even when CLS signal is present 8 - Wrong Cannot be operated until CLS signal is reset. Plausible - it receives an open signal from CLS actuation and is operated from benchboard 1-1 C. Wrong - Pump cannot be stopped from benchboard until CLS signal is reset D. Wrong - Cannot be operated until CLS signal is reset. Plausible - it receives an open signal from CLS actuation and is operated from benchboard 1-1 References ND-91-LP-6
Question #: 48 KA 039.K5.01
Which two (2) of the following four (4) conditions are MOST LIKELY to result in water hammer? 1. Overfilling a steam generator (YELLOW path condition for secondary inventory). 2. Rapidly heating up secondary piping using the MS trip valve bypass valves. 3. Reinitiating feedwater to a steam generator shortly after the feedring is uncovered. 4. Throttling the Condensate Pump discharge valve before securing the Condensate Pump.
References
References
1-0P-FW-001, Rev. 9, Motor Driven AFW Pumps Startup and Shutdown
1-0P-FW-002, Rev. 14, Turbine Driven AFW Pump Startup and Shutdown
Facility examination bank questions MS00010, MS00011, SD00001
ND-88.1-LP-4, Rev. 4, Steam Generators
ND-95.3-LP-43, Rev. 6, FR-H.3, Response to Steam Generator High Level
Choice_a: COMMENT
Conditions 1 and 2 A. Correct.
B. Incorrect. One correct condition (1), one incorrect
condition (4).
C. Incorrect. One correct condition (2), one incorrect
condition (3).
D. Incorrect. Two incorrect conditions (3 and 4).
Choice_b: 1. Correct per ND-95.3-LP-43 and ND-83-LP-10, Rev. 11,
Conditions 1 and 4 Applications of Fluid Phenomena.
2. Correct per Caution statements in 1-0P-FW-001 and -
002.
3. Incorrect. Per ND-88.1-LP-4, the main feed ring
bottom discharge holes are plugged and the ring is
outfitted with J-tubes in order to prevent quick drainage
of the feedring and minimize the chance of water
hammer.
Conditions 2 and 3 4. Incorrect.
A. Correct.
B. Incorrect. One correct condition (1), one incorrect
condition (4).
C. Incorrect. One correct condition (2), one incorrect
condition (3).
D. Incorrect. Two incorrect conditions (3 and 4).
Conditions 3 and 4
References
1-0P-FW-001, Rev. 9, Motor Driven AFW Pumps Startup
and Shutdown
1-0P-FW-002, Rev. 14, Turbine Driven AFW Pump
Question #: 49 KA 059.A4.08 Which ONE of the following correctly describes the operation of the Main Feed Reg Valve MIA station if power to the controller is lost in the associated instrument rack in the relay room? References
ND-93.1-LP-3
Choice_a: COMMENT
The MIA station will shift to manual and can be controlled A - Correct - Controllers that loose power from the Rack
manually from the MCR.. Room still have power supplied from the MCR for
manual operation
B - Incorrect - On loss of power controllers shift state
C - Incorrect - On loss of power controllers shift state
o - Incorrect - This is indicitave of a controller that has
lost power from the MCR not the Rack room
Choice_b:
The MIA station will not automatically shift and output of the
controller will increase to 100 % demand.
Choice_c:
The MIA station will not automatically shift and output of the
controller will decrease to 00/0 demand
Choice_d:
The MIA station will shift to Auto-Hold and can not be controlled
manually from the MCR.
Question #: 50 KA 061 K4.01
Which on of the following correctly describes the normal and backup sources of water supply to the AFW pumps: Norm a 1 _ Backup _ Choice_a: Emergency Condensate Storage tank (1-CN-TK-1) directly to AFWPs* suctions Emergency Makeup Tank (1-CN-TK-3) to AFW booster pumps Choice_b: Emergency Condensate Storage tank (1-CN-TK-1) to AFW booster pumps Fire main to AFW booster pumps Choice_c: Emergency Condensate Storage tank (1-CN-TK-1) directly to AFWPs* suctions Fire main to AFW booster pumps Choice_d: Emergency Condensate Storage tank (1-CN-TK-1) to AFW booster pumps Emergency Makeup Tank (1-CN-TK-3) to AFW booster pumps COMMENT A - Correct B - Incorrect - CN-TK-1 does not use Booster pumps. Firemain does not go to Booster pumps C - Incorrect - Firemain does not go to Booster pumps Plausible - Normal is correct D - Incorrect - CN-TK-1 does not use Booster pumps Plausible Back-up is correct References 1- ES-O.1
Question #: 51 KA 062.A3.05
During a plant tour you observe the following indications on UPS 1A2; Amber Light - Inverter Output Voltage Low All other Amber Lights are Dark Which of the following conditions correctly describes the condition of UPS 1A2? Choice_a: Supplied by MCC-1 H1-1 with the Manual Bypass Switch in Bypass Choice_b: Supplied by MCC-1 H1-1 with the Static Switch aligned to the alternate source. Choice_c: Supplied by MCC-1 H1-2 with the Manual Bypass Switch in Bypass Choice_d: Supplied by MCC-1 H1-2 with the Static Switch aligned to the alternate source. COMMENT A - Correct B - Incorrect - If static switch was aligned the alternate source supplying load light would be lit C. Wrong MCC - 1H1-2 supplies it's normal source D. Wrong MCC - 1H1-2 supplies tt's normal source References ND-90.3-LP-5
Question #: 52 KA 063.A2.01
The following Unit 1 conditions exist: -Unit 1 is at 100 % power, steady state conditions. -The operator notices the DC Ground Detection light is dim. . -During a Main Control Board walk-down, the operator observes that the white light for 1-FW-P-3A, IIIIAllII AFW pump, is out. Which one of the following correctly describes the cause of the white light being out for the plant conditions given above? Choice_a: A hard ground exists on the IIAII DC bus Choice_b: 111 H II bus is de-energized Choice_c: The IIBII DC bus indicates <75 volts DC Choice_d: Operation of the AFWP from the Auxiliary Shutdown Panel has been selected COMMENT A. Correct B. Incorrect - If 1H was deenergized the light on GO would not be dim C. Incorrect If B DC bus was <75 volts B afwp would be affected D. Incorrect - does not have effect on GO References ND-90.3-LP-6, obj A & D, and ND-88.1-LP-9, obj H & I
Question #: 53 KA 064.K3.03
A LOOP has occurred on Unit 1.
-AP-10.07 "Loss of Unit 1 Power" is in progress.
-The #1 EDG is the only source of power to the 1H Bus.
Which one of the following correctly describes why the "A" CCP may not be started, and the reason, for the plant conditions above?
Choice_a:
The "A" CC pump should not be started because the # 1 EDG
could become overloaded if a HI-HI CLS signal was in progress.
Choice_b:
The "A" CC pump should not be started because the # 1 EDG is
not sized to handle the starting current in this plant condition.
Choice_c:
The "A" CC pump can not be started because 15H9, Stub bus
supply breaker, is interlocked to prevent closure when the #1 EDG
is the only source of power to the 1H Bus.
Choice_d:
The'IA" CC pump can not be started because the pump breaker
15H10 is interlocked to prevent closure when # 1 EDG is the sole
source of power to the 1H Bus.
COMMENT
A. Correct, According to the note in AP-1 0.07 Loss of Unit 1 Power": When the EDG is the only source of power to an Emergency
II
Bus, the associated Component Cooling pump should not be in service if a Hi-Hi CLS is in progress.
B. Incorrect - not a starting current issue
C. Incorrect
D. Incorrect
References
1-AP-10.07
Question #: 54 KA 004.K5.15
The plant is operating at 800/0 power when annunciator EK-0703, 'Letdown Ht Ex Tube Outlet Hi Temp," alarms. The operator checks. the temperature and notes it is indicating 145 of. Which ONE of the following correctly describes the effect of this condition on reactor power and the reason? Choice_a: Power level will lower due to the increase in PCS boron concentration. Choice_b: Power level will rise due to the decrease in PCS boron concentration. Choice_c: Power level will lower due to MTC becoming more negative. Choice_d: Power level will rise due to MTC becoming less negative. COMMENT A. Correct. A higher temperature will cause the release of Boron from the mixed bed IX. B. Incorrect. Plausible if candidate believes that the mixed bed is more efficient at removing Boron from the RCS at higher tem peratures. C. Incorrect. Plausible, Power level would go down due to Boron increase, however MTC becomes less negative as boron is added to the core. D. Incorrect. Plausible, however MTC becomes more negative as boron is removed from the core. Memory Knowledge of the operational implications of the following concepts as they apply to the CVCS: Boron and control rod reactivity effects as they relate to MTC 3.3/3.5 References
ND-86.2-LP-8, POWER OPERATIONS
Question #: 55 KA 103.K3.02
Which one of the following correctly describes the required actions lAW Tech Spec 3.8.C if the containment airlock inner door is declared inoperable due to a failed leak test? ' Choice_a: Close the outer door in a maximum of 15 min only if RCS temperature is above 200F Choice_b: Close the outer door in a maximum of 15 min only if RCS temperature is above 350F Choice_c: Close the outer door in a maximum of 1 hour only if RCS temperature is above 200F Choice_d: Close the outer door in a maximum of 1 hour only if RCS temperature is above 350F COMMENT A. Correct lAW TS 3.8.C B Incorrect - TS applicable when temp >200F C. Incorrect - 1 hour requirement for containment integrity is an exception of this condition D. Incorrect - TS applicable when temp >200F References
TS 3.8C
Question #: 56 KA 079.K4.01
Unit 1 is operating at 100 % power Service air header pressure is 85# and decreasing Which one of the following correctly describes the system response for the plant conditions given above with no operator action? Choice_a: The instrument air compressor' (1-IA-C-1) will automatically start Choice_b: The Service air header isolation valve (1-SA-SOV-124) will automatically shut Choice_c: The condensate polishing air compressor (1-CP-AC-1) will automatically start Choice_d: The Condensate polishing service air header regulator (CP-PCV- 103) will automatically shut COMMENT A - Correct - IA is normally supplied from service air - IA compressors automatically start at 90# B - Incorrect -Manually operated valve C - Incorrect -service air is not normally aligned to the CP air system D - Incorrect -service air is not normally aligned to the CP air system References ND-92.1-LP-1
Question #: 57 KA 086.A4.05
What modes of initiation are available for the EHC Deluge system? Choice_a: Manually at the break-glass pull station. Manual push-button from the fire control panel in the MCR. Choice_b: Manually at the break-glass pull station. Automatically by heat detectors. Choice_c: Manual push-button at the deluge valve. Automatically by heat detectors. Choice_d: Manual push-button at the deluge valve. Manual push-button from the fire control panel in the MCR. COMMENT A - Correct - there is no automatic operation of the Deluge system for this area B - Incorrect - there is no automatic operation of the Deluge system for this area C - Incorrect - there is no automatic operation of the Deluge system for this area. Deluge valve operation is manual release of the valve References ND-92.2-LP-1
Question #: 58 KA 002.K6.02
Unit 1 is operating at 25% reactor power during a startup. -RCP 1C Trips. Which one of the following describes the expected steady state RCS flow following the trip of 1C RCP? Choice_a: Loops IIA" and IIB" will indicate 110 %. Loop "C" RCS flow will indicate 20%. Choice_b: Loops IIA" and IIB" will indicate 95%. Loop "C" RCS flow will indicate 0%. Choice_c: Loops "A" and "B II will indicate 95%. Loop "C" RCS flow will indicate 20%. Choice_d: Loops "A" and "B II will indicate 110 %. Loop "C" RCS flow will indicate 0%. COMMENT A- Correct References ND-95.1-LP-3
Question #: 59 KA 029.A3.01
Unit 1 is in Refueling Shutdown with fuel movement in progress. Containment purge is in operation. Unit 2 is at 100% power. Which one of the following correctly describes the automatic response (if any) of the Unit 1 Containment Purge system to a Unit 2 Safety injection signal? Choice_a: Containment Purge Fans trip Dampers to Auxiliary ventilation system shut Choice_b: Containment Purge Fans trip Containment isolation valves (MOV-100A/B/C/D) shut Choice_c: Containment Purge Fans trip Containment isolation valves (MOV-100A/B/C/D) shut Dampers to Auxiliary ventilation system shut Choice_d: Unit 2 81 will not affect the Unit 1 purge system COMMENT A -Correct B - Incorrect - Containment isolations are closed by high rad monitor signal C - Incorrect - Containment isolations are closed by high rad monitor signal D - Incorrect - Purge fans are tripped off during an SI in either plan References ND-88.4-LP-6 Rev. 4, p. 8
Question #: 60 KA 035.K1.01
Unit 1 conditions are as follows: AFW is being supplied by both MDAFW pumps. TDAFW pump is tagged out for maintenance. S/G "C" experiences a major feed line break at the point where the main feed line enters S/G C. Which ONE of the following sets of AFW flow rates are possible based on the given plant conditions? COMBINED FLOW TO S/Gs A & B FLOWTOS/G C Choice_a: 350 gpm 350 gpm
Choice_b:
700 gpm 350 gpm
Choice_c:
700 gpm 700 gpm
Choice_d:
Ogpm 700 gpm COMMENT A. Correct - AFW lines have flow cavitating venturis that limit the flow permissible to a SG with a: ruptured feed line to a maximum of 350 gpm B. Incorrect - MDAFWP are 350GPM C. Incorrect - MDAFWP are 350GPM D. Incorrect - AFW lines have flow cavitating venturis that limit the flow permissible to a SG with a ruptured feed line to a maximum of 350 gpm References
ND-89.3-LP-4
Question #: 61 KA 045.A1.05
Unit 1 was operating at 35% power.. A plant transient has caused the following plant parameter changes with no operator actions: Th increasing Tc increasing PZR Level increasing All S/G levels decreasing Which ONE of the following transients could cause these changes in the plant parameters? Choice_a: Turbine trip Choice_b: Outward rod motion Choice_c: Inward rod motion Choice_d: Main Steam rupture COMMENT A. Correct B. Incorrect - S/G Levels would not be decreasing C. Temperatures would not be increasing O. Tem peratures would not be increasing References
NO 86.2-LP-8 NO-93.3-LP-11
Question #: 62 KA 071.K4.04
A gaseous waste discharge is in progress
Annunciator RMA-C6 ("pROCESS VENT PART ALERT / HIli alarm) just alarmed
The rad monitor is reading above the Alarm setpoint
Which one of the following correctly describes the automatic actions that must be verified lAW ARP 0-RMA-C6?
Choice_a:
Shut GW-FCV-1 01 Decay Tank Bleed Isolation
Choice_b:
Secure CTMT Vacuum Pumps
Choice_c:
Shut GW-HCV-106 (NOUN NAME???)
Choice_d:
Shut GW-PCV-103 WGDT Pressure Control Valve
COMMENT
A. Correct
B. Wrong, Pumps are manually shut off after automatic actions are verified
C. Wrong. Valve is not automatically operated but is manually closed later in ARP
D. Valve is a PCV not a manual isolation valve but isolates the same line as FCV-101
References
ARP O-RMA-C6
Question #: 63 KA 015.K5.10
Which ONE of the following correctly describes why excore nuclear instrumentation must be adjusted over core life? Choice_a: The radial and axial fluxes shift over core life with fuel burnup. Choice_b: "Rod shadowlnq" becomes a greater influence on detector response as boron concentration is reduced. Choice_c: Aging of the detectors and electronic components introduces indication errors. Choice_d: Detector response degrades due to the high temperatures in containment over core life. COMMENT A - Correct B - Incorrect. Operate ARO throughout core life. C, 0 - Incorrect. Not significant factors Knowledge of the operational implications of the following concepts as they apply to the NIS: Ex-core detector operation 2.8/3.0 References ND-93.2-LP-5
Facility verify c and d incorrect.
Question #: 64 KA 001.G2.1.27
The following conditions exist: Reactor power is 45% and stable. Rods are in bank 110 11 position due to a failure in the 1AC power cabinet. One control rod is withdrawn 10 steps. Which ONE of the following statements correctly describes the change (if any) in rod insertion limits and shutdown margin? Choice_a: Rod insertion limits remain the same, shutdown margin remains the same. Choice_b: Rod insertion limits increase, shutdown margin remains the same. Choice_c: Rod insertion limits increase, shutdown margin increases. Choice_d: Rod insertion limits remain the same, shutdown margin increases. COMMENT A. Correct B. Incorrect - Limits and Margin don't change with rod position C. Incorrect - Limits and Margin don't change with rod position O. Incorrect - Limits and Margin don't change with rod position References NO-93.3-LP-3H
Question #: 65 KA 011.A2.10
The pressurizer level control selector switch is in the 459/460 position when a failure causes the following plant events to occur in sequence: (Assume no operator actions taken.) 1) Charging flow decreased to rn inirnurn 2) Pressurizer level decreased 3) Letdown isolated and PZR heaters tripped off 4) Pressurizer level increased until the reactor tripped on pressurizer high level Based on the above conditions, level channel failed _ Choice_a: 459; high Choice_b: 460; low Choice_c: 459; low Choice_d: 460; high COMMENT A. Correct - B Incorrect - Channel 460 is the lower channel and has no effect on level controller Failure of the channel low would cause charging to increase C. Incorrect - Failure of the channel low would cause charging to increase D. Incorrect - Channel 460 is the lower channel and has no effect on level controller References ND-93.3 LP-7
Question #: 66 KA G2.1.1
Which ONE of the following correctly describes the requirements for verifying a change in position to a throttle valve locked in a
throttled position lAW VPAP-1405, Independent, Simultaneous, And Documented Peer Checks Verifications?
Choice_a: Simultaneous Verification of valve position is required Choice_b: Simultaneous Verification of locking device installation is required Choice_c: Independent Verification of valve position is required Choice_d: Independent Verification of locking device installation is required COMMENT A - Correct References VPAP-1405, Rev. 5
Question #: 67 KA G2.1.14
During performance of a surveillance it is discovered that a motor operated valve that is required to be operated is deenergized for breaker maintenance. Which one of the following correctly describes the highest level of perm ission required to manually operate the valve lAW OP- AA-100? Choice_a: Shift Manager Choice_b: Electrical Maintenance Supervisor Choice_c: Manager of Nuclear Operations Choice_d: Unit Supervisor COMMENT A Correct lAW OP-AA-100 B. Incorrect C. Incorrect - MNO permission required for using switches as isolation D. Incorrect - not highest level of permission References
OP-AA-1 OO-REV a
Question #: 68 KA G2.1.7
Given the following plant conditions: - Reactor power is 90 0/0 - Control rod bank 0 is at 180 steps - Rod control is in automatic - The median delta T signal selector output fails high Which one of the following correctly describes an alarm that will annunciate as a result of the median delta T output failure? Choice_a: The 1G-H8 IIRod Bank 0 EXTRA-LOW Limit" alarm will be illuminated. Choice_b: The 1E-C8 nop (D)T II Rx TRIP alarm will be illuminated. Choice_c: The 1E-D8 nOT (D)T II Rx TRIP alarm will be illuminated. Choice_d: The 1G-H5 IIRod Bank A EXTRA-LOW Lirnlt" alarm will be illuminated. COMMENT A. Correct B. Incorrect - median delta T come off upstream of Median selector C. Incorrect - median delta T come off upstream of Median selector D. Incorrect - nAil control rods do not have an input from median Tave References ND-93.3-LP-3
Question #: 69 KA G2.2.22
Which ONE of the following. conditions requires entry into a Technical Specification action statement during Power operation? Choice_a: Tcold is 553 of. Choice_b: "A" diesel generator day fuel tank level contains 209 gallons. Choice_c: Fuel Oil below ground storage tanks contain 3550 gallons total. Choice_d: One Charging Pump is under clearance. COMMENT A. Incorrect. Tcold 552 of B. Correct. T.S. min. is 290 gal C. Incorrect. Min. onsite supply is 35,000 gal. D. Incorrect. Only need one operable High Head pump Knowledge of limiting conditions for operations and safety limits. 3.4/4.1 References T.S.3.16
Question #: 70 KA G2.2.13
Which one of the following correctly describes the safety requirements for using a Failed Open Air Operated Valve as vent/drain points lAW OP-AP-200 Equipment Clearance procedure? Choice_a: Air supply isolated and vented Valve position visually verified Choice_b: Jacking device installed Control panel switch tagged Choice_c: Air supply isolated and vented Control panel switch tagged Choice_d: Jacking device installed Valve position visually verified COMMENT A - Correct B - Incorrect - Tagging control panel switches is not a requirement. Jacking device required when using as an isolation boundary C - Incorrect - Tagging control panel switches is not a requirement o - Incorrect - Jacking device required when using as an isolation boundary References
OP-AA-200-REV 1
Question #: 71 KA G2.2.28
Given the following plant conditions:
* Unit 1 is in a Refueling Shutdown
* The Core is currently em pty
* Core onload is to commence this shift
* The current PCS Boron concentration is 2300 ppm
* Chern istry is adding Boron to increase PCS Boron concentration to 2350 ppm
II
* IIA SRM is out of service
II
* IIB SRM audible count rate is available in the MCR ONLY
* IIBII SRM count rate is 4 cpm
* All other equipment is Operable
Which ONE of the following correctly describes whether or not the MINIMUM requirements to allow fuel transfer operations to
commence this shift are met and the condition(s) necessary to continue with core onload?
Choice_a:
Minimum fuel transfer operating conditions are met.
Prior to the loading the 9th fuel bundle into the core, a minimum
count rate will need to be confirmed.
Choice_b:
Minim urn fuel transfer operating conditions are NOT met.
"A II & "B" SRM Detectors must be operable at this time and have
audible count rate from at least one SRM available in
Choice_c:
Minim urn fuel transfer operating conditions are met.
Prior to the loading the 9th fuel bundle into the core, the IIA" &
IIBII SRM detectors must be in service.
Choice_d:
Minimum fuel transfer operating conditions are NOT met.
IIA" & "B" SRM Detectors must be operable at this time and have
audible count rate from at least one SRM in both the MCR and in
containment.
COMMENT
A - Incorrect. Min fuel transfer conditions are NOT met. Requires 2 SRM detectors operable for chemistry changes.
B - Correct
C - Incorrect. Correct if chern istry changes were not occuring.
D - Incorrect. Audible count rate required in containment, not MCR.
Knowledge of new and spent fuel movement procedures
RO 2.6 SRO 3.5
References
Tech Spec 3.10-2
Question #: 72 KA G2.3.4
Which ONE of the following correctly lists the 10 CFR 20 Whole Body dose limit restriction and the documentation requirements which must be met if the whole body dose limit is to be exceeded? Choice_a: Cannot exceed 3 Rem/Qtr whole body dose. All planned special exposures must be documented. Choice_b: Cannot exceed 25 Rem whole body lifetime dose. Only planned special, exposures above the annual limit need to be documented. Choice_c: Cannot exceed 3 Rem/Qtr whole body dose. Only planned special exposures above the annual limit need to be documented. Choice_d: Cannot exceed 25 Rem whole body lifetime dose. All planned special exposures must be documented. COMMENT A. Incorrect. Only the doses credited over a year are utilized for documenting planned special exposure. B. Incorrect. A person could be at or near the 25 Rem lifetime limit and therefore would require a planned special exposure prior to obtaining 5 Rem in the current year. C. Incorrect. Only the doses credited over a year are utilized for documenting planned special exposure. D. Correct. Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized. RO 2.5 SRO 3.1 References 10 CFR 20
Question #: 73 KA G2.3.2
Operations has a task to be performed in the Auxiliary Building near a20 foot line source that reads 300 mr/hr at (2) feet. Two options exist to complete the assignment: Option 1: Operator A can perform the assignment in 1 hour, working at a distance of (4) feet from the source. Option 2: Operator B can perform the same task, using special extension tooling, in 2 hours working at a distance of (6) feet from the source. Which one of the following correctly describes the option that must be selected for the ALARA plan, and the associated total. personnel exposure? Choice_a: Option 1 with a total dose of 150 mrem. Choice_b: Option 1 with a total dose of 75 mrem. Choice_c: Option 2 with a total dose of 100 mrem. Choice_d: Option 2 with a total dose of 66 mrem. COMMENT A. Correct, B. Incorrect, This is the correct method, but the dose was calculated using the point source method. C. Incorrect, Line source - only 1 hour work D. Incorrect, point source @ 2 hours References ND-81.2-LP-3
Question #: 74 KA G2.4.49
Given the following plant conditions: -Reactor power is 75%. -Feed header pressure has dropped to 790 psig. Annunciator (1H-G8), FW PP DISCH HDR LO PRESS is in alarm. -Both Main Feedwater Pum ps are operating. Which ONE of the following is the required IMMEDIATE action of 1-AP-21.00? Choice_a: Start a third Condensate pump Choice_b: Manually trip the reactor Choice_c: Reduce reactor power to 65% or less Choice_d: Place feedwater regulating valves in manual COMMENT A. Correct - lAW AP-21.00 B. Incorrect - RX Trip only if RX Power is >85% C. Incorrect - Reduction of turbine load is to match steam and feed flows D. Incorrect - Action is from ARP 1H-G8 References AP-21.00
Question #: 75 KA G2.4.11
The following conditions exist: Unit 1 was operating at 100 % power when a control rod dropped 16 hours ago. The Unit is currently operating at 68% power. Which ONE of the following describes the proper method of recovery per AP-1.01 and the reason this method is necessary? Choice_a: Reactor power must be held constant below 75% while the rod is withdrawn at two (2) steps per hour. To prevent Xenon oscillations. Choice_b: Reactor power must be held constant below 75% while the rod is withdrawn at ten (10) steps per hour. To prevent rapid changes in local power densities that could cause DNB. Choice_c: Reactor power must be increased to between 70 - 74% while the rod is withdrawn at ten (10) steps per hour. To prevent rapid changes in local power densities that could cause DNB. Choice_d: Reactor power must be increased to between 70 - 74% while the rod is withdrawn at two (2) steps per hour. To prevent Xenon oscillations. COMMENT A - Correct References ND-93.3-LP-31; AP-1.01
Draft Submittal
(Pink Paper)
Senior Reactor Operator Written Exam
;S tt £Ly 02.!J08-eUJ/ .
IJfUrFT
Question #: 76 KA 007 EG2.4.4 Unit 1 is at 100% power when the following occurs:
* An RCS leak occurred at 1900.
* At 1901, AP-16.00 Immediate Actions are completed and the RO identifies a leak rate exceeding the
capacity of a charging pump.
* At 1902, a manual reactor trip is performed and the team initiates SI.
* At 1903, Annunciator G-B-l, Approach to Saturation Temp Alarm, is received.
It is now 1905 and you have completed reviewing the emergency actions levels per EPIP-l.OI. Which ONE of the following describes the required (1) event classification and (2) the latest time by which the State must be informed? (Reference provided) a. (1) Declare an Alert at 1905.
(2) Notify the State no later than 1930.
b. (1) Declare an Alert 1905.
(2) Notify the State no later than 1920.
c. (1) Declare a Site Area Emergency at 1905.
(2) Notify the State no later than 1930.
d. (1) Declare a Site Area Emergency at 1905.
(2) Notify the State no later than 1920.
Answer (d) Comments a. Incorrect - Plausible if candidate thinks pressurizer level can be maintained with SI and notification
time is 30 minutes after event initiation.
b. Incorrect - Plausible if candidate thinks pressurizer level can be maintained with SI and declaration
time must be based on time of event initiation. Notification time is correct.
c. Incorrect - Correct Declaration, Time is incorrect, but plausible if the candidate believes
notification is based off of 30 minutes after event initiation.
d. Correct - SAE based on RCS leak rate requiring Safety Injection and inability to maintain RCS
subcooling. (EAL TAB-B-3)
Question #: 77 KA 008AA2.26
Given the following Unit 1 conditions: -A PORV and its associated block valve are stuck open. -The reactor is tripped and SI has initiated. -RCS pressure is 1500 psig and lowering. -PZR level indicates 90% and rising. -The crew has just transitioned to E-1, Loss of Reactor or Secondary Coolant. -The STA recommends that one charging pump be secured to prevent the RCS from going solid and losing pressure control. Which ONE of the following correctly describes the action the crew must take for the given plant conditions and the reason for the action?
References
1-E-1 ND-95.3-LP-7 WOG ERG HES11BG Choice_a: COMMENT Do not secure a Charging Pump. SI flow must be maintained to make-up
for mass loss
A - Correct-
B - incorrect - does not meet criteria to secure CCP;
brittle fracture is not a concern at these temperatures in
the primary; still above RTT.
C - incorrect - securing the CCP is not a correct mitigating
Choice_b: strategy
Do not secure a Charging Pump. PZR level instrumentation is inaccurate due
to vapor space accident. o - incorrect - pressure instrumentation is not affected by
vapor space accidents
Choice c:
Securet Charging Pump to minimize mass loss from the RCS.
Choice_d:
Secure 1 Charging Pump to prevent the PZR from going solid and transferring
the steam bubble to the reactor vessel head.
Question #: 78 KA 038EG2.4.6
Given the following plant conditions: -Unit 1 was at 100% power. -A SGTR occurs on S/G B & C. -The crew is perfonning 1-E-3, "STEAM GENERATOR TUBE RUPTURE" and have completed step 8, "INITIATE RCS COOLDOWN" -RCS subcooling is 40°F and rising. -RCS pressure is 1000 psig and increasing. Which ONE of the following correctly describes the EOP mitigating strategy for these conditions?
References
WOG Guidance for ECA-3.1
ECA-3.1
Choice_a: COMMENT
Remain in 1-E-3, "STEAM GENERATOR TUBE RUPTURE II. A. Correct. For multiple tube failures, RCS pressure may
If RCS subcooling does not increase to> 50F prior to RCS temporarily decrease below ruptured SG pressure during
depressurization, transition to 1-ECA-3.1, "SGTR WITH LOSS OF cooldown. However, pressure and subcooling should
REACTOR COOLANT - SUBCOOLED RECOVERYII . quickly increase when the cooldown is completed. The
transition to ECA-3.1 is not necessary if subcooling
increases sufficiently after the cooldown is complete.
Choice_b: B. Incorrect. Transitioning to this procedure would be
Remain in 1-E-3, IISTEAM GENERATOR TUBE RUPTURE II. inappropriate.
If RCS subcooling does not increase to > 50F prior to RCS
depressurization, transition to 1-ECA-3.2, IISGTR WITH LOSS OF C. Incorrect. The transition to ECA-3.1 is not necessary if
REACTOR COOLANT - SATURATED RECOVERYII subcooling increases sufficiently after the cooldown is
complete. This will be analyzed later in step 18.
D. Incorrect. Transitioning to this procedure would be
inappropriate. A temporary loss of subcooling may be
Immediately transition to 1-ECA-3.1, IISGTR WITH LOSS OF indicated with multiple SGTRs.
REACTOR COOLANT - SUBCOOLED RECOVERyll.
Choice_d:
Immediately transition to 1-ECA-3.2, IISGTR WITH LOSS OF
REACTOR COOLANT - SATURATED RECOVERyll.
Question #: 79 KA 054AA2.04
Given the following Unit 1 plant conditions:
- Unit 1 has experienced a loss of all AC power
- ECA-O.O "Loss of All AC Power" is being performed at step 18, "Check Intact SG Levels"
- The Turbine Drive AFW pump is providing equal AFW flow to each S/G
- NR SG Levels are as follows:
A - 10% and rising slowly
B - 85% and rising rapidly
C - 25% and rising slowly
Which ONE of the following correctly describes the actions to be performed lAW ECA-O.O"Loss of All AC Power" prior to
depressurizing S/G's?
References
ECA - 0.0
Choice_a: COMMENT
Isolate steam from the "B" S/G to the TD AFW Pump A - Correct - lAW ECA-O.O step 18
Adjust the "B" S/G PORV lift setpoint to 1035 psig.
B. Incorrect - Securing AFW is not possible without
securing feed to all S/G. Shutting FW-140 will only
secure feed to one header.
C. Incorrect - the requirement is to maintain at least 1
Choice_b: S/G levels>12% or depressurization should be stopped. C
Isolate steam from the "B" S/G to the TD AFW Pump SG is intact and> 12%
Secure AFW to the "B" S/G by shutting 1-FW-140, TD AFW Pump
Discharge to "B" AFW Hdr D. Same as B&C
Choice c:
Raise "A" S/G NR level to a minimum of 12%
Adjust the "B" S/G PORV lift setpoint to 1035 psig.
Choice_d:
Raise "A" S/G NR level to a minimum of 12%
Secure AFW to the "B" S/G by shutting 1-FW-140, TD AFW Pump
Discharge to "B" AFW Hdr
Question #: 80 KA 055EG2.4.30
Given the following Unit 1 conditions: 10:00 AM Unit 1 is 100% reactor power 10:15 AM A loss of all onsite and offsite power for Unit 1 10:25 AM Initial Declaration and Notifications Made 10:30 AM LEOF/CEOF has been activated 10:35 AM Power is restored to Emergency Buses H & J and off site power is restored to all Service Buses Which ONE of the following correctly describes the (1) initial classification of this event, and (2) the required action after the plant is stable, lAW the appropriate EPIP?
(Reference provided) References
EPIP-1.01
EPIP -1.04
Choice_a: COMMENT
(1) Site Area Emergency A -Incorrect - SAE is correct, NRC concurrence is not required.
(2) Have LEOF/CEOF get concurrence from the NRC to terminate the event
B - Correct - Loss of on-site/off-site power yields a SAE (A-2).
C. Incorrect - Alert is due to loss of all ac for less than 15
Choice_b: min but is overridden by SAE of A.2
(1) Site Area Emergency
(2) Send out update notification of change in plant conditions D. Incorrect - Alert is due to loss of all ac for less than 15
min but is overridden by SAE of A.2
Choice_c:
(1) Alert
(2) Have LEOF/CEOF get concurrence from the NRC to terminate the event
Choice_d:
(1) Alert
(2) Send out update notification of change in plant conditions
Question #: 81 'KA WE05EG2.4.6
Given the following plant conditions: - Unit 1 has had a loss of both MFW Pumps - The Reactor was tripped automatically due to a Turbine Trip Signal - The crew has transitioned to 1-ES-O.1 "Reactor Trip Response" - Total AFW flow has been throttled to 300 gpm. - NR SG levels are: - A: 14% and lowering - B: 11% and lowering - C: 8% and lowering - RCS Tave is 548 OFwith steam dumps 8% open. Which ONE of the following correctly states the procedure and action to address the given plant conditions?
References
1-ES-0.1
Status tree F-3 Heat Sink
Choice_a: COMMENT
Remain in 1-ES-0.1 "Reactor Trip Response" A. Correct - ES-O.1 required SG levels to be raised to a
Increase AFW flow to increase all SG Levels to a minimum of 22% NR min of 22% Transition to FR-H.1 not appropriate - no red
lAW ES-0.1 path exists with one SG > 12%NR
B. Incorrect- Transition to FR-H.1 not appropriate - no
red path exists with one SG > 12%NR. Plausible if the
requirements for a red path on heat sink are not known
Choice_b:
GO TO 1-FR-H.1 "Loss of Secondary Heat Sink" C. Incorrect - 33% is the minimum SG control band level
Increase AFW flow to increase all SG Levels to a minimum of 22% NR for normal SG level control. ES-0.1 is the correct procedure.
lAW FR-H.1
D. Incorrect - Transition to FR-H.1 not appropriate - no
red path exists with one SG > 12%NR. Plausible if the
requirements for a red path on heat sink are not known.
33% is a min SG control band level for normal SG level
control.
Remain in 1-ES-O.1 "Reactor Trip Response"
Increase AFW flow to increase all SG Levels to a minimum of 33% NR
lAW ES-0.1
Choice_d:
GO TO 1-FR-H.1 "Loss of Secondary Heat Sink"
Increase AFW flow to increase all SG Levels to a minimum of 33% NR
lAW FR-H.1
Question #: 82 KA WE03EA2.1
Given the following plant conditions: - A LOCA has occurred outside of containment. - The crew is performing 1-ECA-1.2, LOCA Outside Containment. - Efforts to isolate the break were unsuccessful. Which ONE of the following correctly describes the required EOP transition for the given plant conditions?
References
1-ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION
1-E-O, REACTOR TRIP OR SAFETY INJECTIO
Choice_a: COMMENT
1-E-1 , Loss of Reactor or Secondary Coolant A. Incorrect. Plausible because there is a Loss of Reactor
Coolant in progress.
B. Correct. Step 2 RNO of ECA-1 .2 directs operators to
ECA-1.1 if efforts to isolate the leak are not successful.
C. Incorrect. Plausible because this is a normal transition
for long term cooling during a LOCA.
D. Incorrect. Plausible because the goal is to cool and
Choice_b: depressurize the RCS
1-ECA-1.1, Loss of Emergency Coolant Recirculation
Choice_c:
1-ES-1.2, Post LOCA Cooldown and Depressurization
Choice_d:
1-ES-1.1, SI Termination
Question #: 83 KA WE08EG2.4.20
Unit 1 plant conditions are as follows: - The Reactor is tripped - SI has been actuated - All S/G are faulted - All RCP's are secured The crew has transitioned to FR-P.1, "Response To Imminent Pressurizer Thermal Shock Condition", due to a RED PATH in INTEGRITY. While performing step 18, "Depressurlze RCS To Reduce Subcoollnq", PZR level rapidly rose from 30 to 55%. Which ONE of the following correctly describes the cause of the increase in PZR level and action(s) required based on plant conditions given?
References
E-1, LOSS OF REACTOR OR SECONDARY COOLANT
FR-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK
CONDITION
Choice_a: COMMENT
A steam void has formed in the upper head A - Correct lAW FR-P.1
Continue with RCS depressurization
B - Incorrect - depressurization isn't stopped until PZR
level is >69%
C - Incorrect - Accumulators are isolated below 2000#
lAW step 16
Choice_b:
A steam void has formed in the upper head 0- Incorrect - Accumulators are isolated below 2000#
Secure RCS depressurization and establish letdown lAW step 16
Choice_c:
SI Accumulators have injected
Secure RCS depressurization and establish letdown
Choice_d:
SI Accumulators have injected
Continue with RCS depressurization
Question #: 84 KA 061AG2.2.25
Both Units are at 100% power. The fuel handling team is loading a dry cask in the fuel building. 0-RMA-D5 VENT STACK #2 RAD MaN TRBL is illuminated Radiation Monitor 1-VG-RM-131-1 (Vent Stack 2 Monitor) has failed Which ONE of the following correctly describes the action(s) required per Technical Specifications for the plant conditions given above?
References
TRM 3.3.3
TS3.10
ODCM Att5
Choice_a: COMMENT
After placing the fuel assembly in a safe condition, suspend all
irradiated fuel movement in the fuel building. A - Correct lAW TRM 3.3.3 which references the specs of
TS 3.10
B - Plausible - Filtered exhaust is normally aligned during
refueling outages (Le., fuel movement).
C - Plausible - Actions required to verify release stopped and
required by 1-VG-RM-131-1 ARP Procedure.
D- Plausible - 12 hour sampling is required by ODCM
attachment 5 for effluent releases.
Stop fuel movement until filtered exhaust is aligned from the fuel building.
Choice_c:
Immediately trip supply fans associated with the Aux Bldg, Fuel Bldg,
Decon Bldg and Safeguards.
Choice_d:
Sample Vent Stack #2 every 12 hours during irradiated fuel
movement.
Question #: 85 KA 069AA2.02
Plant conditions are as follows:
Unit 1 is perfonning refueling operations
The Refuel pool is filled and the fuel transfer tube isolation valve is open.
Containment Purge is secured while I&C calibrates the Containment particulate and gas radiation monitors RM-RI-159/160.
Both personnel access hatch doors are Operable and open with dedicated individuals assigned to close the hatch if
required.
The MSTVs and the TO AFW Pump steam isolations are closed.
The turbine #2 Main Steam Stop Valve is disassembled for seat repair.
The TO AFW Pump steam admission valve, (PCV-MS-102A) is disassembled.
An electrical penetration has been removed and a blank flange has been bolted over the hole.
Which ONE of the following correctly describes what must be perfonned prior to authorizing irradiated fuel movement lAW OP-
4.13, Fuel Transfer System"?
II
References
Choice_a: COMMENT
Restore radiation monitors RM-RI-159/160 to Operable status and initiate A. Correct. Both doors of the hatch may be left open
Containment Purge. provided at least one door is Operable and capable of
being closed under administrative control, and
Containment Purge is in operation and it must be
capable of being secured by an Operable Containment
Purge Ventilation Isolation system.
Choice_b: B. Incorrect. Disabling the Containment Purge valves will
Disable the Containment Purge Valves and ensure air flow NOT satisfy the requirements for Refueling Ops with the
through the personnel access hatch is to Containment. personnel access doors open.
C. Incorrect. This opening in the main steam header is
downstream of the TO AFW Pump steam isolation valves.
D. Incorrect. An electrical penetration sleeve may be
isolated by a blank flange and satisfy containment
Ensure the TO AFW Pump steam admission valve, (PCV-MS-102A) has isolation.
been reassembled, leak tested, and closed.
Choice_d:
Reinstall the electrical penetration and ensure a satisfactory local
leak rate test has been perfonned.
Question #: 86 KA 013G2.1.23 Given the following plant conditions:
Unit 1 is at 100% power.
A valid Safety Injection signal results in a reactor trip and safety injection.
All systems functioned as designed.
Which one of the following correctly describes the most restrictive requirement to report this event in accordance with VPAP-2802, Notifications and ~~~. . (References Provided) a. A one (1) hour report. b. A four (4) hour report. c. An eight (8) hour report. d. A twenty-four (24) hour report. Answer (b) Comments a. Incorrect - Plausible if candidate thinks that since all systems functioned as designed, it is not reportable. b. Correct- Safety Injection actuation is a 4 hour report. c. Incorrect - Plausible - Reactor Trip and Safety Injection actuation are listed under the 8 hour notification criteria but the 4 hour report is
more limiting.
d. Incorrect - Plausible - a significant change in the normal operation is a statement in VPAP-2802 for 24 notification but the 4 hour report is
more limiting. .
NOTE - Candidates must be provided VPAP-2802 Section 6.3.3 thru 6.3.7.
Question #: 87 KA 039A2.03
Given the following plant conditions:
Unit 2 is at 100% power.
Annunciator 2A-A3 "N16 HIGH' illuminated.
2-MS-RI-290, "A" Steam Generator Radiation Monitor, is 190 GPO and stable.
Air ejector radiation monitor is 170 cpm and stable.
Charging line flow is 90 gpm and stable.
Letdown flow is 105 gpm and stable.
Combined seal return flow is 9 gpm and stable.
Total seal injection flow is 24 gpm and stable.
VCT Level is 50% and stable.
RCS Tave is stable.
Pressurizer level is 53% and stable.
Which ONE of the following describes the required actions for the plant conditions given above and the reason for the actions?
References
2A-A3, ANNUNCIATOR RESPONSE PROCEDURE
Choice_a: COMMENT
Reduce load to < 50% within an hour and be in HSO within the A. Incorrect. 1.S. 3.1-13.C If the primary-to-secondary
following 2 hrs. The Tech Spec primary to secondary leakage limit leakage through all steam generators not isolated from
is being exceeded. the Reactor Coolant System exceeds 1 gpm total and
500 gallons per day through anyone steam generator
not isolated from the Reactor Coolant System, reduce
the leakage rate to within limits within 4 hours or be in
hot shutdown within the next 6 hours and in cold
Choice_b: shutdown within the followinq 30 hours.
Be in HSO within the next 6 hrs and CSO within the followinq 30
hours. The Tech Spec primary to secondary leakage limit is B. Incorrect.
being exceeded.
C. Correct. The high limit alarm correlates to a 150 gpd
leak rate. The reactor must be less than 50% in one hour
and in HSO in the following two hours. This is to meet
the latest guidelines from EPRI concerning SGTLs.
Reduce load to < 50% within an hour and be in HSO within the O. Incorrect.
following 2 hrs. The EPRI guideline for primary to secondary
leakage is being exceeded.
Choice_d:
Reduce load to < 50% within an hour and be in HSO within the
followinq 6 hrs, The EPRI guideline for primary to secondary
leakage is being exceeded.
Question #: 88 KA 062A2.03
Unit 1 is operating at 100% power. A fire in Unit 1 AMSAC UPS and subsequent loss of all site power have just occurred. The UPS has been deeenergized. Station Service switchboards are still deenergized. 125V DC Switchboard 1-1 is deenergized. Which of the following correctly describes the source of power to the unit 1 plant computers, and the maximum time to restore power, for the plant conditions above?
Modification in progress at SPS. This question needs to be raised to the SRO level. References
AO 10.07
Electrical Service system drawings
Choice_a: COMMENT
TSC UPS Battery. Restoration of power is required within 15 min A Correct - Normal power is from station service power.
to prevent loss of power. Alternate power supplies are from the AMSAC and TSC
inverters. WI AMSAC deenergized the TSC battery is
provideing the only source of power. It has a minimum
supply of 15 min.
B. 2 hours supply is the minimum for the Black Battery
Choice_b:
TSC UPS Battery. Restoration of power is required within 2 hours C & D. AP-10.07 places power back though the AMSAC
to prevent loss of power. USP
Choice_c:
U-1 Black Battery. Restoration of power is required within 15 min
to prevent loss of power.
Choice_d:
U-1 Black Battery. Restoration of power is required within 2 hours
to prevent loss of power.
88.
In preparation for returning the Regulating Line Conditioner (RLC) for UPS 1A1 to service, the operator inadvertently opened the inverter output
breaker prior to clearing the tags on the RLC.
Which one of the following describes the consequences of this action?
a. Loss of power to Vital Bus I and IA only. No TS clock since DC charging capability is maintained.
b. Loss of power to Vital Bus I and IA and glLcharging capability to DC Bus1A. 24 hour TS clock to restore DC charging capability or 6
hour clock to HSD.
l of*.... . m Loss..~f. p.9Y.'!~r!9 'y!!a.l..~!J~ IJI ?~9 JIIA 9~.!Y.*.. NC? -r:S. ~1.9c.k .si':1.c~ DQcha.r9 ir:t9.90D?Q.i.ltlY .!~ r.n.~iQt?in.eg .... * ... m" - .' -{ Deleted: ~ I d. Loss of power to Vital Bus III and lilA and illLcharging capability to DC Bus 1A. 24 hour TS clock to restore DC charging capability or 6
hour clock to HSD.
Answer (a)
Comments
a. Correct
b. Incorrect - Plausible - 1A1 UPS supplies DC bus 1A but it also has an alternate charging source. 24 hour TS is plausible since this is the
clock for the EDG battery.
c. Incorrect - Plausible - VB-III/11IAare supplied from UPS 1A2 (via the same 480v power supply as UPS 1A1). Part 2 correct
d. Incorrect - Plausible 1A1 UPS supplies DC bus 1A but it also has an alternate charging source, but VB-III/IliA are supplied from UPS
1A2 (via the same 480v power supply as UPS 1A1). 24 hour TS is plausible since this is the clock for the EDG battery.
Question #: 89 KA 073GG2.1.10
Annunciator 0-RM-M5, "1-CC-RI-105 High" actuates and the control operator detennines that 1-CC-RI-105, CC Radiation Monitor, has failed. Which ONE of the following describes the action required by Technical Specifications when declaring 1-CC-RI-105 inoperable?
References
Ref: Surry ARP 0-RM-M5, 1-CC-RI-105 HIGH
Surry Tech Spec 3.13.C, CCW
ND-88.5-LP-1, CC, Obj G
Choice_a: COMMENT
Secure filling the CC Surge Tank. o Correct
A Incorrect. Plausible as assuming automatic functions occur the
candidate may believe CC should not be added to the system.
S Incorrect. Plausible as overflow of the CC Surge Tank could
yield radiation alarms in the Aux. Building.
C Incorrect. Plausible as 1-CC-RI-1 06 is a redundant Rad.
Monitor.
Choice_b:
Establish the Auxiliary Building Control Area Radiation Monitor
as a compensatory measure.
Choice_c:
Ensure that radiation monitor 1-CC-RI-1 06 is operable.
Choice_d:
Close the Component Cooling Water Surge Tank vent valve.
Question #: 90 KA 078GG2.1.20
Unit one was manually tripped due to a loss of instrument air. Instrument air can not be reestablished. Which of the following correctly describes the mitigation strategy of AP-40.00 Non-recoverable Loss Of Instrument Air?
References
AP-40
Choice_a: COMMENT
Establish natural circulation and place the plant in Cold Shutdown. A Correct - AP-40 directs tripping of RCP due to loss of
CCW. Also directs the CD and boration of the RCS
B. Wrong - Plausible - natural circ needs to be
established but the plant is not held in HSO
C/O Wrong - Forced circulation is not available unless IA
Choice_b: is restored.
Establish natural circulation and place the plant in Hot Shutdown.
Choice c:
Maintain forced circulation and place the plant in Cold Shutdown.
Choice d:
Maintaiil forced circulation and place the plant in Hot Shutdown.
Question #: 91 KA 034A2.01
Given the following plant conditions: Unit 1 is in a Refueling outage Irradiated Fuel is being offloaded from the core to the Fuel Building. A fuel element inadvertently struck the top of another fuel assembly not fully seated in the refuel pool Bubbles were seen coming out from around the base of the fuel element 1-RM-RM-153, Fuel Pit Bridge, increased sharply and then stabilized. No radiation monitors are in alarm. Per O-AP-22.0, FUEL HANDLING ABNORMAL CONDITIONS, which ONE of the following actions is required with regard to movement of the affected fuel assembly?
References
AP-22.0
Choice_a: COMMENT
The Fuel Handling SRO must immediately suspended fuel A. Correct. Fuel movement must be immediately
movement. stopped lAW AP-22.
B. Incorrect. Plausible if student believes that since rad
monitors are not in alarm that an immediate response is
not required.
Choice b: C. Incorrect. Plausible if student believes that rad
The Fuel Handling SRO will suspend fuel movement only after monitors must be in alarm to verify or suspect fuel
getting concurrence from the Shift Manager. damage for AOP entry. Placing the fuel in its
designated storage location would normally be a safe
location.
D. Incorrect. Plausible if student believes that the Fuel
Handling SRO is authorize to continue fuel movement
during conditions that indicate fuel failure without an alarm.
The Fuel Handling SRO will suspend fuel movement once the
fuel assembly is placed in its designated storage location.
Choice_d:
The Fuel Handling SRO can continue fuel movement.
Question #: 92 KA 002G2.4.20 Original question requires detailed knowledge of procedure that is never used and rarely reviewed. A Steam Break has occurred and all S/Gs are faulted. Which ONE of the following indicates the bases for maintaining a minimum flow of 60 gpm in all S/Gs per ECA-2.1, Uncontrolled Depressurization of All Steam Generators? a. This flow rate avoids dryout of the steam generator tubes AND is the minimum verifiable value on the MeR flow meter. b. This flow rate is the lowest at which the MOYs will control AND is the minimum verifiable value on the MCR flow meter. c. This flow rate avoids dryout of the steam generator tubes AND limits thermal stresses to the steam generator feed ring. d. This flow rate is the lowest at which the MOYs will control AND limits thermal stresses to the steam generator feed ring. Answer (a) Comments a. Correct Answer b. Incorrect - 1& part incorrect, AFW MOYs can control less than 60 gpm. 2nd part correct. c. Incorrect - 1& part correct, 2nd part incorrect, SG feed rings are full. d. Incorrect - 1& part incorrect, AFW MOYs can control less than 60 gpm. 2nd part incorrect, SG feed rings are full.
Question #: 93 KA 086A2.04 The Low Pressure C02 system in the normal switchgear room is locked out to prevent actuation due to a design calculation error. A fire breaks out in the normal switchgear room resulting in total destruction of all components/switchgear in the room. In addition to AP-48.00 (Fire), which ONE of the following sets of abnormal procedures will be used to mitigate the resulting plant transient? a. AP-10.07 (Loss of Unit 1 Power) and AP-39 (Natural Circulation). b. AP-10.05 (Loss of Semi-Vital Bus) and AP-40 (Non-Recoverable Loss of Instrument Air). c. AP-10.07 (Loss of Unit 1 Power) and AP-40 (Non-Recoverable Loss of Instrument Air). d. AP-10.05 (Loss of Semi-Vital Bus) and AP-39.00 (Natural Circulation). Answer (a) Comments a. Correct - AP-10.07 for loss of power to the Station Service Buses and AP-39.00 since the RCPs lose power. b. Incorrect - AP-10.05 is incorrect but plausible since the SVB loses power until the EDG loads on the bus. AP-40 incorrect but plausible
since the Station Air Compressors lose power and the Instrument Air Compressors start after EDGs load on the bus.
c. Incorrect - AP-10.07 is correct. AP-40 incorrect but plausible since the Station Air Compressors lose power and the Instrument Air
Compressors start after EDGs load on the bus.
d. Incorrect - AP-10.05 is incorrect but plausible since the SVB loses power until the EDG loads on the bus. AP-39.00 is correct.
Question #: 94 KA G2.1.11
Given the following Unit 1 conditions on 3/8/07: Initial power level is 85%. 1445 - AFD outside the target band but within the acceptable limits 1530 - AFD back within target band 1600 - AFD outside the target band but within the acceptable limits 1630 - Reactor power reduced to less than 50% 1730 - AFD back within target band 1830 - Reactor power is stable at 45% Based on these conditions, what is the EARLIEST time the plant can start a power ascension above 50% on 03/09/07?
References
TS 3.12.B.4.c
Choice_a: COMMENT
1600 A. Correct - TS require 24 hours with less than 1 hr
cumulative penalty.
B. Incorrect - @ 1730 there is no penalty left - .5 min for
every min when RX PWR is less than 50%
C. Incorrect - @ 1630 there is only 30 min penalty left
Choice_b:
1730 D. Incorrect - @ 1500 there is 1.5 hours of penalty left
Choice_c:
1630
Choice_d:
1500
Question #: 95 KA G2.1.20
Unit 1 is operating .at 100% power.
During an engineering review of the MCR Ventilation system, the operability of the 1-VS-E-4E ("E" MCR Chiller) was questioned due to
inadequate seismic supports for the associated service water piping.
The other four MCR Chillers are not impacted by this seismic concern.
OP-AA-102 "0 perability Determination" was initiated by the Shift Manager and it is determined that compensatory measures are required to
ensure proper seismic constraints.
Which one of the following correctly describes the required actions for the conditions given above? References OP-AA-102
Choice_a: COMMENT
Immediately declare the component inoperable. A. Correct - Immediate declaration is required
Restore operability status only after completion of the
Operability Determination. B. Restoration of operability does not require completion
of a RAS (RAS - only NS Comps).
C. Component must be immediately declared
inoperable. Restoration of operability does not require
Choice_b: completion of Operability Determination
Immediately declare the component non-functional.
Restore functionality only after completion of the Reasonable D. Component must be immediately declared inoperable.
Assurance of Safety Determination.
Choice_c:
Declare the component inoperable only if compensatory
measures cannot be implemented within 24 hours.
Restore operability status only after completion of the Operability
Determination.
Choice_d:
Declare the component non-functional only if compensatory
measures cannot be implemented within 24 hours.
Restore functionality only after completion of the Reasonable
Assurance of Safety Determination.
Question #: 96 KA G2.2.6
A change to an Operation's procedure is needed to change the level of use from "information only" to "reference". Which ONE of the following correctly describes this type of procedure change and the lowest approval authority required?
References
DNAP-0502
Choice_a: COMMENT
This is an INTENT change.
The change must be approved by a Senior Reactor Operator and A. Correct. A change in "Level of Use" is an INTENT
a Cognizant B designated individual. change and requires an SRO & Cognizant B designee
review.
B. Incorrect. A change in "Level of Use" is an INTENT
change and requires an SRO & Cognizant B designee
review.
Choice_b:
This is an INTENT change.
The change must be approved by a Senior Reactor Operator and C. Incorrect. A change in '.'Level of Use" is an INTENT
a Cognizant A designated individual. change and requires an SRO & Cognizant B designee
review.
D. Incorrect. A change in "Level of Use" is an INTENT
change and requires an SRO & Cognizant B designee
Choice_c: review.
This is a NON-INTENT change.
The change must be approved by a Reactor Operator and a
Cognizant A designated individual.
Choice_d:
This is a NON-INTENT change.
The change must be approved by
a Reactor Operator and a Cognizant B designated individual.
Question #: 97 KA G2.2.22
Unit one is at 100% power. RWST parameters are as follows; Temperature 50 degrees F Boron concentration 2350ppm Which of the following correctly describes the applicable Limiting Condition of Operation, and the associated Completion Time, for the given plant conditions?
References
TS 3.4
Choice_a: COMMENT
RWSTtemperature is above maximum limit and must be restored A. Correct - lAW TS 3.4.
within 8 hours.
B & D. Wrong- Boron concentration is within spec.
C. Wrong - 24 hour requirement applies to containment
spray subsystem
Choice_b: r
Boron concentration is below minimum specification and must be
restored within 8 hours.
Choice_c:
RWST temperature is above maximum limit and must be restored
within 24 hours.
Choice_d:
Boron concentration is below minimum specification and must be
restored within 24 hours.
Question #: 98 KA G2.3.8
Given the following plant conditions:
Unit 1 is at 100% power.
A WGDT contains 2% Hydrogen and 4% Oxygen
8 WGDT contains 10% Hydrogen and 2% Oxygen
Which ONE of the following correctly describes the action required per Technical Specifications, if any?
References
Choice_a: COMMENT
A WGDTs concentration of Oxygen is too high. Reduce the A. Correct. With the concentration of oxygen in the
Oxygen concentration to < 2%. waste gas holdup system greater than 2% by volume but
less than or equal to 4% by volume, reduce the oxygen
concentration to <2% within 48 hours.
8. Incorrect. With the OXygen concentration at 2% or less
there is no restriction on the Hydrogen limit.
Choice_b:
8 WGDTs hydrogen concentration is too high. Reduce the C. Incorrect. Plausible, Tech Spec states: The
Hydrogen concentration to <4%. concentration of oxygen in the waste gas holdup system
shall be limited to less than or equal to 2% by volume
whenever the hydrogen concentration could exceed 4%
by volume. Thls could be misconstrued to imply this is
correct.
D. Incorrect. Plausible if the candidate believes that the
A & 8 WGDTs are within the Tech Spec limits. No action is values have to be "less than the limit".
required.
Choice_d:
Both A & 8 WGDT Oxygen to Hydrogen % volume concentrations
are to high. The Oxygen and Hydrogen volumes in both WGDTs
should be decreased to below 2% and 4% respectively.
Question #: 99 KA G2.4.1
Given the following plant information: -The Unit was at 100% power when a turbine trip and a loss of power to both 4KV emergency busses occurred. -BOTH emergency diesel generators failed to start. -Reactor power is currently 95% and stable. -SG safety valves are open. Which ONE of the following describes the EOP mitigation strategy for this event?
References
Choice_a: COMMENT
Enter EGA-O.O, Loss Of All AG Power. At step 1, Transition to FR- A. Incorrect. No power to implement FR-S.1.
S.1, Response To Nuclear Power Generation/ATWS, to initiate
emergency boration. B. Incorrect. No power available to initiate emergency
boration.
G. Correct. FR's are monitored only until power is
restored.
Ghoice_b:
Enter E-O, Reactor Trip Or Safety Injection. At step 1, transition to D. Incorrect. E-Oassumes at least one 4KV emergency
FR-S.1, Response To Nuclear Power Generation/ATWS, to bus has power and therefore does not provide direction
initiate emergency boration. to restore power ..
Choice_c:
Enter EGA-O.O, Loss Of All AG Power. Monitor the GSF Status
Trees and upon restoration of power, transition to FR-S.1,
Response To Nuclear Power Generation/ATWS.
Ghoice_d:
Enter E-O, Reactor Trip Or Safety Injection. Monitor CSF Status
Trees and upon restoration of power, transition to FR-S.1,
Response To Nuclear Power Generation/ATWS.
Question #: 100 KA G2.4.28 Site security has reported a saboteur on site and subsequent sabotage of 1-CS-P-1A, "A" Containment Spray Pump. The saboteur did not have an opportunity to damage 1-CS-P-1B, "B" Containment Spray Pump, prior to his capture. Which one of the following correctly describes (1) the Technical Specification LCO Clock and (2) the highest classification of this event? (Reference Provided) a. (1) 72-hour clock
(2) Site Area Emergency
b. (1) 24-hour clock
(2) Site Area Emergency
c. (1) 72-hour clock
(2) Alert
d. (1) 24-hour clock
(2) Alert
Answer (b) Comments a. Plausible if the candidate thinks the containment spray clock is 72-hours (IRS Pump Clock). A site
area emergency is correct.
b. Correct. c. Plausible if the candidate thinks the containment spray clock is 72-hours (IRS Pump Clock). Two
indications on the Alert are met, but the site area emergency overrides this declaration.
d. Containment Spray TS is correct. Part 2 is plausible as two indications on the Alert are met, but the
site area emergency overrides this declaration.
}}