ML17298B131
| ML17298B131 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 10/25/2017 |
| From: | NRC/RGN-II |
| To: | Progress Energy Carolinas |
| References | |
| Download: ML17298B131 (464) | |
See also: IR 05000324/2017301
Text
ES-401, Rev. 10 BWR Examination Outline Form ES-401-1
Facility Brunswick Date of Exam: July 2017
RO K/A 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 2 3 4 * Total
1. 1 4 3 4 3 3 3 20
18 4 3 76
Emergency &
2 1 1 2 1 1 1 79 2 1 34
Abnormal Plant N/A N/A
Evolutions Tier Totals 5 4 6 4 4 4 27
27 6 4 10
10
1 2 2 3 2 2 3 3 3 2 2 2 28
26 3 2 55
2.
2 1 2 1 1 1 1 1 1 1 1 1 10
12 0 2 1 33
Plant
Systems Tier Totals 3 4 4 3 3 4 4 4 3 3 3 38
38 5 3 88
3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7
Categories
3 3 2 2 2 1 2 2
1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO
and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals
in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is
replaced by a K/A from another Tier 3 Category.)
2. The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by +/-1 from that specified in the table
based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
3. Systems/evolutions within each group are identified on the 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 K/A statements.
4. Select topics from as many systems and evolutions as possible; sample every system or evolution
in the group before selecting a second topic for any system or evolution.
5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be
selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
7. *The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics
must be relevant to the applicable evolution or system.
8. On the following pages, enter the K/A 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 K/A catalog, and enter the K/A numbers, descriptions, IRs,
and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
ES-401, REV 10 T1G1 BWR EXAMINATION OUTLINE FORM ES-401-1
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
295001AA1.08 Partial or Complete Loss of Forced 2.5 2.8 Standby liquid control: BWR-1........................
Core Flow Circulation / 1 & 4
295003AA2.01 Partial or Complete Loss of AC / 6 3.4 3.7 Cause of partial or complete loss of A.C. power.......
295004G2.1.30 Partial or Total Loss of DC Pwr / 6 4.4 4.0 Ability to locate and operate components, including local
controls.
295005AK1.01 Main Turbine Generator Trip / 3 4.0 4.1 Pressure effects on reactor power....................
295006AK1.02 SCRAM / 1 3.4 3.7 Shutdown margin.......................................
295016AK3.03 Control Room Abandonment / 7 3.5 3.7 Disabling control room controls.......................
295018G2.4.20 Partial or Total Loss of CCW / 8 3.8 4.3 Knowledge of operational implications of EOP warnings,
cautions and notes.
295019AK2.18 Partial or Total Loss of Inst. Air / 8 3.5 3.5 ADS: Plant-Specific..................................
295021G2.1.7 Loss of Shutdown Cooling / 4 4.4 4.7 Ability to evaluate plant performance and make operational
judgments based on operating characteristics, reactor
behavior, and instrument interpretation.
295023AK3.02 Refueling Acc Cooling Mode / 8 3.4 3.8 Interlocks associated with fuel handling equipment....
295024EK3.04 High Drywell Pressure / 5 3.7 4.1 Emergency depressurization............................
Page 1 of 2 9/7/2016 12:34 PM
ES-401, REV 10 T1G1 BWR EXAMINATION OUTLINE FORM ES-401-1
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
295025EK1.03 High Reactor Pressure / 3 3.6 3.8 Safety/relief valve tailpipe temperature/pressure
relationships........................................
295026EA1.03 Suppression Pool High Water Temp. / 3.9 3.9 Temperature monitoring................................
5
295028EK2.01 High Drywell Temperature / 5 3.7 4.1 Drywell spray: Mark-I&II.............................
295030EA1.01 Low Suppression Pool Wtr Lvl / 5 3.6 3.8 ECCS systems (NPSH considerations): Plant-Specific.............
295031EA2.04 Reactor Low Water Level / 2 4.6 4.8 Adequate core cooling..........................
295037EA2.06 SCRAM Condition Present and Power 4.0 4.1 Reactor pressure......................................
Above APRM Downscale or Unknown
/1
295038EK3.01 High Off-site Release Rate / 9 3.6 4.5 Implementation of site emergency plan.................
600000AK1.01 Plant Fire On Site / 8 2.5 2.8 Fire Classifications by type
700000AK2.03 Generator Voltage and Electric Grid 3.0 3.1 Sensors, detectors, indicators
Distrurbancecs
Page 2 of 2 9/7/2016 12:34 PM
ES-401, REV 10 T1G2 BWR EXAMINATION OUTLINE FORM ES-401-1
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
295009AA2.01 Low Reactor Water Level / 2 4.2 4.2 Reactor water level...................................
295017AK1.03 High Off-site Release Rate / 9 2.7 3.4 Meteorological effects on off-site release............
295020AK3.02 Inadvertent Cont. Isolation / 5 & 7 3.3 3.5 Drywell/containment pressure response...................
295029EK3.03 High Suppression Pool Wtr Lvl / 5 3.4 3.5 Reactor SCRAM.........................................
295032G2.4.11 High Secondary Containment Area 4.0 4.2 Knowledge of abnormal condition procedures.
Temperature / 5
295034EA1.01 Secondary Containment Ventilation 3.8 3.8 Area radiation monitoring system......................
High Radiation / 9
295036EK2.01 Secondary Containment High 3.1 3.2 Secondary containment equipment and floor drain system
Sump/Area Water Level / 5
Page 1 of 1 9/7/2016 12:35 PM
ES-401, REV 10 T2G1 BWR EXAMINATION OUTLINE FORM ES-401-1
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
203000A2.04 RHR/LPCI: Injection Mode 3.5 3.6 A.C. failures
205000K2.01 Shutdown Cooling 3.1 3.1 Pump motors
206000K3.02 HPCI 3.8 3.8 Reactor pressure control: BWR-2,3,4
209001G2.2.22 LPCS 4.0 4.7 Knowledge of limiting conditions for operations and safety
limits.
211000K3.01 SLC 4.3 4.4 Ability to shutdown the reactor in certain conditions
212000A4.16 RPS 4.4 4.4 Manually activate anticipated transient without SCRAM
circuitry/RRCS: Plant-Specific
215003K5.03 IRM 3.0 3.1 Changing detector position
215004K2.01 Source Range Monitor 2.6 2.8 SRM channels/detectors
215005A4.01 APRM / LPRM 3.2 3.1 IRM/APRM recorder
215005K3.07 APRM / LPRM 3.2 3.3 Rod block monitor: Plant-Specific
217000K4.01 RCIC 2.8 2.8 Prevent water hammer: Plant-Specific
Page 1 of 3 9/6/2016 9:26 AM
ES-401, REV 10 T2G1 BWR EXAMINATION OUTLINE FORM ES-401-1
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
218000A1.05 ADS 4.1 4.1 Reactor water level
218000G2.4.34 ADS 4.2 4.1 Knowledge of RO tasks performed outside the main control
room during an emergency and the resultant operational
effects
223002K4.06 PCIS/Nuclear Steam Supply Shutoff 3.4 3.5 Once initiated, system reset requires deliberate operaor
action
239002K6.02 SRVs 3.4 3.5 Air (Nitrogen) supply: Plant-Specific
259002A1.04 Reactor Water Level Control 3.6 3.6 Reactor water level control controller indications
261000K1.06 SGTS 3.0 3.1 High pressure coolant injection system: Plant- Specific
262001A3.04 AC Electrical Distribution 3.4 3.6 Load sequencing
262002A1.02 UPS (AC/DC) 2.5 2.9 Motor generator outputs
262002K1.15 UPS (AC/DC) 2.7 3.0 Stack gas monitors: Plant-Specific
263000A3.01 DC Electrical Distribution 3.2 3.3 Meters, dials, recorders, alarms and indicating lights
264000A2.07 EDGs 3.5 3.7 Loss of off-site power during full-load testing
Page 2 of 3 9/6/2016 9:26 AM
ES-401, REV 10 T2G1 BWR EXAMINATION OUTLINE FORM ES-401-1
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
300000A2.01 Instrument Air 2.9 2.8 Air dryer and filter malfunctions
300000K5.01 Instrument Air 2.5 2.5 Air compressors
400000K6.04 Component Cooling Water 2.8 2.9 Motors
400000K6.07 Component Cooling Water 2.7 2.8 Breakers, relays, and disconnects
Page 3 of 3 9/6/2016 9:26 AM
ES-401, REV 10 T2G2 BWR EXAMINATION OUTLINE FORM ES-401-1
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
201001K2.05 CRD Hydraulic 4.5 4.5 Alternate rod insertion valve solenoids: Plant-Specific
201003G2.1.27 Control Rod and Drive Mechanism 3.9 4 Knowledge of system purpose and or function.
202002K6.05 Recirculation Flow Control 3.1 3.1 Reactor water level
215002A3.03 RBM 3.1 3.1 Alarm and indicating lights: BWR-3,4,5
223001K4.03 Primary CTMT and Aux. 3.7 3.8 Containment/drywell isolation
230000K2.02 RHR/LPCI: Torus/Pool Spray Mode 2.8 2.9 Pumps
234000A1.03 Fuel Handling Equipment 3.4 3.9 core reactivity level
241000A4.18 Reactor/Turbine Pressure Regulator 2.9 2.8 Turbine shell warming: Plant-Specific
245000A2.01 Main Turbine Gen. / Aux. 3.7 3.9 Turbine trip
259001K5.03 Reactor Feedwater 2.8 2.8 Turbine operation: TDRFP's-Only
272000K3.03 Radiation Monitoring 3.2 3.4 Station area radiation monitoring
Page 1 of 2 9/6/2016 9:26 AM
ES-401, REV 10 T2G2 BWR EXAMINATION OUTLINE FORM ES-401-1
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
286000K1.04 Fire Protection 2.6 2.6 D.C. electrical distribution: Plant-Specific
Page 2 of 2 9/6/2016 9:26 AM
ES-401, REV 10 T3 BWR EXAMINATION OUTLINE FORM ES-401-3
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
G2.1.20 Conduct of operations 4.6 4.6 Ability to execute procedure steps.
G2.1.29 Conduct of operations 4.1 4.0 Knowledge of how to conduct system lineups, such as
valves, breakers, switches, etc.
G2.1.39 Conduct of operations 3.6 4.3 Knowledge of conservative decision making practices
G2.2.1 Equipment Control 4.5 4.4 Ability to perform pre-startup procedures for the facility,
including operating those controls associated with plant
equipment that could affect reactivity.
G2.2.6 Equipment Control 3.0 3.6 Knowledge of the process for making changes to
procedures
G2.2.7 Equipment Control 2.9 3.6 Knowledge of the process for conducting special or
infrequent tests
G2.3.15 Radiation Control 2.9 3.1 Knowledge of radiation monitoring systems
G2.3.7 Radiation Control 3.5 3.6 Ability to comply with radiation work permit requirements
during normal or abnormal conditions
G2.4.11 Emergency Procedures/Plans 4.0 4.2 Knowledge of abnormal condition procedures.
G2.4.26 Emergency Procedures/Plans 3.1 3.6 Knowledge of facility protection requirements including
fire brigade and portable fire fighting equipment usage.
Page 1 of 1 9/7/2016 12:37 PM
ES-401, REV 10 SRO T1G1 BWR EXAMINATION OUTLINE FORM ES-401-1
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
295006AA2.03 SCRAM / 1 4.0 4.2 Reactor water level...................................
295021G2.1.25 Loss of Shutdown Cooling / 4 3.9 4.2 Ability to interpret reference materials, such as graphs, curves, tables, etc.
295023AA2.04 Refueling Acc Cooling Mode / 8 3.4 4.1 Occurrence of fuel handling accident..................
295026G2.4.35 Suppression Pool High Water Temp. / 5 3.8 4.0 Knowledge of local auxiliary operator tasks during
emergency and the resultant operational effects
295028G2.1.23 High Drywell Temperature / 5 4.3 4.4 Ability to perform specific system and integrated plant procedures during all
modes of plant operation.
295038EA2.03 High Off-site Release Rate / 9 3.5 4.3 Radiation levels........................................
600000AA2.13 Plant Fire On Site / 8 3.2 3.8 Need for emergency plant shutdown
Page 1 of 1 9/7/2016 12:38 PM
ES-401, REV 10 SRO T1G2 BWR EXAMINATION OUTLINE FORM ES-401-1
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
295008G2.1.7 High Reactor Water Level / 2 4.4 4.7 Ability to evaluate plant performance and make operational judgments
based on operating characteristics, reactor behavior, and instrument
interpretation.
295014AA2.02 Inadvertent Reactivity Addition / 1 3.9 3.9 Reactor period........................................
295029EA2.02 High Suppression Pool Wtr Lvl / 5 3.5 3.6 Reactor pressure......................................
Page 1 of 1 9/7/2016 12:38 PM
ES-401, REV 10 SRO T2G1 BWR EXAMINATION OUTLINE FORM ES-401-1
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
206000A2.17 HPCI 3.9 4.3 HPCI inadvertent initiation: BWR-2,3,4
212000A2.13 RPS 3.8 3.9 Low condenser vacuum: Plant-Specific
239002G2.4.8 SRVs 3.8 4.5 Knowledge of how abnormal operating procedures are
used in conjunction with EOPs.
262002G2.4.11 UPS (AC/DC) 4.0 4.2 Knowledge of abnormal condition procedures.
264000A2.09 EDGs 3.7 4.1 Loss of A.C. power
Page 1 of 1 9/7/2016 12:39 PM
ES-401, REV 10 SRO T2G2 BWR EXAMINATION OUTLINE FORM ES-401-1
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
204000A2.02 RWCU 3.2 3.2 Pressure control valve failure: LP-RWCU
214000G2.1.25 RPIS 3.9 4.2 Ability to interpret reference materials such as graphs,
monographs and tables which contain performance data.
233000A2.11 Fuel Pool Cooling/Cleanup 2.9 3.2 Fuel pool gate seal high flow
Page 1 of 1 9/7/2016 12:39 PM
ES-401, REV 10 SRO T3 BWR EXAMINATION OUTLINE FORM ES-401-3
KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:
G2.1.43 Conduct of operations 4.1 4.3 Ability to use procedures to determine the effects on
reactivity of plant changes
G2.1.5 Conduct of operations 2.9 3.9 Ability to locate and use procedures related to shift
staffing, such as minimum crew complement, overtime
limitations, etc.
G2.2.11 Equipment Control 2.3 3.3 Knowledge of the process for controlling temporary
design changes.
G2.3.11 Radiation Control 3.8 4.3 Ability to control radiation releases
G2.3.4 Radiation Control 3.2 3.7 Knowledge of radiation exposure limits under normal or
emergency conditions
G2.4.30 Emergency Procedures/Plans 2.7 4.1 Knowledge of events related to system operations/status
that must be reported to internal orginizations or outside
agencies.
G2.4.35 Emergency Procedures/Plans 3.8 4.0 Knowledge of local auxiliary operator tasks during
emergency and the resultant operational effects
Page 1 of 1 9/7/2016 12:40 PM
ES-301 Administrative Topics Outline Form ES-301-1
Facility: Brunswick Steam Electric______________________ Date of Examination: _____________
Examination Level: RO Operating Test Number: __________
Administrative Topic (see Note) Type Describe activity to be performed
Code*
Perform DC Ground Isolation For Bus P, N, And
P/N Per OP-51
Conduct of Operations R, D
263000 A4.04
Ability to manually operate and/or monitor in the
control room - ground detection circuit
Importance 3.0/3.2
Evaluate Jet Pump Performance Per 0PT-13.1,
LOT-ADM-JP-002-02
Conduct of Operations R, D 2.1.23
Ability to perform specific system and integrated
plant procedures in all modes of plant
operations
Importance 4.3/4.4
Evaluate proposed temporary change
Equipment Control R, D
GEN 2.2.11
Knowledge of the process for controlling
temporary design changes
Importance 2.5/3.4
Determine TEDE While Working in a High
Airborne Area
Radiation Control R, N 2.3.4
Knowledge of Radiation Exposure Limits under
normal or emergency conditions.
Importance 3.2/3.7
N/A N/A
NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they
are retaking only the administrative topics (which would require all five items).
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom
(D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1; randomly selected)
Gty
ES-301 Administrative Topics Outline Form ES-301-1
Facility: Brunswick Steam and Electric__________________ Date of Examination: _____________
Examination Level: SRO Operating Test Number: __________
Administrative Topic (see Note) Type Describe activity to be performed
Code*
Determine Reportability Requirements
2.1.25
Conduct of Operations R, N Ability to interpret reference materials, such as
graphs, curves tables, ect.
Importance 3.9/4.2
Evaluate Jet Pump Performance Per 0PT-13.1,
LOT-ADM-JP-002-02
Conduct of Operations R, D 2.1.23
Ability to perform specific system and integrated
plant procedures in all modes of plant operations
Importance 4.3/4.4
Determine Post-Maintenance Testing
Requirements
Equipment Control R, N 2.2.21
Knowledge of pre- and post-maintenance
operability requirements.
Importance 2.9/4.1
Determine TEDE While Working in a High
Airborne Area
Radiation Control R, N 2.3.4
Knowledge of Radiation Exposure Limits under
normal or emergency conditions.
Importance 3.2/3.7
Classify an Event
2.4.29
Emergency Plan R, N Knowledge of the emergency plan.
Importance 3.1/4.4
NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they
are retaking only the administrative topics (which would require all five items).
- Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom
(D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes)
(N)ew or (M)odified from bank ( 1)
(P)revious 2 exams ( 1; randomly selected)
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2
Facility: Brunswick Steam and Electric_______________ Date of Examination: _____________
Exam Level: RO SRO-I SRO-U Operating Test No.: ______________
Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U
System / JPM Title Type Code* Safety
Function
a. Shift Running CRD pumps, by a shaft shear on the running A, N, S 1
pump.
b. SBGT System Operations to Reduce Humidity S, D 9
c. Place FW master controller in Service per GP-02 - master A, S, D 2
Controller Fails.
d. Place RHR In Suppression Pool Cooling Per AOP-36.2 S, E, EN, D 5
e. Restarting RCIC after AUTO initiation and Turbine Trip using A, D, E, EN, 4
the Hard Card - Controller Failure S
f. Perform Mechanical Over speed Trip Test of the main A, L, N, S 3
turbine IAW OPT-40.2.6, Turbine fails to trip by 1998 RPM
and must be manually tripped
g. Rod Worth Minimizer Functional Test - Failure To Enforce A, D, E, L, S 7
Blocks
h. Reduce RPV Water Level Using RWCU To Radwaste D, S 2
In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
i. 6.3.14 Transfer of Recirc VFD-CONT-UPS-A(B) from N, R 6
Inverter Operation to Maintenance Bypass
j. Restoring Seal Purge Flow with Pump Running - Seal D, R 1
Leakage Abnormal
k. Setting Service Air Dryer Sweep Value to Zero D, E, R 8
- All RO and SRO-I control room (and in-plant) systems must be different and serve different
safety functions; all five SRO-U systems must serve different safety functions; in-plant
systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO / SRO-I / SRO-U
A)lternate path 4-6 / 4-6 / 2-3
(C)ontrol room
(D)irect from bank 9/8/4
(E)mergency or abnormal in-plant 1/1/1
(EN)gineered safety feature 1 / 1 / 1 (control room system)
(L)ow-Power / Shutdown 1/1/1
(N)ew or (M)odified from bank including 1(A) 2/2/1
(P)revious 2 exams 3 / 3 / 2 (randomly selected)
(R)CA 1/1/1
(S)imulator
1. Unit Two is operating at 100% power, when the 2A Recirc Pump trips.
Which one of the following completes both statements below?
Annunciator A-05 (4-8), OPRM TRIP ENABLED, (1) be in alarm.
IAW 2AOP-04.0, Low Core Flow, reactor power must be reduced to less than (2)
power.
A. (1) will
(2) 50%
B. (1) will
(2) 60%
C. (1) will NOT
(2) 50%
D. (1) will NOT
(2) 60%
Page: 1 of 193 5/4/2017
K/A:
295001 Partial or Complete Loss of Forced Core Flow Circulation
AA1 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS
OF FORCED CORE FLOW CIRCULATION: (CFR: 41.7 / 45.6)
06 Neutron monitoring system
RO/SRO Rating: 3.3/3.4
Tier 1 / Group 1
K/A match: The Tier 1 aspect being tested is the Single Loop abnormal evolution (Partial Loss of Forced
Core Flow) of monitoring power (Nuetron Monitoring System) as it is manually reduced.
Pedigree: New
Objective: LOI-CLS-LP-302-C, Obj. 4
Q Reference: 2AOP-04.0, A-05(4-8)
Ref provided: None
Cog Level: Fundamental
Explanation: The trip of the recirc pump will lower core flow to less than 60% which would cause the
OPRM TRIP ENABLED alarm. IAW the AOP the action is to lower power to less than 50%
which prevents a trip based on THI.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect because - First part is correct. Second part is plausible because this is the power
level that the plant is lowered to for a loss of a different pump. (Reactor Feed Pump)
Choice C: Incorrect because- First part is plausible because the student may think this is set for less
than 50% flow. Second part is correct.
Choice D: Incorrect because - First part is plausible because the student may think this is set for less
than 50% flow. Second part is plausible because this is the power level that the plant is
lowered to for a loss of a different pump. (Reactor Feed Pump).
SRO Basis: N/A
Page: 2 of 193 5/4/2017
2. An electrical transient occurred with the following annunciators in alarm:
UA-15 (2-4), BUS E1 TO SUB E5 BKR TRIP
UA-15 (3-4), SUB E5 XFMR TEMP HIGH/480V GROUND
A field operator has reported local transformer winding temperature on Bus E5 is
reading 150°C
Which one of the following completes both statements below?
Bus E5 has de-energized due to (1) .
0AOP-36.1, Loss of any 4160V Buses or 480V E-buses, provides direction to cross-tie
480V buses E5 and (2) .
A. (1) a transformer fault
(2) E6
B. (1) a transformer fault
(2) E7
C. (1) high transformer temperature
(2) E6
D. (1) high transformer temperature
(2) E7
Page: 3 of 193 5/4/2017
K/A:
295003 Partial or Complete Loss of A.C. Power
AA2 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE
LOSS OF A.C. POWER : (CFR: 41.10 / 43.5 / 45.13)
01 Cause of partial or complete loss of A.C. power
RO/SRO Rating: 3.4/3.7
Tier 1 / Group 1
K/A match: The applicant must determine the cause of the loss of power to the E5 480V bus.
Pedigree: New
Objective: LOI-CLS-LP-50.2, Obj. 7.
Q Reference: 0AOP-36.1, UA-15(3-4)
Ref provided: None
Cog Level: Fundamental
Explanation: The temperature listed is within the normal operating temperature for a transformer. The trip
setpoint is 210 °C. OAOP-36.1 provides direction to cross-tie E5 to E6 only.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect because - First part is correct. Second part is plausible because cross tying an
emergency bus from the same division would be correct if conducted IAW 0AOP-36.1 at
the 4160V level (i.e. E1 to E3).
Choice C: Incorrect because- First part is plausible because the significant alarm is a high temperature
or fault alarm. Second part is correct.
Choice D: Incorrect because - First part is plausible because the significant alarm is a high temperature
or fault alarm. Second part is plausible because cross tying an emergency bus from the same
division would be correct if conducted IAW 0AOP-36.1 at the 4160V level (i.e. E1 to E3).
SRO Basis: N/A
Page: 4 of 193 5/4/2017
3. Unit One is at 100% power and the CRS has directed entry into 0AOP-39.0, Loss of
DC Power, due to DC Switchboard 1B Breaker GL6, Feeder for 125 VDC Panel 1B,
tripping open.
Which one of the following completes both statements below?
The (1) DC control power supply for DG2 has been de-energized.
IAW 0AOP-39.0, Emergency Diesel Generator DC control power transfer from normal
to alternate supply is performed at the (2) .
A. (1) normal
(2) DC Distribution Panels
B. (1) normal
(2) DG2 Excitation Control Panel
C. (1) alternate
(2) DC Distribution Panels
D. (1) alternate
(2) DG2 Excitation Control Panel
Page: 5 of 193 5/4/2017
K/A:
295004 Partial or Complete Loss of D.C. Power
G2.1.30 Ability to locate and operate components, including local controls. l (CFR: 41.7 / 45.7)
RO/SRO Rating: 4.4/4.0
Tier 1 / Group 1
K/A match: The K/A is met because the applicant must have knowledge of which DC panels supply
normal and alternate DC control power for each diesel generator and the method used to
swap control power supplies IAW 0AOP-39.0. This requires local operation.
Pedigree: New
Objective: LOI-CLS-LP-302-G, Obj. 3d. Given plant conditions and any of the following AOPs, determine
the required Supplemental Actions: AOP-39.0, Loss of DC Power.
Q Reference: 0AOP-39.0, Loss of DC Power, Rev. 042; 0OI-50, 125/250 and 24/48 VDC Electrical Load
List, Rev. 64
Ref provided: None
Cog Level: Fundamental
Explanation: Part 1: The normal control power supply for DG-2 is DC Distribution Panel1B. Part 2:
Control power for the diesel generators is shifted between the normal and alternate supply
by opening and closing breakers in the Excitation Control Panel in accordance with 0AOP-3.
Distractor Analysis:
Choice A: Part 1 is correct, see explanation. Part 2 is incorrect but plausible because Normal and
Alternate power supplies are swapped at the individual panels in other cases.
Choice B: Correct, See explanation
Choice C: Part 1 is incorrect because the alternate control power supply for DG-2 is DC Distribution
Panel 2B. Part 2 is incorrect but plausible because Normal and Alternate power supplies are
swapped at the individual panels in other cases.
Choice D: Part 1 is incorrect because the alternate control power supply for DG-2 is DC Distribution
Panel 2B. Part 2 is correct, see explanation.
SRO Basis: N/A
Page: 6 of 193 5/4/2017
4. Unit One is at 104 MWe IAW 0GP-05, Unit Shutdown.
Subsequently, the Unit One Bypass Valves have failed as-is.
Which one of the following completes both statements below?
The lowest pressure that would cause A-05 (3-5), REACTOR VESS HI PRESS,
is (1) .
If the Emergency Trip System Trip pushbutton is depressed IAW 1OP-26, Turbine
System Operating Procedure, the initial plant response is a(n) (2) in core power.
A. (1) 1050
(2) decrease
B. (1) 1050
(2) increase
C. (1) 1060
(2) decrease
D. (1) 1060
(2) increase
Page: 7 of 193 5/4/2017
K/A:
295005 Main Turbine Generator Trip
AK1 Knowledge of the operational implications of the following concepts as they apply to MAIN
TURBINE GENERATOR TRIP: (CFR: 41.8 to 41.10)
01 Pressure effects on reactor power
RO/SRO Rating: 4.0/4.1
Tier 1 / Group 1
K/A match: The KA is matched by having the applicant evaluate parameters during a plant shutdown that
result in the automatic bypass of a reactor trip signal with the failure of the steam dumps to
operate, which results in a pressure spike and increase in reactor power when the turbine is
tripped.
Pedigree: New
Objective: LOI-CLS-LP-026.2, Obj 15b
Reference: 0GP-05 Unit Shutdown rev180, 1OP-26 R96, 1APP-A-05 R76
Ref Provided: None
Cog Level: High
Explanation: The setpoint for A-05 3-5 [Reactor Vess Hi Press] is 1050 psig. The Turb CV Fast Clos/SV
Trip is bypassed so when the Emergency Trip System Trip pushbutton is depressed with the
failure of the bypass valves to open reactor pressure will increase, voids will collapse and
reactor power will increase.
Distractor Analysis:
Choice A: Incorrect because - The Turb CV Fast Clos/SV Trip is bypassed so when the Emergency
Trip System Trip pushbutton is depressed with the failure of the bypass valves to open
reactor pressure will increase, voids will collapse and reactor power will increase. It is
plausible as reactor power could remain unchanged if the bypass valves functioned or
reactor power could decrease if the bypass valves failed open, additionally the applicant
must evaluate the stem of the question to determine the reactor trip on turbine trip is
bypassed for the given plant conditions.
Choice B: Correct, See explanation.
Choice C: Incorrect because- The setpoint for A-05 3-5 [Reactor Vess Hi Press] is 1050 psig, 1060 psig
is plausible as that is the high pressure trip setpoint The Turb CV Fast Clos/SV Trip is
bypassed so when the Emergency Trip System Trip pushbutton is depressed with the failure of
the bypass valves to open reactor pressure will increase, voids will collapse and reactor power
will increase. It is plausible as reactor power could remain unchanged if the bypass valves
functioned or reactor power could decrease if the bypass valves failed open, additionally the
applicant must evaluate the stem of the question to determine the reactor trip on turbine trip is
bypassed for the given plant conditions.
Choice D: Incorrect because - The setpoint for A-05 3-5 [Reactor Vess Hi Press] is 1050 psig, 1060 psig
is plausible as that is the high pressure trip setpoint.
SRO Basis: N/A
Page: 8 of 193 5/4/2017
5. Following an automatic reactor scram the crew is evaluating whether the reactor will
remain shut down under all conditions without boron.
Which one of the following completes both statements below?
IAW 0OI-37.13, EOP Cautions and Tables Basis Document, the Maximum Subcritical
Banked Withdrawal Position (MSBWP) is position (1) .
If the MSBWP is met for all control rods, then LEP-02, Alternate Control Rod Insertion,
Section 2.2, Control Rod Insertion Verification, (2) required to be performed.
A. (1) 00
(2) is
B. (1) 00
(2) is NOT
C. (1) 02
(2) is
D. (1) 02
(2) is NOT
Page: 9 of 193 5/4/2017
K/A:
295006 SCRAM
AK1 Knowledge of the operational implications of the following concepts as they apply to SCRAM:
(CFR: 41.8 to 41.10)
RO/SRO Rating: 3.4/3.7
Tier 1 / Group 1
K/A match: Table Q-1, Shutdown Without Boron, is the determination that shutdown margin is satisfied
following any scram, and provides specific conditions based on control rod position
combinations or Reactor Engineering determination that the reactor is and will remain shut
down under all conditions without boron.
The Tier 1 aspect of the K/A is being met because Table Q-1 is associated with the SCRAM
procedure; the operational implication of shutdown margin is that the required rod position
is/is not met and the SCRAM procedure requirements for LEP-01, Section 2.2, Control Rod
Insertion Verification.
Pedigree: New
Objective: LOI-CLS-LP-300-J, Obj. 3
Q Reference: 0OI-37.13, Section 5.19; 0OI-37.3, Section 5.3, EOP-01-RSP, and 0EOP-01-LEP-02
Ref Provided: None
Cog Level: Fundamental
Explanation: 0OI-37.13, Section 5.19 states that the MSBWP is position 00. Step RSP-8 of the Reactor
Scram (flowchart) Procedure requires performing LEP-02; LEP-02 Step 7 directs when all
control rods are in (position 00), performance of Section 2.2, Control Rod Insertion
Verification is required.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Plausible because the title of LEP-02 is alternate control rod insertion; the applicant may
not know that LEP-02 is used for each and every scram. Incorrect because RSP, Step 8
directs performance of LEP-02 even when Table Q-1 requirements are met.
Choice C: Plausible because Table Q-1 includes special case where ten rods may be at position 02.
Incorrect because 0OI-37.13, Section 5.19 defines the Brunswick MSBWP is position 00.
Choice D: Plausible because Table Q-1 includes special case where ten rods may be at position 02 and
because the title of LEP-02 is alternate control rod insertion. Incorrect because 0OI-37.13,
Section 5.19 defines the Brunswick MSBWP is position 00, and because RSP, Step 8 directs
performance of LEP-02 even when Table Q-1 requirements are met.
SRO Basis: N/A
Page: 10 of 193 5/4/2017
6. Which one of the following identifies why ASSD procedures direct the Normal/Local
switches for the RHR equipment to be placed in Local?
A. To ensure automatic operation of RHR remains available.
B. To prevent spurious operation of control room RHR control circuits.
C. To prevent overloading DGs during spurious operation of RHR pumps.
D. To ensure divisional separation of RHR equipment required for safe shutdown.
K/A:
295016 Control Room Abandonment
AK3 Knowledge of the reasons for the following responses as they apply to CONTROL ROOM
ABANDONMENT: (CFR: 41.5 / 45.6)
03 Disabling control room controls
RO/SRO Rating: 3.5/3.7
Tier 1 / Group 1
K/A match: The K/A is met because the applicant needs to exhibit knowledge of why control room
controls are disabled during Control Room Abandonment.
Pedigree: New
Objective: LOI-CLS-LP-304, Obj 6
Q Reference: 0ASSD-2, Control Building, Rev 57; 0AOP-32.0 Rev 57
Ref Provided: None
Cog Level: Fundamental
Explanation: The controls for all Control Room functions are transferred to local control at the MCCs. At
that time, control circuit power to these specific components is redirected through different
control fuses. They remove valve position indication from the Control Room and eliminate
most valve interlocks. Specific functions are then available only locally.
Distractor Analysis:
Choice A: Incorrect - This is incorrect because the isolate switches actually prevents auto operation
of the system. This is plausible if the applicant believes the fire could affect auto actuation.
Choice B: Correct, See explanation.
Choice C: Incorrect - This is incorrect because even though spurious operation of the pumps may occur,
DG overload is not the reason they are isolated. This is plausible if the applicant thinks a fire
can spuriously start pumps.
Choice D: Incorrect - This is incorrect because these switches only serve to isolate the components from
the control room. This is plausible if the applicant thinks a fire in the Control Room could affect
train separation.
SRO Basis: N/A
Page: 11 of 193 5/4/2017
7. A reactor scram was inserted on Unit Two due to a complete loss of RBCCW.
IAW 0AOP-16.0, RBCCW System Failure, which one of the following identifies the
maximum time the CRD Pumps are allowed to be operated?
A. 1.5 minutes
B. 10 minutes
C. 20 minutes
D. 30 minutes
K/A:
295018 Partial or Complete Loss of Component Cooling Water
G2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.
CFR: 41.10 / 43.5 / 45.13)
RO/SRO Rating: 3.8/4.3
Tier 1 / Group 1
K/A match: Applicant knowledge of CRD pump operating restriction following a loss of cooling water
supply contained within 0AOP-16, is required to answer this question.
Pedigree: 2010 NRC Question
Objective: LOI-CLS-LP-008, Obj. 8f.
Q Reference: 0AOP-16.0 Rev. 31, 2OP-02 Rev 168, 2OP-21 Rev 93
Ref provided: None
Cog Level: Fundamental
Explanation: Loss of RBCCW will result in elevated CRD pump component temperatures which could
lead to CRD pump failure. Both pumps should be tripped if a total loss of RBCCW occurs,
but may be run for up to 20 minutes without RBCCW Cooling if directed by the SRO for rod
insertions or RPV level control.
Distractor Analysis:
Choice A: Incorrect but plausible because Reactor Recirculation pumps must be shut down within 90
seconds (1.5 minutes) of a loss of both seal injection and seal cooling flow.
Choice B: Incorrect but plausible because Reactor Recirculation pumps are allowed to operate for a
maximum of 10 minutes with no RBCCW cooling flow.
Choice C: Correct, See explanation.
Choice D: Incorrect but plausible because 30 minutes is the required Drywall cooldown time prior to
RBCCW pump restart.
SRO Basis: N/A
NRC 2010 Exam Question
Page: 12 of 193 5/4/2017
Page: 13 of 193 5/4/2017
8. Both units are at 100% power with Unit Two PNS loads being supplied by Unit One
PNS IAW 0OP-46, Instrument and Service Air System Operating Procedure.
0AOP-20.0, Pneumatic (Air/Nitrogen) System Failures, has been entered due to a
rupture on the Unit One PNS header outside of primary containment.
IAW the procedures above, which one of the following completes both statements
below?
To secure the PNS cross-tie alignment (1) PNS-CS-5804B control switch(es)
is(are) required to be in CLOSE position.
With PNS header pressure at 0 psig, a pneumatic supply to the Unit One ADS valves
(2) be aligned from the Backup Nitrogen System.
A. (1) either Units'
(2) requires manual action to
B. (1) either Units'
(2) will automatically
C. (1) both Units'
(2) requires manual action to
D. (1) both Units'
(2) will automatically
K/A:
295019 Partial or Complete Loss of Instrument Air
AK2 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR
and the following: (CFR: 41.7 / 45.8)
18 ADS
RO/SRO Rating: 3.5/3.5
Tier 1 / Group 1
K/A match: The PNS system contains no dedicated K/A category in NUREG-1123. Additionally, SRV
pneumatics can be supplied from IAN, PNS, and B/U Nitrogen. Applicant knowledge of PNS
system control and automatic actions of SRV pneumatic supply in the event of (complete)
PNS loss is required to answer this question.
Pedigree: Mod 2012 NRC
Objective: LOI-CLS-LP-046-A, Obj. 7d
Q Reference: 0AOP-20.0, Rev. 46
Ref Provided: None
Cog Level: Fundamental
Explanation: The PNS systems can be cross-tied, both units have a control switch for this. In the event
that header pressure degrades the operator would be expected to close the cross-tie valve
(Unit 1 only in this case). The cross-tie valve may be opened with only one switch in open
but requires both switches in the closed position to be closed. The Backup Nitrogen System
can supply pneumatics to SRV Accumulators when low pressure is sensed on the
Page: 14 of 193 5/4/2017
pp y p p
pneumatic header. PNS header pressure less than 95 psig is the initiation setpoint for the
backup nitrogen system.
Distractor Analysis:
Choice A: Incorrect because - Placing both Units in CLOSED is required to effect an isolation of
PNS. This is plausible as multiple switches are provided for and would require
manipulation if Unit 2 PNS was supplying Unit 1 loads. Manual action is not required to
align the backup nitrogen system, this is plausible because non-interruptible instrument air
can supply drywell pneumatics but requires manual action.
Choice B: Incorrect because- Placing both Units in CLOSED is required to effect an isolation of
PNS. This is plausible as multiple switches are provided for and would require
manipulation if Unit 2 PNS was supplying Unit 1 loads.
Choice C: Incorrect because - Manual action is not required to align the backup nitrogen system, this is
plausible because non-interruptible instrument air can supply drywell pneumatics but requires
manual action.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 15 of 193 5/4/2017
9. Alternate shutdown cooling using SRV's has been established. Plant conditions are:
RHR Loop A in torus cooling
RHR Loop B injecting to the reactor
Reactor pressure 115 psig
SRV B21-F013G open
It becomes desired to make an adjustment to lower the cool down rate.
[Reference Provided]
Which one of the following completes the statement below?
This can be accomplished by closing B21-F013G and opening SRV:
A. B21-F013A.
B. B21-F013J.
C. B21-F013K.
D. B21-F013L.
Page: 16 of 193 5/4/2017
K/A:
295021 Loss of Shutdown Cooling
G2.1.25 Conduct of Operations: Ability to interpret reference materials, such as graphs, curves, tables,
etc. (CFR: 41.10 / 43.5 / 45.12)
RO/SRO Rating: 3.9/4.2
Tier 1/ Group 1
K/A match: Must be able to interpret a table in the AOP for Loss of Shutdown Cooling.
Pedigree: Modified 2014 NRC exam
Objective: LOI-CLS-LP-302-L, Obj 3 -Given plant conditions and AOP-15.0, Loss of Shutdown Cooling,
determine the required supplementary actions. (LOCT)
Q Reference: 0AOP-15.0, Rev. 31
Ref provided: 0AOP-15.0, Rev 31, Pg 17
Cog Level: Fundamental
Explanation: Per table in 0AOP-15.0. SRV K is in block below SRV G in the column for RHR Pump B/D
and therefore opening SRV K would lower the cooldown rate.
Distractor Analysis:
Choice A: It is plausible because SRV A would be picked if using the column for RHR A/C but should not
be used as this is not the loop that is injecting.
Choice B: It is plausible because SRV J is below SRV G but should not be picked as the cooldown rate
will be equivalent per note above table.
Choice C: Correct, See explanation.
Choice D: It is plausible because SRV L is in block above SRV G but is wrong because it would increase
cooldown rate.
SRO Basis: N/A
Page: 17 of 193 5/4/2017
10. A Unit Two core offload was in progress when a seismic event occurred.
Subsequently, A-04 (6-7), FUEL POOL COOLING ALARM, alarms
IAW 0AOP-38.0, Loss of Fuel Pool Cooling, which one of the following completes both
statements below?
An entry condition into SCCP will first be required when Spent Fuel Pool water
temperature exceeds (1) .
0AOP-38.0 directs the use of (2) to maintain Unit Two Spent Fuel Pool Water
Temperature within a specified band.
A. (1) 125°F
(2) Supplemental Spent Fuel Pool Cooling System
B. (1) 150°F
(2) Supplemental Spent Fuel Pool Cooling System
C. (1) 125°F
(2) Alternate Decay Heat Removal System Primary Loop
D. (1) 150°F
(2) Alternate Decay Heat Removal System Primary Loop
Page: 18 of 193 5/4/2017
K/A:
295023 Refueling Accidents
AK2 Knowledge of the interrelations between REFUELING ACCIDENTS and the following:
(CFR: 41.7 / 45.8)
02 Fuel pool cooling and cleanup system
RO/SRO Rating: 2.9/3.2
Tier 1 / Group 1
K/A match: The following fuel pool cooling and cleanup system design basis function support mitigation
of refueling accidents: maintain the spent fuel water at a temperature equal to or less than
150°F. Knowledge of the entry conditions to 0EOP-03-SCCP could be reached for this
casualty and actions taken to mitigate the high temperature of the fuel pool in accordance
with 0AOP-38.0 are specific to the unit based on available systems.
Pedigree: New
Objective: LOI-CLS-LP-013, Obj. 10b, 10e
Q Reference: SD-13 rev 11, 0AOP-38.0 rev 33, 2APP-04 6-7, 0EOP-03-SCCP rev 10
Ref Provided: None
Cog Level: High
Explanation: With a core offload in progress and the fuel pool gates removed entry into 0EOP-03-SCCP
is not required until fuel pool water exceeds 150 °F. For Unit 2, 0AOP-38.0 directs the use of
ADHR system to maintain temperature in the event of a loss of Spent Fuel Pool Cooling.
Distractor Analysis:
Choice A: Incorrect because - Entry into 0EOP-03-SCCP is not required at 125°F given the plant
conditions. Plausible because if the fuel pool gates are not removed entry into
0EOP-03-SCCP is required at 125 °F. The second part is plausible because for Unit 1,
0AOP-38.0 directs the use of Supplemental Spent Fuel Pool Cooling system to maintain
temperature, Alternate Decay Heat Removal system Primary Loop is plausible given that
is a directed action for Unit 2
Choice B: Incorrect because - Part 1 is correct. The second part is plausible because for Unit 1,
0AOP-38.0 directs the use of Supplemental Spent Fuel Pool Cooling system to maintain
temperature, Alternate Decay Heat Removal system Primary Loop is plausible given that
is a directed action for Unit 2.
Choice C: Incorrect because - Entry into 0EOP-03-SCCP is not required at 125 °F given the plant
conditions. Plausible because if the fuel pool gates are not removed entry into 0EOP-03-SCCP
is required at 125 °F. Teh second part is correct.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 19 of 193 5/4/2017
11. Which one of the following, if violated, would result in an Emergency Depressurization
IAW the PC/P leg of PCCP?
A. SRV Tailpipe Level Limit Curve
B. PSP, Pressure Suppression Pressure, graph
C. HCTL, Heat Capacity Temperature Limit, graph
D. PCPL-A, Primary Containment Pressure Limit, graph
Page: 20 of 193 5/4/2017
K/A:
295024 High Drywell Pressure
EK3 Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL
PRESSURE: (CFR: 41.5 / 45.6)
04 Emergency depressurization
RO/SRO Rating: 3.7/4.1
Tier 1 / Group 1
K/A match: Applicant knowledge of the overall mitigative strategy employed during performance of the
PC/P leg of PCCP as it relates to emergency depressurization is required to answer this
question.
Pedigree: New
Objective: LOI-CLS-LP-300-L, Obj. 5-7
Q Reference: 0EOP-02-PCCP rev. 12; 0EOP-01-UG rev 67
Ref Provided: None
Cog Level: High
Explanation: Entering the UNSAFE region of the PSP graph is the reason for performance of an ED while
performing the PC/P leg of PCCP. This is part of the overall mitigative strategy used while
conducting the PC/P leg of PCCP.
Distractor Analysis:
Choice A: Incorrect because - The SRV Tailpipe Level Limit is a PCCP limiting condition during High
Torus Water Level conditions (T/L leg). Although exceeding the numerical value of the
SRV Tailpipe limit does result in corresponding entry into the UNSAFE region of the PSP
graph (+6), this is an incorrect answer choice as the stated question requests the reason
for ED IAW the PC/P leg of PCCP.
Choice B: Correct, See explanation.
Choice C: Incorrect because - Exceeding the limits of the Heat Capacity Temperature Limit requires an
ED IAW the Torus Temperature (T/T) leg of PCCP, not the PC/P leg of PCCP.
Choice D: Incorrect because - Exceeding the limits of PCPL-A IAW the PC/P leg of PCCP requires
operator action to vent containment IAW 0EOP-01-SEP-01. While entry into the UNSAFE
region of PCPL-A could result in corresponding UNSAFE region entry of PSP, this remains an
incorrect answer choice as the stated question requests the reason for ED IAW the PC/P leg
of PCCP. Additionally, based on the given conditions provided, PSP exceedance would occur
before PCPL-A exceedance.
SRO Basis: N/A
Page: 21 of 193 5/4/2017
12. Initial Conditions:
A spurious Group 1 isolation occurred on Unit One
Pressure peaked at 1135 psig before decreasing
Pressure decreased to 870 psig before rising
Current Conditions:
SRVs are being manually operated IAW the EOPs
Reactor power is 13.8%
Reactor pressure is stable at 1000 psig
Torus pressure is 5.3 psig
Which one of the following completes both statements below?
(1) SRVs are currently open to maintain reactor pressure stable at 1000 psig.
For the currently open SRVs, the SRV tailpipe temperature recorder, 1-B21-TR614,
will read approximately (2) .
A. (1) Two
(2) 300°F
B. (1) Two
(2) 550°F
C. (1) Four
(2) 300°F
D. (1) Four
(2) 550°F
Page: 22 of 193 5/4/2017
K/A:
295025 High Reactor Pressure
K1 Knowledge of the operational implications of the following concepts as they apply to HIGH
REACTOR PRESSURE: (CFR: 41.8 to 41.10)
03 Safety/relief valve tailpipe temperature/pressure relationships
RO/SRO Rating: 3.6/3.8
Tier 1 / Group 1
K/A match: The proposed question tests the applicants knowledge of a high reactor pressure emergency
evolution (Group 1/ATWS), Step RC/P-3, and the SRV tailpipe temperature/pressure
relationship.
Pedigree: Hatch 2007
Objective: LOI-CLS-LP-020 Obj 15 H: Given plant conditions, predict how ADS/SRVs will be affected by
the following:MISV Closure
Q Reference: SD-25, SD-20, A-3, 1-1
Ref provided: Steam Tables
Cog Level: High
Explanation: Eleven SRVs exist at BNP, each valve has ~ 830,000 lbm/hr capacity; 100% rated steam
flow is ~12.781 Mlbm/hr; the lift set point for 4 SRVs is 1130 psig (4 other SRVs lift at 1140
psig, and 3 others lift at 1150 psig). Initially 4 SRVs lifted because reactor pressure peaked
at 1135 psig. In order to accommodate 13.8% power (1.76 Mlb/hr), two SRVs must remain
open. The isenthalpic throttling process results in tailpipe temperature much less than
saturation temperature for 1000 psig. With a given 5.3 psig (corresponding to ~ 20 psia), a
saturation temperature of exiting steam from an open SRV would read approximately 300°F.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Plausible because 550°F is the saturation temperature for 1000 psig reactor pressure.
Incorrect because the isenthalpic throttling process results in tailpipe temperature at torus
pressure.
Choice C: Plausible because 4 SRVs initially opened (four amber lights). Incorrect because only 2 SRVs
are open to stabilize pressure at 13.8% power
Choice D: Plausible because 4 SRVs initially opened (four amber lights) and because 550°F is the
saturation temperature for 1000 psig reactor pressure. Incorrect because the operator
manually opened SRVs to stabilize reactor pressure at 950 psig, which is two SRVs given
13.8% power. Also incorrect because tailpipe temperature will be saturation temperature for
torus pressure.
SRO Basis: N/A
Page: 23 of 193 5/4/2017
13. Unit Two is at 75% power with HPCI testing in progress IAW 0PT-9.2, HPCI System
Operability Test.
Suppression Pool temperatures were recorded at the following times:
Time Temperature
1300 97° F
1500 107° F
Which one of the following completes both statements below?
The earliest time blue bar annunciator UA-12 (5-4), SPTMS DIV I BULK WTR TEMP
SETPOINT TS1, was at (1) .
At the time selected above, entry into PCCP (2) required.
A. (1) 1300
(2) is
B. (1) 1300
(2) is NOT
C. (1) 1500
(2) is
D. (1) 1500
(2) is NOT
Page: 24 of 193 5/4/2017
K/A:
295026 Suppression Pool High Water Temperature
EA1 Ability to operate and/or monitor the following as they apply to SUPPRESSION POOL HIGH
WATER TEMPERATURE: (CFR: 41.7 / 45.6)
03 Temperature monitoring
RO/SRO Rating: 3.9/3.9
Tier 1 / Group 1
K/A match: The K/A is matched by having testing in progress that adds heat to the suppression pool but
where the RHR system is not being used correctly to maintain the temperature. The operator
is given temperature information and asked when other indication (alarm) in the control room
will alert the operator to an abnormal suppression pool average temperature condition. To
test the tier 1 aspect of the K/A the operator must evaluate the conditions and determine that
entry into the EOP for primary containment control is not warranted with testing in progress.
Pedigree: New
Objective: LOI-CLS-LP-300-L, Obj 2
Q Reference: 2APP-UA-12 rev 28, 0EOP-02-PCCP rev 12
Ref Provided: None
Cog Level: High
Explanation: The alarm setpoint for the UA-12 5-4 [SPTMS DIV I BULK WTR TEMP SETPOINT] is 95°F
so at 1300 this would be the earliest time this would be in alarm and since testing is in
progress the EOP for primary containment control is entered at 105°F, not 95°F, so entry is
not required.
Distractor Analysis:
Choice A: Incorrect because - Entry is not required at 1300 since with testing in progress the EOP
entry criteria has not been reached, plausible if the applicant incorrectly evaluates the
stem of the question (i.e. no testing in progress) or incorrectly believe entry is required at
95°F.
Choice B: Correct, See explanation.
Choice C: Incorrect because - The alarm setpoint is 95°F which occurred prior to 1300, plausible if the
applicant does not know the setpoint or believes that the setpoint is 105°F (the entry condition
for the EOP with testing in progress). Entry is not required at 1300 since with testing in
progress the EOP entry criteria has not been reached, plausible if the applicant incorrectly
evaluates the stem of the question (i.e. no testing in progress) or incorrectly believe entry is
required at 95°F.
Choice D: Incorrect because - The alarm setpoint is 95°F which occurred prior to 1300, plausible if the
applicant does not know the setpoint or believes that the setpoint is 105°F (the entry condition
for the EOP with testing in progress).
SRO Basis: N/A
Page: 25 of 193 5/4/2017
14. Following a small break LOCA on Unit Two, drywell sprays have been secured.
Current plant conditions are:
Drywell pressure 7 psig
RPV water level 187 inches
Reactor pressure 300 psig
Torus water level -20 inches
Drywell average air temperature 280°F and rising
[Reference Provided]
Which one of the following completes both statements below?
IAW PCCP the first drywell temperature threshold that requires initiation of drywell
sprays is before reaching (1) .
IAW 0EOP-01-SEP-02, Drywell Spray Procedure, drywell sprays (2) allowed to be
re-initiated.
A. (1) 300° F
(2) are
B. (1) 300° F
(2) are NOT
C. (1) 340° F
(2) are
D. (1) 340° F
(2) are NOT
K/A:
295028 High Drywell Temperature
EK2 Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following:
(CFR: 41.7 / 45.8)
01 Drywell spray: Mark-I&II
RO/SRO Rating: 3.7/4.1
Tier 1 / Group 1
K/A match: The proposed question tests the applicants knowledge of BNPs emergency operating
procedures (EOPs) associated with high drywell temperature (i.e., PCCP DW/T and SEP-02),
including the drywell spray requirement.
Pedigree: Hatch 2007
Objective: LOI-CLS-LP-300-K, Obj. 4
Q Reference: 0EOP-02-PCCP, DW/T; 0EOP-01-SEP-02
Ref Provided: 0EOP-01-UG, Attachment 5, Drywell Spray Initiation Limit
Cog Level: High
Explanation: PCCP, Step DWT-7 and DWT-9 require initiating DW Spray BEFORE drywell temperature
Page: 26 of 193 5/4/2017
exceeds 300°F; SEP-02, Step 9.f requires re-verification of DWSIL and torus level less than
21 inches before re-initiating drywell spray. Since current torus level is less than 21 inches,
drywell spray is allowed to be re-initiated.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Plausible because the current drywell average air temperature and drywell pressure are in
the SAFE region of the DWSIL Curve and because torus level is less than 21 inches.
Choice C: Plausible because PCCP Step DWT-11 requires emergency depressurization at 340°F, and
because the current drywell average air temperature and drywell pressure are in the SAFE
region of the DWSIL Curve. Incorrect because the fill-in-the-blank temperature asks for the
lowest required temperature.
Choice D: Plausible because PCCP Step DWT-11 requires emergency depressurization at 340°F.
Incorrect because the fill-in-the-blank temperature asks for the lowest required temperature.
SRO Basis: N/A
Page: 27 of 193 5/4/2017
Page: 28 of 193 5/4/2017
15. Following a LOCA on Unit Two, plant conditions are as follows:
Reactor water level 50 inches and rising
Reactor pressure 140 psig
Torus temperature 220°F
Torus pressure 10.5 psig
Torus level -43 inches
2A CS pump flow 5000 gpm
2B RHR pump flow 7000 gpm
[Reference Provided]
Which one of the following completes both statements below?
2A Core Spray pump (1) operating within NPSH limits.
2B RHR pump (1) operating within NPSH limits.
A. (1) is
(2) is
B. (1) is
(2) is NOT
C. (1) is NOT
(2) is
D. (1) is NOT
(2) is NOT
K/A:
295030 Low Suppression Pool Water Level
EA1 Ability to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL
WATER LEVEL: (CFR: 41.7 / 45.6)
01 ECCS systems (NPSH considerations)
RO/SRO Rating: 3.6/3.8
Tier 1 / Group 1
K/A match: Must know when and why to stop or leave running ECCS systems when threatened with a
loss of NPSH due to low torus water level.
Pedigree: Modified from 2012 NRC Exam
Objective: LOI-CLS-LP-300-B, Obj. 12
Q Reference: 0EOP-01-UG rev 67
Ref Provided: OEOP-01-UG, Attachments 8 & 9
Cog Level: High
Explanation: The key to this question is the Note at the bottom of each NPSH limit graph:
Torus level at -43 is exactly 1 foot below -31, so the candidate needs to subtract 0.5 psig
from the given torus pressure of 10.5 psig, and get 10 psig. Therefore, the curve to use on
Page: 29 of 193 5/4/2017
each graph is the 10 psig curve.
Plotting the 2A CS Pump at 5000 gpm and 220F gives a point above the 10 psig curve,
which is unsatisfactory. So the 2A CS Pump is NOT operating within its NPSH limit.
Plotting the 2B RHR Pump at 7000 gpm and 220F gives a point on the 10 psig curve, so it
hasnt exceeded the limit..
Distractor Analysis:
Choice A: Incorrect because - the RHR pump is operating within NPSH limits, so it is not true that
neither pump is operating within limits. Plausible if a candidate: 1) mis-plots one or both
pump operating points, or 2) doesnt apply the Note, and thinks that the operating points
need to be above an interpolated 10.5 psig curve.
Choice B: Incorrect because - the CS Pump is operating within limits. Plausible if a candidate does
not understand which side of the curve the operating point needs to be. (He would think
that the RHR pump was okay, but the CS pump was not.).
Choice C: Correct, See explanation.
Choice D: Incorrect because - the RHR pump is operating outside its NPSH limit. Plausible if 1) a
candidate does not apply the Note, and compares the RHR pump operating point with an
interpolated 10.5 psig curve, or 2) thinks that on or above the line is the safe region.
SRO Basis: N/A
Page: 30 of 193 5/4/2017
16. IAW 0EOP-01-UG, Users Guide, which one of the following identifies the name and
indicated reactor water level value associated with Low Level 4 (LL 4) for Unit One at a
reactor pressure of 0 psig?
A. Minimum Zero Injection RPV Level / -45 inches
B. Minimum Zero Injection RPV Level / -27.5 inches
C. Minimum Steam Cooling RPV Level / -45 inches
D. Minimum Steam Cooling RPV Level / -27.5 inches
K/A:
295031 Reactor Low Water Level
EA2 Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER
LEVEL: (CFR: 41.10/43.5/45.13)
04 Adequate core cooling
RO/SRO Rating: 4.6/4.8
Tier 1 / Group 1
K/A match: Answering this question requires applicant knowledge of the designation used and reactor
water level specific to low level 4 on Unit 1. Reactor water levels associated with low level 4
are a method of adequate core cooling.
Pedigree: New
Objective: LOI-CLS-LP-300-B Obj 7: Define all EOP terms per the EOP definitions list in EOP-01-UG.
Q Reference: 0EOP-01-UG, Rev 67
Ref provided: None
Cog Level: Fundamental
Explanation: In accordance with 0EOP-01-UG, Users Guide, low level 4 is named the Minimum Steam
Cooling RPV Level. When Unit 1 reactor pressure is < 60 psig, low level 4 is designated at
-27.5 inches. The distractors for this question relate to low level 5 (Minimum Zero Injection
RPV Level and -45 inches).
Distractor Analysis:
Choice A: Incorrect because low level 4 is named the Minimum Steam Cooling RPV Level. When
Unit 1 reactor pressure is < 60 psig, low level 4 is designated at -27.5 inches.
Choice B: Incorrect because low level 4 is named the Minimum Steam Cooling RPV Level.
Choice C: Incorrect because when Unit 1 reactor pressure is < 60 psig, low level 4 is designated at
-27.5 inches.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 31 of 193 5/4/2017
17. Unit Two has experienced an ATWS with the following conditions:
SLC has been initiated
Control Rods 22-23 and 10-19 are at position 48
10 control rods are at position 02
All other control rods are at position 00
The following sequence of events has occurred in the order listed:
Event 1 Control Rod 22-23 is at position 00
Event 2 APRM downscale indication lights are lit
Event 3 Control Rod 10-19 is at position 02
Event 4 SLC has injected into the RPV to a tank level of 0%
IAW the ATWS procedure, which one of the following identifies the earliest event
where actions to cooldown and depressurize the RPV may proceed?
A. Event 1
B. Event 2
C. Event 3
D. Event 4
Page: 32 of 193 5/4/2017
K/A:
295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown
EA2 Ability to determine and/or interpret the following as they apply to SCRAM CONDITION
PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:
(CFR: 41.10 / 43.5 / 45.13)
06 Reactor pressure
RO/SRO Rating: 4.0/4.1
Tier 1 / Group 1
K/A match: With ATWS conditions present and the reactor unable to be shutdown without boron under all
conditions, actions to cooldown and depressurize are not permitted until either SLC is
completely injected or the reactor is shutdown with no boron. Answering this K/A requires
knowledge of the overall mitigative strategy of the ATWS procedure concerning the start of
reactor depressurizaton.
Pedigree: New
Objective: LOI-CLS-LP-300-E, Obj. 16b
Q Reference: 2EOP-01-ATWS (Rev 1), 0OI-37.5 (Rev 15)
Ref Provided: None
Cog Level: High
Explanation: The earliest event where conditions are met to cooldown and depressurize occurs at event 4
when SLC has injected into the RPV to a tank level of 0%. At no time are the conditions
within Table Q-1 [shutdown without boron] met. This is in accordance with Step RC/P 7.
Distractor Analysis:
Choice A: Incorrect because - Table Q-1 conditions are not met: 10 rods are withdrawn to position
02 AND one rod is at position 48. Plausible: either 10 rods may be withdrawn to position 2
OR one rod at position 48 would be acceptable for shutdown without boron, the applicant
could incorrectly evaluate the conditions and requirements.
Choice B: Incorrect because- Table Q-1 conditions are not met: 10 rods are withdrawn to position
02 AND one rod is at position 48. Plausible: the applicant could incorrectly evaluate
reactor power as <2% as a possible path to proceed, RC/Q-6 uses the statement can
reactor power be determined to be below 2%.
Choice C: Incorrect because - Table Q-1 conditions are not met: 11 rods are withdrawn to position 02.
Plausible: either 10 rods may be withdrawn to position 2 OR one rod at position 48 would be
acceptable for shutdown without boron, the applicant could incorrectly evaluate the conditions
and requirements.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 33 of 193 5/4/2017
18. Unit Two has entered RRCP due to fuel failure which has resulted in a high offsite
release. The crew is currently realigning ventilation IAW 0OP-37, Control Building
Ventilation System Operating Procedure, as directed by RRCP.
Which one of the following completes both statements below?
In order to complete the alignment of the Control Building Mechanical Equipment
Room Vent Fans, the crew will manipulate the appropriate control switch(es)
on (1) .
This ventilation realignment is performed to maintain habitability of the (2) .
A. (1) Unit Two ONLY
(2) Control Room ONLY
B. (1) Unit Two ONLY
(2) Control Building Envelope
C. (1) BOTH Units
(2) Control Room ONLY
D. (1) BOTH Units
(2) Control Building Envelope
Page: 34 of 193 5/4/2017
K/A:
295038 High Off-Site Release Rate
EK3 Knowledge of the reasons for the following responses as they apply to HIGH OFF-SITE
RELEASE RATE: (CFR: 41.5 / 45.6)
03 Control room ventilation isolation: Plant-Specific
RO/SRO Rating: 4.0/4.1
Tier 1 / Group 1
K/A match: Given an in-progress High Offsite Release, the applicant is expected to demonstrate
knowledge of performance with reasoning for Control Room ventilation isolation as directed
by 0EOP-04-RRCP.
Pedigree: New
Objective: LOI-CLS-LP-300-N, Obj. 4b
Q Reference: 0EOP-04-RRCP, Radioactivity Release Control, Rev. 21
0OP-37, Control Building Ventilation System Operating Procedure, Rev. 62,
0OI-37.10, Radioactivity Release Control Procedure Basis Document, Rev. 12.
Ref Provided: None
Cog Level: Fundamental
Explanation: In accordance with the 0OP-37 NOTE concerning CB MER switch operation during a
manual startup of the CB Emergency Recirculation System, operation of the CB MER
switches on BOTH Units are required place these fans in the procedurally required lineup.
Additionally, as specified in 0OI-37.10, with a Fuel Failure in progress, these actions are
performed to maintain habitability of the Control Room ONLY. Both answer choices originate
from correct implementation of 0EOP-04-RRCP, specifically Step RRCP-2 (second
override).
Distractor Analysis:
Choice A: Incorrect because - in order to place the Control Building ventilation in Emergency
Recirculation Mode, both Units Control Building Mechanical Equipment Room Vent Fans
control switches must be simultaneously placed in OFF.
Choice B: Incorrect because - in order to place the Control Building ventilation in Emergency
Recirculation Mode, both Units Control Building Mechanical Equipment Room Vent Fans
control switches must be simultaneously placed in OFF. Control Building ventilation is
placed in Emergency Recirculation Mode to maintain habitability of the Control Room
ONLY.
Choice C: Correct, See explanation.
Choice D: Incorrect because - Control Building ventilation is placed in Emergency Recirculation Mode to
maintain habitability of the Control Room ONLY.
SRO Basis: N/A
Page: 35 of 193 5/4/2017
19. The following sequence of reports have been received on Unit One:
0800 A report of smoke from the Service Water Building.
0807 Fire Brigade members have commenced an attack on a minor electrical fire
on the Service Water Building pump 20 EL area.
0810 Fire Brigade members have transitioned to using a hose reel and are
applying a water fog to the fire.
0815 The fire is out with re-flash watch stationed.
Which one of the following completes both statements below?
IAW 0PFP-PBAA, Power Block Auxiliary Areas Pre-Fire Plans (SW, RW, AOG, TY,
EY, PDC, DGS, MCP), the appropriate fire extinguisher to use to attack the fire at 0807
would be a(n) (1) extinguisher.
IAW 0PFP-013, General Fire Plan, the earliest time that the Reactor Operator is
required to request "911" off-site assistance is at (2) .
A. (1) AFFF
(2) 0807
B. (1) AFFF
(2) 0810
C. (1) CO2
(2) 0807
D. (1) CO2
(2) 0810
Page: 36 of 193 5/4/2017
K/A:
600000 Plant Fire On Site
AK1 Knowledge of the operation applications of the following concepts as they apply to Plant Fire On
Site:
01 Fire Classifications by type
RO/SRO Rating: 2.5/2.8
Tier 1 / Group 1
K/A match: The applicant needs to display knowledge that AFFF is not an appropriate fire extinguishing
agent around electrical equipment and that outside assistance is required whenever a fire
hose is required to be used for fire suppression.
Pedigree: New
Objective: LOI-CLS-LP-041, Obj. 4
Q Reference: 0PBFP-PBAA, Rev 31, 0PFP-013, General Fire Plan. Rev. 48.
Ref Provided: None
Cog Level: Fundamental
Explanation: Per 0PBFP-PBAA, the 20 level, the appropriate fire extinguishers are CO2 or ABC type
only. AFFF is an AB extinguisher. Any use of a site hose to suppress a fire requires
immediate notification of off-site assistance.
Distractor Analysis:
Choice A: Incorrect because - The first part is incorrect because AFFF is not to be used near
electrical equipment. The second part is incorrect because use of a hose requires that
off-site assistance be requested.
Choice B: Incorrect because - The first part is incorrect because AFFF is not to be used near
electrical equipment. The second part is correct.
Choice C: Incorrect because - The first part is correct. The second part is incorrect because use of a
hose requires that off-site assistance be requested.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 37 of 193 5/4/2017
20. Initial Conditions:
Unit One is operating at 95% power
A grid alert has been issued due to high electric demand
Subsequently:
UA-06 (1-2), GEN UNDER FREQ RELAY, alarms
Grid frequency, as read on 1-GEN-FM-736, is 59.1 Hz
Which one of the following completes both statements below?
Generator megawatts (1) automatically increase toward the EHC pressure set
limit.
IAW 0AOP-22.0, Grid Instability, a reactor manual scram is required (2) .
A. (1) will
(2) immediately
B. (1) will
(2) within five minutes
C. (1) will NOT
(2) immediately
D. (1) will NOT
(2) within five minutes
Page: 38 of 193 5/4/2017
K/A:
700000 Generator Voltage and Electric Grid Disturbances
AK2 Knowledge of the interrelations between GENERATOR VOLTAGE AND ELECTRIC GRID
DISTURBANCES and the following: (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)
03 Sensors, detectors, indicators
RO/SRO Rating: 3.0/3.1
Tier 1 / Group 1
K/A match: This question tests the applicants knowledge of the function of the 81 relay and determines
their knowledge of manual scram criteria for a grid disturbance
Pedigree: New
Objective: LOI-CLS-LP-302-G2, Obj. 3
Q Reference: UA-06 (1-2) Rev 44, 0AOP-22.0 Rev 27.
Ref Provided: None
Cog Level: High
Explanation: At 59.8 Hz, Generator Frequency Relay 81 will actuate causing the alarm UA-06 (1-2) and
picking up load until the limits of the pressure set. At 59.1 Hz, operation is allowed for up to 5
minutes.
Distractor Analysis:
Choice A: Incorrect because - The first part is correct. The second part is wrong because operation
at this frequency is allowed for up to 5 minutes. It is plausible because the manual scram
criteria is at 58.1 Hz.
Choice B: Correct, See explanation.
Choice C: Incorrect because- The first part is incorrect because at 59.8 Hz, Generator Frequency Relay
81 will actuate causing the alarm UA-06 (1-2) and picking up load until the limits of the
pressure set. It is plausible if the applicant assumes the alarm comes in before the auto action.
The second part is wrong because operation at this frequency is allowed for up to 5 minutes. It
is plausible because the manual scram criteria is at 58.1 Hz.
Choice D: Incorrect because - The first part is incorrect because at 59.8 Hz, Generator Frequency Relay
81 will actuate causing the alarm UA-06 (1-2) and picking up load until the limits of the
pressure set. It is plausible if the applicant assumes the alarm comes in before the auto action.
The second part is correct.
SRO Basis: N/A
Page: 39 of 193 5/4/2017
21. Which one of the following identifies the value of TAF with Unit One reactor pressure
at 30 psig?
A. -7.5 inches
B. -10 inches
C. -17.5 inches
D. -20 inches
K/A:
295009 Low Reactor Water Level
AA2 Ability to determine and/or interpret Reactor water level as they apply to LOW REACTOR
WATER LEVEL: (CFR: 41.10/43.5/45.13)
01 Reactor water level
RO/SRO Rating: 4.2/4.2
Tier 1 / Group 2
K/A match: This question elicits applicant knowledge of the relationship between layout of the TAF graph
(procedurally used during LOW REACTOR WATER LEVEL conditions) and reactor water level
instrumentation.
Pedigree: New
Objective: LOI-CLS-LP-300-D Obj 6: Given EOP Users Guide Caution 1, Reference Leg Area Drywell
Temperature, RPV pressure,and the indicated reactor water level, determine if reactor water
leve is above TAF, Low Level 4 (LL4), Low Level 5 (LL5), or cannot be determined (LOCT).
Q Reference: 0EOP-01-NL rev 27
Ref provided: None
Cog Level: Fundamental
Explanation: In accordance with 0EOP-01-NL, EOP/SAMG Numerical Limits and Values, Attachment 1,
U1 RWL at TAF, with reactor pressure decrease to 0 psig, a standard value of -7.5 inches is
used for TAF.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect because - For reactor pressure at 60 psig and drywell temperature above 200 F
this would be correct.
Choice C: Incorrect because- For reactor pressure at 0psig and drywell temperature <200F this could be
interpolated.
Choice D: Incorrect because - For reactor pressure at 60 psig and drywell temperature <200F this would
be correct.
SRO Basis: N/A
Page: 40 of 193 5/4/2017
22. Which one of the following completes both statements below?
Entry into RRCP is first required when the offsite gaseous release rate is above
the (1) EAL as specified in RRCP Table M-4.
IAW AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment, the Sea
Breeze Effect can (2) the projected dose by a factor of up to 2.5.
A. (1) Alert
(2) increase
B. (1) Alert
(2) decrease
C. (1) Site Area Emergency
(2) increase
D. (1) Site Area Emergency
(2) decrease
Page: 41 of 193 5/4/2017
K/A:
295017 High Off-Site Release Rate
AK1 Knowledge of the operational implications of the following concepts as they apply to HIGH
OFF-SITE RELEASE RATE : (CFR: 41.8 to 41.10)
03 Meteorological effects on off-site release
RO/SRO Rating: 2.7/3.4
Tier 1 / Group 2
K/A match: The question meets the K/A by testing the knowledge of the effect of the Sea Breeze
(meteorlogical effect) on the dose assessment.
Pedigree: New
Objective: LOI-CLS-LP-301-A, Obj. 6 - Determine data required for offsite dose projection in accordance
with AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment, and PEP-03.6.1,
Release Estimates Based Upon Stack/Vent Readings.
Q Reference: AOP-05.4, AD-EP-ALL-0202
Ref Provided: None
Cog Level: Fundamental
Explanation: IAW AOP-05.4 if the release exceeds SAE then RRCP is to be entered. IAW RRCP the
crew is to monitor site boundary dose IAW AD-EP-ALL-0202. this procedure defines Sea
Breeze effect. "Coastal wind circulation pattern that develops during the daytime, when the
land temperature is hotter than the ocean, causing surface winds to blow inland from the
sea. The consequence of a release traveling inland and then returning back to sea can
increase the projected dose by a factor of up to 2.5. The Sea Breeze Effect can exist with
wind directions from 16° - 269°.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: The first part is correct. The second part could be thought to be true if the sea breze effect
is thought to only provide movement of the concentration away from the land, thereby
reducing the effect.
Choice C: The first part question is incorrect, actual entry conditions for RRCP are offsite gaseous
release rates at the ALERT (RA1) level. Entry conditions at the SAE level is plausible, first of
all, because entry at those levels is actually required (just at the lower level). The RRCP
procedure contains specific steps that are tied to the SAE EAL (RRCP-7 and RRCP-8).
Furthermore, Protective Action Recommendations (PARs) are only required at the highest
level (General Emergency) EAL, which makes it plausible that an applicant may believe that
the procedure would only need to be entered at one level below the level required by PARs.
The second part is correct.
Choice D: The first part question is incorrect, actual entry conditions for RRCP are offsite gaseous
release rates at the ALERT (RA1) level. Entry conditions at the SAE level is plausible, first of
all, because entry at those levels is actually required (just at the lower level). The RRCP
procedure contains specific steps that are tied to the SAE EAL (RRCP-7 and RRCP-8).
Furthermore, Protective Action Recommendations (PARs) are only required at the highest
level (General Emergency) EAL, which makes it plausible that an applicant may believe that
the procedure would only need to be entered at one level below the level required by PARs.
The second part could be thought to be true if the sea breze effect is thought to only provide
movement of the concentration away from the land, thereby reducing the effect.
SRO Basis: N/A
Page: 42 of 193 5/4/2017
23. Unit One is operating at 100% power when the 1A RPS MG set trips.
Which one of the following completes both statements below?
The 1-CAC-AQH-1264-3, Reactor Bldg Vent Noble Gas Monitor, effluent flow rate
will (1) .
Entry into SCCP based on reactor building negative pressure, (2) required.
A. (1) lower
(2) is
B. (1) lower
(2) is NOT
C. (1) rise
(2) is
D. (1) rise
(2) is NOT
Page: 43 of 193 5/4/2017
K/A:
295020 Inadvertent Containment Isolation
AK3 Knowledge of the reasons for the following responses as they apply to INADVERTENT
CONTAINMENT ISOLATION: (CFR: 41.5/45.6)
02 Drywell/containment pressure response
RO/SRO Rating: 3.3/3.5
Tier 1 / Group 2
K/A match: Justification for K/A match: The inadvertent containment isolation portion of the K/A is being
met based on the RPS MG set trip; even though an RPS MG set trip has no effect on drywell
pressure, it does have an effect on the containment (secondary) pressure response. The
applicants knowledge of the REASON for the secondary containment response is being
indirectly tested based on their knowledge that the RPS MG set trip isolates Rx Bldg HVAC
and causes the SBGT to auto-start, which maintains negative pressure. The Tier 1 aspect is
an abnormal evolution knowledge of SCCP entry condition for secondary containment negative
pressure.
Pedigree: New
Objective: LOI-CLS-LP-300-M Obj 3: Given plant conditions, determine if the Secondary Containment
Control procedure Should be entered (LOCT).
Q Reference: SD-3, OP-03, SD-37.1, UA-05 (6-7), 0OI-37.9
Ref provided: None
Cog Level: Fundamental
Explanation: The RPS MG Set causes a Rx Bldg Ventilation Isolation, which will stop the flow to
CAC-1264; the UA-05, 6-7 (Rx Bldg Static Press Diff Low) annunciator may initially alarm;
however, the SBGT auto-start will maintain the Rx Bldg Diff Pressure negative; entry to
SCCP is not required based on the annunciator procedure guidance that says entry only
required if pressure cannot be maintained negative.
Distractor Analysis:
Choice A: Incorrect because entry to SCCP is not required since building pressure will be maintained
negative by the SBGT; plausible because the Rx Bldg Ventilation fans will trip following
the RPS Bus loss.
Choice B: Correct, See explanation.
Choice C: Incorrect because flow to CAC-1264 will lower when the Rx Bldg Ventilation trips, and because
SCCP entry is not required; plausible because the applicant may misconstrue where the SBGT
flow is directed - not directed to the Rx Bldg Ventilation, directed to the Main Stack, and
because the UA-05, 6-7 annunciator may initially alarm when the HVAC isolates.
Choice D: Incorrect because flow to CAC-1264 will lower when the Rx Bldg Ventilation trips, plausible
because the applicant may misconstrue where the SBGT flow is directed - not directed to the
Rx Bldg Ventilation, directed to the Main Stack.
SRO Basis: N/A
Page: 44 of 193 5/4/2017
24. Which one of the following completes the statement below?
IAW PCCP, with torus water level rising, a reactor scram is required before level first
reaches the:
A. SRV Tailpipe
B. Torus Downcomer Vent opening
C. Suppression Chamber to Drywell Vacuum Breakers
D. Reactor Building to Suppression Chamber Vacuum Breakers
K/A:
295029 High Suppression Pool Water Level
EK3 Knowledge of the reasons for the following responses as they apply to HIGH SUPPRESSION
POOL WATER LEVEL : (CFR: 41.5 / 45.6)
03 Reactor SCRAM
RO/SRO Rating: 3.4/3.5
Tier 1 / Group 2
K/A match: Must have knowledge of why the reactor must be scrammed as Suppression Pool level rises.
Pedigree: New
Objective: LOI-CLS-LP-300-M Obj 8a: Given plant conditions and the Secondary Containment Control
Procedure, determine if any of the following are required: Manual reactor scram (LOCT)
Q Reference: 0EOP-02-PCCP, Rev. 12; 0EOP-01-UG, Rev. 67
Ref provided: None
Cog Level: High
Explanation: In accordance with 0EOP-01-UG, which defines SRV Tailpipe Level Limit as the highest
torus level at which opening an SRV will not result in exceeding the capability of the tail pipe
and quencher supports (Attachment 11). Attachment 11 shows the limit being +6 for all
RPV pressures; i.e., its always +6.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: INCORRECT because the Torus downcomer opening (which is usually below normal
Suppression Pool water level) is affected by a Torus low level condition at - 5.5 feet.
(Reference PCCP Step T/L-17).
Choice C: INCORRECT because the SC to DW vacuum breakers would be covered at +21, not the +6
that requires a scram. And at this level the PCCP T/L direction is to terminate DW sprays, not
scram. Plausible because it is the next level value called out in the T/L leg.
Choice D: INCORRECT because the RB to SC vacuum breakers are at elevation -6. Plausible because
it is a penetration a candidate might recall, and there could be a +/-6 misconception.
SRO Basis: N/A
Page: 45 of 193 5/4/2017
25. Unit Two is operating at 100% power when UA-3 (6-2), RX BLDG VENT TEMP HIGH
alarms.
Which one of the following completes both statements below?
The reactor building ventilation system (1) tripped and isolated.
This alarm indicates that the (2) are approaching their environmental qualification
temperatures.
A. (1) has
(2) RB Ventilation radiation monitors
B. (1) has
(2) RB Isolation Damper solenoid valves
C. (1) has NOT
(2) RB Ventilation radiation monitors
D. (1) has NOT
(2) RB Isolation Damper solenoid valves
Page: 46 of 193 5/4/2017
K/A:
295032 High Secondary Containment Area Temperature
G2.4.11 Knowledge of abnormal condition procedures. (CFR: 41.10 / 43.5 / 45.13)
RO/SRO Rating: 4.0/4.2
Tier 1 / Group 2
K/A match: The question tests the knowledge of automatic functions which are verified in accordance
with the alarm response procedure and the knowledge of the component that is protected by
this isolation.
Pedigree: New
Objective: LOI-CLS-LP-037.1, Obj. 4
Q Reference: 2 APP UA-03 6-2 rev. 58, SD-11 Rev 10
Ref Provided: None
Cog Level: Fundamental
Explanation: Reactor building vent temperature high will trip and isolate the reactor building ventilation
system and start the standby gas treatment trains. The purpose of the alarm and isolation of
RB Vent is that the rad monitors are EQ only up to this temperature (135°F).
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect because - the first part is correct. The second part is plausible because solenoid
valves are the limiting EQ in the Drywell.
Choice C: Incorrect because - Reactor building vent temperature high will trip and isolate the reactor
building ventilation system and start the standby gas treatment trains.
Plausible because this actuation will occur on any one of the following:
Low Reactor Water Level #2, High Drywell Pressure, Main Stack High Radiation, Reactor
Building Ventilation Exhaust High Temperature, and Reactor Building Ventilation Exhaust High
Radiation. The applicant may fail to recognize this particular condition as causing the
actuation. The second part is correct.
Choice D: Incorrect because - Reactor building vent temperature high will trip and isolate the reactor
building ventilation system and start the standby gas treatment trains.
Plausible because this actuation will occur on any one of the following:
Low Reactor Water Level #2, High Drywell Pressure, Main Stack High Radiation, Reactor
Building Ventilation Exhaust High Temperature, and Reactor Building Ventilation Exhaust High
Radiation. The applicant may fail to recognize this particular condition as causing the
actuation. The second part is plausible because solenoid valves are the limiting EQ in the
Drywell.
SRO Basis: N/A
Page: 47 of 193 5/4/2017
26. Unit Two is operating at 100% power with the following indications noted:
Annunciator UA-03 (2-3), RX BLDG
ROOF VENT RAD HIGH is in alarm.
Annunciator UA-03 (2-7), AREA RAD RX
BLDG HIGH is in alarm.
The 2-CAC-AQH-1264-3, Reactor Bldg
Vent Noble Gas Monitor is indicating 50
mr/hr and slowly rising.
Area Radiation Monitor Panel (P-600),
Channel 30, is indicating as shown.
Which one of the following completes both
statements below?
The Reactor Building Ventilation System
(1) currently isolated.
ARM Channel 2-30 is reading (2) .
A. (1) is
(2) 15 MR/HR
B. (1) is
(2) 50 MR/HR
C. (1) is NOT
(2) 15 MR/HR
D. (1) is NOT
(2) 50 MR/HR
Page: 48 of 193 5/4/2017
K/A:
295034 Secondary Containment Ventilation High Radiation
EA1 Ability to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT
VENTILATION HIGH RADIATION: (CFR: 41.7 / 45.6)
01 Area radiation monitoring system
RO/SRO Rating: 3.8/3.8
Tier 1 / Group 2
K/A match: The question tests the knowledge of automatic functions which are verified in accordance
with the alarm response procedure and the knowledge of the component that is protected by
this isolation.
Pedigree: New
Objective: LOI-CLS-LP-037.1, Obj. 4
Q Reference: 1APP-UA-03 2-3 and 2-7, Rev 61, 0AOP-5.4, Rev 1
Ref Provided: None
Cog Level: Fundamental
Explanation: Several Secondary Containment Ventilation high radiation alarms would cause the
ventilation system to trip and isolate. This one does not. The high noble gas alarm could
conceivably come from the SFP Cooling system. The reading for this monitor is 50 mr/hr.
Distractor Analysis:
Choice A: Incorrect because - Reactor building vent exhaust radiation will isolate not the roof vent.
The second part is plausible because of the log scale the previous division are by 1
increment, if this is applied to the current log scale then this would be correct.
Choice B: Incorrect because - Reactor building vent exhaust radiation will isolate not the roof vent.
The second part is correct.
Choice C: Incorrect because - the first part is correct. The second part is plausible because of the log
scale the previous division are by 1 increment, if this is applied to the current log scale then
this would be correct.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 49 of 193 5/4/2017
27. Initial Conditions:
Unit One has had a complete loss of RBCCW
Following a manual reactor scram, the Scram Discharge Volume (SDV) ruptured
Current conditions:
Drywell pressure 2.3 psig
Drywell average temperature 190°F
Rx Bldg 20 ft south temperature 195°F
1-UA-12 (1-3), SOUTH CS RM FLOOD LEVEL HI-HI, is in alarm
1-UA-12 (1-4), SOUTH RHR RM FLOOD LEVEL HI-HI, is in alarm
Which one of the following identifies the operator action required by SCCP?
A. Reset RPS to isolate the primary system discharge.
B. Commence a reactor cooldown not to exceed 100°F/hr
C. Rapidly depressurize to the main condenser irrespective of cooldown rate
D. Perform an emergency depressurization
Page: 50 of 193 5/4/2017
K/A:
295036 Secondary Containment High Sump/Area Water Level
EK2 Knowledge of the interrelations between SECONDARY CONTAINMENT HIGH SUMP/AREA
WATER LEVEL and the following: (CFR: 41.7 / 45.8)
01 Secondary containment equipment and floor drain system
RO/SRO Rating: 3.1/3.2
Tier 1 / Group 2
K/A match: Candidate must have knowledge of emergency procedure requirements for a given a high
water level condition in Secondary Containment.
Pedigree: BNP 2014 Q59
Objective: LOI-CLS-LP-300-M Obj 6d: Given plant conditions and the Secondary Containment Control
Procedure, determine if any of the following have been exceeded: Maximum safe operating
water level (LOCT)
Q Reference: SCCP, Rev. 10
Ref provided: None
Cog Level: Fundamental
Explanation: The SDV rupture (an RCS leak) combined with the inability of the sump pumps to keep up,
has led to two areas above the max safe limits (as indicated by the flood level Hi-Hi alarms).
This requires emergency depressurization per SCPP block SCCP-10.
Distractor Analysis:
Choice A: Incorrect because RPS cannot be reset due to the high DW pressure, so RCS leakage
from the SDV cannot be stopped. Plausible because the SCCP flowpath attempts to stop
RCS leakage.
Choice B: Incorrect because with two areas at max safe level, the proper response is to emergency
depressurize. Plausible because its the correct action with just one area at max safe
level.
Choice C: Incorrect because its no longer a correct action with two areas at max safe level. Plausible
because this would be an option before the second area reach its max safe level (the
Anticipatory Depressurization step, SCCP-8), or if the candidate doesnt think of the SDV as
being a primary system.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 51 of 193 5/4/2017
28. Unit One is in MODE 4 with the following conditions:
Loop A RHR is in Standby, aligned for LPCI Injection
Loop B RHR is in Shutdown Cooling with the 1B RHR Pump in service
Subsequently, 1A MG set trips.
Which one of the following completes both statements below?
The Shutdown Cooling valve that has closed is the (1) .
Subsequently, if level lowers to LL2, the E11-F015A, Inboard Injection Valve, (2)
require operator action to open the valve.
A. (1) 2E11-F008, RHR S/D Cooling Suction Isolation Valve - Outboard
(2) will NOT
B. (1) 2E11-F009, RHR Shutdown Cooling Suction Isolation Valve - Inboard
(2) will NOT
C. (1) 2E11-F008, RHR S/D Cooling Suction Isolation Valve - Outboard
(2) will
D. (1) 2E11-F009, RHR Shutdown Cooling Suction Isolation Valve - Inboard
(2) will
Page: 52 of 193 5/4/2017
K/A:
203000 RHR/LPCI: Injection Mode (Plant Specific)
A2 Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT
SPECIFIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
04 A.C. failures
RO/SRO Rating: 3.5/3.6
Tier 2 / Group 1
K/A match: Test item requires applicants to predict the impact of the RPS MG Set 1A loss on the
RHR/LPCI Injection Loop A, and also requires the applicant to use RO knowledge to
determine if the F015A will require manual action to open on a low level condition.
Pedigree: New
Objective: LOI-CLS-LP-017 Obj 9: Given an RHR pump or valve, list the interlocks, permissives and/or
automatic actions associated with the RHR pump or valve, including setpoints. (LOCT).
Q Reference: 1OP-03, Attachment 2, Tech Spec 3.5.2
Ref provided: None
Cog Level: High
Explanation: Loss of RPS 1A causes a closure signal to F015A when F008 & F009 are open. Since
Shutdown Cooling was in service on Loop 2, the F015A receives a close signal , but it will
re-open automatically if either the F008 or F009 Close, and the F009 does close on an RPS
A loss.
Distractor Analysis:
Choice A: Part 1 is incorrect, RPS B will close the F008 valve. The second part is correct.
Choice B: Correct, See explanation.
Choice C: Part 1 is incorrect, RPS B will close the F008 valve. The second part is plausible because the
valve does get a close signal on the loss of RPS.
Choice D: Part 1 is correct. The second part is plausible because the valve does get a close signal on the
loss of RPS.
SRO Basis: N/A
Page: 53 of 193 5/4/2017
29. Which one of the following identifies the 4160V power supply to RHR Pump 1C?
A. E1
B. E2
C. E3
D. E4
K/A:
205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)
K2 Knowledge of electrical power supplies to the following: (CFR: 41.7)
01 Pump motors
RO/SRO Rating: 3.1/3.1
Tier 2 / Group 1
K/A match: To match the Tier 2 aspect of this K/A concerning RHR Shutdown Cooling Mode power
supplies to pump motors, the applicant is required to select the appropriate power supply to
the 1C RHR pump.
Pedigree: Modified from 2016 NRC Exam
Objective: LOI-CLS-LP-017, OBJ 17.a.
Q Reference: 1OP-17 Rev 129
Ref provided: None
Cog Level: Fundamental
Explanation: RHR pump 1C is powered from 4160V Emergency Bus 1E. As seen on 1OP-17/rev. 129,
Attachment 1A on page 5 of 5.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect because - this is not the power supply to RHR pump 1C, plausible because it is
the power supply to RHR pump 1D.
Choice C: Incorrect because- this is not the power supply to RHR pump 1C, plausible because it is the
power supply to RHR pump 1A.
Choice D: Incorrect because - this is not the power supply to RHR pump 1C, plausible because it is the
power supply to RHR pump 1B.
SRO Basis: N/A
2016 NRC Exam Question
Page: 54 of 193 5/4/2017
Page: 55 of 193 5/4/2017
30. Unit Two is in the EOPs with the following conditions:
HPCI is operating post-scram in pressure control mode with the HPCI Flow
Controller in Automatic
RPV level, temperature, and pressure are stable at expected values
Subsequently, E41-F012, Minimum Flow Bypass To Suppression Pool Valve, fails
full-open
Which one of the following identifies RPV pressure response after new steady-state
conditions are reached, and why?
A. Pressure will be lower, because the flow controller will sense lower flow and speed
up the HPCI turbine.
B. No change in pressure, because the flow controller is not affected by E41-F012
inadvertently opening.
C. No change in pressure, because the HPCI turbine speed controller will maintain a
constant turbine RPM.
D. Pressure will be higher, because the flow controller will sense higher flow and slow
down the HPCI turbine.
Page: 56 of 193 5/4/2017
K/A:
206000 High Pressure Coolant Injection System
K3 Knowledge of the effect that a loss or malfunction of the HIGH PRESSURE COOLANT
INJECTION SYSTEM will have on following: (CFR: 41.7 / 45.4)
02 Reactor pressure control: BWR-2,3,4
RO/SRO Rating: 3.8/3.8
Tier 2 / Group 1
K/A match: Candidate must have sufficient system knowledge to determine how a malfunction of HPCI
will affect reactor pressure control.
Pedigree: New
Objective: LOI-CLS-LP-019 Obj 22b: Given plant conditions, predict how a loss or malfunction of the
HPCI System will affect the following: Reactor pressure (LOCT)
Q Reference: SD-19, High Pressure Coolant Injection System, Rev. 24
Ref provided: None
Cog Level: High
Explanation: F012 opening robs from flow transmitter (in CST discharge path). Flow controller sees
sensed flow less than automatic setpoint, which develops an error signal, so V9 is directed
to open more, admitting more steam to HPCI turbine. RPV pressure decreases until error
signal is reduced to zero (which happens when additional steam flow compensates for F012
flow). New steady-state condition reached with RPV pressure less than where it started.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect because the flow transmitter is downstream of the F012 branch line, so the flow
signal would be affected, as discussed above for the correct answer. Plausible if a
candidate doesnt know where flow is sensed, and assumes it captures the extra flow
through F012. As it would, for instance, if the flow transmitter were at the pump discharge,
as is common.
Choice C: Incorrect because the flow controller will in fact change speed of the turbine. Plausible
because this is what would happen if the controller were in Manual.
Choice D: Incorrect because the flow controller will sense lower flow and speed up the turbine. Plausible
if the candidate thinks the controller works opposite from what it does.
SRO Basis: N/A
Page: 57 of 193 5/4/2017
31. Unit One is in MODE 5.
The spent fuel storage pool gates are removed.
RHR Loop B is INOPERABLE
IAW Tech Spec 3.5.2, ECCS-Shutdown, which one of the following completes both
statements below?
(1) Core Spray subsystem(s) is(are) required to be OPERABLE.
A water level greater than or equal to a minimum of (2) over the top of the reactor
pressure vessel flange is required to exit the mode of APPLICABILITY for LCO 3.5.2.
A. (1) one
(2) 19 feet 11 inches
B. (1) one
(2) 21 feet 10 inches
C. (1) two
(2) 19 feet 11 inches
D. (1) two
(2) 21 feet 10 inches
Page: 58 of 193 5/4/2017
K/A:
209001 Low Pressure Core Spray System
G2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.2 / 45.2)
RO/SRO Rating: 4.0/4.7
Tier 2 / Group 1
K/A match: This question meets the K/A because it tests knowledge of LCO 3.5.2 above the line
information.
Pedigree: New
Objective: LOI-CLS-LP-018, Obj. 17.
Q Reference: TS 3.5.2, 3.77 Amendment No. 203
Ref provided: None
Cog Level: Fundamental
Explanation: RHR consists of two subsystems. Low Pressure Core Spray consists of two subsystems.
LCO 3.5.2 requires two of the four subsystems to be operable when within the mode of
applicability. Since one RHR loop (LPCI subsystem) is operable, only one low pressure core
spray subsystem is required to be operable meet the LCO 3.5.2 requirement to have two
subsystems operable. The second part is correct because it matches the LCO 3.5.2
applicability statement for MODE 5.
Distractor Analysis:
Choice A: Incorrect because - The first part is correct. See explanation above. The second part is
incorrect, but plausible, because it is the water level over the top of irradiated fuel
assemblies listed in LCO 3.7.7.
Choice B: Correct, See explanation.
Choice C: Incorrect because-The first part is incorrect because, with one RHR subsystem being
operable, only one core spray subsystem would be required to be operable to meet the LCO
3.5.2 requirement to have two subsystems operable. The second part is incorrect. See
explanation above.
Choice D: Incorrect because-The first part is incorrect because, with one RHR subsystem being
operable, only one core spray subsystem would be required to be operable to meet the LCO
3.5.2 requirement to have two subsystems operable. The second part is incorrect, but
plausible, because it is the water level over the top of irradiated fuel assemblies listed in LCO
SRO Basis: N/A
Page: 59 of 193 5/4/2017
32. Unit One is in MODE 3 when a steam line rupture occurs in the drywell.
One minute following the rupture, the following conditions exist:
Drywell pressure 10 psig
Reactor pressure 360 psig
Reactor water level 55 inches
Which one of the following completes both statements below?
The E21-F004A, Outboard Injection Vlv, and E21-F005A, Inboard Injection Vlv,
are (1) .
The shutoff head of the Core Spray pumps is approximately (2) .
A. (1) closed
(2) 200 psig
B. (1) closed
(2) 300 psig
C. (1) open
(2) 200 psig
D. (1) open
(2) 300 psig
Page: 60 of 193 5/4/2017
K/A:
209001 Low Pressure Core Spray System
A1 Ability to predict and/or monitor changes in parameters associated with operating the LOW
PRESSURE CORE SPRAY SYSTEM controls including: (CFR: 41.5 / 45.5)
04 Reactor pressure
RO/SRO Rating: 3.7/3.7
Tier 2 / Group 1
K/A match: The applicant is expected to monitor changes in reactor pressure associated with operating
the Core Spray System injection valves, 1E21-F004A/B and 1E21-F005A/B.
Pedigree: New
Objective: LOI-CLS-LP-018 Obj 7: Given plant conditions, determine if the Core Spray System should
automatically initiate. (LOCT). Obj. 11 Given a Core Spray System valve, list any interlocks/
automatic actions associated with the valve
Q Reference: SD-17, Residual Heat Removal System (RHR), Rev. 19, SD-18, Core Spray System, Rev. 6
Ref provided: None
Cog Level: High
Explanation: When the reactor pressure drops below 410 psig, Core Spray system injection valves,
1E21-F004A/B and 1E21-F005A/B, open. With Reactor pressure at 360 psig, the Core
Spray pumps do not inject into the reactor vessel since the pump shutoff head is 290 psig.
202 psig is the shutoff head for RHR pumps.
Distractor Analysis:
Choice A: Incorrect because - 202 psig is the shutoff head for RHR pumps. When the reactor
pressure drops below 410 psig, Core Spray system injection valves, 1E21-F004A/B and
1E21-F005A/B, open. With Reactor pressure at 360 psig, the Core Spray pumps do not
inject into the reactor vessel since the pump shutoff head is 290 psig.
Choice B: Incorrect because - When the reactor pressure drops below 410 psig, Core Spray system
injection valves, 1E21-F004A/B and 1E21-F005A/B, open.
Choice C: Incorrect because - 202 psig is the shutoff head for RHR pumps. With reactor pressure at 360
psig, the Core Spray pumps do not inject into the reactor vessel since the pump shutoff head
is 290 psig.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 61 of 193 5/4/2017
33. Unit Two was operating at 100% power with MCC 2XG tagged out.
The following sequence of events occur:
Time Event
1302 An ATWS occurs on Unit Two
2-C41-CS-S1, SLC PUMPS A & B, is placed in PUMP B RUN position
1303 SLC PUMP 2A red light OUT
SLC PUMP 2B red light LIT
SLC A SQUIB VALVE CONTINUITY yellow light OUT
SLC B SQUIB VALVE CONTINUITY yellow light LIT
1305 2-C41-CS-S1, SLC PUMPS A & B, is placed in PUMP A&B RUN position.
Which one of the following completes both statements below?
At 1304, the SLC system (1) injecting into the reactor vessel.
At 1306, the SLC system (2) .
A. (1) is
(2) flow rate into the reactor vessel has increased
B. (1) is NOT
(2) has begun to inject into the reactor vessel
C. (1) is
(2) flow rate into the reactor vessel has NOT increased
D. (1) is NOT
(2) flow rate into the reactor vessel has NOT increased
K/A:
211000 Standby Liquid Control System
K3 Knowledge of the effect that a loss or malfunction of the STANDBY LIQUID CONTROL SYSTEM
will have on following: (CFR: 41.7 / 45.4)
01 Ability to shutdown the reactor in certain conditions
RO/SRO Rating: 4.3/4.4
Tier 2 / Group 1
K/A match: Given an operationally valid set of plant conditions, the applicant must first determine whether
or not the Standby Liquid Control System (SLC) is injecting into the Reactor Vessel or not. The
ability of SLC to shutdown the Reactor in certain conditions is directly related to the ability of
the SLC to inject into the Reactor. Therefore, the first part of the question matches the K/A.
Furthermore, with the second part of the question the applicant must additionally determine,
based on the same conditions, whether an operator action of re-positioning the SLC control
switch will affect the SLC flow rate. Again, SLC flow rate (or lack thereof) is directly
proportional to the ability of the SLC to shutdown the Reactor with various time constraints. For
this specific question, the successful applicant will determine that the SLC is not injecting
under either circumstance, and therefore the SLC system is ultimately not able to shutdown
the Reactor under the given plant conditions. The K/A is therefore met under both question
parts.
Page: 62 of 193 5/4/2017
Pedigree: Modified from 2007 NRC Exam
Objective: LOI-CLS-LP-005, Obj. 6
Q Reference: SD-17, Residual Heat Removal System (RHR), Rev. 19, SD-18, Core Spray System, Rev. 6
Ref provided: None
Cog Level: High
Explanation: Based on the conditions provided at time=1300, MCC 2XG is de-energized and tagged out.
The applicant must determine that the 2XG outage impacts the A train of SLC; specifically,
the A squib valve has no power and will not fire due to MCC 2XG being de-energized. At
time=1304, the applicant can determine that the B squib valve did NOT fire correctly
because the yellow continuity light is LIT, although the 2B SLC pump is running following
the SLC control switch being placed in PUMP B RUN. Therefore, at time=1304 (first part
question), the 2B SLC pump is recirculating through the relief valve and there is no SLC
flow into the reactor. Placing the SLC control switch to the PUMP A&B RUN position will
not have any impact on the situation due to the de-energized MCC 2XG. Accordingly, at
time=1306 (second part question), the SLC system is still not injecting into the Reactor
Vessel.
Distractor Analysis:
Choice A: Incorrect because - Both parts of distractor A are incorrect. For the first part question,
the SLC system is NOT injecting into the Vessel because, although the 2B SLC pump is
running, the B squib valve did NOT fire (open) as designed, as indicated by the yellow
continuity light remaining LIT for the B train squib valve. The second part question is also
incorrect, because taking the SLC control switch to the PUMP A&B RUN position does
not increase flow rate due to MCC 2XG being de-energized; no power is available to the
A train of SLC. Plausible because - The applicant may believe that SLC is injecting at
time=1306. The applicant may further believe that actuation of the opposite train would
increase SLC flow rate if the applicant does not recall that MCC 2XG powers the A SLC
train, and it would not actuate in the absence of electrical power.
Choice B: Incorrect because - The first part of distractor B is correct; the SLC system is not
injecting. The second part is incorrect; taking the SLC control switch to the PUMP A&B
RUN position does not cause SLC injection to begin, because MCC 2XG is de-energized;
no power is available to the A train of SLC. Plausible because - The first part of
distractor B is correct. The applicant may believe that actuation of the opposite train of
SLC would cause injection to begin if the applicant does not recall that MCC 2XG powers
the A SLC train. Therefore, distractor B is plausible.
Choice C: Incorrect because - The first part of distractor C is incorrect. The SLC system is NOT
injecting into the Reactor Vessel because the B SLC squib valve did NOT fire (open) as
designed. The 2B SLC pump is actually running with discharge pressure greater than Reactor
pressure, as required to establish conditions to inject, but with the squib valve closed the SLC
system is just dumping the flow rate through the relief valve. The second part answer of
distractor C is correct; the flow rate into the Reactor Vessel has NOT increased (and in fact
remains at zero flow rate into the Reactor. Plausible because - The applicant may believe,
incorrectly, that the B SLC train is injecting into the Vessel, because the 2B SLC pump is
running with discharge pressure greater than Reactor pressure; these are indeed indications
that would be present with injection occurring, whilst and at the same time correctly inferring
from the 2XG outage that the A train of SLC is unavailable. Therefore, distractor C is
plausible.
Choice D: Correct, See explanation.
SRO Basis: N/A
2007 NRC Exam Question
Page: 63 of 193 5/4/2017
Page: 64 of 193 5/4/2017
34. The OATC is directed to manually initiate ARI.
The 2-C12-CS-5560, ARI Auto/Manual Initiation Switch, is mechanically bound in the
AUTO position.
The 2-C12-CS-5561, ARI Initiation Key-lock Switch, is placed in the TRIP position.
Which one of the following completes the statement below?
(1) of the control rods inserted because (2) of the ARI solenoid valves have
repositioned.
A. (1) All
(2) all
B. (1) some
(2) ONLY half
C. (1) None
(2) none
D. (1) None
(2) ONLY half
Page: 65 of 193 5/4/2017
K/A:
212000 Reactor Protection System
A4 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
16 Manually activate anticipated transient without SCRAM circuitry/RRCS:
RO/SRO Rating: 4.4/4.4
Tier 2 / Group 1
K/A match: Applicant must understand that the ARI circuitry requires both manual switches to work to
manually initiate ARI
Pedigree: Modified from 2007 NRC Exam
Objective: LOI-CLS-LP-003, Obj. 16a
Q Reference: SD-03, Rev 12
Ref provided: None
Cog Level: High
Explanation: The contacts for CS-5560 and 5561 are in series. This requires both switches to be used to
manually initiate an ARI.
Distractor Analysis:
Choice A: Incorrect because - This is incorrect because the contacts for CS-5560 and 5561 are in
series. This requires both switches to be used to manually initiate an ARI. This could be
chosen by the applicant if he believes that the ARI switch circuitry works like the RPS
circuitry.
Choice B: Incorrect because - This is incorrect because the contacts for CS-5560 and 5561 are in
series. This requires both switches to be used to manually initiate an ARI. This could be
chosen by the applicant if he believes that the solenoids that do operate, will vent
associated HCUs. Plausibility is also met if the applicant thinks that the ARI controls
design prevents a single failure from keeping the rods from being scrammed.
Choice C: Correct, See explanation.
Choice D: Incorrect because - This is incorrect because the contacts for CS-5560 and 5561 are in
series. This requires both switches to be used to manually initiate an ARI. This could be
chosen by the applicant if he believes that some solenoids would open, but single failure
criteria would have the design not scram any rods. Plausibility is also met if the applicant
knows that ARI controls are not safety related, therefore needing two distinct actions to cause
a scram, but mistakenly believes that the switches feed right into the solenoid circuitry.
SRO Basis: N/A
Page: 66 of 193 5/4/2017
35. A startup is being conducted on Unit Two with the following conditions:
The reactor is critical with SRM/IRM overlap having just been confirmed
All SRMs are reading between 5x103 and 1x104 cps
All IRMs are reading between 35 and 45 on Range 1
The operator has inadvertently selected both the SRMs and the IRMs for withdraw
Which one of the following Control Rod Blocks will be the first automatic action to
occur as the detectors are withdrawn?
A. SRM Downscale
B. IRM Downscale
C. SRM Retract Permissive
D. IRM Detector Not Full In
Page: 67 of 193 5/4/2017
K/A:
215003 Intermediate Range Monitor (IRM) System
K5 Knowledge of the operational implications of the following concepts as they apply to
INTERMEDIATE RANGE MONITOR (IRM) SYSTEM : (CFR: 41.5 / 45.3)
03 Changing detector position
RO/SRO Rating: 3.0/3.1
Tier 2 / Group 1
K/A match: Applicant is required to apply knowledge of the Source and Intermediate Range NI
instrumentation Rod Block circuit to deduce plant impact from an inadvertent NI detector
withdrawal.
Pedigree: Modified Browns Ferry 2015-302 NRC Exam
Objective: LOI-CLS-LP-009-A, Obj. 3a- List the SRM/IRM system signals/conditions that will cause the
following actions and the conditions under which each is bypassed: Rod Blocks (LOCT)
Q Reference: SD-09.1, rev 9
Ref provided: None
Cog Level: High
Explanation: Provided no other Block conditions are met, upon the given condition of initiation of
SRM/IRM detector movement, a Control Rod Block is expected. This would be true for both
detectors.
Distractor Analysis:
Choice A: Incorrect - Plausible as this is an SRM Rod Block and could eventually occur as detectors
continue to be withdrawn. Cannot be construed as a correct answer as the Not Full In
rod block would occur well in advance of an SRM downscale (ref given SRM indications).
Choice B: Incorrect - Plausible as this is an IRM Rod Block. Would not occur with the given conditions
as this Rod Block is auto-bypassed when IRMs are on Range 1.
Choice C: Plausible as this is an SRM Rod Block and could eventually occur as detectors continue to be
withdrawn. Cannot be construed as a correct answer as the Not Full In rod block would occur
well in advance of the Retract Permissive count rate (ref given SRM indications).
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 68 of 193 5/4/2017
36. Which one of the following completes the statement below?
Unit Two SRM C High Voltage Power Supply (HVPS), instrumentation, and trip units
are powered from +/-24 VDC System Panel:
A. 21A
B. 23A
C. 22B
D. 24B
K/A:
215004 Source Range Monitor (SRM) System
K2 Knowledge of electrical power supplies to the following: (CFR: 41.7)
01 SRM channels/detectors
RO/SRO Rating: 2.6/2.8
Tier 2 / Group 1
K/A match: This proposed question simply asks for straightforward memory level recognition of the
correct electrical power supply for the C SRM HVPS, instrumentation, and trip units.
Pedigree: New
Objective: LOI-CLS-LP-009-A, Obj. 2o- State the purpose and/or function of the following components
pertaining to the SRM and IRM Systems as applicable: Power Supplies.
Q Reference: SD-09.1, rev 9
Ref provided: None
Cog Level: High
Explanation: System Description (SD)-09.1, NEUTRON MONITORING SYSTEM (STARTUP AND
INTERMEDIATE RANGE), revision 9, states the following: The SRM and IRM Systems
receive power from the +/- 24 VDC system. Panels 21A(23A) supply channels A/C unit 1 and
2 respectively; and Panels 22B(24B) supply channels B/D unit 1 and 2 respectively to
power the detector HVPS, instrumentation, and trip units.
Distractor Analysis:
Choice A: Incorrect because - Panel 23A, not 21A, is the power supply to the Unit 2 C SRM.
Plausible because - Panel 21A is the power supply to the Unit 1 C SRM.
Choice B: Correct, See explanation.
Choice C: Incorrect because - Panel 23A, not 22B, is the power supply to the Unit 2 C SRM. Plausible
because - Panel 22B is the power supply to the Unit 1 B SRM.
Choice D: Incorrect because - Panel 23A, not 24B, is the power supply to the Unit 2 C SRM. Plausible
because - Panel 24B is the power supply to the Unit 2 D SRM.
SRO Basis: N/A
Page: 69 of 193 5/4/2017
37. Unit Two is operating at 100% power.
Which one of the following identifies a plant transient that will cause the APRM ODA
displays to automatically shift to the stability screen?
A. Control rod 46-15 drifts fully into the core.
B. Recirculation Pump A inadvertently speeds up.
C. 2-FW-V120, FW HTRS 4&5 BYP VLV, opens.
D. Recirculation Pump A trips.
K/A:
215005 Average Power Range Monitor/Local Power Range Monitor System
A4 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)
01 IRM/APRM recorder
RO/SRO Rating: 3.2/3.1
Tier 2 / Group 1
K/A match: Applicant is required to deduce expected APRM recorder response to plant conditions.
Pedigree: Modified - Brunswick 2008-302 Exam Q14
Objective: LOI-CLS-LP-009.6, Obj. 24g- Given PRNMS settings during normal or abnormal operation,
predict the effect on the following:APRM indications following a Recirc Pump or RFP trip.
Q Reference: COLR Cycle 22, rev 0
Ref provided: None
Cog Level: High
Explanation: APRM ODA's automatically shift to the stability screen when the OPRM trip enabled
setpoint is reached (Reactor Power >25% concurrent with RR flow <60%). A RR Pump trip
is the only transient listed (taken separately from the others) which results in flow lowering
(negative reactivity transient) with subsequent entry into the OPRM enabled region.
Distractor Analysis:
Choice A: Incorrect - Negative reactivity transient resulting in lower power. Flow is not affected
thereby avoiding entry into the OPRM enabled region.
Choice B: Incorrect - Positive reactivity transient resulting in higher power. Flow is varied but in a way
that avoids entry into the OPRM enabled region.
Choice C: Incorrect - Positive reactivity transient resulting in higher power. Flow is not affected thereby
avoiding entry into the OPRM enabled region.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 70 of 193 5/4/2017
38. Unit One is operating at 100% power with APRM 1 bypassed.
Subsequently, a critical self-test fault occurs on APRM 3.
Which one of the following completes both statements below?
RBM Channel A reference signals are currently from (1) .
RBM Channel B reference signals are currently from (2) .
A. (1) APRM 2
(2) APRM 2
B. (1) APRM 2
(2) APRM 4
C. (1) APRM 4
(2) APRM 2
D. (1) APRM 4
(2) APRM 4
Page: 71 of 193 5/4/2017
K/A:
215005 Average Power Range Monitor/Local Power Range Monitor
K3 Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE
MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following:
(CFR: 41.7 / 45.4)
07 Rod block monitor
RO/SRO Rating: 3.2/3.3
Tier 2 / Group 1
K/A match: Each RBM channel designates a hierarchy of normal and alternate APRM channels to use as
their reference APRM channel. The alternate channels are used in hierarchical order when the
preferred channels are not available. With a channel bypassed and a subsequent fault on an
APRM channel the applicant is asked to state the effect on the rod block monitor system.
Pedigree: New
Objective: LOI-CLS-LP-09.6, Obj. 13a
Q Reference: SD-09.6 rev 12
Ref provided: None
Cog Level: High
Explanation: Rod block monitor (RBM) channel A is using data from APRM channel 4 due to APRM 1
bypassed and a critical self test failure on APRM channel 3. RBM channel B is using data
from APRM channel 2.
Distractor Analysis:
Choice A: Incorrect because - RBM channel A is using data from APRM channel 4 (the second
alternate). Plausible as APRM channel 2 is the normally assigned as the reference
channel to RBM B
Choice B: Incorrect because - RBM channel A is using data from APRM channel 4 (the second
alternate). Plausible as APRM channel 2 is the normally assigned as the reference
channel to RBM B. RBM channel B is using data from APRM channel 2. Plausible as the
applicant could incorrectly believe that the primary reference channel is APRM channel 4
(the first alternate)
Choice C: Correct, See explanation.
Choice D: Incorrect because - RBM channel B is using data from APRM channel 2. Plausible as the
applicant could incorrectly believe that the primary reference channel is APRM channel 4 (the
first alternate).
SRO Basis: N/A
Page: 72 of 193 5/4/2017
39. Which one of the following completes both statements below?
Demineralized Water Keep Fill System is connected to the RCIC System (1) .
Following a Loss of Offsite Power to both Units, the (2) will assure RCIC System
protection against water hammer.
A. (1) suction line
(2) 2-E51-V70, RCIC Keepfill Station Bypass Valve
B. (1) suction line
(2) 2-E51-F014, RCIC Pump Discharge Check Valve
C. (1) discharge line
(2) 2-E51-V70, RCIC Keepfill Station Bypass Valve
D. (1) discharge line
(2) 2-E51-F014, RCIC Pump Discharge Check Valve
K/A:
217000 Reactor Core Isolation Cooling System (RCIC)
K4 Knowledge of REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) design feature(s)
and/or interlocks which provide for the following: (CFR: 41.7)
01 Prevent water hammer
RO/SRO Rating: 3.2/3.3
Tier 2 / Group 1
K/A match: Applicant is required to apply RCIC system knowledge of Keep Fill system interrelationship
and knowledge of plant response following a LOOP to address RCIC system water hammer
protection to fully answer this question.
Pedigree: New
Objective: LOI-CLS-LP-031.2, Obj. 6
Q Reference: UFSAR, Rev. 22 (6.3.2.6)
Ref provided: None
Cog Level: High
Explanation: As specified in the UFSAR, ECCS keepfill is attached to the RCIC system discharge lines
with the primary function of eliminating discharge line voiding (and subsequent water
hammer effects). Following the given LOOP, the Demineralized Water Pumps will lose
power and will be unavailable to maintain keepfill system pressures to ECCS systems. While
operation of the RCIC keepfill bypass valve is the normal method which operators are
directed to pursue to restore keepfill system pressure, for the given condition, manipulation
of this bypass valve would be ineffective in maintaining protection against water hammer.
Per the UFSAR, ECCS pump discharge check valves are credited with preventing voiding
effects during a loss of keep fill to permit continued use of ECCS systems (if they are
required).
First-part discussion: Suction line piping is plausible because an applicant can theorize
that water pressure would find its way through the centrifugal pump to the discharge piping,
thus keeping both sides of the pump full and pressurized.
Second-part discussion: While operation of the RCIC keepfill bypass valve is the normal
method which operators are directed to pursue to restore keepfill system pressure, for the
Page: 73 of 193 5/4/2017
protection against water hammer. RCIC maintains full auto initiation capability when the site
suffers a LOOP (although its use is not specified for this question).
Distractor Analysis:
Choice A: Incorrect because - Suction line piping is plausible because an applicant can theorize that
water pressure would find its way through the centrifugal pump to the discharge piping,
thus keeping both sides of the pump full and pressurized. Demineralized Water Pumps
will lose power and will be unavailable to maintain keepfill system pressures to ECCS
systems. While operation of the RCIC keepfill bypass valve is the normal method which
operators are directed to pursue to restore keepfill system pressure, for the given
condition, manipulation of this bypass valve would be ineffective in maintaining protection
against water hammer.
Choice B: Incorrect because - Suction line piping is plausible because an applicant can theorize that
water pressure would find its way through the centrifugal pump to the discharge piping,
thus keeping both sides of the pump full and pressurized. The second part is correct.
Choice C: Incorrect because - The first part is correct. Demineralized Water Pumps will lose power and
will be unavailable to maintain keepfill system pressures to ECCS systems. While operation of
the RCIC keepfill bypass valve is the normal method which operators are directed to pursue to
restore keepfill system pressure, for the given condition, manipulation of this bypass valve
would be ineffective in maintaining protection against water hammer.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 74 of 193 5/4/2017
40. Unit Two was operating at 100% power when a rupture in the Recirculation Loop B
occurred.
RHR Pump A Running with discharge pressure of ~225 psig
RHR Pump D OFF
RHR Pumps B / C Unavailable
CS Pump A Running with discharge pressure of ~360 psig
CS Pump B Unavailable
Reactor water level 110 inches and lowering rapidly
ADS NOT inhibited
Which one of the following predicts the response of the ADS system?
A. ADS will initiate in 83 seconds, the timer is now timing.
B. ADS will initiate in 83 seconds after RHR Pump D is started.
C. ADS will initiate in 83 seconds after reactor water level lowers to 105 inches.
D. ADS will initiate in 83 seconds after reactor water level lowers to 45 inches.
Page: 75 of 193 5/4/2017
K/A:
218000 Automatic Depressurization System
A1 Ability to predict and/or monitor changes in parameters associated with operating the
AUTOMATIC DEPRESSURIZATION SYSTEM controls including: (CFR: 41.5 / 45.5)
01 Reactor water level
RO/SRO Rating: 4.1/4.1
Tier 2 / Group 1
K/A match: This question requires the students to determine the response of ADS with lowering level.
Pedigree: New
Objective: LOI-CLS-LP-020, Obj. 11 -Given plant conditions, determine if an automatic initiation of ADS
should occur.
Q Reference: 2APP-A-03, rev 58.
Ref provided: None
Cog Level: High
Explanation: With level lowering an a CS pump running ADS will initiate after 83 seconds of LL3 (45
inches). LPCI requires two pumps in one loop.
Distractor Analysis:
Choice A: Incorrect because the timer has not started until level lowers to LL3.
Choice B: Incorrect because starting the D pump does not provide a Loop of RHR running. Only one
pump in each loop.
Choice C: Incorrect because this is LL2 not LL3.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 76 of 193 5/4/2017
41. With both units at 100% power, a Main Control Room evacuation has been initiated
IAW 0AOP-32.0, Plant Shutdown from Outside Control Room.
Immediate actions were NOT performed by the crew prior to evacuation.
Which one of the following completes the statement below IAW 0AOP-32.0?
(1) SRVs will be cycled remotely to maintain reactor pressure less than a
maximum of (2) .
A. (1) ADS
(2) 950 psig
B. (1) ADS
(2) 1050 psig
C. (1) Non-ADS
(2) 950 psig
D. (1) Non-ADS
(2) 1050 psig
Page: 77 of 193 5/4/2017
K/A:
218000 Automatic Depressurization System
G2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the
resultant operational effects. (CFR: 41.10 / 43.5 / 45.13)
RO/SRO Rating: 4.2/4.1
Tier 2 / Group 1
K/A match: The applicant is expected to know the RO tasks that will be performed outside the control
room during an emergency and understand the resultant operational effects on ADS.
Pedigree: New
Objective: LOI-CLS-LP-302E Obj. 6: Given plant conditions and entry into 0AOP-32.0, Plant Shutdown
From Outside Control Room, explain the basis for a specific caution, note, or series of
procedure steps (LOCT).
Q Reference: 0AOP-32.0, Plant Shutdown from Outside Control Room, Rev. 57; SD-20, Automatic
Depressurization System (ADS), Rev 3
Ref provided: None
Cog Level: High
Explanation: Only SRVs B, E and G (Non-ADS) can be operated from the Remote Shutdown Panel. In
accordance with 0AOP-32.0, the overall mitigative strategy concerning pressure control prior
to initiation of a cooldown is to control reactor pressure less than 1050 psig.
Distractor Analysis:
Choice A: Incorrect because - only SRVs B, E and G (Non-ADS) can be operated from the Remote
Shutdown Panel. 950 psig corresponds to an SRV pressure control measure contained
within RVCP.
Choice B: Incorrect because -Only SRVs B, E and G (Non-ADS) can be operated from the Remote
Shutdown Panel.
Choice C: Incorrect because - 950 psig corresponds to an SRV pressure control measure contained
within RVCP.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 78 of 193 5/4/2017
42. With Unit Two operating at 50% power, which one of the following completes both
statements below?
High Steam Flow sensed in a minimum of (1) steam line(s) will result in a full
MSIV closure signal.
If a full MSIV closure signal were received and the initiating condition had cleared, the
MSIV isolation signal logic (2) be reset with the MSIV control switches in the
OPEN position.
A. (1) one
(2) can
B. (1) one
(2) can NOT
C. (1) two
(2) can
D. (1) two
(2) can NOT
K/A:
223002 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off
K4 Knowledge of PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY
SHUT-OFF design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)
06 Once initiated, system reset requires deliberate operator action
RO/SRO Rating: 3.4/3.5
Tier 2 / Group 1
K/A match: This question satisfies the K/A statement by asking the applicant to know the detectors
required to receive a full MSIV isolation on Hi Stm Flow with knowledge of performing the
manual reset of a full MSIV isolation.
Pedigree: Modified from 2015 NRC Exam
Objective: LOI-CLS-LP-012, obj 8
Q Reference: SD-12 rev11, SD-25 rev15
Ref provided: None
Cog Level: High
Explanation: High steam flow sensed in any 1 line completes the logic for high steam flow in RUN
isolation of the MSIVs. MSIVs that have received an isolation signal may be reset with the
trip signal cleared by depressing isolation reset pushbuttons with all MSIV control switches in
the CLOSED position.
Distractor Analysis:
Choice A: Incorrect because - The isolation signal cannot be reset with the MSIV control switches in
OPEN. Plausible because the control switches will be in the OPEN position without
operator action, the applicant could mistakenly believe that depressing the reset
pushbuttons would be sufficient to reset the isolation logic, which would be successful in
resetting either the A/C or B/D if a partial signal was received.
Page: 79 of 193 5/4/2017
g /C / p g
Choice B: Correct, See explanation.
Choice C: Incorrect because - With the mode switch in Run high steam flow sensed in any 1 line
completes the logic to isolate the MSIVs. Plausible because with the mode switch in
STARTUP high steam flow sensed in any 2 lines completes the logic to isolate the MSIVs. The
isolation signal cannot be reset with the MSIV control switches in OPEN. Plausible because
the control switch will be in the OPEN position without operator action, the applicant could
mistakenly believe that depressing the reset pushbuttons would be sufficient to reset the
isolation logic, which would be successful in resetting either the A/C or B/D if a partial signal
was received.
Choice D: Incorrect because - With the mode switch in Run high steam flow sensed in any 1 line
completes the logic to isolate the MSIVs. Plausible because with the mode switch in Startup
high steam flow sensed in any 2 lines completes the logic to isolate the MSIVs.
SRO Basis: N/A
2015 NRC Exam Question
Page: 80 of 193 5/4/2017
Page: 81 of 193 5/4/2017
43. Unit One has been at 20% reactor power for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
Which one of the following completes both statements below?
The pneumatic supply for the Safety Relief Valves will transition to Backup Nitrogen
supply if a complete loss of (1) supply occurs.
The Safety Relief Valve accumulators are designed to ensure an adequate
pneumatic supply to cycle each valve (2) time(s).
NOTE:
RNA, Reactor Non-Interruptible Air
PNS, Pneumatic Nitrogen System
A. (1) RNA
(2) one
B. (1) RNA
(2) at least five
C. (1) PNS
(2) one
D. (1) PNS
(2) at least five
Page: 82 of 193 5/4/2017
K/A:
239002 Relief/Safety Valves
K6 Knowledge of the effect that a loss or malfunction of the following will have on the
RELIEF/SAFETY VALVES : (CFR: 41.7 / 45.7)
02 Air (Nitrogen) supply: Plant-Specific
RO/SRO Rating: 3.4/3.5
Tier 2 / Group 1
K/A match: The K/A is met because knowledge of the pneumatic supply to the Safety Relief Valves
(SRVs) given specific plant conditions and the effect of losing all normal pneumatic supplies
(PNS, RNA, and Backup Nitrogen) would have on the ability to operate the SRVs is required
to answer the question.
Pedigree: New
Objective: LOI-CLS-LP-020, Obj. 15a- Given plant conditions, predict how ADS/SRVs will be affected by
the following: Loss of Non-Interruptible Air to the Drywell (LOCT)
Q Reference: SD-20 Rev. 3, 0OP-46 Rev. 179, TS 3.6.3.1 Amendment No. 203, SD-25 Rev. 15
Ref provided: None
Cog Level: High
Explanation: PNS is required to be aligned to the drywell pneumatic loads within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of raising
power to greater than or equal to 15%. If the normal pneumatic supply is lost, the SRV
accumulators ensure a supply of air adequate to provide a minimum of five cycles or hold
the valve open for 30 minutes if it is already open.
Distractor Analysis:
Choice A: Incorrect because - RNA may only be aligned if power is <15%. The SRV accumulators
will supply enough air for 5 cycles of the SRVs. This is plausible because the MSIV
accumulators will supply enough air for one stroke. See explanation for correct answer
above.
Choice B: The SRV accumulators will supply enough air for 5 cycles of the SRVs. This is plausible
because the MSIV accumulators will supply enough air for one stroke. See explanation for
correct answer above.
Choice C: RNA may only be aligned if power is <15%. See explanation for correct answer above.
Choice D: Correct, See explanation.
SRO Basis:
N/A
Page: 83 of 193 5/4/2017
44. A Unit One startup is in progress IAW 0GP-02, Approach to Criticality and
Pressurization of the Reactor.
Operators are currently performing 1OP-32, Condensate and Feedwater System
Operating Procedure, Section 6.1.5, Reactor Feed pump Operation from Idle Speed to
Injection at Low Pressure Conditions.
Current plant conditions are:
RFPT A Speed Controller in MANUAL
RFPT speed at 2550 rpm
RFPT A discharge pressure 100 psig higher than reactor pressure
Feedwater Control Mode Select Switch in 1 ELEM
RFPT A MAN/DFCS control switch in MAN
IAW 1OP-32, Section 6.1.5, which one of the following completes both statements
below?
When the RFPT A MAN/DFCS control switch is selected to DFCS, the (1) will
control RFPT speed.
Following the DFCS selection performed above, if the DFCS control signal lowers to
2400 rpm, the (2) will maintain RFPT speed.
A. (1) C32-SIC-R600, Master RFPT Speed/Reactor Level Controller
(2) Woodward 5009 digital control
B. (1) C32-SIC-R600, Master RFPT Speed/Reactor Level Controller
(2) C32-SIC-R601A, RFPT A Speed Controller
C. (1) C32-SIC-R601A, RFPT A Speed Controller
(2) Woodward 5009 digital control
D. (1) C32-SIC-R601A, RFPT A Speed Controller
(2) C32-SIC-R601A, RFPT A Speed Controller
Page: 84 of 193 5/4/2017
K/A:
259002 Reactor Water Level Control System
A1 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR
04 Reactor water level control controller indications
RO/SRO Rating: 3.6/3.6
Tier 2 / Group 1
K/A match: Test item tests the applicants ability to predict how a change in MAN/DFCS switch position
affects the RWLC system, and also tests the applicants knowledge of how the DFCS works
when rpm decreases below 2450 rpm.
Pedigree: New
Objective: LOI-CLS-LP-32.2 Obj 5a: Describe the operation of the DFCS in the following operating
modes: Master Level Control Mode (auto and manual).
Q Reference: 1OP-32, Section 6.1.5
Ref provided: None
Cog Level: High
Explanation: As long as the RFPT A Speed Controller remains in MANUAL, the Master Controller will not
be controlling RFPT speed; Also, the Woodward controls take over if RFPT speed control
signal drops below 2450 rpm when DFCS is selected.
Distractor Analysis:
Choice A: Incorrect because at this point in the procedure, the Master Controller will not control
RFPT A Speed because the RFPT A Controller is still in MANUAL. Plausible because
placing the feedpumps MAN/DFCS switch to the DFCS position could be misconstrued
as transferring control to the Master Controller.
Choice B: Incorrect because at this point in the procedure, the Master Controller will not control
RFPT A Speed because the RFPT A Controller is still in MANUAL. Also incorrect because
when the DFCS speed control signal drops below 2450 rpm, the Woodward controls take
over. Plausible because placing the feedpumps MAN/DFCS switch to the DFCS position
could be misconstrued as transferring control to the Master Controller. Also plausible
because if the Master Controller were in effect, and DFCS speed control signal lowered
below 2450 rpm, the RFPT could be misconstrued as being back in control again.
Choice C: Correct, See explanation.
Choice D: Incorrect because - at this point in the procedure, the Master Controller will not control RFPT
A Speed because the RFPT A Controller is still in MANUAL. Plausible because placing the
feedpumps MAN/DFCS switch to the DFCS position could be misconstrued as transferring
control to the Master Controller. Also plausible because RFPT A Controller remains in MAN,
even after the DFCS is selected.
SRO Basis: N/A
Page: 85 of 193 5/4/2017
45. Which one of the following completes both statements below?
Upon receipt of a SBGT initiation signal, 2-SGT-V8 and 2-SGT-V9 (1) receive an
automatic actuation signal.
These valves (2) provide a flow path for the HPCI Barometric Condenser Vacuum
Pump discharge to the SBGT suctions.
NOTE:
2-SGT-V8, Post LOCA Vent
2-SGT-V9, Post LOCA Vent
A. (1) will
(2) do
B. (1) will
(2) do NOT
C. (1) will NOT
(2) do
D. (1) will NOT
(2) do NOT
Page: 86 of 193 5/4/2017
K/A:
261000 Standby Gas Treatment System
AA2 Knowledge of the physical connections and/or causeeffect relationships between STANDBY
GAS TREATMENT SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
02 High pressure coolant injection system
RO/SRO Rating: 3.0/3.1
K/A match: Applicant is asked for automatic SBGT initiation response of SGT-V8 and SGT-V9 in addition
to the discharge flowpath relationship between the HPCI Barometric Condenser and the
SBGT system.
Pedigree: New
Objective: LOI-CLS-LP-10.0
Obj. 3 Draw one-line diagram of the SBGT System, including major flow paths,
components, and connections with other systems
Obj. 4 Given plant conditions determine if SBGTs should have initiated.
Q Reference: SD-10, Rev. 7, 2OP-10, Rev. 81
Ref provided: None
Cog Level: Fundamental
Explanation: A SBGT initiation signal provides no input for automatic operation of the SGT-V8 and
SGT-V9 MOVs. When in a standby alignment (i.e. during normal plant configuration) these
valves are in their normal OPEN position. Operator action is required to manipulate these
valves.
The HPCI Barometric Condenser exhaust is among several flowpaths served by the
SGT-V8 and SGT-V9.
Distractor Analysis:
Choice A: Incorrect - The first half distractor is plausible due to the presence of several SBGT
system valves that do receive either 1) open signals upon SBGT actuation (SBGT
alignment BFVs) or 2) close signals upon SBGT actuation (Purge Fan BFVs). The
second part is correct.
Choice B: Incorrect - The first half distractor is plausible due to the presence of several SBGT
system valves that do receive either 1) open signals upon SBGT actuation (SBGT
alignment BFVs) or 2) close signals upon SBGT actuation (Purge Fan BFVs). The
second half distractor is plausible due to the multitude of flowpaths that can be aligned
through the SGT-V8 and SGT-V9 which include a Drywell, Torus, and RCIC Barometric
Condenser discharge path. Plausibility is enhanced due to the naming of the SGT-V8 and
SGT-V9.
Choice C: Correct, See explanation.
Choice D: Incorrect - The first half is correct. The second half distractor is plausible due to the multitude
of flowpaths that can be aligned through the SGT-V8 and SGT-V9 which include a Drywell,
Torus, and RCIC Barometric Condenser discharge path. Plausibility is enhanced due to the
naming of the SGT-V8 and SGT-V9.
SRO Basis: N/A
Page: 87 of 193 5/4/2017
46. Unit Two was operating at 100% power when a large break LOCA occurred.
Which one of the following identifies the load sequence on Bus E4?
A. Core Spray then RHR
B. RHR then Core Spray
C. Core Spray, RHR and then the Fire Pump
D. RHR, Core Spray and then the Fire Pump
K/A:
262001 A.C. Electrical Distribution
A3 Ability to monitor automatic operations of the A.C. ELECTRICAL DISTRIBUTION including:
(CFR: 41.7 / 45.7)
04 Load sequencing
RO/SRO Rating: 3.4/3.6
Tier 2 / Group 1
K/A match: This question matches the K/A because it tests the operators ability to describe the proper
loads and sequence that automatically occur for a given E bus following plant events that
initiate the load sequence.
Pedigree: New
Objective: LOI-CLS-LP-039, Obj. 8
Q Reference: SD-50.1 rev 23
Ref provided: None
Cog Level: High
Explanation: In the event of a LOCA on a Unit with or without a loss of off-site power, the loads
associated with the LOCA will sequentially load on the E-Bus. The sequential loading is as
follows: Ten seconds after the time reference the 2B RHR Pump starts, Fifteen seconds
after the time reference the 2B Core Spray Pump starts. The Fire pump is normally fed from
E2.
Distractor Analysis:
Choice A: Incorrect because - The core spray pump sequences on after the RHR pump, plausible
as the applicant could mistake the correct sequence..
Choice B: Correct, See explanation.
Choice C: Incorrect because- The core spray pump sequences on after the RHR pump, plausible as the
applicant could mistake the correct sequence. The Fire Pump will start in on bus E2 (the
normal power source), the applicant could mistakenly think the Fire Pump started on bus E4
(the alternate power source) if this were the case it would be last to start.
Choice D: Incorrect because - The Fire Pump will start in on bus E2 (the normal power source), the
applicant could mistakenly think the Fire Pump started on bus E4 (the alternate power source)
if this were the case it would be last to start.
SRO Basis: N/A
Page: 88 of 193 5/4/2017
47. Which one of the following completes both statements below?
The Stack Radiation Monitor is capable of being provided with Uninterruptible Power
Supply (UPS) power using a/an (1) transfer device.
With the Stack Rad Monitor Power Transfer Switches in the NORM position,
(2) UPS is supplying power to the Stack Radiation Monitor.
A. (1) automatic
(2) Unit One
B. (1) automatic
(2) Unit Two
C. (1) manual
(2) Unit One
D. (1) manual
(2) Unit Two
Page: 89 of 193 5/4/2017
K/A:
262002 Uninterruptable Power Supply (A.C./D.C.)
K1 Knowledge of the physical connections and/or causeeffect relationships between
UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) and the following: (CFR: 41.2 to 41.9 / 45.7
to 45.8)
15 Stack gas monitors: Plant-Specific
RO/SRO Rating: 2.7/3.0
Tier 2 / Group1
K/A match: The applicant is expected to know the physical connections between UPS and stack gas
monitors.
Pedigree: BNP 2014 Q59
Objective: LOI-CLS-LP-052 Obj 5: Given plant conditions, determine the lineup of the primary UPS, the
Standby UPS, and their reserve sources. (LOCT)
Q Reference: SD-52, 120 VAC Electrical System, Rev. 4; 2-OP-52, 120 Volt AC UPS, Emergency, and
Conventional Electrical Systems Operating Procedure, Rev 60
Ref provided: None
Cog Level: Fundamental
Explanation: Stack Radiation Monitor is capable of being provided with power by the UPS of either unit
using a manual transfer device. Stack Rad Monitor Power Transfer Switches in NORM
position indicates that Unit 2 UPS is supplying power to the Stack Radiation Monitor.
Distractor Analysis:
Choice A: Incorrect because - Stack Radiation Monitor is capable of being provided with power by
the UPS of either unit using a manual transfer device. Stack Rad Monitor Power Transfer
Switches in NORM position indicates that Unit 2 UPS is supplying power to the Stack
Radiation Monitor.
Choice B: Incorrect because - Stack Radiation Monitor is capable of being provided with power by
the UPS of either unit using a manual transfer device.
Choice C: Incorrect because - Stack Rad Monitor Power Transfer Switches in NORM position indicates
that Unit 2 UPS is supplying power to the Stack Radiation Monitor.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 90 of 193 5/4/2017
48. Which one of the following completes the statement below?
Upon a loss of 125VDC Distribution Panel 3A, a loss of the (1) flow controller
indication and annunciation of (2) will occur.
A. (1) HPCI
(2) A-01 (1-6), CORE SPRAY SYS I LOGIC PWR FAILURE
B. (1) HPCI
(2) A-04 (1-8), STM LINE LOW PRESS A
C. (1) RCIC
(2) A-01 (1-6), CORE SPRAY SYS I LOGIC PWR FAILURE
D. (1) RCIC
(2) A-04 (1-8), STM LINE LOW PRESS A
Page: 91 of 193 5/4/2017
K/A:
26300 D.C. Electrical Distribution
A3 Ability to monitor automatic operations of the D.C. ELECTRICAL DISTRIBUTION including:
(CFR: 41.7 / 45.7)
01 Meters, dials, recorders, alarms, and indicating lights
RO/SRO Rating: 3.2/3.3
Tier 2 / Group 1
K/A match: The applicant is expected to know the resulting indications and alarms from a loss of a D.C.
bus.
Pedigree: New
Objective: LOI-CLS-LP-051, Obj. 4 and 10
Q Reference: 0AOP-39, Loss of DC Power, R42; 1OP-51, R74; 1APP-A-01, R57; 1APP-A-04, R58
Ref provided: None
Cog Level: High
Explanation: Loss of 125V DC Distribution Panel 3A causes control room annunciator, Core Spray Sys I
Logic Pwr Failure Annunciator (A-01, 1-6) and loss of indication to the HPCI flow controller.
Loss of 125V DC Distribution Panel 11A/12A results in control room annunciator Steam Line
Low Press A Annunciator (A-04, 1-8). Loss of 125V DC Distribution Panel 3B/4B results in
loss of indication to the RCIC flow controller.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect because - Loss of 125V DC Distribution Panel 11A/12A results in control room
annunciator Steam Line Low Press A Annunciator (A-04, 1-8).
Choice C: Incorrect because - Loss of 125V DC Distribution Panel 3B/4B results in loss of indication to
the RCIC flow controller.
Choice D: Incorrect because - Loss of 125V DC Distribution Panel 11A/12A results in control room
annunciator Steam Line Low Press A Annunciator (A-04, 1-8). Loss of 125V DC Distribution
Panel 3B/4B results in loss of indication to the RCIC flow controller.
SRO Basis: N/A
Page: 92 of 193 5/4/2017
49. DG2 is running in Control Room Manual during performance of the 24-month load test
portion of 0PT-12.2B, No. 2 Diesel Generator Monthly Load Test.
DG2 is loaded to 3700 kW.
Subsequently, Unit Two experiences a Loss of Offsite Power (LOOP)
Which one of the following completes both statements below?
Bus E2 (1) remained energized throughout the LOOP.
When restoring off-site power back onto Bus E2, prior to transferring DG2 from AUTO
to MANUAL, DG2 frequency will be required to be (2) .
A. (1) has
(2) raised
B. (1) has
(2) lowered
C. (1) has NOT
(2) raised
D. (1) has NOT
(2) lowered
Page: 93 of 193 5/4/2017
K/A:
264000 Emergency Generators (Diesel/Jet)
A2 Ability to (a) predict the impacts of the following on the EMERGENCY GENERATORS
(DIESEL/JET) ; and (b) based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
07 Loss of off-site power during full-load testing
RO/SRO Rating: 3.5/3.7
Tier 2 / Group 1
K/A match: The question is asking the applicant to predict the impact on the E2 bus after a LOOP that
occurred during a DG2 load test. Then the applicant is asked to determine frequency control
actions for follow-on DG2 operations.
Pedigree: Modified from 2012 BNP NRC Q30
Objective: LOI-CLS-LP-039, Obj 10a-Given plant conditions, determine the effects of the following
events on EDG availability and operation: A LOOP during full load EDG testing (LOCT)
Q Reference: AOI-CLS-LP-39, Obj. 03e; 0PT-12.2B, Rev 107
Ref provided: None
Cog Level: High
Explanation: Any auto start signal when in the manual mode will cause the EDG controls to automatically
revert to isochronous mode. If this occurs during load testing and an E bus under-voltage
condition exists, the EDG will automatically trip and then tie back onto the bus in the
isochronous mode. The speed control rheostat will be set at the higher load. Therefore the
EDG will tie onto the bus at an elevated frequency which requires lowering prior to
paralleling with offsite power.
Distractor Analysis:
Choice A: Plausible because this could be correct depending on given plant conditions and nature of the
affected casualty on DG2.
Choice B: Plausible because this could be correct depending on given plant conditions and nature of the
affected casualty on DG2.
Choice C: Plausible because this could be correct depending on given plant conditions and nature of the
affected casualty on DG2.
Choice D: Correct Answer, see explanation.
SRO Basis: N/A
Page: 94 of 193 5/4/2017
50. Unit Two is operating at 100% power with the following conditions:
Unit One and Unit Two Service Air Systems are cross-tied
2D Service Air Compressor is the lead compressor
2A Air Dryer in service
Subsequently,
Control Power for the 2A Air Dryer is lost
2-UA-1 (5-3) AIR DRYER 2A TROUBLE is in ALARM
Which one of the following completes both statements below?
The loss of control power to Air Dryer 2A will result in (1) .
IAW 2-APP-UA-01 (5-3), the crew is required to (2) .
A. (1) continued air flow through one tower of Air Dryer 2A
(2) place Service Air Dryer 1B in service and remove 2A from service
B. (1) continued air flow through one tower of Air Dryer 2A
(2) consider swapping 1D Service Air Compressor into lead
C. (1) 2-SA-PV-5067, Service Air Dryer Bypass Pressure Control Valve, opening
(2) place Service Air Dryer 1B in service and remove 2A from service
D. (1) 2-SA-PV-5067, Service Air Dryer Bypass Pressure Control Valve, opening
(2) consider swapping 1D Service Air Compressor into lead
Page: 95 of 193 5/4/2017
K/A:
300000 Instrument Air System (IAS)
A2 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b)
based on those predictions, use procedures to correct, control, or mitigate the consequences of
those abnormal operation: (CFR: 41.5 / 45.6)
01 Air dryer and filter malfunctions
RO/SRO Rating: 2.9/2.8
Tier 2 / Group 1
K/A match: Question meets the K/A because it requires knowledge of the effect a loss of control power to
the air dryer has on the instrument air system and the APP requirements for the associated
alarm.
Pedigree: New
Objective: LOI-CLS-LP-46-A, Obj. 6g- Given plant conditions, determine if the following automatic actions
should occur: (LOCT) Air Dryer bypass.
Q Reference: U2 APP UA-01 5-3 Rev. 83,
Ref provided: None
Cog Level: High
Explanation: Part 1: In accordance with APP UA-01 5-3, Air Dryer 2A Trouble, if control power is lost or
interrupted, the air dryer will fail safe, providing air flow through one tower. Part 2: The
actions of APP UA-01 5-3 require placing Air Dryer 1B in service if it is available.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Part 1 is correct, see explanation. Part 2 is plausible because if Air Dryer 1B is not
available the APP requires considering swapping the 1D Service Air Compressor to the
lead compressor.
Choice C: Part 1 is incorrect because the air dryer bypass valve would only open if air pressure lowers to
98 psig. This could happen if the loss of control power caused a loss of flow through the air
dryer. This is plausible because a note in 0AOP-20 states In service air compressor high
discharge pressure (Unit 1: greater than or equal to 125 psig, Unit 2: greater than or equal to
130 psig) or relief valves lifting could be an indication of air dryer high differential pressure
potentially caused by power failures resulting in valves in the flow path failing closed..... With
continued air flow through the dryer this should not occur. Part 2 is correct, see explanation.
Choice D: Part 1 is incorrect because the air dryer bypass valve would only open if air pressure lowers to
98 psig. This could happen if the loss of control power caused a loss of flow through the air
dryer. This is plausible because a note in 0AOP-20 states In service air compressor high
discharge pressure (Unit 1: greater than or equal to 125 psig, Unit 2: greater than or equal to
130 psig) or relief valves lifting could be an indication of air dryer high differential pressure
potentially caused by power failures resulting in valves in the flow path failing closed..... With
continued air flow through the dryer this should not occur. Part 2 is plausible because if Air
Dryer 1B is not available the APP requires considering swapping the 1D Service Air
Compressor to the lead compressor.
SRO Basis:
N/A
Page: 96 of 193 5/4/2017
51. Which one of the following completes the statement below?
With a loss of the Station Air Compressors, RNA-SV-5481, Division II Backup N2 Rack
Isolation Valve, will automatically actuate when instrument air pressure first lowers
to less than:
A. 105 psig
B. 100 psig
C. 98 psig
D. 95 psig
K/A:
300000 Instrument Air System (IAS)
K5 Knowledge of the operational implications of the following concepts as they apply to the
INSTRUMENT AIR SYSTEM: (CFR: 41.5 / 45.3)
01 Air compressors
RO/SRO Rating: 2.5/2.5
Tier 2 / Group 1
K/A match: The applicant is expected to determine which instrument air pressure corresponds to
automatic actuation of an instrument air component upon a loss of plant air compressors.
Pedigree: New
Objective: LOI-CLS-LP-46-A, Obj. 6c- Given plant conditions, determine if the following automatic actions
should occur: (LOCT) Nitrogen Backup Initiation.
Q Reference: 0AOP-20.0, rev 46
Ref provided: None
Cog Level: Fundamental
Explanation: In accordance with 0AOP-20.0 Section 3.0, the Division I and II Backup N2 Rack Isolation
Valves (RNA-SV-5481 and RNA-SV-5482) open when pressure lowers to 95 psig.
Distractor Analysis:
Choice A: Incorrect, but plausible because the service air header automatically isolates at 105 psig.
Choice B: Incorrect, but plausible because the interruptible instrument air header would be isolated
manually if conditions were met and pressure lowered to less than 100 psig.
Choice C: Incorrect, but plausible because the air dryer bypass valve begins to open if pressure lowers to
98 psig.
Choice D: Correct, See explanation.
SRO Basis:
N/A
Page: 97 of 193 5/4/2017
52. Unit One is operating at 100% power when a malfunction in a running Conventional
Service Water (CSW) pump motor causes CSW system pressure to start lowering.
Which one of the following completes both statements below?
The highest CSW system pressure that will cause an automatic start signal to be
generated for the standby CSW pump is (1) .
If CSW system pressure remains below the pressure setpoint selected above for a
minimum of (2) , then SW-V3(V4), SW TO TBCCW HXS OTBD(INBD) ISOL, will
reposition to their throttled positions.
A. (1) 65 psig
(2) 30 seconds
B. (1) 65 psig
(2) 70 seconds
C. (1) 40 psig
(2) 30 seconds
D. (1) 40 psig
(2) 70 seconds
Page: 98 of 193 5/4/2017
K/A:
400000 Component Cooling Water System (CCWS)
K6 Knowledge of the effect that a loss or malfunction of the following will have on the CCWS: (CFR:
41.7 / 45.7)
05 Motors
RO/SRO Rating: 2.8/2.9
Tier 2 / Group 1
K/A match: Given an operationally valid set of plant conditions involving a proposed malfunction of
running CSW pump motors, the applicant will demonstrate knowledge of the effect of this
malfunction by correctly recognizing the automatic plant response to this condition.
Pedigree: Brunswick NRC 2010-301 RO Q63
Objective: LOI-CLS-LP-43, Obj. 6b- Given plant conditions, predict whether any of the following pumps
should start: Conventional Service Water Pumps.
Q Reference: SD-43, 0AOP-19.0, revision 26
Ref provided: None
Cog Level: Fundamental
Explanation: Part 1: The standby pump selected to the conventional service water header starts at 40
psig. Part 2: IF conventional service water header pressure remains less than 40 psig for
70 seconds, THEN SW-V3 (SW To TBCCW HXs Otbd Isol), closes to a throttled position
and SW-V4 (SW To TBCCW HXs Inbd Isol), closes to a throttled position.
Distractor Analysis:
Choice A: Part 1 is incorrect,but plausible because RCC pumps auto-start setpoint is 65 psig. Part 2
is incorrect but plausible because D/G cooling valves swap to the opposite unit after low
pressure for 30 seconds.
Choice B: Part 1 is incorrect,but plausible because RCC pumps auto-start setpoint is 65 psig. Part 2
is correct, see explanation.
Choice C: Part 1 is correct, see explanation. Part 2 is incorrect but plausible because D/G cooling valves
swap to the opposite unit after low pressure for 30 seconds.
Choice D: Correct, See explanation.
SRO Basis:
N/A
Page: 99 of 193 5/4/2017
53. Unit Two is operating at 100% power with the following RBCCW switch positions:
Subsequently:
E3 Feeder Breaker (Master Breaker), 2-2D-AD1, inadvertently trips
DG3 fails to auto-start
Which one of the following completes both statements below three minutes after the
breaker trip?
RBCCW Pump 2B (1) running.
RBCCW Pump 2C (1) running.
A. (1) is
(2) is
B. (1) is
(2) is NOT
C. (1) is NOT
(2) is
D. (1) is NOT
(2) is NOT
Page: 100 of 193 5/4/2017
K/A:
400000 Component Cooling Water System (CCWS)
K6 Knowledge of the effect that a loss or malfunction of the following will have on the CCWS: (CFR:
41.7 / 45.7)
07 Breakers, relays, and disconnects
RO/SRO Rating: 2.7/2.8
Tier 2 / Group 1
K/A match: Given an operationally valid set of plant conditions involving a breaker malfunction and an
automatic start failure of a D/G, the applicant will demonstrate knowledge of the effect these
conditions will have on the Reactor Building Closed Cooling Water (RBCCW) pumps.
Pedigree: Brunswick NRC 2012-301 RO Q63
Objective: LOI-CLS-LP-21, Obj. 9a- Given plant conditions, determine how RBCCW is affected by the
following: Loss of AC power.
Q Reference: SD-21, Reactor Building Closed Cooling Water System, revision 10;0OI-50.3 Rev 62
Ref provided: None
Cog Level: High
Explanation: Part 1: The loss of E3 does not result in a loss of power to the 2B RBCCW pump. It does
lose power to the start pressure switch, but since it is already running this has no effect
on the pump. Therefore, the 2B pump will be running 3 minutes after the transient.
Part 2: The loss of E3 results in RBCCW pumps 2A and 2C without power so the 2C
would not be running 3 minutes after the transient.
Distractor Analysis:
Choice A: Part 1 is correct, see explanation. Part 2 is incorrect but plausible the candidate may
believe that the 2B and 2C pump are effected similarly. For instance in reality the 2A and
2C pumps are effected similarly by the loss of E3.
Choice B: Correct, See explanation.
Choice C: Part 1 is incorrect, but plausible because it does lose its auto start capability. Part 2 is
incorrect but plausible the candidate may believe that the 2B and 2C pump are effected
similarly. For instance in reality the 2A and 2C pumps are effected similarly by the loss of E3.
Choice D: Part 1 is incorrect, but plausible because it does lose its auto start capability. Part 2 is correct,
see explanation
SRO Basis: N/A
Page: 101 of 193 5/4/2017
54. Which one of the following completes the statement below?
Alternate Rod Insertion (ARI) solenoids are powered from (1) and (2) to
actuate.
A. (1) 120 VAC
(2) de-energize
B. (1) 120 VAC
(2) energize
C. (1) 125 VDC
(2) de-energize
D. (1) 125 VDC
(2) energize
Page: 102 of 193 5/4/2017
K/A:
201001 Control Rod Drive Hydraulic System
K2 Knowledge of electrical power supplies to the following: (CFR: 41.7)
05 Alternate rod insertion valve solenoids
RO/SRO Rating: 3.1/3.1
Tier 2 / Group 1
K/A match: Candidate must know the generic power supply (AC or DC) to the ARI valve solenoids.
Pedigree: Peachbottom 2008 NRC Exam
Objective: LOI-CLS-LP-017, OBJ 17.a.
Q Reference: SD-03, RPS with ARI, Rev. 12
Ref provided: None
Cog Level: Fundamental
Explanation: Several solenoids are associated with reactor trip systems: the normal Scram Pilot valves,
the Scram Discharge Volume vent and drain valves, the Backup Scram valves, and the ARI
valves. They are supplied by either 120VAC or 125VDC, and some de-energize to actuate,
while others energize to actuate:
Valve Power Actuation
Scram Pilot 120VAC de-energize
SDV vents/drains 120VAC de-energize
Backup Scram 120VDC energize
D is CORRECT: the ARI valves are supplied by 125VDC Panels 11A & 12A. Normally
de-energized, they energize to actuate.
120VAC is plausible because that is what supplies the normal Scram Pilot and SDV
vent/drain solenoids. Note that the RPS Trip Cabinets operate on 125VDC, but its 120VAC
that actually holds the Scram Pilot solenoids closed. This DC-to-AC configuration might lead
an applicant to rule out DC for ARI.
De-energize is plausible because thats what the Scram Pilot and SDV vent/drain solenoids
do. Applicants might reason that ARI, being a backup system, would deenergize to actuate,
or fail safe, like EDG air start solenoids typically do. They might further reason that the
normal reactor trip system would be designed to only trip the reactor when you meant to
(energize-to-actuate), and not be susceptible to say, a blown fuse, and therefore the backup
ARI system would be opposite.
SRO Basis: N/A
Page: 103 of 193 5/4/2017
55. Which one of the following identifies attributes of the control rod drive (CRD)
mechanism systems purpose?
1. Each control rod (blade) is positioned and supported by a separate positioning
device, and a failure of any positioning device will not affect the operation of any
other positioning device.
2. The CRD mechanisms provide motive force for positioning the control rod
blades to control reactor power during normal and scram conditions.
3. The control rod blades provide excess negative reactivity to shut down the
reactor from any normal operation or accident condition at the most reactive time
in core life.
4. In the event of a scram, the CRD system provides sufficiently rapid control rod
(blade) insertion so that no fuel damage results from any abnormal operating
A. 1 and 2
B. 2 and 3
C. 3 and 4
D. 4 and 1
Page: 104 of 193 5/4/2017
K/A:
G2.1.27 Knowledge of system purpose and/or function. (CFR: 41.7)
RO/SRO Rating: 3.9/4.0
Tier 2 / Group 2
K/A match: The applicant is expected to know the purposes of the control rod drive mechanism system in
order to answer the question.
Pedigree: New
Objective: LOI-CLS-LP-008.1 Obj 1: State the purpose of the Control Rod Drive (CRD) Mechanism
Q Reference: SD-08.1, Control Rod Drive (CRD) Mechanism, Rev. 6
Ref provided: None
Cog Level: Fundamental
Explanation: The CRD mechanisms provide motive force for positioning the control rod blades to control
reactor power during normal and scram conditions. The control rod blades provide control of
reactor power for shaping of both axial and radial flux (power) profiles, and to provide excess
negative reactivity to shutdown the reactor from any normal operation or accident condition
at the most reactive time in core life. These statements are CRD system purposes.
Distractor Analysis:
Choice A: Incorrect because - option 1 is incorrect. Each control rod (blade) is positioned and
supported by a separate positioning device, and a failure of any positioning device will not
affect the operation of any other positioning device. This factual statement is a design
basis consideration for the CRD system and not a system purpose.
Choice B: Correct, See explanation.
Choice C: Incorrect because - option 4 is incorrect. In the event of a scram, the CRD system provides
sufficiently rapid control rod (blade) insertion so that no fuel damage results from any
abnormal operating transient. This factual statement is part of the design basis for the CRD
system and not a system purpose.
Choice D: Incorrect because - option 1 & 4 are both incorrect since these statements are part of the
CRD system design basis and not system purpose.
SRO Basis: N/A
Page: 105 of 193 5/4/2017
56. Unit One is at 100% power with I/C technicians performing a reactor water level
instrument surveillance.
Subsequently, I/C inadvertently causes a downscale failure of the 1-B21-LTS-N031A-4
and C-4, Level Transmitter Slave Trip Units which results in the following:
A-1 (1-8), RHR SYS I ACTUATED alarming
RHR Pumps 1A and 1C running
Which one of the following completes both statements below?
The 1-B32-F031A, (Recirculation) Pump A Disch Vlv, (1) received an isolation
signal.
The purpose of the 1-B32-F031A isolation signal is to ensure (2) .
A. (1) has
(2) RHR pumps inject into the core region through the jet pumps to provide
maximum core cooling
B. (1) has
(2) any recirculation line break will not prevent re-flooding the reactor core to at
least the height of the jet pump throat inlets (2/3 active core height)
C. (1) has NOT
(2) RHR pumps inject into the core region through the jet pumps to provide
maximum core cooling
D. (1) has NOT
(2) any recirculation line break will not prevent re-flooding the reactor core to at
least the height of the jet pump throat inlets (2/3 active core height)
Page: 106 of 193 5/4/2017
K/A:
202002 Recirculation Flow Control System
K6 Knowledge of the effect that a loss or malfunction of the following will have on the
RECIRCULATION FLOW CONTROL SYSTEM : (CFR: 41.7 / 45.7)
05 Reactor water level
RO/SRO Rating: 3.1/3.1
Tier 2 / Group 2
K/A match: Given a failure of the A and C LL3 (+45) Level transmitters, the applicant is expected to
provide an expected system response and actuation signal purpose.
Pedigree: New
Objective: LOI-CLS-LP-002, Obj. 13- Explain why the Reactor Recirculation Pump discharge valves have
an automatic closure feature.
Q Reference: SD-02, REACTOR RECIRCULATION SYSTEM, Rev 20
Ref provided: None
Cog Level: Fundamental
Explanation: Part 1: Although a repositioning of the F031A will result from indication of a LOCA signal
(level LOCA signal provided), actuation of the F031A will only occur when reactor
pressure drops below 310 psig.
Part 2: The purpose of the F031A isolation signal is to ensure RHR pumps inject into the
core region through the jet pumps to provide maximum core cooling.
Distractor Analysis:
Choice A: Part 1 is plausible since a LOCA signal does cause the isolation signal, however that
signal is 310 psig and not LL3. Part 2 is correct, see explanation.
Choice B: Part 1 is plausible since a LOCA signal does cause the isolation signal, however that
signal is 310 psig and not LL3. Part 2 is plausible because the statement is factual, on its
own, however it is not the purpose of the isolation signal.
Choice C: Correct, See explanation.
Choice D: Part 1 is correct, see explanation. Part 2 is plausible because the statement is factual, on its
own, however it is not the purpose of the isolation signal.
SRO Basis:
N/A
Page: 107 of 193 5/4/2017
57. Unit One is at 50% power with control rod 02-19 currently selected when the following
Rod Block Monitor (RBM) timeline of events occur:
1000: RBM B indicates receipt of a non-critical self-test fault
1005: RBM A flux indicates 2% and stable
1010: RBM A watchdog timer has timed out
1015: RBM B critical self-test fault is received
There are no rod movements in progress.
Which one of the following identifies the earliest time annunciators A-05 (2-2), ROD
OUT BLOCK, and A-06 (4-7), RBM DOWNSCALE/ TROUBLE, will BOTH be in alarm?
A. 1000
B. 1005
C. 1010
D. 1015
Page: 108 of 193 5/4/2017
K/A:
215002 Rod Block Monitor System
A3 Ability to monitor automatic operations of the ROD BLOCK MONITOR SYSTEM including: (CFR:
41.7 / 45.7)
03 Alarm and indicating lights: BWR-3,4,5
RO/SRO Rating: 3.1/3.1
Tier 2 / Group 2
K/A match: This question satisfies the K/A statement by asking the applicant to identify causes of the rod
out block and rod block monitor trouble given abnormal conditions associated with the
operation of the rod block monitor system.
Pedigree: New
Objective: LOI-CLS-LP-007, Obj. 11a-Describe the possible cause(s) and required operator actions for
the following alarms: A-5 2-2, Rod Out Block (LOCT).
Q Reference: SD-07 rev 11, 1APP-A-05 rev 76, 1APP-A-06 rev 70
Ref provided: None
Cog Level: High
Explanation: An RBM downscale received in conjunction with a non-critical self-test fault will result in
concurrent receipt of the annunciators listed.
Distractor Analysis:
Choice A: Incorrect because - A non-critical self-test fault of the rod block monitor [RBM] channel B
will cause a RBM trouble alarm. Plausible because the applicant my incorrectly link a rod
block to this failure, however a critical self-test fault is required to cause a rod block.
Choice B: Correct, See explanation.
Choice C: Incorrect because- A rod withdrawal block will be generated for a watchdog timer that has
timed out. A non-critical self-test fault of the rod block monitor [RBM] channel B will cause a
RBM trouble alarm. Although conditions exist at this time for receipt of both annunciators, this
is not the earliest time.
Choice D: Incorrect because - While the RBM Channel B critical self-test fault would result in these
conditions this is not the earliest time this has occurred.
SRO Basis:
N/A
Page: 109 of 193 5/4/2017
58. Which one of the following completes both statements below?
The TIP system isolation signal is from the (1) Primary Containment Group
Isolation Logic.
The TIP Shear Valve (2) a Primary Containment Isolation Valve (PCIV).
A. (1) Inboard
(2) is
B. (1) Inboard
(2) is NOT
C. (1) Outboard
(2) is
D. (1) Outboard
(2) is NOT
Page: 110 of 193 5/4/2017
K/A:
223001 Primary Containment System and Auxiliaries
K4 Knowledge of PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES design feature(s) and/or
interlocks which provide for the following: (CFR: 41.7)
03 Containment/drywell isolation
RO/SRO Rating: 3.7/3.8
Tier 2 / Group 2
K/A match: This question matches the K/A because it tests knowledge of PCIS logic and PCIV
designation.
Pedigree: New
Objective: LOI-CLS-LP-09.5 Obj 3b: Describe the relationships between the TIP System and the
following: Primary Containment Isolation System.
Q Reference: System Description SD-09.5 Rev. 7, SD-12 Rev. 11
Ref provided: None
Cog Level: High
Explanation: Traversing In-Core Probes automatically retract from Inboard logic only. The TIP Shear
Valve is a PCIV.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect because - First part is correct. See explanation for choice B. Second part is
incorrect, but plausible. If used, the TIP Shear Valve would be used to isolate the
containment path, however, it is designated a PCIV.
Choice C: First part is incorrect, but plausible. For some group isolations, the inboard and outboard logics
can accomplish separate functions. Second part is correct. See explanation for choice A.
Choice D: First and second parts are incorrect. Both parts are plausible.
SRO Basis: N/A
Page: 111 of 193 5/4/2017
59. Fuel movements are in progress on Unit One with the following signal/conditions:
Reactor MODE switch is in REFUEL
Main hoist is full up
NO hoists are loaded
All rods are NOT full in
BRIDGE NEAR/OVER REACTOR VESSEL (LS1) is actuated
Which one of the following completes both statements below?
ROD BLOCK INTERLOCK #1 (1) actuated.
FUEL HOIST INTERLOCK (2) actuated.
A. (1) is
(2) is
B. (1) is
(2) is NOT
C. (1) is NOT
(2) is
D. (1) is NOT
(2) is NOT
Page: 112 of 193 5/4/2017
K/A:
234000 Fuel Handling Equipment
A1 Ability to predict and/or monitor changes in parameters associated with operating the FUEL
HANDLING EQUIPMENT controls including: (CFR: 41.5 / 45.5)
03 core reactivity level
RO/SRO Rating: 3.4/3.9
Tier 2 / Group 2
K/A match: Given an operationally valid set of plant conditions, the applicant will be able to predict which
fuel handling interlocks that involve core reactivity are, or are not, actuated as indicated by the
appropriate interlock lights LIT or NOT LIT.
Pedigree: New
Objective: LOI-CLS-LP-058.1, Obj. 21a- Given conditions associated with refueling operations, determine
when the following refueling interlock(s) should be in effect:Rod block (LOCT)
Q Reference: SD-58.1, Refuel Bridge/Refueling Interlocks, rev. 5
Ref provided: None
Cog Level: Fundamental
Explanation: Part 1: The Rod Block Interlock is NOT met and therefore the yellow ROD BLOCK
INTERLOCK #1 light is NOT LIT. With the REACTOR MODE switch in REFUEL, with
the Main Grapple Not Full Up or the Any Hoist Loaded signal would also need to be
actuated for the control rod block to be active.
Part 2: The Fuel Hoist Interlock is met and therefore the light is LIT; the two signals that
are required for this interlock are the All Rods Not In and Bridge Over Core; because
both of these signals are present in the stem the interlock is actuated.
Distractor Analysis:
Choice A: Part 1 is incorrect, but plausible because the interlock would be lit if the Reactor MODE
switch was in STARTUP instead of REFUEL, as stated in the stem. Part 2 is correct,
see explanation.
Choice B: Part 1 is incorrect, but plausible because the interlock would be lit if the Reactor MODE
switch was in STARTUP instead of REFUEL, as stated in the stem. Part 2 is incorrect,
but plausible becausemost fuel movement interlocks contain more than two conditions for
the interlock to become activated.
Choice C: Correct, See explanation.
Choice D: Part 1 is correct, see explanation. Part 2 is incorrect, but plausible becausemost fuel
movement interlocks contain more than two conditions for the interlock to become
activated.cause the statement is factual, on its own, however it is not the purpose of the
isolation signal.
SRO Basis:
N/A
Page: 113 of 193 5/4/2017
60. Unit One is in MODE 2 with the following conditions:
HP Turbine Shell Warming is in progress IAW
1OP-26,Turbine System Operating Procedure.
The Chest/Shell Warming potentiometer has
just been rotated as shown.
The Chest/Shell Warming Selector OFF light
has gone out.
IAW 1OP-26, which one of the following
completes both statements below?
Steam (1) being admitted to the HP Turbine.
The Chest/Shell Warming potentiometer must be rotated in the (2) direction to
raise HP Turbine shell pressure.
A. (1) is
(2) clockwise
B. (1) is
(2) counter-clockwise
C. (1) is NOT
(2) clockwise
D. (1) is NOT
(2) counter-clockwise
Page: 114 of 193 5/4/2017
K/A:
241000 Reactor Turbine Pressure Regulating System
A4 Ability to manually operate and/or monitor in thecontrol room:(CFR: 41.7 / 45.5 to 45.8)
18 Turbine shell warming: Plant-Specific
RO/SRO Rating: 2.9/2.8
Tier 2 / Group 2
K/A match: Question meets the K/A because it tests knowledge of the required position of the Chest/Shell
Warming potentiometer for steam to be admitted to the HP turbine shell and the direction of
rotation required to raise HP Turbine shell pressure.
Pedigree: New
Objective: LOI-CLS-LP-026 Obj 16b: State the following as they apply to Turbine Shell Warming:
Required steam pressure to commence Shell Warming.
Q Reference: 1OP-26 Rev. 96 (Turbine System Operating Procedure), Lesson plan LOI-CLS-LP-026
Rev. 8 (Main Turbine, Gland Seal and MSR Systems)
Ref provided: None
Cog Level: High
Explanation: IAW 1OP-26, steam is not admitted to the HP turbine until the Chest/Shell Warming
potentiometer indicates greater than 2.9 units. The Chest/Shell Warming potentiometer is
rotated clockwise to raise HP Turbine shell pressure.
Distractor Analysis:
Choice A: Incorrect because - First part is incorrect and plausible. Second part is correct. See
explanation for choice B.
Choice B: Correct, See explanation..
Choice C: Incorrect because- First and second parts are incorrect. Both parts are plausible. See
explanation for choice B.
Choice D: Incorrect because - First part is correct. Second part is incorrect and plausible. See
explanation for choice B.
SRO Basis: N/A
Page: 115 of 193 5/4/2017
61. Unit Two is at 20% power with main turbine roll in progress IAW 2OP-26, Turbine
System Operating Procedure. Turbine speed is 900 RPM and slowly rising.
The following turbine journal bearing vibration readings are observed on TSI-XR-640:
Bearing #1 5 mils Bearing #6 10 mils
Bearing #2 5 mils Bearing #7 11 mils
Bearing #3 6 mils Bearing #8 13 mils
Bearing #4 7 mils Bearing #9 11 mils
Bearing #5 8 mils Bearing #10 10 mils
Which one of the following:
1) predicts the plant response and
2) identifies required operator action(s), if any, IAW 2OP-26?
A. (1) An automatic turbine trip will occur.
(2) Manually scram the reactor.
B. (1) An automatic turbine trip will occur.
(2) The reactor is NOT required to be manually scrammed.
C. (1) An automatic turbine trip will NOT occur.
(2) Manually scram the reactor and then trip the turbine.
D. (1) An automatic turbine trip will NOT occur.
(2) Manually trip the turbine, the reactor is NOT required to be manually scrammed.
Page: 116 of 193 5/4/2017
K/A:
245000 Main Turbine Generator and Auxiliary Systems
A2 Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND
AUXILIARY SYSTEMS ; and (b) based on those predictions, use procedures to correct, control,
or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
01 Turbine trip
RO/SRO Rating: 3.7/3.9
Tier 2 / Group 2
K/A match: The applicant is expected to understand the impact of high vibrations on the turbine generator
system and understand procedure requirements in order to answer the question.
Pedigree: Last used on the 2012 NRC Exam
Objective: LOI-CLS-LP-026, Obj. 28n - Given plant conditions, predict the effect that the following will
have on the Main Turbine, Gland Seal, and Moisture Reheater System: Main Turbine Hi
Vibration
Q Reference: APP-UA-23 6-1, Turbine Vibration High, Rev. 72, 1OP-26, Turbine System Operating
Procedure, Rev. 96.
Ref provided: None
Cog Level: High
Explanation: TSI is normally disarmed during turbine operation (>23% power with the turbine online), but
is armed for turbine roll. When turbine RPM is between 801 - 1400 RPM, the trip setpoints
are 12 mils for bearings 1-8 and 10 mils for 9 & 10. This requires an immediate turbine trip
per OP-26, Section 5.4.2. If reactor power is greater than 26% then a reactor scram will
occur when the turbine is tripped. Operator actions call for scramming the reactor first, then
tripping the turbine. In the described conditions, a scram is not required since power is
below 26%.
Distractor Analysis:
Choice A: Plausible because if power was greater than 26% then a reactor scram would be required.
Choice B: Correct Answer, see explanation.
Choice C: Plausible because TSI is normally bypassed (but in this case it is not for the startup) and a
turbine trip only is required.
Choice D: Plausible because TSI is normally bypassed (but in this case it is not for the startup) and a
scram is not required at less than 26% power.
SRO Basis: N/A
Page: 117 of 193 5/4/2017
62. Which one of the following completes both statements below?
DC Distribution Panel 10A provides control power for Condensate Pumps (1) .
4160 VAC Bus 2D supplies power to Condensate Pump (2) .
A. (1) 2A and 2C ONLY
(2) 2A
B. (1) 2A and 2C ONLY
(2) 2B
C. (1) 2A, 2B and 2C
(2) 2A
D. (1) 2A, 2B and 2C
(2) 2B
K/A:
256000 Reactor Condensate System
K2 Knowledge of electrical power supplies to the following: (CFR: 41.7)
01 System pumps
RO/SRO Rating: 2.7/2.8
Tier 2 / Group 2
K/A match: The applicant is expected to know the AC and DC power sources for condensate pumps.
Pedigree: New
Objective: LOI-CLS-LP-32 Obj 6a: State the power supplies for the following components: Condensate
Pumps.
Q Reference: System, Rev. 21, 0OI-50, 125/250 and 24/48 VDC Electrical Load List, Rev. 64.
Ref provided: None
Cog Level: Fundamental
Explanation: DC Distribution Panel 10A provides control power for all three condensate pumps. 4 KV bus
2D supplies power to condensate pumps 2A and 2C.
Distractor Analysis:
Choice A: Incorrect because - DC Distribution Panel 10A provides control power for all three
condensate pumps.
Choice B: Incorrect because - DC Distribution Panel 10A provides control power for all three
condensate pumps. 4 KV bus 2D supplies power to condensate pump 2A.
Choice C: Correct, See explanation.
Choice D: Incorrect because - 4 KV bus 2D supplies power to condensate pump 2A.
SRO Basis: N/A
Page: 118 of 193 5/4/2017
63. IAW 1OP-32, Condensate and Feedwater System Operating Procedure, which one of
the following completes the statement below?
Reactor power is limited to a maximum continuous power of (1) with one Reactor
Feed Pump in service to limit (2) .
A. (1) 60%
(2) pump runout
B. (1) 80%
(2) pump runout
C. (1) 60%
(2) turbine exhaust drain casing level
D. (1) 80%
(2) turbine exhaust drain casing level
Page: 119 of 193 5/4/2017
K/A:
259001 Reactor Feedwater System
K5 Knowledge of the operational implications of the following concepts as they apply to REACTOR
FEEDWATER SYSTEM : (CFR: 41.5 / 45.3)
03 Turbine operation: TDRFP's-Only
RO/SRO Rating: 2.8/2.8
Tier 2 / Group 2
K/A match: This question meets the K/A because it tests knowledge of the number of Reactor Feed
Pumps that must be operated in the Condensate and Feedwater System based on Reactor
Feed Pump turbine limitations. The Turbine topic is being met because the reactor feed
pump is a turbine driven pump.
Pedigree: New
Objective: LOI-CLS-LP-302-I Obj 4a: Given plant conditions and any of the following AOPs, determine the
required supplementary actions: AOP-23.0, Condensate/Feedwater System Failure (LOCT).
Q Reference: 1OP-32 Rev. 180 (Condensate and Feedwater System Operating Procedure)
Ref provided: None
Cog Level: Fundamental
Explanation: The Reactor Feed Pumps are designed to operate up to 80% reactor power, but maximum
continuous reactor power is limited to 60% by turbine exhaust drain casing level. See
attached information from 1OP-32, Condensate and Feedwater System Operating
Procedure.
Distractor Analysis:
Choice A: Incorrect because -The first part is correct. See explanation for choice C. The second part
is incorrect but plausible, because pump runout is frequently limiting when discussing the
maximum capacity of a pump
Choice B: Incorrect because- The first part is incorrect but plausible, because the pumps are
designed to operate up to 80% power. The second part is incorrect but plausible, because
pump runout is frequently limiting when discussing the maximum capacity of a pump
Choice C: Correct, See explanation.
Choice D: Incorrect because - The first part is incorrect but plausible, because the pumps are designed
to operate up to 80% power. The second part is correct. See explanation for choice C.
SRO Basis: N/A
Page: 120 of 193 5/4/2017
64. Consider a loss of AC power supply to the Area Radiation Monitors (ARMs) in the
following locations:
Unit One Turbine Building control access corridor
Unit One Feedwater Heater Bay access corridor
Service Building electrical equipment room E/W corridor
Which one of the following completes the statement below?
The ARM local display will indicate (1) which (2) produce a MCR alarm.
A. (1) downscale
(2) will
B. (1) downscale
(2) will NOT
C. (1) upscale
(2) will
D. (1) upscale
(2) will NOT
Page: 121 of 193 5/4/2017
K/A:
272000 Radiation Monitoring System
K3 Knowledge of the effect that a loss or malfunction of the RADIATION MONITORING System will
have on following: (CFR: 41.5 / 45.3)
03 Station area radiation monitoring
RO/SRO Rating: 3.2/3.4
Tier 2 / Group 2
K/A match: The applicant is expected to know the effect that a malfunction of the radiation monitoring
system has on station area radiation monitoring.
Pedigree: New
Objective: LOI-CLS-LP-011.1 Obj 8: Describe the effect that a loss or malfunction of the Area Radiation
Monitoring System will have on plant operations.
Q Reference: APP UA-03 1-7, Annunciator Panel Procedure, Rev. 61, SD-11.1, Area Radiation
Monitoring System, Rev. 7.
Ref provided: None
Cog Level: Fundamental
Explanation: A loss of AC power supply to the Area Radiation Monitors (ARMs) located in the Core spray
pump room 1B, Unit 1 TIP room and Unit 1 spent fuel pool cooling system room will result in
the ARM local display indicating downscale and producing a MCR alarm.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect because - a loss of the AC power supply will result in a MCR alarm.
Choice C: Incorrect because - a loss of the AC power supply will result in the local display indicating
downscale.
Choice D: Incorrect because - a loss of the AC power supply will result in the local display indicating
downscale and a MCR alarm.
SRO Basis: N/A
Page: 122 of 193 5/4/2017
65. Which one of the following identifies the operation of the Diesel Fire Pump starting
system due to an auto start signal?
A. Power from two banks of batteries is used simultaneously to crank the engine.
B. Only one bank of batteries is normally selected, and if the engine does not start
after 15 seconds of cranking, the starting controller automatically selects the other
bank.
C. Only one bank of batteries is normally selected, and if the engine does not start
after 15 seconds of cranking, an operator must locally select the other bank.
D. Only one bank of batteries is physically connected to the starting controller at any
given time. The other bank is an installed spare and would require Electrical
Maintenance to land leads.
Page: 123 of 193 5/4/2017
K/A:
286000 Fire Protection System
K1 Knowledge of the physical connections and/or causeeffect relationships between FIRE
PROTECTION SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
04 D.C. electrical distribution: Plant-Specific
RO/SRO Rating: 2.6/2.6
Tier 2 / Group 2
K/A match: The Diesel Fire Pump is part of the fire protection system. It has two banks of lead-acid
storage batteries which provide DC power to crank the diesel engine. While not part of the
plants DC electrical distribution system per se, it is a DC subsystem that supports fire
protection. And its more complex than simply one battery dedicated to starting the diesel fire
pump; there are two banks of batteries with requisite cabling, and a controller which
dynamically selects which source to use.
Pedigree: New
Objective: LOI-CLS-LP-041 Obj 18: Given plant conditions, predict the response of the Fire Suppression
and Fire Detection Systems. (LOCT).
Q Reference: SD-41, Fire Suppression Systems, Rev. 11
Ref provided: None
Cog Level: Fundamental
Explanation: From SD-41, 3.2.3 Diesel Fire Pump, p. 29: Cranking begins when a signal to run is
established; cranking stops when the engine is running. If the engine fails to start after
approximately 15 seconds of cranking, the controller will interrupt the cranking circuit for
approximately 15 seconds, shift to the opposite battery and crank the engine for
approximately 15 seconds.
Distractor Analysis:
Choice A: Incorrect because the start controller does not use both battery banks simultaneously, but
rather sequentially as discussed above. Plausible if the applicant is not aware of the relays
inside the engine control panel that select which battery bank is used.
Choice B: Correct, See explanation.
Choice C: Incorrect because the switching to the other battery bank is automatic, not manual. Plausible
because when performing a manual start, the operator has to select which battery bank to use
by placing the control switch in Manual A or Manual B. An applicant who is not
knowledgeable of the auto-start functionality might believe that the control switch would have
to be manipulated for a failed auto-start..
Choice D: Incorrect because both battery banks are connected to the control panel. Plausible because
sometimes there are installed spare components that require mechanical and/or electrical
alignments to place in service.
SRO Basis: N/A
Page: 124 of 193 5/4/2017
66. IAW AD-HU-ALL-0004, Procedure and Work Instruction Use and Adherence, which
one of the following completes both statements below concerning Reference Use
procedures?
The procedure (1) required to be at the jobsite.
Signoff steps are required to be signed (2) .
A. (1) is
(2) as frequently as practical
B. (1) is
(2) after each step is performed
C. (1) is NOT
(2) as frequently as practical
D. (1) is NOT
(2) after each step is performed
K/A:
G2.1.20 Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12)
RO/SRO Rating: 4.6/4.6
Tier 3
K/A match: This question requires the applicant to assess if a Reference Use procedure is required at the
jobsite and when the steps are signed off in the procedure.
Pedigree: New
Objective: LOI-CLS-LP-201C Obj 15a: Describe the following as they apply to AD-HU-ALL-0004,
Procedure and Work Instruction Use and Adherence: Responsibility of the procedure user
Q Reference: AD-HU-ALL-0004, Procedure and Work Instruction Use and Adherence, Rev. 5
Ref provided: None
Cog Level: Fundamental
Explanation: Reference Use Procedures are required to be at the jobsite but the steps do not have to be
signed off at the completion of each step.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect because - the first part is correct and the second part is for a Continuous Use
procedure.
Choice C: Incorrect because - the first part is for an Info Use procedure and the second part is correct.
Choice D: Incorrect because - the first part is for an Info Use procedure and the second part is for a
Continuous Use procedure.
SRO Basis: N/A
Page: 125 of 193 5/4/2017
67. Operators are placing RHR Loop B in Standby. Some MOVs in this loop have black
dots on the associated RTGB control switches and are required to be in the closed
position.
Which one of the following identifies the proper sequence for valve position verification
for these MOV's IAW 2OP-17, Residual Heat Removal System Operating Procedure?
A. An AO must be dispatched to check the valve position locally at the valve using
local position indication.
B. The breaker is turned on, the operator checks valve position using red/green
indicating lights, then the breaker is turned off.
C. An AO must be dispatched to check the valve position locally by attempting to
manually close the valve using the valve hand wheel.
D. The breaker is turned on, the operator strokes the valve open, then closed to
ensure proper operation of red/green indicating lights, then the breaker is turned off.
K/A:
G2.1.29 Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.
(CFR: 41.10 / 45.1 / 45.12)
RO/SRO Rating: 4.1/4.0
K/A match: this question tests the knowledge of how to verify a de-energized valve per the procedure.
Pedigree: Bank
Objective: LOI-CLS-LP-201-C
Obj. 15 Describe the following as they apply to AD-HU-ALL-0004, Procedure and Work
Instruction Use and Adherence: a. Responsibility of the procedure user
Q Reference: 2OP-17
Ref provided: None
Cog Level: Fundamental
Explanation: IAW 2OP-17 the breaker is turned on, the valve is verified to be in the correct position and
then the breaker is turned back off. The vlave can be verified using the position indicating
lights (red open / green closed lights)
Distractor Analysis:
Choice A: Incorrect - These valves are in hi rad areas and do not need to be verified locally.
Choice B: Correct, See explanation.
Choice C: Incorrect - These valves are in hi rad areas and do not need to be verified locally.
Choice D: Incorrect - The valves do not have to be cycled in order to determine proper position.
SRO Basis: N/A
Page: 126 of 193 5/4/2017
68. A reactor startup is in progress on Unit Two IAW 0GP-02, Approach to Criticality and
Pressurization of the Reactor.
Reactor power is currently in the Source Range advancing to the POAH
Following a notch rod withdrawal of Control Rod 30-31, the Operator at the Controls
observes a stable reactor period of 30 seconds.
Which one of the following identifies the required action(s) IAW 0GP-02?
A. Fully insert ALL control rods.
B. Re-insert control rod 30-31 to obtain a stable period between 80 to 180 seconds.
C. Commence rod insertion until at least ten control rods have been fully inserted past
the step where the 30 second period was observed.
D. Stop control rod withdrawal and verify period remains stable at 30 seconds or
greater. Control rod insertion is NOT required.
K/A:
G2.1 Conduct of Operations
39 Knowledge of conservative decision making practices. (CFR: 41.10 / 43.5 / 45.12)
RO/SRO Rating: 3.6/4.3
Tier 3
K/A match: Ensures that applicants understand what the conservative procedurally required actions for a
short period would be.
Pedigree: Modified from Previous NRC: Browns Ferry 2015 Question 67
Objective: LOI-CLS-LP-307-A, Obj B03-State the shortest stable reactor period allowed during a reactor
startup in accordance with GP-02. (LOCT)
Q Reference: 0GP-02, rev 110
Ref provided: None
Cog Level: Fundamental
Explanation: 0GP-02 section 6.2 caution requires restoring the period to 80 to 180 seconds if the period
starts approaching 30 seconds.
Distractor Analysis:
Choice A: Plausible for a short period startup. RE may need to evaluate.
Choice B: Correct Answer, see explanation.
Choice C: Plausible because this is the action required if the period approaches 12 seconds.
Choice D: Plausible if the applicant believes this is a safe period.
SRO Basis: N/A
Page: 127 of 193 5/4/2017
69. Unit One is in MODE 4 with the following tests being performed IAW 0GP-01,
PreStartup Checklist.
Section 6.6, Rod Drift Alarm Test
Section 6.7, RMCS Timer Test
Which one of the following completes both statements below?
Section 6.6 requires the operator to manipulate the (1) .
Section 6.7 (2) require the operator to select a control rod.
A. (1) Rod Movement Control Switch to the OUT NOTCH position
(2) does
B. (1) Rod Movement Control Switch to the OUT NOTCH position
(2) does NOT
C. (1) Emergency Rod In/Rod Override Control Switch to the EMERG ROD IN
position
(2) does
D. (1) Emergency Rod In/Rod Override Control Switch to the EMERG ROD IN
position
(2) does NOT
Page: 128 of 193 5/4/2017
K/A:
G2.2.1 Ability to perform pre-startup procedures for the facility, including operating those controls
associated with plant equipment that could affect reactivity. (CFR: 41.5 / 41.10 / 43.5 / 43.6 / 45.1)
RO/SRO Rating: 4.56/4.4
Tier 3
K/A match: The proposed test item tests the applicants ability to operate the RMCS controls (controls
associated with plant equipment that could affect reactivity) in accordance with the pre-startup
procedure.
Pedigree: New
Objective: LOI-CLS-LP-07 Obj 8e: Explain how the Reactor Manual Control System operates to
accomplish the following: Emergency rod insertion
Q Reference: 0GP-01, Section 6.6 and 6.7
Ref provided: None
Cog Level: Fundamental
Explanation: Step 3.a in the Rod Drift Alarm Test requires the operator to notch INSERT a rod using
EMERG IN while the Rod Drift Alarm Test Switch is being held in the TEST position. Since
the stem says that all rods are inserted, there is no need to withdraw the control rod to its
original position. Step 1 of the RMCS Timer Test requires selecting the first rod to be
withdrawn in accordance with GP-10.
Distractor Analysis:
Choice A: Incorrect because Step 3.a in the Rod Drift Alarm Test requires the operator to notch
INSERT a rod using EMERG IN while the Rod Drift Alarm Test Switch is being held in the
TEST position. Plausible because using the EMERG IN switch wont result in any rod
movement since all rods are inserted, and applicant could misconstrue that actual rod
movement is required.
Choice B: Incorrect because Step 3.a in the Rod Drift Alarm Test requires the operator to notch
INSERT a rod using EMERG IN while the Rod Drift Alarm Test Switch is being held in the
TEST position. Also incorrect because the RMCS Timer Test does require selecting a
control rod. Plausible because using the EMERG IN switch wont result in any rod
movement since all rods are inserted, and applicant could misconstrue that actual rod
movement is required. Also plausible because applicants could misconstrue that Timer
Test does not require control rod selection because no actual rod movement is required.
Choice C: Correct, See explanation.
Choice D: Incorrect because the RMCS Timer Test does require selecting a control rod. Plausible
because applicants could misconstrue that Timer Test does not require control rod selection
because no actual rod movement is required.
SRO Basis: N/A
Page: 129 of 193 5/4/2017
70. Which one of the following completes both statements below IAW AD-HU-ALL-0004,
Procedure and Work Instruction Use and Adherence?
During execution of a procedure, if a procedure problem is identified that is an obvious
typographical or editorial error as defined in Attachment 2, Field Editorial Corrections,
then a formal procedure revision (1) required prior to completing the procedure.
Supervisor approval (2) required before continuing execution of the procedure.
A. (1) is
(2) is
B. (1) is
(2) is NOT
C. (1) is NOT
(2) is
D. (1) is NOT
(2) is NOT
Page: 130 of 193 5/4/2017
K/A:
G2.2.1 Knowledge of the process for making changes to procedures. l (CFR: 41.10 / 43.3 / 45.13)
RO/SRO Rating: 3.0/3.6
Tier 3
K/A match: This question requires the applicant to know the process for making changes to procedures in
accordance with AD-HU-ALL-0004.
Pedigree: New
Objective: LOI-CLS-LP-201-C
Obj. 15 Describe the following as they apply to AD-HU-ALL-0004, Procedure and Work
Instruction Use and Adherence: a. Responsibility of the procedure user
Q Reference: AD-HU-ALL-0004, Procedure and Work Instruction Use and Adherence
Ref provided: None
Cog Level: Fundamental
Explanation: Per AD-HU-ALL-0004, QC Hold points may be marked as NA and procedural step
completion out of sequence is not applicable to Information Use procedures. Both actions
relate to making changes to procedures.
Distractor Analysis:
Choice A: Incorrect because - the first part might be thought that all revisions must be processed
and the second part is correct.
Choice B: Incorrect because - the first part might be thought that all revisions must be processed
and the second part might be thought to be true since the revision is minor in nature.
Choice C: Correct, See explanation.
Choice D: Incorrect because - the first part is correct and the second part might be thought to be true
since the revision is minor in nature.
SRO Basis: N/A
Page: 131 of 193 5/4/2017
71. Unit Two is shutdown with crew performing 0GP-06, Cold Shutdown to Refueling
(Head Unbolted).
IAW AD-OP-ALL-0106, Conduct of Infrequently Performed Tests or Evolutions (IPTE),
which one of the following identifies the individual who can exercise oversight of the
evolution for Entry into Lowered Inventory conditions on Unit Two?
A. Refuel Floor Supervisor
B. Health Physics Supervisor
C. Shift Manager
D. Maintenance Manager
Page: 132 of 193 5/4/2017
K/A:
G2.2.07 Knowledge of the process for conducting special or infrequent tests. (CFR: 41.10 / 43.3 / 45.13)
RO/SRO Rating: 2.9/3.6
Tier 3
K/A match: This question requires the applicant to determine which of the listed individuals can perform
duties as an IPTE Manager during Lowered Inventory conditions. This question applies to
AD-OP-ALL-0106, Conduct of Infrequently Performed Tests or Evolutions.
Pedigree: New
Objective:
Q Reference: AD-OP-ALL-0106, Conduct of Infrequently Performed Tests or Evolutions, Rev. 2,
0GP-06, Cold Shutdown to Refueling (Head Unbolted), Rev. 42
Ref provided: None
Cog Level: Fundamental
Explanation: An IPTE Manager is a line manager, senior to the Shift Manager, responsible for oversight of
a specific IPTE. IPTE Managers may include the Plant Manager, Department Managers, or
other personnel senior to the Shift Manager who have been designated by the Plant
Manager.
Distractor Analysis:
Choice A: Incorrect because - the Refuel Floor Supervisor is not senior to the Shift Manager. While
this supervisor may be involved in the Lowered Inventory conditions evolution, they could
not be designated as the IPTE Manager.
Choice B: Incorrect because - the Health Physics Supervisor is not senior to the Shift Manager.
While this supervisor may be involved in the Lowered Inventory conditions evolution, they
could not be designated as the IPTE Manager.
Choice C: Incorrect because - the Shift Manager is specifically excluded from participation as an IPTE
Manager. While this supervisor would be involved in the Lowered Inventory conditions
evolution, they could not be designated as the IPTE Manager.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 133 of 193 5/4/2017
72. Which of the following radiation monitoring systems utilize ion chamber detectors?
1. Main Steam Line Radiation Monitoring System
2. Condenser Off-Gas Radiation Monitoring System
3. Main Stack Radiation Monitoring System
A. 1 ONLY
B. 1 and 2 ONLY
C. 2 and 3 ONLY
D. 1, 2 and 3
Page: 134 of 193 5/4/2017
K/A:
G2.2.1 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable
survey instruments, personnel monitoring equipment, etc. (CFR: 41.12 / 43.4 / 45.9)
RO/SRO Rating: 2.9/3.1
Tier 3
K/A match: This question matches the K/A because it tests the applicants knowledge of the type of
radiation monitoring systems as it applies to detector types used in plant operations.
Pedigree: New
Objective: LOI-CLS-LP-011 Obj 4a: Describe the interrelationship between the Process Radiation
Monitoring System and the following systems: Main Steam System.
Q Reference: SD-11.0 Rev. 10
Ref provided: None
Cog Level: Fundamental
Explanation: Each of the four Instrument Channels (A, B, C, and D) of the Main Steam Line Radiation
Monitoring System consists of an ion chamber detector. All three channels of the
Condenser Off-Gas Monitoring System uses an ion chamber detector.
Distractor Analysis:
Choice A: Incorrect because - The Condenser Off-Gas Radiation Monitoring System also consists
of an ion chamber detector. Plausible because the plant utilizes 3 types of radiation
detectors to monitor for gamma radiation (Geiger-Mueller detector, an ion chamber
detector, and a scintillation detector).
Choice B: Correct, See explanation.
Choice C: Incorrect because- The Main Stack Radiation Monitoring System utilizes scintillation detectors.
Plausible because the plant utilizes 3 types of radiation detectors to monitor for gamma
radiation (Geiger-Mueller detector, an ion chamber detector, and a scintillation detector).
Choice D: Incorrect because - The Main Stack Radiation Monitoring System utilizes scintillation
detectors. Plausible because the plant utilizes 3 types of radiation detectors to monitor for
gamma radiation (Geiger-Mueller detector, an ion chamber detector, and a scintillation
detector).
SRO Basis: N/A
Page: 135 of 193 5/4/2017
73. You have been directed to open a valve in the RCA for the performance of a hotspot
flush.
The valve is located 6 feet above the RCA floor.
It will initiate flow that will flush a 1 R/hr hotspot through a pipe in an adjacent area.
Radiation levels in the valve area are 20 mR/hour.
The valve area is NOT posted as an Alpha or Airborne Radioactivity Area.
Beta/Gamma contamination levels in the valve area are 2500 dpm/100 cm2.
IAW PD-RP-ALL-0001, Radiation Worker Responsibilities, which one of the following
completes the statement below?
Use of the Self-Briefing process is NOT allowed because:
A. the height of the valve is too high.
B. of the flushing operation being performed.
C. radiation levels in the valve area are too high.
D. contamination levels in the valve area are too high.
Page: 136 of 193 5/4/2017
K/A:
G2.3.07 Ability to comply with radiation work permit requirements during normal or abnormal
conditions.(CFR: 41.12 / 45.10)
RO/SRO Rating: 3.5/3.6
Tier 3
K/A match: This question evaluates the knowledge required to be applied in order to determine the type
of brief that is required to meet Radiological Work Permit requirements.
Pedigree: New
Objective:
Q Reference: PD-RP-ALL-0001 (Radiation Worker Responsibilities)
Ref provided: None
Cog Level: Comprehension
Explanation: Section 5.4.3.9 of PD-RP-ALL-0001states, RP may permit Radworkers to use the
Self-Briefing process with the following radiological conditions and activity restrictions:
Section 5.4.3.9.b of PD-RP-ALL-0001states, in part, The following Activity restrictions which
may change radiological conditions apply to worker Self-Briefings:
- No system operation, such as opening a valve that may allow radioactive material to
transfer through a pipe or cause elevated dose rates in the work area or an adjacent
area.
Distractor Analysis:
Choice A: Incorrect because- Section 5.4.3.9 of PD-RP-ALL-0001states, RP may permit
Radworkers to use the Self-Briefing process with the following radiological conditions and
activity restrictions: Section 5.4.3.9.b of PD-RP-ALL-0001states, in part, The following
Activity restrictions which may change radiological conditions apply to worker
Self-Briefings: No use of ladders, scaffolds, or man-lift equipment to access overhead
areas above 7 feet.
Choice B: Correct, See explanation.
Choice C: Incorrect because- Section 5.4.3.9 of PD-RP-ALL-0001states, RP may permit Radworkers to
use the Self-Briefing process with the following radiological conditions and activity restrictions:
Section 5.4.3.9.a of PD-RP-ALL-0001states, in part, The following Area restrictions apply to
worker Self-Briefings: (2) No entry or work in: Radiation Areas greater than 25 mrem/hr
Choice D: Incorrect because- Section 5.4.3.9 of PD-RP-ALL-0001states, RP may permit Radworkers to
use the Self-Briefing process with the following radiological conditions and activity restrictions:
Section 5.4.3.9.a of PD-RP-ALL-0001states, in part, The following Area restrictions apply to
worker Self-Briefings: (2) No entry or work in: Contaminated Areas greater than 10,000
dpm/100 cm2
SRO Basis: N/A
Page: 137 of 193 5/4/2017
74. Unit One was operating at 100% power when inadvertent MSIV closure results in RPV
pressure peaking at 1135 psig.
Subsequently:
The reactor failed to scram.
The reactor mode switch is in SHUTDOWN.
ARI has been initiated.
Reactor power is indicating 4%.
Which one of the following completes both statements below?
(1) Safety Relief Valves have opened automatically.
The operating crew (2) required to trip the recirculation pumps
A. (1) Four
(2) is
B. (1) Four
(2) is NOT
C. (1) Eight
(2) is
D. (1) Eight
(2) is NOT
K/A:
G2.4.11 Knowledge of abnormal condition procedures. (CFR: 41.10 / 43.5 / 45.13)
RO/SRO Rating: 4.0/4.2
Tier 3
K/A match: The applicant is expected to apply knowledge of abnormal condition procedures
(1EOP-01-ATWS) to determine the number of SRVs that auto opened and the criteria for
tripping recirculation pumps.
Pedigree: Modified from 2015 NRC Exam
Objective: LOI-CLS-LP-020, Obj. 9
LOI-CLS-LP-003, Obj. 24
Q Reference: 0OI-37.4, Reactor Vessel Control Procedure Basis Document, Rev 11,
0OI-37.5, ATWS Basis Document, Rev. 15,
SD-20, Automatic Depressurization System (ADS), Rev. 3,
SD-03, Reactor Protection System Including Alternate Rod Injection System, Rev. 12.
Ref provided: None
Cog Level: High
Explanation: Scram should have occurred at 1060 psig. SRV opening is 4@1130 psig, 4@1140 psig and
3@1150 psig; therefore 4 SRVs are open. Recirculation Pump trips are required based on
exceeding 2% RTP during the ATWS.
Distractor Analysis:
Page: 138 of 193 5/4/2017
y
Choice A: Correct, See explanation.
Choice B: Incorrect because - although Recirculation pumps are not required to be tripped based on
pressure, they are required to be tripped based on exceeding 2% RTP during the ATWS.
Choice C: Incorrect because - SRV opening is 4@1130 psig, 4@1140 psig and 3@1150 psig; therefore
4 SRVs are open.
Choice D: Incorrect because - SRV opening is 4@1130 psig, 4@1140 psig and 3@1150 psig; therefore
4 SRVs are open. Recirculation Pump trips are required based on exceeding 2% RTP during
the ATWS.
SRO Basis: N/A
Page: 139 of 193 5/4/2017
75. A fire has been confirmed in the Unit Two Cable Spread Room.
Which one of the following completes both statements below?
The Unit Two Cable Spread Room has an installed (1) fire suppression system.
IAW 0PFP-013, General Fire Plan, if the above referenced fire suppression system is
not effective in extinguishing the fire, the Fire Brigade (2) permitted to use a
firehose to attack the fire.
A. (1) carbon dioxide
(2) is
B. (1) carbon dioxide
(2) is NOT
C. (1) water sprinkler
(2) is
D. (1) water sprinkler
(2) is NOT
Page: 140 of 193 5/4/2017
K/A:
G2.4 Emergency Procedures / Plan
26 Knowledge of facility protection requirements, including fire brigade and portable fire fighting
equipment usage. (CFR: 41.10 / 43.5 / 45.12)
RO/SRO Rating: 3.1/3.6
Tier 3
K/A match: Candidate must have knowledge of an installed fire suppression system, and a requirement
from the General Fire Plan for applying water to potentially-energized electrical equipment.
Pedigree: New
Objective: LOI-CLS-LP-041, Obj 7a - Identify the areas/components that are provided with the following
types of fire protection: Carbon Dioxide
Q Reference: 0PFP-013, General Fire Plan, Rev. 48; 0PFP-CB, Control Building Pre-Fire Plans, Rev. 13
Ref provided: None
Cog Level: Fundamental
Explanation: Part 1: The Unit 2 Cable Spread Room has an installed water sprinkler system. Part 2: The
General Fire Plan allows firewater to be put on an electrical fire if certain conditions are met.
For <10kV, as is the case in the Cable Spread Room, a fog nozzle must be used, and a
minimum of 10 of distance. The Fire Preplan specifies the fog nozzle and a distant attack.
Distractor Analysis:
Choice A: Part 1 is plausible becausethe HPCI Pump Room has a CO2 system. Also, a candidate might
reason that you wouldnt want a water sprinkler system in a room with a lot of energized
cabling and equipment (many plants have CO2 or Halon systems in these areas). Part 2 is
plausible it is correct, see explanation.
Choice B: Part 1 is plausible becausethe HPCI Pump Room has a CO2 system. Also, a candidate might
reason that you wouldnt want a water sprinkler system in a room with a lot of energized
cabling and equipment (many plants have CO2 or Halon systems in these areas). Part 2 is
plausible because it is generally believed that water and electricity dont mix, so a candidate
might reason that it would be a personnel hazard if the Fire Brigade were to apply water to the
fire. Additionally, the presence of ABC extinguishers as well as wheeled dry chemical
extinguisher in the area provides a counter-balance if the applicant believes that a fire hose
cannot be used.
Choice C: Correct Answer, see explanation.
Choice D: Part 1 is plausible it is correct, see explanation. Part 2 is plausible because it is generally
believed that water and electricity dont mix, so a candidate might reason that it would be a
personnel hazard if the Fire Brigade were to apply water to the fire. Additionally, the presence
of ABC extinguishers as well as wheeled dry chemical extinguisher in the area provides a
counter-balance if the applicant believes that a fire hose cannot be used.
SRO Basis: N/A
Page: 141 of 193 5/4/2017
76. Which one of the following completes the statement below IAW 0OI-37.4, Reactor
Vessel Control Procedure Basis Document?
0EOP-01-SEP-11, Alternative Source Term Actions, is first required if RPV level drops
to (1) which is associated with a fuel temperature of (2) .
A. (1) LL4
(2) 1500°F
B. (1) LL4
(2) 1800°F
C. (1) LL5
(2) 1500°F
D. (1) LL5
(2) 1800°F
Page: 142 of 193 5/4/2017
K/A:
295031 Reactor Low Water Level
EA2 Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER
LEVEL: (CFR: 41.10 / 43.5 / 45.13)
04 Adequate core cooling
RO/SRO Rating: 4.6/4.8
K/A match: The K/A is met at the SRO-only level because procedure basis knowledge of actions with
regard to reactor low water level and adequate core cooling is required to answer the
question. This knowledge is above the major mitigating strategy of the procedure.
Pedigree: New
Objective: LOI-CLS-LP-300D
Obj. 11 Explain the basis for LL4, Minimum Steam Cooling Reactor Water Level.
Q Reference: 0OI-37.4 R11
Ref provided: None
Cog Level: Fundamental
Explanation: Per the RC/L-3 third override, Alternative Source Term (AST) actions are required if RPV
level drops to LL-4. LL-4 is used as the initiation point for the Alternative Source Term
analyses actions since the fuel temperature (1500°F) associated with this RPV level is just
below that at which cladding perforation is initiated (1600°F based on document
EAS-62-1088, Brunswick Steam Electric Plant Loss-of-Coolant Accident Engineering
Analysis Description).
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect because - Second part is incorrect. Second part is plausible because it is the
temperature that steam cooling will maintain.
Choice C: Incorrect - The first part is plausible because RVCP has actions for LL5. The second part is
correct.
Choice D: Incorrect - The first part is plausible because RVCP has actions for LL5. Second part is
plausible because it is the temperature that steam cooling will maintain.
SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal,
abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
Page: 143 of 193 5/4/2017
77. Unit One is in MODE 4 with the following initial conditions:
RHR Pump A running in the Shutdown Cooling Mode
Reactor average temperature 174°F
Reactor vessel level 187 inches
Current conditions:
A RHR Pump tripped and neither pump in that loop can be restarted.
A-3 (3-9), RHR A/B DISCH & SUCT HDR PRESS HI, has just alarmed
IAW 0AOP-15.0, Loss of Shutdown Cooling, which one of the following completes both
statements below?
A PCIS Group 8 isolation signal (1) occurred.
The CRS will first direct (2) .
A. (1) has
(2) a reduction of reactor pressure IAW 0GP-05, Unit Shutdown
B. (1) has
(2) shifting of shutdown cooling loops IAW 1OP-17, Residual Heat Removal
System Operating Procedure
C. (1) has NOT
(2) a reduction of reactor pressure IAW 0GP-05, Unit Shutdown
D. (1) has NOT
(2) shifting of shutdown cooling loops IAW 1OP-17, Residual Heat Removal
System Operating Procedure
Page: 144 of 193 5/4/2017
K/A:
295021 Loss of Shutdown Cooling
G2.1.7 Ability to evaluate plant performance and make operational judgments based on operating
characteristics, reactor behavior, and instrument interpretation.(CFR: 41.5 / 43.5 / 45.12 / 45.13)
RO/SRO Rating: 4.4/4.7
Tier 1 / Group 1
K/A match: Candidate must know which procedure to select given plant conditions. This also makes it an
SRO-level question.
Pedigree: New
Objective: LOI-CLS-LP-017, Obj 14-Given plant conditions, determine if a Shutdown Cooling isolation
should have occurred. (LOCT)
Q Reference: 0AOP-15.0 rev 31, 1APP-A-03 rev 58
Ref provided: None
Cog Level: High
Explanation: A PCIS Group 8 isolation signal does not occur at the setpoint of the APP-A-03 3-9 [RHR
A/B DISCH & SUCT HDR PRESS HI] alarm, pressure would have to rise to 130.8 psig for
the isolation. Without a group 8 isolation signal 0AOP-15.0 [Loss of Shutdown Cooling
directs the operator to shift shutdown cooling loops in accordance with 1OP-17.
Distractor Analysis:
Choice A: Incorrect because - A PCIS Group 8 isolation signal does not occur at the setpoint of the
APPA-03 3-9 [RHR A/B DISCH & SUCT HDR PRESS HI] alarm, this is plausible as the
applicant could incorrectly believe that the alarm is at the isolation setpoint. Reducing reactor
pressure in accordance with 0GP-05 [Unit Shutdown] is not required to restore shutdown
cooling because a PCIS 8 isolation did not occur, this is plausible because if a PCIS 8 isolation
were to occur thiswould be the correct procedure selection for the condition.
Choice B: Incorrect because- Reducing reactor pressure in accordance with 0GP-05 [Unit Shutdown] is
not required to restore shutdown cooling because a PCIS 8 isolation did not occur, this is
plausible because if a PCIS 8 isolation were to occur this would be the correct procedure
selection for the conditionuld be correct depending on given plant conditions and nature of the
affected casualty on DG2.
Choice C: Incorrect because - A PCIS Group 8 isolation signal does not occur at the setpoint of the
APPA- 03 3-9 [RHR A/B DISCH & SUCT HDR PRESS HI] alarm, this is plausible as the
applicant could incorrectly believe that the alarm is at the isolation setpoint.
Choice D: Correct Answer, see explanation.
SRO Basis: Candidate must know which procedure to select given plant conditions. This also makes it an
SRO-level question. Assessment of facility conditions and selection of appropriate
procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
Page: 145 of 193 5/4/2017
78. Both units are operating at 100% power with the following conditions on Unit One:
A 61 BTH Dry Shielded Canister (DSC) containing 60 spent fuel assemblies, is
inside a Transfer Cask. This Transfer Cask is in the Unit One Reactor Building
Airlock with both airlock doors closed.
The dose rate at the top of the Transfer Cask was discovered to be 180 mr/hr at 30
cm. Due to a human performance error, the water shield (located between the inner
and outer shell of the Transfer Cask) was only partially filled.
As a corrective action, the Transfer Cask water shield was filled with demineralized
water. When the dose rate was subsequently measured, the top of the Transfer
Cask had lowered to ~2 mr/hr.
The crew referred to 0AOP-05, Radioactive Spills, High Radiation, and Airborne
Activity, and 0AOP-41.0, Independent Fuel Storage Installation Abnormal Events.
No emergency declaration was required.
[Reference Provided]
IAW 0OI-01.07, Notifications, which one of the following identifies the latest time that
the NRC is required to be notified?
A. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
B. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
C. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
D. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
Page: 146 of 193 5/4/2017
K/A:
295023 Refueling Accidents
AA2 Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS:
(CFR: 41.10 / 43.5 / 45.13)
04 Occurrence of fuel handling accident
RO/SRO Rating: 3.4/4.1
Tier 1 / Group1
K/A match: The 10CFR55.43(b)(5) SRO topic (Assessment of facility conditions and selection of
appropriate procedures during normal, abnormal, and emergency situations) is being met
because the applicants are required to interpret the ISFSI event and determine the required
notification per 0OI-1.07. The K/A is being met because the ISFSI event is, in a sense, a
refueling accident and/or a fuel handling accident.
Pedigree: BNP 2014 Q59
Objective: LOI-CLS-LP-201-D1 Obj 6: Given plant conditions and an event, determine any applicable
reporting requirements per 0OI- 01.07, Notifications. (LOCT)
Q Reference: 0OI-1.07, SD-65, AOP-41.0
Ref provided: 0OI-1.07, Attachment 1 and Transnuclear Technical Specifications (Attachment A,
Amendment 10) pages A-136 through A141 (TN TS 1.2.11 through 1.2.11e)
Cog Level: High
Explanation: The transfer cask (which is classified as equipment important to safety) suffered an event
that disabled the transfer casks water shield, which is used to prevent exposures that could
exceed regulatory limits. Item #4.2 would be invoked per 0OI-01.07
Distractor Analysis:
Choice A: Incorrect because event did not involve criticality or the loss of special nuclear material.
Plausible because there is a one hour reportability Item #1.3 associated with ISFSI, and it
could be misconstrued to mean that the fill water was lost. Also plausible because it could
be misconstrued as a radiation exposure greater than 100 mr/hr when transported outside
the protected area.
Choice B: Incorrect because event did not involve an emergency where the Tech Specs or license
conditions were departed from. Plausible because there is an ISFSI 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report for
action Item #2.6 for ISFSI, and because it could be misconstrued to mean that the Tech
Specs for the transfer cask were not met. Also plausible because it could be misconstrued
as a radiation exposure greater than 100 mr/hr when transported outside the protected
area.
Choice C: Incorrect because - the event did not involve a transfer cask defect (occurred due to human
error), and because the event did not involve a reduction in effectiveness of the ISFSI
confinement system. Plausible because there is an ISFSI 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report for an ISFSI safety
defect (Item #3.7), and another 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report for ISFSI reduction in effectiveness of a cask
confinement system (Item 3.8). Also plausible because it could be misconstrued as a radiation
exposure greater than 100 mr/hr when transported outside the protected area.
Choice D: Correct, See explanation.
SRO Basis: N/A
Page: 147 of 193 5/4/2017
79. Unit Two was at 100% power when an ATWS occurred with the following plant
conditions:
Reactor power 2%
Control Rods Being inserted with RMCS
Torus water temperature 109.5°F and rising
SLC Pumps A and B both tripped
Which one of the following completes both statements below?
IAW LEP-03, Alternate Boron Injection, SLC injection will first be attempted
using (1) .
In using the system identified above, it will take (2) to achieve hot shutdown boron
weight as compared to injection with the SLC System.
A. (1) Section 2.1, CRD SLC Tank Injection
(2) shorter
B. (1) Section 2.1, CRD SLC Tank Injection
(2) longer
C. (1) Section 2.2, HPCI/RCIC SLC Tank Injection
(2) shorter
D. (1) Section 2.2, HPCI/RCIC SLC Tank Injection
(2) longer
Page: 148 of 193 5/4/2017
K/A:
295026 Suppression Pool High Water Temperature
G2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational
effects. (CFR: 41.10 / 43.5 / 45.13)
RO/SRO Rating: 3.8/4.0
K/A match: The question requires the applicant to indicate where the AO will be aligning locally to inject
boron due to high torus water temperature. Then they will determine the operational effect of
the alternate flowpath vs the normal flowpath.
Pedigree: New
Objective: LOI-CLS-LP-300J
Obj. 10 Given plant condtions determine which method of Alternate Boron Injection is
appropriate in accordance with EOP-01-LEP-03.
Q Reference: 0EOP-01-LEP-03 R30, ATWS Procedure R1
Ref provided: None
Cog Level: High
Explanation: SLC injection occurs in the RC/Q leg of the ATWS procedure in advance of Torus Water
Temperature exceeding 110°F (when power determined to be below 2%). With the given
conditions, The first question relies on applicant procedural knowledge to specify where
local AO activities will first be attempted to inject the contents of the SLC tank. Per
2EOP-01-ATWS, LEP-03 is to be used to place CRD in service prior to using HPCI/RCIC.
The second question tests the operators knowledge of the flowrate of the flowpath. The
normal flowrate is ~80 gpm while the CRD system flowrate is ~11 gpm which will require a
longer period of time to achieve HSBW.
Distractor Analysis:
Choice A: Incorrect - The first part is correct. The second part is plausible because the CRD System
flowrate is larger than the SLC system flowrate.
Choice B: Correct, See explanation.
Choice C: Incorrect - The first part is plausible because under different conditions this would be the
preferred flowpath. The second part is plausible because the CRD System flowrate is larger
than the SLC system flowrate.
Choice D: Incorrect - The first part is plausible because under different conditions this would be the
preferred flowpath. The second part is correct.
SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal,
abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
Page: 149 of 193 5/4/2017
80. A small break LOCA occurred on Unit Two.
The following conditions currently exist:
Drywell temperature 270°F
Drywell pressure 5.0 psig
Torus pressure 2.8 psig
Torus level 5 inches
Reactor pressure 395 psig
Containment Hydrogen Monitors CAC-AT-4409 & 4410 are not available.
Chemistry has been called but has not yet sampled the drywell.
[Reference Provided]
Which one of the following procedures provides the required actions that mitigate these
plant conditions?
A. 0EOP-01-SEP-02, Drywell Spray Procedure
B. 0EOP-01-SEP-03, Torus Spray Procedure
C. 0EOP-01-SEP-05, Primary Containment Purging
D. 0EOP-01-SEP-10, Circuit Alteration Procedure (Section 2.1, Defeating Drywell
Cooler LOCA Lockout)
Page: 150 of 193 5/4/2017
K/A:
295028 High Drywell Temperature
G2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant
operation. (CFR: 41.10 / 43.5 / 45.2 / 45.6)
RO/SRO Rating: 4.3/4.4
Tier 1 / Group1
K/A match: Candidate must know which procedure to select given plant conditions.
Pedigree: BNP 2007-301 Q80
Objective: LOI-CLS-LP-300-L Obj 11: Given the Primary Containment Control Procedure, which steps
have been completed and plant parameters, determine the required operator actions. (LOCT)
Q Reference: OEOP-01-SEP-03, Torus Spray Procedure, rev. 12
Ref provided: 0EOP-01-UG Attachment 5, Drywell Spray Initiation Limit graph
Cog Level: High
Explanation: For the given conditions, Torus Spray will be used to mitigate the plant conditions. Drywell
spray in this condition is in the UNSAFE region of the Drywell Spray Initiation Curve so the
drywell cannot be sprayed. The drywell cannot be purged because there is no indication of
H2 at this time. The drywell coolers cannot be restarted with drywell LOCA conditions. The
drywell is still below its design temperatures.
Distractor Analysis:
Choice A: Incorrect because drywell conditions fall in the UNSAFE region of the Drywell Spray
Initiation Curve, so the drywell cannot be sprayed. Plausible because 5 psig and 270F is in
the Unsafe region on the left side, but just barely. A candidate could misread the DSIL
graph, or fail to refer to it and just assume its safe to spray.
Choice B: Correct, See explanation.
Choice C: Incorrect because drywell cannot be purged with containment sample results unavailable at
this time. Plausible because it would be a correct thing to do, and a candidate might forget the
requirement in SEP-05 to Confirm sample results allow purging.
Choice D: Incorrect because drywell coolers can only be used if LOCA conditions do NOT exist in the
drywell, but here drywell pressure is above 1.7 psig concurrent with low reactor pressure.
Plausible if candidate forgets the SEP-10 requirement that LOCA conditions NOT exist in
containment.
SRO Basis: N/A
Page: 151 of 193 5/4/2017
81. A release is occurring with the following plant conditions:
1-D12-RR-4548-4, TB Vent Effluent Rad Level, is reading 2.195 E+6 Ci/sec
2-D12-RR-4548-4, TB Vent Effluent Rad Level, is reading 1.463 E+4 Ci/sec
2-D12-RR-4548-3-2-1, TB Vent Effluent Rad Level, is reading 4.025 E+0 Ci/cc
2-VA-FT-3358, TB Air Flow, is INOPERABLE
2-VA-FT-7554, TB Total Flow, is reading 7,400 scfm
U2 Turbine Building Ventilation System is in recirc mode
[Reference Provided]
Which one of the following identifies the highest emergency action level (EAL)
classification that is required for this event?
A. Unusual Event
B. Alert
C. Site Area Emergency
D. General Emergency
K/A:
295038 High Off-Site Release Rate
EA2 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE
RATE : (CFR: 41.10 / 43.5 / 45.13)
03 Radiation levels
RO/SRO Rating: 3.5/4.3
Tier 1 / Group 1
K/A match: The candidate is required to evaluate data for a given radiation release and reference
0PEP-03.6.1 [RELEASE ESTIMATES BASED UPON STACK/VENT READINGS] for source
term calculation to then compare those to the EALs for rad effluent. Based on this
comparison the candidate must make the correct EAL designation.
Pedigree: Modified from 2014 NRC Exam
Objective: LOI-CLS-LP-301-A, Obj. 6
Q Reference: 0PEP-03.6.1, Rev 16 0PEP-02.1 Rev. 53
Ref Provided: 0PEP-03.6.1, 0PEP-02.1
Cog Level: High
Explanation: Per Attachment 3 the estimated release is calculated as follows:
Monitor reading (Ci/cc) X Flow (95,500) X Conversion factor (472)
4.025 E+0 X 15,000 X 472 = 2.94 E+7 Ci/sec
The threshold for a General Emergency is a value greater than 1.07 E+8 Ci/sec.
Distractor Analysis:
Choice A: Incorrect because - while the value for 2-D12-RR-4548-4 exceeds the threshold for an
Unusual Event, this is not the highest EAL classification given the conditions, plausible
because this is the EAL classification for 2-D12-RR-4548-4.
Page: 152 of 193 5/4/2017
Choice B: Incorrect because- while the value for 1-D12-RR-4548-4 exceeds the threshold for an
Alert, this is not the highest EAL classification given the conditions, plausible because this
is the EAL classification for 1-D12-RR-4548-4 if the applicant fails to evaluate PEP-03.6.1
Choice C: Correct, See explanation.
Choice D: Incorrect because - this is not the highest EAL classification given the conditions, plausible
because if 95000 scfm (valuefor once thru) is used the estimated release calculation will
exceed the General Emergency.
SRO Basis: Radiation hazards that may arise during normal and abnormal situations, including
maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)]
Question from 2014 NRC Exam
Page: 153 of 193 5/4/2017
82. Unit One and Unit Two were both operating at 100% power when a large fire in the
north section of the Unit Two Reactor Building occurs.
Current Conditions:
Both units remain at 100% power
Unit Two CRS determines that entry into the ASSD procedures is required
NOTE:
0ASSD-01, Alternative Safe Shutdown Procedure Index
2ASSD-05, Reactor Building North
Which one of the following completes both statements below?
Once procedure 0ASSD-01 is entered, the Unit Two CRS is required to (1) .
The Unit Two CRS will enter 2ASSD-05 and (2) 0ASSD-01.
A. (1) always MANUALLY SCRAM Unit Two reactor if power is greater than 2%
(2) exit
B. (1) always MANUALLY SCRAM Unit Two reactor if power is greater than 2%
(2) concurrently perform
C. (1) make a determination if Unit Two shutdown is required, or not, based on an
assessment of the situation
(2) exit
D. (1) make a determination if Unit Two shutdown is required, or not, based on an
assessment of the situation
(2) concurrently perform
K/A:
600000 Plant Fire On Site
AA2 Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE:
(CFR: 41.10 / 43.5 / 45.13)
13 Need for emergency plant shutdown
RO/SRO Rating: 3.2/3.8
Tier 1 / Group 1
K/A match: Given an operationally valid set of plant conditions involving an onsite fire which requires
entry into the alternative safe shutdown (ASSD) procedure set, the SRO applicant will need to
demonstrate knowledge on the SRO-only level of the procedural requirements of 0ASSD-01,
Alternative Safe Shutdown Procedure Index. Both parts of this 2x2 question involve
SRO-only knowledge; the first part question is the specific match to the K/A language of
determining and interpreting a need for emergency plant shutdown, which is a job function
specific to the SRO at Brunswick; and the second part question is also related to the K/A.
Pedigree: New
Objective: LOI-CLS-LP-304, Obj. 12
Q Reference: 0ASSD-01 Rev 41
Ref Provided: None
Page: 154 of 193 5/4/2017
Cog Level: High
Explanation: Under section 3.0, OPERATOR ACTIONS, of procedure 0ASSD-01, step 3.3 states: IF
the fire is in an ASSD fire area on either unit, THEN the Unit CRS will assess the situation
considering the following: [and then provides a list of bulleted criteria/considerations].
Because the given conditions of the question postulate a fire in Unit 2 Reactor Building
North, the fire is in an ASSD location and step 3.3 is required to be implemented. Later in the
procedure, step 3.5 states: IF the Unit CRS determines that reactor shutdown is required
based on Step 3.3... , and then gives various criteria and direction to manually SCRAM the
reactor. Based on these two steps, C(1) and D(1) are correct, in that 0ASSD-01 requires the
Unit [2, in this case] CRS make a determination if Unit 2 shutdown is required, or not, based
on an assessment of the situation per the criteria and considerations given in step 3.3.
Based on the location of the ASSD fire (U2 Reactor Building), the operators will continue to
step 3.5.3, which states: IF the fire is NOT in the Control Building, THEN ENTER the
applicable ASSD procedure AND EXECUTE concurrently with this procedure. Accordingly,
answers B(2) and D(2) are correct answers to the second part question
Distractor Analysis:
Choice A: Incorrect because - First part question is incorrect: although a manual SCRAM is very
likely, it is NOT always required by procedure 0ASSD-01; per the discussion above, given
the conditions listed in the question, what is always required is that the Unit CRS perform
an assessment of the prevailing plant conditions to determine if an emergency unit
shutdown is required. Second part question is also incorrect: 0ASSD-01 directs operators
to perform [2ASSD-05] concurrently.
Plausible because - First part question is plausible because 0ASSD-01 always requires a
manual SCRAM if the Unit CRS determines that the ability to confirm reactor power less
than 2% is in jeopardy. Second part question is plausible because 0ASSD-01 directs the
operator to EXIT this procedure [0ASSD-01] AND ENTER 0ASSD-02, Control Building, if
the fire is in the Control Building fire area and control room evacuation is required (step
3.5.2.e).
Choice B: Incorrect because - First part question is incorrect: although a manual SCRAM is very
likely, it is NOT always required by procedure 0ASSD-01; per the discussion above, given
the conditions listed in the question, what is always required is that the Unit CRS perform
an assessment of the prevailing plant conditions to determine if an emergency unit
shutdown is required. Second part distractor is correct.
Plausible because - First part question is plausible because 0ASSD-01 always requires a
manual SCRAM if the Unit CRS determines that the ability to confirm reactor power less
than 2% is in jeopardy
Choice C: Incorrect because - First part question is correct. Second part question is incorrect: 0ASSD-01
directs operators to perform [2ASSD-05] concurrently.
Plausible because - Second part question is plausible because 0ASSD-01 directs the operator
to EXIT this procedure [0ASSD-01] AND ENTER 0ASSD-02, Control Building, if the fire is in
the Control Building fire area and control room evacuation is required (step 3.5.2.e).
Choice D: Correct, See explanation.
SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal,
abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
Page: 155 of 193 5/4/2017
83. Unit Two is performing SBO and has just entered ELAP conditions.
RPV water level is 190 inches and being controlled with RCIC
Reactor pressure is 480 psig and being controlled with SRVs
Which one of the following completes the statement below?
IAW SBO, the Unit CRS will direct the defeat of (1) per (2) .
A. (1) RCIC Automatic Logic (does not include turbine exhaust isolation/trip and
manual turbine trip)
(2) 0EOP-01-SBO-02, BLACKED OUT UNIT INITIAL ACTIONS
B. (1) RCIC Automatic Logic (does not include turbine exhaust isolation/trip and
manual turbine trip)
(2) 0EOP-01-SEP-10, CIRCUIT ALTERATION PROCEDURE
C. (1) ONLY the RCIC High RPV Level Closure of E51-F045
(2) 0EOP-01-SBO-02, BLACKED OUT UNIT INITIAL ACTIONS
D. (1) ONLY the RCIC High RPV Level Closure of E51-F045
(2) 0EOP-01-SEP-10, CIRCUIT ALTERATION PROCEDURE
Page: 156 of 193 5/4/2017
K/A:
295008 High Reactor Water Level
2.1 Conduct of Operations
G2.1.7 Ability to evaluate plant performance and make operational judgments based on operating
characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13)
RO/SRO Rating: 4.4/4.7
Tier 1 / Group 2
K/A match: This meets the K/A because it tests the applicants ability to make an operational judgement
based on plant conditions (ELAP) which will allow for an expanded level band by defeat of
trips (inlcuding high water level),isolations, and initiation signals.
Pedigree: New
Objective: LOI-CLS-LP-303-A, Obj. 05 - Given plant conditions, control room indications, and
EOP-01-SBO, Station Blackout Procedure, determine the required operator actions. (LOCT)
Q Reference: 2EOP-01-SBO;0EOP-01-SEP-10;0EOP-01-SBO-02;0OI-37.14
Ref Provided: None
Cog Level: High
Explanation: IAW SBO, operation of the steam driven injection system (RCIC) is critical during Phase 1 of
an ELAP. Therefore, IAW SBO-24 RCIC automatic logic is defeated per EOP-01-SEP-10.
This defeats protective logic trips,isolations, and intiation signals to allow for expanded level
and pressure bands.
Distractor Analysis:
Choice A: Part 1 is correct, see explanation. Part 2 is incorrect but is plausible because SBO-02
does contain procedural steps for the suction swap and defeat of temperature isolations of
RCIC.
Choice B: Correct, See explanation.
Choice C: Part 1 is plausible because SRV operation would lead to the need to defeat the RCIC high
RPV level 45 closure and in RVCP you are directed to defeat as necessary, however it is
incorrect since SBO directs defeat of automatic logic for reasons given in the explanation. Part
2 is incorrect but is plausible because SBO-02 does contain procedural steps for the suction
swap and temperature isolations of RCIC.
Choice D: Part 1 is plausible because SRV operation would lead to the need to defeat the RCIC high
RPV level 45 closure, however it is incorrect since SBO directs defeat of automatic logic for
reasons given in the explanation. Part 2 is correct, see explanation.
SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal,
abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
Page: 157 of 193 5/4/2017
84. Unit Two was in MODE 2 withdrawing rods with SRM D bypassed.
Subsequently, an inadvertant ECCS initiation occurred resulting in off-scale high
readings for reactor period on all SRM channels.
After the ECCS initiation was secured, SRM readings are as follows:
Power Period
SRM A 710 cps sec
SRM B 770 cps sec
SRM C 690 cps Off-scale high
SRM D 670 cps sec
Which one of the following completes both statements below?
The number of currently OPERABLE SRM channels is (1) .
IAW Tech Spec 3.3.1.2 Bases, SRM Instrumentation, a minimum of (2) SRMs are
required to perform a reactor startup.
A. (1) two
(2) two
B. (1) two
(2) three
C. (1) three
(2) two
D. (1) three
(2) three
Page: 158 of 193 5/4/2017
K/A:
295014 Inadvertent Reactivity Addition
AA2 Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY
ADDITION : (CFR: 41.10 / 43.5 / 45.13)
02 Reactor period
RO/SRO Rating: 3.9/3.9
K/A match: This question asks the applicant to determine the end-state status of SRMs based on
indications provided following an inadvertent reactivity addition. The given SRM period status
requires applicant determination of OPERABILITY IAW TS Bases.
Pedigree: New
Objective: LOI-CLS-LP-009-A Obj 12. Given plant conditions, determine whether minimum Technical
Specifications requirements associated with the Source Range and Intermediate
Range Monitoring Systems are met. (LOCT)
Q Reference: TS 3.3.1.2, amend 243; TS 3.3.1.2 Bases, rev 57; SD-09.1, rev 9; SD-03, rev 12
Ref provided: None
Cog Level: High
Explanation: Initially, with SRM D bypassed and no change of bypass state provided, SRM D is and
remains INOPERABLE throughout the timeline of events offered in this question.
Additionally, following the inadvertent reactivity addition SRM C is indicated to be
INOPERABLE based on a loss of ability to monitor reactor period. The basis for this
INOPERABILITY is found in the TS 3.3.1.2 Bases. For the first half question, based on the
INOPERABILITIES discussed above, only two SRMs remain OPERABLE. For the second
half question, IAW TS 3.3.1.2 and with the plant in Mode 2 (given information) and IRMs <
range 3, three SRM channels are required to conduct a reactor startup. With the SRMs at
~700 cps the IRMs would be on range 1.
Distractor Analysis:
Choice A: Incorrect - The second half answer is incorrect. The second half answer is plausible as
this would be the correct number of SRMs required to be OPERABLE if the plant was in
Modes 3, 4, or 5
Choice B: Correct, See explanation.
Choice C: Incorrect -Both half answers are incorrect
Choice D: Incorrect - The first half answer is incorrect. The first half answer is plausible as this would be
correct if only the status of SRM D (currently bypassed and therefore INOPERABLE) were
accounted for in determination of this answer choice (i.e. disregarding SRM C Period
indication status).
SRO Basis:10 CFR Part 55 Content: 41(1) and (7) and 43 (5)
Page: 159 of 193 5/4/2017
85. Following a scram on Unit One, the following conditions currently exist:
Control rods 10 between 04 and 22; all other rods are fully inserted.
Reactor pressure 400 psig and stable
Reactor water level 15 inches and stable;
High pressure injection ONLY RCIC available and injecting
Drywell spray In service on Loop A RHR;
Drywell pressure 5 psig and slowly lowering
Torus level 6 inches and slowly rising
Torus pressure 4 psig and slowly lowering
Which one of the following completes both statements below?
0EOP-01-SEP-15, Anticipate Emergency Depressurization, (1) allowed to be
entered.
0EOP-01-EDP, Emergency Depressurization, (2) required to be entered.
A. (1) is
(2) is
B. (1) is
(2) is NOT
C. (1) is NOT
(2) is
D. (1) is NOT
(2) is NOT
Page: 160 of 193 5/4/2017
K/A:
295029 High Suppression Pool Water Level
EA2 Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL
WATER LEVEL : (CFR: 41.10 / 43.5 / 45.13)
02 Reactor pressure
RO/SRO Rating: 3.5/3.6
Tier 1 / Group 2
K/A match: The proposed test item tests the applicants ability to determine that emergency
depressurization is required based on high torus level; the test item also tests the applicants
ability to interpret whether anticipation of emergency depressurization is allowed.
Pedigree: New
Objective: LOI-CLS-LP-300-E, Obj 16c-Given plant conditions and the Anticipated Transient Without
Scram Procedure, determine if any of the following are appropriate or required: Emergency
Depressurization (LOCT)
Q Reference: PCCP Step T/L-8, 9; 0OI-37.8, RVCP
Ref provided: None
Cog Level: High
Explanation: Anticipatory Depressurization (Step T/L-8 in PCCP) is only allowed in the RC/P leg of RVCP
only; it is not an option during an ATWS. Emergency Depressurization is required based on
Torus Level cannot be maintained below + 6 (Step T/L-9 in PCCP).
Distractor Analysis:
Choice A: Incorrect because Anticipatory Depressurization (Step T/L-8 in PCCP) is only allowed in the
RC/P leg of RVCP only; it is not an option during an ATWS. Plausible because Step T/L-8 in
PCCP requires Anticipatory Depressurization to prevent heating up containment.
Choice B: Incorrect because- Anticipatory Depressurization (Step T/L-8 in PCCP) is only allowed in the
RC/P leg of RVCP only; it is not an option during an ATWS. Also because Emergency
Depressurization is required based on Torus Level cannot be maintained below + 6 (Step
T/L-9 in PCCP). Plausible because Step T/L-8 in PCCP requires Anticipatory Depressurization
to prevent heating up containment. Also plausible because by Emergency Depressurizing,
RCIC (the only available high pressure injection source) will no longer be available.
Choice C: Correct Answer, see explanation.
Choice D: Incorrect because Emergency Depressurization is required based on Torus Level cannot be
maintained below + 6 (Step T/L-9 in PCCP). Plausible because by Emergency
Depressurizing, RCIC (the only available high pressure injection source) will no longer be
available.
SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal,
abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
Page: 161 of 193 5/4/2017
86. Unit Two was at 100% power when a transient occurred that caused an auto initiation
of HPCI. The following annunciators were received:
A-1 (4-2), HPCI TURB BRG OIL PRESS LO
A-1 (1-2), HPCI OIL TANK LEVEL HI/LO
The RB AO reports an unisolable oil leak on the HPCI system.
Which one of the following completes both statements below?
If the HPCI oil system completely depressurizes, the final position of 2-E41-V8, HPCI
Turbine Stop Valve, will be (1) .
When evaluating Tech Spec 3.5.1, ECCS - Operating, for a subsequent plant startup,
use of LCO 3.0.4.b (2) applicable to HPCI.
A. (1) open
(2) is
B. (1) closed
(2) is
C. (1) open
(2) is NOT
D. (1) closed
(2) is NOT
K/A:
206000 High Pressure Coolant Injection System
A2 Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT
INJECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control,
or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
15 Loss of control oil pressure: BWR-2,3,4
RO/SRO Rating: 3.4/3.5
K/A match: Applicant must evaluate given conditions and annunciators provided to determine status of
the HPCI Turbine Stop Valve following an oil leak. Following this determination, the applicant
is asked to indicate whether LCO 3.0.4.b is applicable to LCO 3.5.1 TS action statements for
HPCI.
Pedigree: New
Objective: LOI-CLS-LP-19
Obj. 3 Given plant conditions, predict how the HPCI System will respond to the following
events: f. Loss of lube oil
Obj. 25 Given plant conditions, associated with the HPCI system, determine the required
actions: b. to be taken in accordance with Technical Specifications or the TRM.
Q Reference: SD-19, Rev. 24; Technical Specifications, Amendment No. 233
Ref provided: None
Cog Level: High
Explanation: First-part question
Page: 162 of 193 5/4/2017
On HPCI, the Turbine Stop Valve (a normally closed valve) is repositioned open upon
system initiation and requires sufficient control oil pressure to maintain the valve open. On
RCIC, the Steam Supply Valve (a normally closed valve) is repositioned open upon system
initiation and requires DC power availability for completion of valve strokes. In the event of
a loss of motive force, the HPCI turbine stop valve will close while the RCIC Steam Supply
Valve will fail as is. The HPCI/RCIC steam supply difference noted is aided in plausibility for
the open distractor choice due to the given automatic initiation condition.
Second-part question
LCO 3.0.4.b, which concerns use of risk assessments when considering mode change
applicability, is specifically identified as being not applicable for HPCI. Plausibility is assured
as other TS systems can invoke the requirements of LCO 3.0.4.b for mode changes. While
the knowledge elicited is above-the-line TS knowledge it concerns an application of LCO
3.0.4.b which is outside of RO TS LOK..
Distractor Analysis:
Choice A: First part plausible because the HPCI turbine stop valve will close while the RCIC Steam
Supply Valve will fail as is. The HPCI/RCIC steam supply difference noted is aided in
plausibility for the open distractor choice due to the given automatic initiation condition.
Second part is plausible because other TS systems can invoke the requirements of LCO
3.0.4.b for mode changes.
Choice B: First part is correct. Second part is plausible because other TS systems can invoke the
requirements of LCO 3.0.4.b for mode changes.
Choice C: First part plausible because the HPCI turbine stop valve will close while the RCIC Steam
Supply Valve will fail as is. The HPCI/RCIC steam supply difference noted is aided in
plausibility for the open distractor choice due to the given automatic initiation condition.
Second part is correct.
Choice D: Correct, See explanation.
SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Page: 163 of 193 5/4/2017
87. For the past three months, Unit Two has been operating at 100% power when main
condenser vacuum begins lowering at a steady 2 inches Hg per minute.
Which one of the following completes both statements below?
A reactor scram will be caused first by the closure of the (1) .
Assuming a constant rate of lowering vacuum and 15 minutes after the reactor scram,
control of RPV pressure will be procedurally maintained IAW (2) .
A. (1) turbine stop valves
(2) Reactor Scram Procedure
B. (1) main steam isolation valves
(2) Reactor Scram Procedure
C. (1) turbine stop valves
(2) Reactor Vessel Control Procedure
D. (1) main steam isolation valves
(2) Reactor Vessel Control Procedure
Page: 164 of 193 5/4/2017
K/A:
212000 High Off-Site Release Rate
A2 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM ; and
(b) based on those predictions, use procedures to correct, control, or mitigate the consequences
of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
13 Low condenser vacuum
RO/SRO Rating: 3.8/3.9
Tier 2 / Group 1
K/A match: This question requires the applicant to determine the cause of an RPS scram from a lowering
condenser vacuum condition. The applicant is required to apply procedure knowledge to
address RPV pressure control with the low condenser vacuum condition
Pedigree: New
Objective: LOI-CLS-LP-026, Obj. 4j
Q Reference: 0AOP-37.0, Rev. 39, 1EOP-01-RSP, Rev. 16, 1EOP-01-RVCP, Rev. 10.
Ref Provided: None
Cog Level: High
Explanation: If condenser vacuum lowers to 22.4 Hg, then the main turbine trips. TSV and TCVs being <
90% open generates a reactor scram. RPV pressure is controlled by RSP following the
scram by ensuring automatic BPV actuation. For the given situation and upon MSIV closure,
RVCP requires entry due to High Reactor Pressure (1060 psig) and transition of the plant to
SRVs for pressure control.
Distractor Analysis:
Choice A: Incorrect because - Main turbine bypass valves can be utilized as a pressure control
mechanism in RSP, however, these valves get an automatic close signal when condenser
vacuum reaches 7 Hg. With MSIVs already closed at 10 Hg, and unavailability of main
turbine bypass valves, RPV pressure will be maintained using SRVs in accordance with
guidance provided within RVCP.
Choice B: Incorrect because - If condenser vacuum lowers to 22.4 inches Hg, then, the main turbine
trips. TSV and TCVs being < 90% open generates a reactor scram. If condenser vacuum
lowers to 10 inches Hg, then the MSIVs receive a close signal. Main turbine bypass valves
can be utilized as a pressure control mechanism in RSP, however, these valves get an
automatic close signal when condenser vacuum reaches 7 Hg. With MSIVs already
closed at 10 Hg, and unavailability of main turbine bypass valves, RPV pressure will be
maintained using SRVs in accordance with guidance provided within RVCP.
Choice C: Correct, See explanation.
Choice D: Incorrect because - if condenser vacuum lowers to 22.4 inches Hg, then, the main turbine
trips. If the turbine trips and reactor power is greater than 26%, then, an automatic reactor
scram occurs.
SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal,
abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
Page: 165 of 193 5/4/2017
88. Unit One is operating at 100% power when the following occurs:
A-3 (1-1), SAFETY OR DEPRESS VLV LEAKING, is in alarm
A-3 (1-10), SAFETY/RELIEF VALVE OPEN, is in alarm
The RO has reported that 1-B21-F013A, Auto Relief A, has a lit amber indication
Which one of the following completes both statements below?
PCCP actions (1) be performed concurrently with 0AOP-30.0, Safety/Relief Valve
Failures, actions.
Upon determination that 1-B21-F013A cannot be closed (fuses were pulled),
the Unit CRS will direct crew entry into (2) Procedure.
A. (1) will
(2) Reactor Vessel Control
B. (1) will NOT
(2) Reactor Vessel Control
C. (1) will
(2) Reactor Scram
D. (1) will NOT
(2) Reactor Scram
Page: 166 of 193 5/4/2017
K/A:
239002 Safety Relief Valves
G2.4.08 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
(CFR: 41.10 / 43.5 / 45.13)
RO/SRO Rating: 3.8/4.5
K/A match: As written question requires applicant knowledge of AOP usage and EOP network transition
to successfully answer.
Pedigree: New
Objective: LOI-CLS-LP-302-M
Obj. 4 Given plant conditions determine the required supplementary actions IAW:
b. 0AOP-30.0, Safety/Relief Valve Failures
Q Reference: 0AOP-30, Safety/Relief Valve Failures, Rev. 20
Ref provided: None
Cog Level: Fundamental
Explanation: As specified in 0AOP-30.0, actions taken following entry into PCCP will be performed
concurrently with actions taken in 0AOP-30.0. Upon determination that the open SRV
cannot be closed, the Unit CRS will direct a procedure transition to RSP (and insertion of a
Distractor Analysis:
Choice A: Incorrect because - Second half answer is incorrect. AOP-30 specifically directs
procedure transition to RSP. Entry into RSP will be short, as entry criteria is already met
for entry into RVCP (due to PCCP entry), however, the EOP procedure transition remains
RSP to RVCP.
Choice B: Incorrect - Both answer choices incorrect. See explanations above.
Choice C: Correct, See explanation.
Choice D: Incorrect - First half answer is incorrect. AOP-30 specifically directs performance of PCCP
actions concurrent with AOP-30..
SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal,
abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
Page: 167 of 193 5/4/2017
89. Unit Two was at 100% power, when the following timeline occurs:
Time 1300 A LOOP occurs
Time 1302 E3 and E4 lose power and remain de-energized
Time 1325 Power is restored to E8 via FLEX DG2 and E7 remains de-energized
Which one of the following completes both statements below?
The CRS will direct de-energization of the Primary UPS IAW (1) .
As required by the SBO Coping Analysis the latest that this action must be completed
is by time (2) .
A. (1) 0AOP-36.1, Loss of Any 4160V Buses or 480V E-Buses
(2) 1332
B. (1) 0AOP-36.1, Loss of Any 4160V Buses or 480V E-Buses
(2) 1402
C. (1) 0EOP-01-SBO-10, Battery Load Stripping
(2) 1332
D. (1) 0EOP-01-SBO-10, Battery Load Stripping
(2) 1402
Page: 168 of 193 5/4/2017
K/A:
262002 Uninterruptable Power Supply (A.C./D.C.)
G2.4.11 Knowledge of abnormal condition procedures. (CFR: 41.10 / 43.5 / 45.13)
RO/SRO Rating: 4.0/4.2
K/A match: Given an operationally valid set of plant conditions involving procedure 0AOP-12.0, LOSS
OF UNINTERRUPTIBLE POWER SUPPLY (UPS), the SRO applicant will demonstrate
knowledge of the content of AOP-12.0 by correctly identifying the procedure required to
mitigate a loss of the Reactor Manual Control System (RMCS)which is powered from
UPSand correctly identifying where the SRO would find detailed procedure steps that direct
the electrical restoration.
Pedigree: New
Objective: LOI-CLS-LP-039
Obj. 21 Given plant conditions associated with the Emergency Diesel Generators, determine
the required actions:
b. to be taken in accordance with Technical Specifications or the TRM. (SRO only)
Q Reference: 0EOP-01-SBO
Ref provided: None
Cog Level: Comprehensive
Explanation: With the battery charger not energized the procedure require load stripping (de-energizing
the primary UPS) actions. The critical time requirement identified in the SBO procedure for
this to happen is 60 minutes.
Distractor Analysis:
Choice A: Incorrect - Although 36.1 has actions for load stripping the plant conditions indicate a
station blackout exists. The 30 minute requirement is another critical action identified in
the station blackout procedure.
Choice B: Incorrect - Although 36.1 has actions for load stripping the plant conditions indicate a station
blackout exists. The 60 minute requirement is correct.
Choice C: Incorrect - This is the correct procedure to use. The 30 minute requirement is another critical
action identified in the station blackout procedure.
Choice D: Correct, See explanation
SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal,
abnormal, and emergency situations. [10 CFR 55.43(b)(5)] SRO-only level is met in
accordance with Attachment 2 of ES-401 in that this question requires the applicant to :
-assess plant conditions and select a procedure to mitigate, and it also requires the
applicant to demonstrate knowledge of when to implement attachments and appendices,
including how to coordinate these items with procedure steps.
Page: 169 of 193 5/4/2017
90. Unit One was operating at 100% power when the following sequence of events
occurred:
0900 DG2 running loaded performing 0PT-12-2B, No. 2 Diesel Monthly Load Run
0905 UA-16 (1-1), DG-2 TO BUS E2 BKR TRIP, alarms
0910 The CRS is evaluating Tech Spec 3.8.1, AC Sources - Operating, entry
criteria. DG1, DG3, DG4, and the Supp DG have been determined to be
Which one of the following completes both statements below?
At 0906, DG2 (1) trip and restart.
Tech Spec 3.8.1 (2) required to be entered.
A. (1) did
(2) is
B. (1) did
(2) is NOT
C. (1) did NOT
(2) is
D. (1) did NOT
(2) is NOT
Page: 170 of 193 5/4/2017
K/A:
264000 Emergency Generators (Diesel/Jet)
A2 Ability to (a) predict the impacts of the following on the EMERGENCY GENERATORS
(DIESEL/JET); and (b) based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
06 Opening normal and/or alternate power to emergency bus
RO/SRO Rating: 3.4/3.4
Tier 2 Group 1
K/A match: The applicants must determine the status of DG2 based upon the conditions provided and
then determine if credit can be taken for the SUPP-DG when evaluating entry into TS 3.8.1.
Pedigree: New
Objective: LOI-CLS-LP-039
Obj. 3 Given plant conditions, determine if EDGs will automatically start.
Obj. 4 Given plant conditions, determine if EDGs will trip:
b. After a manual start
Obj. 21 Given plant conditions associated with the Emergency Diesel Generators, determine
the required actions:
b. to be taken in accordance with Technical Specifications or the TRM. (SRO only)
Q Reference: 1APP-UA-16, rev 29; SD-39, rev 20; TS amend 264; TS Bases, rev 84
Ref provided: None
Cog Level: High
Explanation: The DG will trip upon the opening of its output breaker. It will subsequently receive an auto
start signal (based on E-Bus undervoltage), but will be prevented from starting due to
existing electrical lockout. The SUPP-DG cannot be used when making a determination of
TS 3.8.1 applicability. Availability of the SUPP-DG is used when making the determination
for appropriate completion time following an EDG INOPERABILITY (ref Condition D),
however, its status has no bearing on entry into TS 3.8.1..
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect - The second half answer is incorrect. Answer choice is plausible considering that
the SUPP-DG is addressed in TS LCO 3.8.1, Condition D, when making the determination
for EDG INOPERABILITY.
Choice C: Incorrect - The first half answer is incorrect. Answer choice is plausible as this would be the
correct answer in a situation where the EDG was relied upon to re-power its associated E-bus
following a loss of off-site power. Based on the given condition of a fault associated with the
output breaker, the EDG would not restart based on a generator electrical lockout (not subject
to one-time reset).
Choice D: Incorrect - Both question parts are incorrect.
SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Page: 171 of 193 5/4/2017
91. Which one of the following completes both statements below concerning the RWCU
system area ventilation high differential temperature Group 3 isolation?
This signal results in a RWCU isolation (1) a 28.5 minute time delay.
If the appropriate RWCU valve(s) fail(s) to isolate for this Group 3 signal, due to valve
failure(s), the CRS will enter TS 3.6.1.3 Condition(s) (2) .
A. (1) with
(2) A ONLY
B. (1) with
(2) A and B
C. (1) without
(2) A ONLY
D. (1) without
(2) A and B
Page: 172 of 193 5/4/2017
K/A:
204000 Reactor Water Cleanup System
A2 Ability to (a) predict the impacts of the following on the REACTOR WATER CLEANUP SYSTEM ;
and (b) based on those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations:(CFR: 41.5 / 45.6)
14 System high temperature
RO/SRO Rating: 3.2/3.2
Tier 2 Group 2
K/A match: The applicants must predict the appropriate isolation for a group 3 temperature isolation of
RWCU, and this knowledge is required to determine the appropiate TS which mitigates a
failure for this isolation.
Pedigree: New
Objective: LOI-CLS-LP-012 Obj. 15 Given plant conditions, determine whether minimum Technical
Specifications requirements associated with the Primary Containment Isolation System are
met. (LOCT).
Q Reference: TS 3.6.1.3; SD-12
Ref provided: TS 3.6.1.3
Cog Level: High
Explanation: Part 1: The group 3 isolation signal for RWCU system area ventilation high differential
temperature results in the immediate closure of the G31-F001 and F004. Part 2: Since a
failure of RWCU to isolate on this signal would equate to a failure of 2 PCIVs, TS 3.6.1.3
Condtion A and B would both be entered.
Distractor Analysis:
Choice A: Incorrect - Part 1 is incorrect but plausible since the group 3 isolation signal for
differential flow has a 28.5 minute time delay. Part 2 is incorrect, but plausible since a
failure of isolation for the group 3 signal for NRHX outlet temp would require only the
failure of the F004, and therefore only require entry into TS 3.6.1.3 Condition A.
Choice B: Incorrect - Part 1 is incorrect but plausible since the group 3 isolation signal for differential
flow has a 28.5 minute time delay. Part 2 is correct, see explanation.
Choice C: Incorrect - Part 1 is correct, see explanation. Part 2 is incorrect, but plausible since a failure of
isolation for the group 3 signal for NRHX outlet temp would require only the failure of the F004,
and therefore only require entry into TS 3.6.1.3 Condition A.
Choice D: Correct, See explanation.
SRO Basis: This question meets SRO only due to requiring the applicant to determine the TS condition
for a Group 3 isolation failure. Facility operating limitations in the TS and their bases. [10
CFR 55.43(b)(2)]
Page: 173 of 193 5/4/2017
92. Unit One is operating at 100% power. All rod position information on Panel P603, the
Process Computer and ERFIS, has been lost for control rod 02-19. Repairs are NOT
expected within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The crew entered 0AOP-02.0, Control Rod Malfunction/Misposition, and are attempting
to determine rod position in the RPIS cabinet via reed switch continuity measurements.
The operator determines that continuity exists only between input terminals S and H.
[Reference Provided]
Which one of the following completes both statements below?
IAW 0AOP-02.0, the Equivalent Control Rod Position is (1) .
IAW Tech Spec 3.1.3, Control Rod Operability, control rod 02-19 (2) OPERABLE.
A. (1) 36
(2) is
B. (1) 36
(2) is NOT
C. (1) 47
(2) is
D. (1) 47
(2) is NOT
Page: 174 of 193 5/4/2017
K/A:
214000 Rod Position Information System
G2.1.25 Conduct of Operations: Ability to interpret reference materials, such as graphs, curves, tables,
etc. (CFR: 41.10 / 43.5 / 45.12)
RO/SRO Rating: 3.9/4.2
Tier 2 / Group 2
K/A match: The proposed test item requires the applicant to use 1-FP-50228-SH001 to identify the
meaning of continuity on Terminals S & H.
Pedigree: New
Objective: LOI-CLS-LP-007, Obj 15-Given plant conditions, determine whether given plant conditions
meet minimum Technical Specification requirements associated with the Reactor Manual
Control System. (LOCT)
Q Reference: 0AOP-02.0, Attachment 1, Tech Spec 3.1.3
Ref provided: 1-FP-50228-SH001 upper left portion from 0AOP-02.0, Attachment 1 (see Notes)
Cog Level: High
Explanation: Using the reference drawing, the equivalent rod position is determined by adding the tens
input (S Terminal is 30) and the units input (H Terminal is 6); the bases for Tech Spec
3.1.3.1 says that an alternate method may be used to satisfy SR 3.1.3.1.
Distractor Analysis:
Choice A: Correct Answer, see explanation.
Choice B: Incorrect because the equivalent rod position method results were continuity only on two
Terminals, this means that the equivalent rod position is trustworthy, and the Tech Spec Bases
indicated that an alternate method of determining rod position can be used to fulfill SR 3.1.3.1.
Plausible because the stem says that all rod position information in the main control room is
lost.
Choice C: Incorrect because using the reference drawing, the equivalent rod position is determined by
adding the tens input (S Terminal is 30) and the units input (H Terminal is 6), which is 36.
Plausible because the S Terminal and H Terminal, if incorrectly read, can be interpreted as 47.
Choice D: Incorrect because using the reference drawing, the equivalent rod position is determined by
adding the tens input (S Terminal is 30) and the units input (H Terminal is 6), which is 36. Also
incorrect because Tech Spec Bases states that an alternate method of determining rod
position can be used to fulfill SR 3.1.3.1. Plausible because the S Terminal and H Terminal, if
incorrectly read, can be interpreted as 47. Also plausible because the stem says that all rod
position information in the main control room is lost.
SRO Basis: The SRO aspect is being met because the applicant must have knowledge of the
allowances of Tech Spec Bases for SR 3.1.3.1 (control rod position) for alternate means of
control rod position. Assessment of facility conditions and selection of appropriate
procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
Page: 175 of 193 5/4/2017
93. Unit One is in MODE 5, with the following conditions:
1-A-04 (5-7), FUEL POOL GATE RX WELL SEAL in alarm
Fuel Pool Gate Seal Pressure is 22 psig and lowering slowly.
NOTE: 0OP-46, Instrument and Service Air System Operating Procedure
Which one of the following completes both statements below?
1-A-04 (5-7) is in alarm due to flow through 1-G41-FSH-N003, Fuel Pool Gate and
Reactor Well Seal Leak Flow Switch, which exceeded a minimum of (1) .
IAW A-04 (5-7), the Control Room Supervisor will next direct (2) to the Fuel Pool
Gate Seals.
A. (1) 1.7 gpm
(2) aligning Nitrogen bottles IAW 0OP-46
B. (1) 1.7 gpm
(2) aligning the Nitrogen Backup System IAW OP-46
C. (1) 5 gpm
(2) aligning Nitrogen bottles IAW 0OP-46
D. (1) 5 gpm
(2) aligning the Nitrogen Backup System IAW OP-46
Page: 176 of 193 5/4/2017
K/A:
233000 Fuel Pool Cooling and Clean-up
A2 Ability to (a) predict the impacts of the following on the FUEL POOL COOLING AND CLEAN-UP ;
and (b) based on those predictions, use procedures to correct, control, or mitigate the
consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)
11 Fuel pool gate seal high flow
RO/SRO Rating: 2.9/3.2
Tier 2 / Group 2
K/A match: The applicant is expected to know the high seal flow setpoint which results in a control room
annunciator and use procedures to mitigate the consequences of high spent fuel pool seal
leak flowrate.
Pedigree: New
Objective: LOI-CLS-LP-013 Obj 1: Under all plant conditions operate the Fuel Pool Cooling system in
accordance with station procedures.
Q Reference: 1APP-A-04, Annunciator Procedure for Panel A-04, Rev. 58, 0-OP-46, Instrument and
Service Air System Operating Procedure, Rev. 179
Ref provided: None
Cog Level: High
Explanation: Spent Fuel Pool Gate Rx Well Seal (A-04 5-7) alarm is received when flow through flow
switch, G41-FSH-N003, located on 50 ft elevation of Reactor Building, exceeds 5 gpm. APP
A-04 5-7 states that, if fuel pool gate seal is leaking, then the following steps need to be
performed: a. Ensure fuel pool gate seal pressure is between 25 and 30 psig. c. If
instrument air is not available, then transfer seal supply to the nitrogen bottle supply in
accordance with 0OP-46.
Distractor Analysis:
Choice A: Incorrect because - Spent Fuel Pool Gate Rx Well Seal (A-04 5-7) alarm is received when
flow through flow switch, G41-FSH-N003, located on 50 ft elevation of Reactor Building,
exceeds 5 gpm. A-1 (6-7) has a 1.7 gpm setpoint. If instrument air is not available, then
transfer seal supply to the nitrogen bottle supply in accordance with 0OP-46
Choice B: Incorrect because - Spent Fuel Pool Gate Rx Well Seal (A-04 5-7) alarm is received when
flow through flow switch, G41-FSH-N003, located on 50 ft elevation of Reactor Building,
exceeds 5 gpm. A-1 (6-7) has a 1.7 gpm setpoint. Nitrogen is used as the backup supply
but not from the Nitrogen Backup system.
Choice C: Correct, See explanation.
Choice D: Incorrect because - APP A-04 5-7 states that, if fuel pool gate seal is leaking and seal
pressure is not between 25 and 30 psig, then, ensure instrument air valve lineup is correct with
instrument air available. Nitrogen is used as the backup supply but not from the Nitrogen
Backup system.
SRO Basis: N/A
Page: 177 of 193 5/4/2017
94. Unit Two is operating at 60% power with feedwater heaters 4A and 5A bypassed.
Feedwater temperature is 378°F.
A portion of 2OP-32, Condensate and Feedwater System Operating Procedure,
Attachment 6, is included below:
Which one of the following completes the statements below IAW 0OI-01.01, BNP
Conduct of Operations Supplement?
The Unit CRS (1) required to implement the thermal limit penalties for FHOOS,
(Feedwater Heater Out of Service).
Entry into TS 3.0.3 (2) required if final feedwater temperature is less than the
110.3°F Reduced FFWT table value.
A. (1) is
(2) is
B. (1) is
(2) is NOT
C. (1) is NOT
(2) is
D. (1) is NOT
(2) is NOT
K/A:
G2.1.43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor
coolant system temperature, secondary plant, fuel depletion, etc. (CFR: 41.10 / 43.6 / 45.6)
RO/SRO Rating: 4.1/4.3
Tier 3
K/A match: Applicant is required to use the final feedwater temperature reduction attachment to
determine if the effect of the feedwater reduction is severe enough on reactivity to require
implementation of thermal limit penalties.
Pedigree: Modified from 2016 BNP NRC Exam (Q95)
Objective: CLS-LP-032 #27: Given plant conditions and Tech Specs, including the Bases, TRM, ODCM,
and COLR, determine whether given plant conditions meet minimum Tech Spec
requirements associated with the Condensate and Feedwater System.
Page: 178 of 193 5/4/2017
Q Reference: 0OI-01.01; 2OP-32
Ref provided: None
Cog Level: High
Explanation: Part 1 Final FW temp of 378°F is less than the nominal FW temp for 60% power
(387.6°F), but greater than the Nominal FW Temp Reduced 10°F value of 377.6°F.
Therefore the thermal limit penalties for FHOOS do not need to be implemented.
Part 2 There are no operating limits specified in the COLR for operation beyond 110.3°F
Final FW Temp, so since thermal limits cannot be verified to be within the limits in the
COLR, entry into the Actions of LCOs 3.2.1, 3.2.2, and 3.2.3 is required. These LCOs
require thermal limits be restored within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and if theyre not, then reduce power to
<23% within 4 more hours.
Distractor Analysis:
Choice A: First part is plausible in that an applicant could think that the thermal limit penalties for
FHOOS do need to be implemented if they think that being below the Nominal FW Temp
drives that determination. Additionally, even if an applicant recalls that its the Reduced
10°F value that necessitates thermal limit penalties, they could misread the table and
compare the given FW temp of 378°F to the 61% power entry for Nominal FW Temp
Reduced 10°F of 379.0°F and therefore arrive at the incorrect answer. Second part is
plausible because if the applicant believed that TS 3.2.1 / 3.2.2 / 3.2.3 were entered if
below the Reduced 10°F limit, and that if you got to the 110.3°F Reduced limit then
there was no Tech Spec to go to, so 3.0.3 would apply.
Choice B: First part is plausible in that an applicant could think that the thermal limit penalties for
FHOOS do need to be implemented if they think that being below the Nominal FW Temp
drives that determination. Additionally, even if an applicant recalls that its the Reduced
10°F value that necessitates thermal limit penalties, they could misread the table and
compare the given FW temp of 378°F to the 61% power entry for Nominal FW Temp
Reduced 10°F of 379.0°F and therefore arrive at the incorrect answer. Second part is
correct.
Choice C: First part is correct. Second part is plausible because if the applicant believed that TS 3.2.1 /
3.2.2 / 3.2.3 were entered if below the Reduced 10°F limit, and that if you got to the 110.3°F
Reduced limit then there was no Tech Spec to go to, so 3.0.3 would apply.
Choice D: Correct, See explanation
SRO Basis: Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Question requires that the applicant determines whether the TS thermal limits should incur a
penalty. In addition, it also requires that the candidate determines whether LCO 3.0.3 applies
for a given condition
from BNP 2016 NRC question 95
Page: 179 of 193 5/4/2017
Page: 180 of 193 5/4/2017
95. Which one of the following completes both statements below?
IAW 0OI-01.01, BNP Conduct of Operations Supplement, the minimum required
number of Auxiliary Operators for manning a shift at BNP is (1) .
IAW Tech Spec 5.2.2, Facility Staff, to accommodate unexpected absences, the total
number of non-licensed operators may be less than the requirement above for a period
of time not to exceed (2) , provided immediate action is taken to restore the shift
crew composition to within the minimum requirements.
A. (1) three
(2) one hour
B. (1) three
(2) two hours
C. (1) nine
(2) one hour
D. (1) nine
(2) two hours
Page: 181 of 193 5/4/2017
K/A:
G2.1 Conduct of Operations
05 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime
limitations, etc. (CFR: 41.10 / 43.5 / 45.12)
RO/SRO Rating: 2.9/3.9
Tier 3
K/A match: The proposed test item tests the applicants knowledge of minimum crew complement Tech
Spec requirements
Pedigree: 2016 BNP NRC Q94
Objective: LOI-CLS-LP-200-B Obj.12.-Identify conditions and limitations in the facility license.
Q Reference: Tech Spec 5.2.2.a and 5.2.2.c
Ref provided: None
Cog Level: Fundamental
Explanation: Verbatim requirement from Tech Spec 5.2.2.a and 5.2.2c and 0OI-01.01, BNP Conduct of
Operations Supplement.
Distractor Analysis:
Choice A: Incorrect because Tech Spec 5.2.2.c allows 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to be below the minimum crew
complement. Plausible because AD-OP-ALL-1000, Conduct of Ops, Section 5.5.19.4, Short
Term Relief, states that whenever a watch station is relieved for greater than ONE HOUR, this
information shall be entered in a Narrative Log Program, a formal turnover and shift turnover
sheet will be completed, including the logs signed over. Plausible because TS 5.2.2 requires 3
AOs for both Units which does not take into account ASSD and Fire Brigade requirements..
Choice B: Incorrect plausible because TS 5.2.2 requires 3 AOs for both Units which does not take into
account ASSD and Fire Brigade requirements.
Choice C: Incorrect because Tech Spec 5.2.2.c allows 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to be below the minimum crew
complement. Plausible because 0OI-01.01, BNP Conduct of Ops Supplement, Section 5.3,
Operations Shift Staffing, requires a minimum of nine aux operators. Also plausible because
AD-OP-ALL-1000, Conduct of Ops, Section 5.5.19.4, Short Term Relief, states that whenever
a watch station is relieved for greater than ONE HOUR, this information shall be entered in a
Narrative Log Program, a formal turnover and shift turnover sheet will be completed, including
the logs signed over.
Choice D: Correct Answer, see explanation
SRO Basis: The SRO portion of the question is associated with the required action when an unexpected
absence occurs. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Page: 182 of 193 5/4/2017
96. Which one of the following completes the statements below concerning Temporary
Alterations in Support of Maintenance (TAISOMs) IAW 0OI-01.01, BNP Conduct of
Operations Supplement?
Disabling an annunciator by pulling the annunciator card in response to a failed
pressure transmitter in an effort to avoid a nuisance alarm, (1) an example of a
TAISOM.
A disabled annunciator controlled as a TAISOM requires a 10CFR 50.59 review if it is
expected to be in effect for greater than (2) .
A. (1) is
(2) 14 days
B. (1) is
(2) 90 days
C. (1) is NOT
(2) 14 days
D. (1) is NOT
(2) 90 days
Page: 183 of 193 5/4/2017
K/A:
G2.2.11 Knowledge of the process for controlling temporary design changes. (CFR: 41.10 / 43.3 / 45.13)
RO/SRO Rating: 2.3/3.3
Tier 3
K/A match: The applicant is expected to apply knowledge of the Temporary Alteration in Support of
Maintenance (TAISOM) process as it applies to temporary design changes (the disabling of
Pedigree: New
Objective: LOI-CLS-LP-201-D Objective 1.m
Q Reference: 0OI-01.01, BNP Conduct of Operations Supplement, Rev 77;
0PLP-22, Temporary Alteration Control, Rev. 24
Ref provided: None
Cog Level: fundamental
Explanation: As stated in 0OI-01.01, a TAISOM does not include nonconforming or degraded conditions
intended to be restored back to their as-designed condition. A TAISOM would not be used to
control an annunciator disabled for a nuisance alarm received due to a failed pressure
transmitter. Additionally, a 10CFR 50.59 review is required when a TAISOM is expected to
be in effect for greater than 90 days.
Distractor Analysis:
Choice A: Incorrect because - A TAISOM would not be used to control an annunciator disabled for a
nuisance alarm received due to a failed pressure transmitter. A 10CFR 50.59 review is
required when a TAISOM is expected to be in effect for greater than 90 days. 14 days is
plausible as an answer choice since it relates to control of a temporary alteration as
discussed in 0PLP-22.
Choice B: Incorrect because- A TAISOM would not be used to control an annunciator disabled for a
nuisance alarm received due to a failed pressure transmitter.
Choice C: Incorrect because - A 10CFR 50.59 review is required when a TAISOM is expected to be in
effect for greater than 90 days. 14 days is plausible as an answer choice since it relates to
control of a temporary alteration as discussed in 0PLP-22.
Choice D: Correct, See explanation.
SRO Basis: Facility licensee procedures required to obtain authority for design and operating changes in
the facility. [10 CFR 55.43(b)(3)]
Page: 184 of 193 5/4/2017
97. Unit Two is at 100% power
A BSEP Radioactive Liquid Release Permit is being completed for discharge of the
Unit Two Saltwater Release Tank via the General Electric Radiation Monitor IAW
0OP-6.4, Discharging Radioactive Liquid Effluents to the Discharge Canal
2-G16-FIT-N057, Liquid Radwaste Effluent Flow Measurement Device, is
Which one of the following completes both statements below?
IAW the ODCM, the maximum release rate is determined so that (1) limits are not
exceeded after dilution in the discharge canal.
IAW 0OP-6.4, release of the Unit Two Saltwater Release Tank (2) .
A. (1) 10 CFR 100, Reactor Site Criteria
(2) may be authorized if ODCM compensatory actions are implemented
B. (1) 10 CFR 100, Reactor Site Criteria
(2) is NOT allowed unless 2-G16-FIT-N057 is OPERABLE
C. (1) 10 CFR 20, Standards for Protection Against Radiation
(2) may be authorized if ODCM compensatory actions are implemented
D. (1) 10 CFR 20, Standards for Protection Against Radiation
(2) is NOT allowed unless 2-G16-FIT-N057 is OPERABLE
K/A:
G2.3.11 Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10)
RO/SRO Rating: 3.8/4.3
Tier 3
K/A match: The question meets the K/A at the SRO level because it tests ODCM knowledge of
instrumentation requirements and limits that must be met to control the radiation release
Pedigree: Modified from 2012 NRC Exam
Objective: LOI-CLS-LP-6.3, Obj.08.a
Q Reference: ODCM (BSEP) Rev. 37, 0OP-6.4 Rev.74
Ref provided: None
Cog Level: fundamental
Explanation: In accordance with the BSEP Off-Site Dose Calculation Manual (ODCM), the maximum
release rate is determined so that 10 CFR 20 limits are not exceeded after dilution in the
discharge canal. In accordance with Attachment 13 of 0OP-6.4, ODCM 7.3.1, Radioactive
Liquid Effluent Monitoring Instrumentation, contains compensatory requirements if the Liquid
Radwaste Radioactivity Effluent Monitor or Liquid Radwaste Effluent Flow Measuring Device
is INOPERABLE..
Distractor Analysis:
Page: 185 of 193 5/4/2017
Choice A: Incorrect because - First part is incorrect. Second part is correct. See explanation for B
and C.
Choice B: Incorrect because- First and second parts are incorrect. First part is plausible because
the SJAE radiation monitor setpoint is based on not exceeding a small fraction of 10 CFR
100 limits. The second part is plausible because it would be reasonable to believe that a
release should not be approved without required instrumentation being OPERABLE.
Choice C: Correct, See explanation.
Choice D: Incorrect because - First part is correct. Second part is incorrect. See explanation for B and C.
SRO Basis: Radiation hazards that may arise during normal and abnormal situations, including
maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)]
From the 2012 NRC Exam:
Page: 186 of 193 5/4/2017
98. Initial Conditions:
An emergency was declared, and the EOF was fully staffed and activated.
A gaseous release was determined to be in progress.
Current Conditions:
One offsite dose assessment team, already dispatched, is estimated to receive an
exposure in excess of 10 CFR 20 limits.
Another onsite activity requires additional operator support to protect a large
population.
Mr. Wilmington, the most senior maintenance technician available, has
volunteered to participate in this activity where he will receive an estimated
Mr. Wilmington has received all required briefings, has completed the
Emergency Exposure Authorization Form, and is fully aware of the risks
involved.
Two hours ago, Mr. Wilmington received an actual emergency exposure of
1.75 REM TEDE performing a previous emergency activity.
There has not been time to perform an exposure evaluation following Mr.
Wilmingtons first emergency exposure.
IAW 0PEP-03.7.6, Emergency Exposure Controls, which one of the following
completes both statements below?
The (1) is responsible for authorizing the offsite dose assessment team to receive
exposures in excess of 10 CFR 20 limits.
Mr. Wilmington (2) be authorized to perform the activity required for the protection
of a large population and receive the second emergency exposure.
A. (1) Emergency Response Manager
(2) can
B. (1) Emergency Response Manager
(2) can NOT
C. (1) Site Emergency Coordinator
(2) can
D. (1) Site Emergency Coordinator
(2) can NOT
K/A:
G2.3.04 Knowledge of radiation exposure limits under normal or emergency conditions.
(CFR: 41.12 / 43.4 / 45.10)
RO/SRO Rating: 3.2/3.7
Tier 3
K/A match: Given an operationally valid set of plant conditions, this question requires the SRO applicant
to demonstrate knowledge of radiation limits under emergency conditions. Only SRO licensed
individuals could potentially stand as interim Site Emergency Coordinators, who are
Page: 187 of 193 5/4/2017
responsible for authorizing emergency doses in excess of 10CFR20 limits; therefore, the
proposed question meets SRO-only requirements and can be linked to 10CRD55.43.b(4).
Specifically, the SRO applicant needs to demonstrate knowledge, firstly, concerning the one
exception in which the Site Emergency Coordinator is not responsible for authorizing
emergency doses in excess of 10CFR20 limits. Furthermore, in the second part of the
question, the SRO applicant demonstrates knowledge that an individual is only permitted one
emergency exposure per emergencytherefore, even though the fictitious Mr. Wilmington
would otherwise meet all other requirements to be authorized, the fact that he has already
received one emergency exposure that same day disqualifies him from receiving another.
The K/A is matched on the SRO-only level on both the first part question and second part
question.
Pedigree: New
Objective: CLS-LP-102-A, Obj. 11
Q Reference: 0PEP-03.7.6, EMERGENCY EXPOSURE CONTROLS, rev 4.
Ref provided: None
Cog Level: High
Explanation: Section 4.1 of 0PEP-03.7.6, Emergency Exposure Controls, states as follows: The
Emergency Response Manager is responsible for authorization of exposures in excess of
10CFR20 limits and approval of the administration of potassium iodide (KI) for station ERO
personnel performing offsite functions. Therefore, given the conditions present in the
question stem, the Emergency Response Manager is responsible for authorizing the offsite
dose assessment team to receive an emergency dose and is the correct answer to the first
part question. Section 5.3.1.3 of 0PEP-03.7.6 states the following: Individuals receiving
emergency exposures are restricted from further occupational radiation exposure pending
the outcome of exposure observations and medical surveillance. Therefore, because Mr.
Wilmington has already received an emergency does and has yet to receive emergency
dose medical evaluations, he is prohibited from being authorized to receive an additional
emergency dose in excess of 10CFR20 limits.
Distractor Analysis:
Choice A: Incorrect because - The first part of distractor A is correct; the Emergency Response
Manager is the appropriate authorizing official for the offsite dose assessment teams
under the given conditions. The second part of distractor D is incorrectas per the
above discussion, Mr. Wilmington has already received one emergency exposure and is
therefore prohibited from receiving another (until the required medical
evaluations/surveillances are performed, which have not happened as stated in the
question stem).
Plausible because - The first part of distractor A is correct. The second part question is
plausible because, aside from the previous emergency exposure, every other requirement
is met for Mr. Wilmington. Specifically, plausibility of this distractor is enhanced by
reference to the individuals age, the fact that he has volunteered, the fact that he has
received all required briefings, and the fat that his proposed overall dose for both activities
- 10.90 REM - is less than the limit (25 REM TEDE) for an emergency activity required to
protect a large population.
Choice B: Correct, See explanation.
Choice C: Incorrect because - Both parts of this question are incorrect. First part question is incorrect
because the procedure specifies that the Emergency Response Manager, not the Site
Emergency Coordinator, is the responsible authority given the conditions in the question.
Second part question is also incorrectas per the above discussion, Mr. Wilmington has
already received one emergency exposure and is therefore prohibited from receiving another
(until the required medical evaluations/surveillances are performed, which have not happened
as stated in the question stem).
Plausible because - Section 4.2 of 0PEP-03.7.6 stated the following: The Site Emergency
Coordinator is responsible for authorization of exposures in excess of 10CFR20 limits and
Page: 188 of 193 5/4/2017
approval of the administration of potassium iodide (KI) for station ERO personnel performing
onsite functions. Furthermore, section 4.2.1 of 0PEP-03.7.6 stated the following: The Site
Emergency Coordinator is also responsible for authorization for personnel performing offsite
functions whenever the EOF has not been activated. In other words, the first part answer of
this distractor is plausible because it is the correct answer in every other potential case, except
the one exception listed in the procedure. Furthermore, there are many more potential cases
for the Site Emergency Coordinator to authorize emergency exposure limits that it is for the
Emergency Response Manager. Finally, as a general rule, the Site Emergency Coordinator is
normally the highest possible authority during an emergency activationtherefore, it is
plausible that this position would be responsible to authorize the emergency dose in these
conditions. The second part question is plausible because, aside from the previous emergency
exposure, every other requirement is met for Mr. Wilmington. Specifically, plausibility of this
distractor is enhanced by reference to the individuals age, the fact that he has volunteered,
the fact that he has received all required briefings, and the fact that his proposed overall dose
for both activities - 10.90 REM - is less than the limit (25 REM TEDE) for an emergency
activity required to protect a large population.
Choice D: Incorrect because - First part distractor is incorrect; see above discussions.
Plausible because - Second part distractor is correct. See above discussions for plausibility
explanation of first part distractor..
SRO Basis: Radiation hazards that may arise during normal and abnormal situations, including
maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)]
Page: 189 of 193 5/4/2017
99. Unit One reactor steam dome pressure momentarily reached 1350 psig during a
pressure control transient.
[Reference Provided]
Which one of the following completes the statement below?
IAW 0OI-01.07, Notifications, Attachment 1, Reportability Evaluation Checklist, a
reportable event has occurred due to Item #:
A. 1.1
B. 2.1
C. 3.1
D. 3.2
K/A:
G2.4.30 Knowledge of events related to system operation/status that must be l reported to internal
organizations or external agencies, such as the State, l the NRC, or the transmission system
operator. (CFR: 41.10 / 43.5 / 45.11)
RO/SRO Rating: 2.7/4.1
Tier 3
K/A match: The question matches the K/A at the SRO level by determining the required notification time
based on reportable events that occur at the site.
Pedigree: New
Objective: LOI-CLS-LP-201-D1 Objective 6
Q Reference: 0OI-01.07, Tech Specs (2.0 safety limits), NUREG-1022
Ref provided: 0OI-01.07, Attachment 1, Reportability Evaluation Checklist
Cog Level: High
Explanation: Safety limit 2.1.2 was exceeded, a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report is required IAW 0OI-01.07 item # 2.1. Note
1 of Item #2.1 clarifies invoking this requirement in the event of a Safety Limit violation.
Distractor Analysis:
Choice A: Incorrect because - a departure from Technical Specifications is not outlined in the
question stem. Plausible if the applicant incorrectly evaluates the given condition as a
deviation of Technical Specifications in accordance with 10 CFR 50.54(X).
Choice B: Correct, See explanation.
Choice C: Incorrect because- the event specified results in a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reportable event based on
exceeding a safety limit. Based on NUREG-1022 discussion concerning the criteria of Item
3.1, it is noted that items reportable under 50.72(b)(2) do not also require reporting under
50.72(b)(3). Plausible if the applicant incorrectly evaluates the condition as an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report
requirement for serious degradation of a principal safety barrier.
Choice D: Incorrect because - the event specified results in a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reportable event based on
exceeding a safety limit. Based on NUREG-1022 discussion concerning the criteria of Item
Page: 190 of 193 5/4/2017
exceeding a safety limit. Based on NUREG-1022 discussion concerning the criteria of Item
3.1, it is noted that items reportable under 50.72(b)(2) do not also require reporting under
50.72(b)(3). Plausible if the applicant incorrectly evaluates the condition as an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report
requirement for being in an unanalyzed condition.
SRO Basis: Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Page: 191 of 193 5/4/2017
100. Following an earthquake, the Unit Two RB AO has reported a lowering level in the
spent fuel pool.
Which one of the following completes both statements below IAW SCCP?
The Unit Two CRS will direct the AO to perform (1) to stage the fire water spent
fuel pool spray equipment.
Spent fuel pool sprays are commenced before spent fuel pool level reaches a
minimum of (2) .
A. (1) 0EOP-01-SEP-12, Fuel Pool Level Control
(2) 17 feet 3 inches
B. (1) 0EOP-01-SEP-12, Fuel Pool Level Control
(2) 19 feet 11 inches
C. (1) 0EDMG-002, Portable Sprays and Enhanced Ventilation Under Conditions of
Extreme Damage
(2) 17 feet 3 inches
D. (1) 0EDMG-002, Portable Sprays and Enhanced Ventilation Under Conditions of
Extreme Damage
(2) 19 feet 11 inches
Page: 192 of 193 5/4/2017
K/A:
G2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational
effects. (CFR: 41.10 / 43.5 / 45.13)
RO/SRO Rating: 3.8/4.0
Tier 3
K/A match: The applicant is tasked with identifying the correct procedure to direct AO actions for
alignment of Spent Fuel Pool Spray and the resultant effect of lowering Spent Fuel Pool level
with respect to initiating spray.
Pedigree: New
Objective: LOI-CLS-LP-300-M, Obj. 8e Given plant conditions and the Secondary Containment Control
Procedure, determine if any of the following are required: Supplemental Spent Fuel Pool per
SEP-12 and EDMG-002.
Q Reference: SCCP, rev 10; SEP-12, Rev 1; EDMG-002, rev 1; TS amend 233
Ref provided: None
Cog Level: Fundamental
Explanation: Field actions to stage equipment for use during Fuel Pool Spray is carried out IAW SEP-12.
The commencement of Fuel Pool Spray begins when Fuel Pool water level reaches a
minimum of 17 feet 3 inches.
Distractor Analysis:
Choice A: Correct, See explanation.
Choice B: Incorrect because - First half answer is correct. Second half answer is incorrect. Plausible
as 19 11 corresponds to the TS minimum level associated with Spent Fuel Pool water
level.
Choice C: Incorrect because- First half answer is incorrect. While EDMG-002 contains the actions for
lining up the Fire Truck for fuel pool spray, the actions to stage equipment for use is contained
in SEP 12. Plausible as both EOP contingency procedures contain procedural sections
relevant to performance of spent fuel pool spray. Second part is correct.
Choice D: Incorrect because - First half answer is incorrect. While EDMG-002 contains the actions for
lining up the Fire Truck for fuel pool spray, the actions to stage equipment for use is contained
in SEP 12. Plausible as both EOP contingency procedures contain procedural sections
relevant to performance of spent fuel pool spray. Second half answer is incorrect. Plausible as
19 11 corresponds to the TS minimum level associated with Spent Fuel Pool water level.
SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal,
abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
Page: 193 of 193 5/4/2017
Appendix D Scenario Outline Form ES-D-1
Facility: Brunswick Scenario No.: NRC-1 Rev 1 Op-Test No.: 2017-301
Developer: J. Viera
Technical Review: Validators: _____________________________
__________________________ _____________________________
Facility Representative: _____________________________
__________________________
Initial Conditions: Unit 2 Reactor Power is 75% RTP.
Turnover: Increase Reactor Power to 80% using Recirculation Flow in accordance with 0GP-12.
Event Malf. Event Event
No. No. Type* Description
Increase Reactor Power via recirculation to 80% RTP IAW 0GP-12
1 n/a R (ATC)
C (ATC) Recirculation Pump 2A seal failure
2 1
TS (SRO) (2APP-A-06, 2APP-A-4, 0AOP-14.0, 2OP-02, TS 3.4.1)
Remove one RFP from service
3 n/a N (BOP)
Failure of the Idled RFP Speed Controller in MAN
4 2 I (BOP)
(2OP-32)
Failure of 2-G31-TS-N008 upscale with failure of 2-G31-F004 to
5 3, 4 I (ATC)
automatically isolate (2-APP-A-02, 4-6)
C (BOP) Failure of EFCV 2-B21-IV-2455
6 5
TS (SRO) (2APP-UA-24, 0OI-44, TS 3.6.1.3, TRM)
Spurious Reactor SCRAM/ATWS (25% power)
7 6 M
Terminate and Prevent RPV injection within 120 seconds of ATWS
T (BOP)
initiation (0OI-37.5, Section 5.4)
Inadvertent MSIV closure
8 7 n/a
Failure of SLC Control Switch to initiate SLC pumps
9 8 n/a
10 9 C (ATC) Inadvertent trip of running CRD pump(s) (0EOP-01-LEP-02)
T (ATC) Lower Power < 5% before exceeding HCTL (0OI-37.8, Section 5.6)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, Critical (T)ask
Event 1: Increase Reactor Power via recirculation to 80% RTP IAW 0GP-12
- From an initial power of 75% RTP, the operating crew is directed to increase
reactor power to 80%. Expect applicants to use a provided copy of 0GP-12 and
2OP-02, Section 6.1.3.
- Once reactor power is steady at 80%, the Chief Examiner can move to Event 2.
Event 2: Recirculation Pump 2A seal failure
(2APP-A-06, 2APP-A-4, 0AOP-14.0, 2OP-02, TS 3.4.1)
- The ATC operator will respond to receipt of annunciator 2-A-6 (6-3), PUMP A
SEAL STAGING FLOW HIGH/LOW. 2 minutes following failure of the inner seal,
the outer seal will fail causing receipt of annunciator 2-A-6 (5-3), OUTER SEAL
LEAKAGE FLOW DETECTION HI, and 2-A-4 (1-2), DRYWELL FLR DR SUMP
LVL HI. The ATC is expected to diagnose and report failure of both the inner and
outer RR pump seals.
and transition to 2OP-02, Section 6.3.1, Transferring from Two-Pump, Two-Loop
Operation to One-Pump, One-Loop Operation in conjunction with SRO direction.
- Insertion of Control Rods may be required following the securing of RR Pump 2A
depending on N/F map location. [Ensure starting reactor power is altered to
minimize or prevent adjustment of control rod position.]
- Once the plant is stable in single loop with TSs addressed or upon Chief
Examiner direction if the applicant elects not to pursue TSs at this time, the Chief
Examiner can move to Event 3.
Event 3: Remove one RFP from service
choice). Ensure a cue is present for the simulator booth to provide this direction if
required by the Chief Examiner.
- Once the desired idle RFP reaches 1000 rpm, Event 4 is automatically triggered.
Event 4: Failure of the Idled RFP Speed Controller in MAN
(2OP-32)
- The BOP operator will respond to an unexpected speed increase of the idled
RFP speed controller. The BOP is expected to diagnose the failure of the idled
RFP speed controller and trip the pump in accordance with 2OP-32, Section
6.2.1, Step 27.
- Once the idled RFP has been tripped, the Chief Examiner can move to Event 5.
Event 5: Failure of 2-G31-TI-R607, RWCU TEMP, indication upscale
(2-APP-A-02)
- The ATC operator will respond to upscale failure of 2-G31-TS-N008 following
receipt of annunciator 2-A-2 (4-6). The applicant is expected to diagnose failure
of the temperature switch and respond to failure of the 2-G31-F004 to
automatically close. Ensure booth operator instructions specify that mechanical
maintenance will examine the causes of these failures with no time estimate for
completion.
- Once diagnosis of 2-G31-TS-N008 and closure of the 2-G31-F004 are complete,
the Chief Examiner can move to Event 6.
Event 6: Failure of EFCV 2-B21-IV-2455
(2APP-UA-24, 0OI-44, TS 3.6.1.3, TRM)
- The BOP operator will respond to receipt of annunciator 2-UA-24 (1-4). The BOP
operator is expected to confirm EFCV status as closed by visual indication and to
report potential TS impact per APP. The ATC operator is expected to confirm
EFCV status by observing downscale indication on 2-B21-LI-R604A, REACTOR
WATER LEVEL (P-601). Ensure the simulator booth reports no abnormal
secondary containment building conditions when called during BOP completion
of 0OI-44.
- The SRO is expected to enter TS LCO 3.6.1.3, Condition C (PCIV designation
found in TRM table 3.6.1.3-1, page 3).
- If the SRO applicant does not pursue a TS call at this time, provide the following
report from mechanical maintenance to reset the EFCV. If the SRO applicant
addresses TSs and is complete, ensure simulator booth has instruction to direct
crew to perform a reset of the EFCV. Insert a Chief Examiner cue for this
decision point.
the EFCV. Once action to reset the EFCV is complete, the Chief Examiner can
move to Event 7.
Event 7: Spurious Reactor SCRAM/ATWS (25% power)
- Following initiation of the spurious Reactor SCRAM, the crew is expected to
enter and carry out actions in 2EOP-01-RSP. Expect the ATC operator to report
ATWS conditions. Expect the crew to transition to 2EOP-01-ATWS with the BOP
operator maintaining directed level/pressure bands. (Ensure inclusion of hard
card actions)
- Expect the ATC operator to transition to 0EOP-01-LEP-02 for control rod
insertion. The ATC is expected to pursue Sections 2.2, 2.4, and 2.5 (requires
CRD pumps). Expect the ATC to request jumper insertion for bypass of
automatic SCRAM signal per Section 2.3 (simulator operator will receive ATC
report for jumper installation, jumpers not installed for this scenario).
injection. (Critical Task to be performed with 120 seconds of ATWS initiation per
0OI-37.5, Section 5.4) Ensure time blocks are available in the scenario guide to
track applicant times for performance of this critical task.
Event 8: Inadvertent MSIV closure
closure are different to require unsuccessful bypass of automatic SCRAM signal).
Expect the crew to enter 0EOP-02-PCCP.
Event 9: Failure of SLC Control Switch to initiate SLC pumps
SLC pumps, efforts are unsuccessful. Expect SRO to direct SLC injection using
0EOP-01-LEP-03 (simulator operator will receive ATC/BOP report for field
actions, actions not taken for this scenario).
Event 10: Inadvertent trip of running CRD pump(s)
- Following the full insertion of the first Control Rod during performance of 0EOP-
01-LEP-02, all running CRD Pumps inadvertently trip. The ATC operator is
expected to recognize this malfunction and restart CRD (either pump can be
successfully started).
- Expect the ATC to continue rod insertion IAW 0EOP-01-LEP-02 and to lower
power < 5% before meeting Emergency Depressurization criteria on HCTL
exceedance (0EOP-02-PCCP). (Critical Task is met if HCTL is not exceeded
during control rod insertion to < 5% power, due to not challenging the mitigation
strategy employed for protection of containment) (0OI-37.8, Section 5.6)
- Examiner evaluation of this critical task in the simulator is required to ensure
HCTL is challenged during rod insertion for use as a Critical Task.
Once reactor power is < 5% the Chief Examiner can terminate this scenario.
Appendix D Scenario Outline Form ES-D-1
Facility: Brunswick Scenario No.: NRC-2 Rev 0 Op-Test No.: 2017-301
Developer: J. Viera
Technical Review: Validators: _____________________________
__________________________ _____________________________
Facility Representative: _____________________________
__________________________
Initial Conditions: Unit 2 is pressurized at the POAH. Following recent testing the 2B RHR Loop is in
Torus Cooling.
Turnover: Place the Master RFPT Controller to AUTO in accordance with 0GP-02, Section 6.3, Step
46. Once complete, continue power ascension using control rod withdrawal in accordance with 0GP-
03, Section 6.1, Step 5.
Event Malf. Event Event
No. No. Type* Description
Place the Master RFPT Controller to AUTO
1 n/a N (BOP)
(0GP-02)
I (ATC) RWM inoperable during Reactor Startup.
2 1
TS (SRO) (0GP-03, 0GP-10, 2OP-07, 2APP-A-05, 0GP-11, TS 3.3.2.1)
Increase Reactor Power using control rod withdrawal.
3 n/a R (ATC)
Respond to annunciator 2-UA-03 (4-3), OFF GAS FILTER DIFF-
4 2 C (BOP)
HIGH (2APP-UA-03, 2OP-30)
I (ATC) Respond to downscale failure of 2-B21-LT-N031D-3
5 3
TS (SRO) (2APP-A-03, 2OP-17, TS 3.3.5.1, TRM)
Respond to annunciator 2-UA-06 (4-1), BUS COMM B OVCT DIFF
6 4 C (BOP)
L/O, with breaker failure (2APP-UA-06)
SDV Leak (LOCA) with trip of running RFP and CRD Pumps
7 5, 6, 7 M
(2EOP-01-RSP, 2EOP-01-RVCP, 0EOP-03-SCCP)
8, 9,
8 n/a Failure of HP injection sources (2EOP-01-RVCP)
10
Restore RPV injection using a single RFP prior to TAF
9 T (ATC)
ED when > 2 areas above max safe
10 T (BOP)
(2APP-UA-12, 0EOP-03-SCCP, 0EOP-01-EDP)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, Critical (T)ask
Event 1: Place the Master RFPT Controller to AUTO
(0GP-02, Section 6.3, Step 46)
- The scenario should start with plant conditions established to perform 0GP-02,
Section 6.3, Step 46. If necessary, a pre-marked up copy of 0GP-02 should be
available.
- The crew will move to Event 2 by continuing the reactor startup.
Event 2: RWM inoperable during reactor startup
(0GP-03, 0GP-10, 2OP-07, 2APP-A-05, 0GP-11, TS 3.3.2.1)
- The applicant crew will continue a reactor startup using a copy of 0GP-03,
Section 6.1, pre-marked to Step 5. Appropriate rods already pulled will be
provided on a pre-marked up copy of 0GP-10, using the A2X(?) sequence.
- Following withdrawal of the second rod, the ATC will respond to annunciation of
2-A-5 (5-2) with the RWM ODA displaying UNKNOWN (mid-display will be
blank with no soft key selections available).
- If consulted as Reactor Engineer or Shift Management, notify the crew that they
are to continue the reactor startup.
- It is expected that the operating crew will transition to using 0GP-11 (from 0GP-
10) to continue the reactor startup. When plant management permission is
requested per 0GP-11 (4.1) to bypass the RWM, ensure permission is given. The
ATC is expected to bypass the RWM prior to continuing the startup.
- The SRO is expected to enter TS LCO 3.3.2.1, Condition C.
- The crew will move to Event 3 by continuing the reactor startup.
Event 3: Reactor Startup
- The operating crew will continue to withdraw rods to raise reactor power. Once a
sufficient power change has been observed, the Chief Examiner can transition to
Event 4.
- It is not desired to proceed to the point of placing the MODE switch in RUN (15%
Event 4: Respond to annunciator 2-UA-03 (4-3), OFF GAS FILTER DIFF-HIGH
(2APP-UA-03, 2OP-30, Section 6.3.3)
- The BOP operator will respond to receipt of annunciator 2-UA-03 (4-3) and is
expected to perform 2OP-30, Section 6.3.3, Placing the Off Gas Standby Filter in
Service (as directed by the APP). Ensure instructions exist for booth operator to
report completion of Unit 1 steps (2 - 4) and Unit 2 field steps (5 - 6) when
prompted by applicant. Per Chief Examiner direction, this event can be run
concurrently with Event 5.
Event 5: Respond to downscale failure of 2-B21-LT-N031D-3
(2-APP-A-03, 2OP-17, TS 3.3.5.1, TRM)
- Ensure that a 2OP-17, Section 6.1.9, procedure in progress is available for the
applicant crew at scenario start (Step 14 should remain open). Ensure that Torus
temperature at scenario start is high but < 90oF.
- The ATC operator will respond to receipt of annunciator 2-A-03 (3-2), AUTO
DEPRESS RELAYS ENERG, caused by failure of 2-B21-LT-N031D-3. The ATC
operator is expected to secure Torus cooling per APP direction in accordance
with 2OP-17, Section 6.2.3.
- Following shutdown of RHR pumps with corresponding clearing of annunciator 2-
A-03 (3-2), the operating crew is expected to diagnose the cause of the
annunciator as being due to a failure of 2-B21-LT-N031D-3 with an effect of ADS
Trip System A inoperability. (Devices contained within APP and B RHR Loop
pumps secured)
- The SRO is expected to enter TS LCO 3.3.5.1 Conditions A and E. ADS
Instrumentation designation can be found in TRM Table 3.3.5.1-1.
- Per Chief Examiner direction, this event can be run concurrently with Event 4.
Once Torus Cooling with the B RHR Loop is secured with TSs addressed or
upon Chief Examiner direction if the applicant elects not to pursue TSs at this
time, the Chief Examiner can move to Event 6.
Event 6: Respond to annunciator 2-UA-06 (4-1), BUS COMM B OVCT DIFF L/O, with
breaker failure
- The BOP operator will respond to receipt of annunciator 2-UA-06 (4-1) and is
expected to diagnose failure of 2-COM-B-AA2, SAT TO BUS COMMON B, to trip
as expected. The applicant is expected to trip this breaker and de-energize bus
Common B.
- There are no TS associated with de-energizing this bus. Expect receipt of
multiple alarms upon de-energizing this bus.
- Once bus COMMON B is de-energized, the Chief Examiner can move to Event
7.
Event 7: Scram Discharge Volume (SDV) Leak (LOCA) with trip of running Reactor
Feed Pump (RFP) and Control Rod Drive (CRD) Pumps
(2EOP-01-RSP, 2EOP-01-RVCP, 0EOP-03-SCCP)
automatic reactor SCRAM is expected when level lowers < 166. At the time of
the SDV leak, the running RFP and CRD Pumps will inadvertently trip.
- It is expected for the SRO to enter 2EOP-01-RSP.
water level < 166 based on indications &/or alarms. It is expected for the BOP
operator to report receipt of annunciator(s) 2-UA-12 (2-1) NORTH CS RM
FLOOD LEVEL HI (after a period of time also expect (2-2), NORTH RHR RM
FLOOD LEVEL HI)
- It is expected for the SRO to transition to both 2EOP-01-RVCP and 0EOP-03-
SCCP.
Event 8: Failure of HP injection sources
one (or all) of the following sources (Table L-1), all attempts will be unsuccessful.
o CRD Pumps - Pump control switch actuation will be unsuccessful
o RCIC - Will not actuate on LL-2 and control switch actuation will be
unsuccessful
o HPCI - Will not actuate on LL-2 and control switch actuation will be
unsuccessful
- It is not expected for the SRO to invoke availability of Table L-2 alternate
injection subsystems. If any of these systems are requested, the simulator
operator will acknowledge the report. Any attempted use of SLC injection will be
unsuccessful.
o SLC Pumps - Pump control switch actuation will be unsuccessful
Event 9: Restore RPV injection using a single RFP prior to Top of Active Fuel (TAF)
- If maintenance investigation of the tripped RFP is requested, the simulator
operator will acknowledge the request but provide no follow-up report.
RFP or with injection of the idle RFP. (Critical Task is met if RPV water level is
maintained > TAF with a RFP in service following initiation of Event 7, due to not
challenging the RPV level control mitigation strategy)
- Chief Examiner discretion will be utilized to deliver the second area Max Safe
water level limit. This is to permit sufficient time for applicants to attempt RPV
injection prior to ED criteria.
Event 10: ED when > 2 areas above max safe
(2APP-UA-12, 0EOP-03-SCCP, 0EOP-01-EDP)
- The BOP operator is expected to continue performance of 2APP-UA-12 actions
for received annunciators (no actions apart from field reports expected).
- If any crew member requests field actions to perform a SCRAM reset, the
simulator operator will acknowledge the report. (no actions taken to reset the
- The operating crew may elect to Anticipate Emergency Depressurization. There
will be no failures associated with crew performance of Anticipation.
- Chief Examiner discretion will be utilized to deliver the second area Max Safe
water level limit. This is to permit sufficient time for applicants to attempt RPV
injection prior to ED criteria.
second area Max Safe water level limit is reached. (Critical Task is met when the
crew performs an Emergency Depressurization upon receipt of 2 area water
levels above Max Safe limits)
Depressurization.
Once the applicant crew has initiated an Emergency Depressurization in accordance
with 0EOP-01-EDP the Chief Examiner can terminate this scenario.
Appendix D Scenario Outline Form ES-D-1
Facility: Brunswick Scenario No.: NRC-3 Rev 0 Op-Test No.: 2017-301
Developer: J. Viera
Technical Review: Validators: _____________________________
__________________________ _____________________________
Facility Representative: _____________________________
__________________________
Initial Conditions: Unit 2 is operating at 90% RTP. HPCI is currently under clearance with return-to-
service expected on the next shift. Level Instrument N026B is currently INOP due to I/C maintenance.
Turnover: Place the 2A SJAE in FULL LOAD in accordance with 2OP-30, Section 6.3.1.
Event Malf. Event Event
No. No. Type* Description
Place the 2A SJAE in FULL LOAD
1 n/a N (BOP)
(2OP-30)
C(ATC) Inward Rod Drift
2 1, 2
TS (SRO) (2APP-A-05, 0AOP-02.0, 2OP-07, TS 3.1.3)
Respond to receipt of 2-UA-14 (5-1), DEMIN WATER XFR PUMP
3 3 C (BOP)
HDR PRESS LOW/PWR LOSS (2APP-UA-14)
C (ATC) Respond to receipt of 2-A-3 (3-6), CORE SPRAY A HI PRESS
4 4, 5
TS (SRO) VALVE LEAK (2APP-A-3, TS 3.6.1.3, TS 3.5.1, TRM)
Failure of 2-HTOG-V2
5 6 C (BOP)
(2APP-UA-44, 0AOP-37.0, 2OP-30)
Lowering condenser vacuum due to failure of 2-HTOG-V2
6 n/a R (ATC)
(0AOP-37.0)
Small piping break (Steam Space LOCA)
7 7 M
(0AOP-14.0, 2APP-A-05, 2EOP-01-RSP)
Degraded RPV Level indicators
8 8 C (ATC)
(2EOP-01-RSP, 0EOP-02-PCCP, 0EOP-01-RXFP)
9 T (All) Transition to 0EOP-01-RXFP (0EOP-01-RXFP)
10 9 C (BOP) Only one SRV available for depressurization (0EOP-01-RXFP)
Rapid depressurization required to lower RPV pressure to within
11 T (BOP)
100 psig of Torus pressure (0EOP-01-RXFP)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, Critical (T)ask
Event 1: Place the 2A SJAE in FULL LOAD
(2OP-30, Section 6.3.1)
expected to perform 2OP-30, Section 6.3.2.
- Once the 2A SJAE is in FULL LOAD, the Chief Examiner can move to Event 2.
Event 2: Inward Rod Drift
(2APP-A-05, 0AOP-02.0, 2OP-07, TS 3.1.3)
- Once initiated, a low Rod Worth Control Rod will begin to slowly drift in. The ATC
operator is expected to respond to receipt of annunciator 2-A-5 (3-2), ROD
DRIFT, and take action per the APP to arrest the rod motion using a momentary
WITHDRAW signal using the RMCS IAW APP Step 2.a.1. It is expected for
correct operator action to be successful in arresting the rod motion. There are no
TS entry criteria met following the ROD DRIFT with successful arrest.
- It is expected for the operating crew to enter 0AOP-02.0 (no operator actions
required). When the simulator staff is contacted for further direction as Reactor
Engineering per 0AOP-02.0, inform the applicant crew to fully insert the rod. A
pre-marked 0GP-10 rod pull sheet will be required to deliver to applicants.
(NOTE: Applicants may elect to attempt rod motion IAW 2APP-A-05 (3-2), Step
5.a.3, without RE consultation.)
- Once rod motion is attempted, rod motion will not be possible. Operators will
attempt actions to insert rod IAW 2OP-07, Section 6.3.3. All attempts at rod
insertion will be unsuccessful.
3.1.3, Condition A for this rod.
- Once TSs are addressed or upon Chief Examiner direction if the applicant elects
not to pursue TSs at this time, the Chief Examiner can move to Event 3.
Event 3: Respond to receipt of 2-UA-14 (5-1), DEMIN WATER XFR PUMP HDR
PRESS LOW/PWR LOSS
- Once initiated, a DEMIN HDR low pressure condition will initiate with a failure of
the AUTO pump to start. The pump running at scenario start will remain in
operation.
- Expect BOP response IAW annunciator 2-UA-14 (5-1) to manually start the pump
selected to the AUTO position. Once this pump is started, system pressure will
return to an expected two-pump operation value.
- Once this annunciator is clear with two pumps in operation, the Chief Examiner
can move to Event 4. Consideration should be given to performing Events 3 and
4 concurrently to eliminate applicant overlap potential.
Event 4: Respond to receipt of 2-A-3 (3-6), CORE SPRAY B HI PRESS VALVE LEAK
(2APP-A-3, TS 3.6.1.3, TS 3.5.1, TRM)
- The ATC operator is expected to respond to this annunciator and carry out
actions IAW 2APP-A-3. Once pressure has been relieved and the applicant
performs APP Step 5, the 2-E21-F004B valve breaker will trip open on thermal
overload (valve will remain fully closed). (Per facility conduct of operations, one
time reset of thermal overload is not permitted unless plant is in emergency)
- Expect SRO to identify entry into TS 3.5.1, Conditions A and E and TS 3.6.1.3,
Condition A. (Require facility input to determine if TS 3.6.1.3, Condition B is
required due to original F005B/F006B seat leakage.) Ref TRM Table 3.6.1.3-2 for
PCIV assignment.
- Once TSs are addressed or upon Chief Examiner direction if the applicant elects
not to pursue TSs at this time, the Chief Examiner can move to Event 5.
Consideration should be given to performing Events 3 and 4 concurrently to
minimize applicant overlap potential.
Event 5: Failure of 2-HTOG-V2
(2APP-UA-44, 0AOP-37.0, 2OP-30)
- Expect BOP operator to respond to receipt of annunciator 2-UA-44 (5-1), SJAE
DISCHARGE PRESS HIGH, and identify inadvertent closure of 2-HTOG-V2.
- Upon determination that condenser vacuum is lowering, expect the SRO to direct
entry into 0AOP-37.0.
transition to 2OP-30, Sections 6.3.2 and 6.3.7 to shutdown the 2A SJAE.
- Events 5 and 6 are concurrent events. Once plant conditions are stable with the
2B SJAE in service in FULL LOAD with a power reduction > 5% power
completed, the Chief Examiner can move to Event 7.
Event 6: Lowering condenser vacuum due to failure of 2-HTOG-V2
(0AOP-37.0)
- With indications of lowering condenser vacuum, expect the ATC operator to carry
out immediate actions of 0AOP-37.0 to lower reactor power IAW 0ENP-24.5 to
maintain condenser vacuum.
- The ATC operator is expected to pursue either a recirculation flow reduction or a
manual runback to lower power.
- Ensure that 22.4 inches of vacuum (or the critical parameter established by the
SRO) is not challenged and that power reduction offers sufficient time for the
BOP operator to restore the 2B SJAE to FULL LOAD.
- Events 5 and 6 are concurrent events. Once plant conditions are stable with the
2B SJAE in service in FULL LOAD with a power reduction > 5% power
completed, the Chief Examiner can move to Event 7.
Event 7: Small piping break (Steam Space LOCA)
(0AOP-14.0, 2APP-A-05, 2EOP-01-RSP)
- When entered, a small piping break on Wide Range Level instrument N026A
(Steam Space LOCA) will commence. Operators will receive indications of
degrading conditions in the Drywell, specifically:
o Rising DW pressure and temperature indication
o Receipt of annunciator 2-A-5 (5-5), PRI CTMT HI/LO PRESS
temperature indications and insert a manual reactor SCRAM based on High
Drywell pressure. Operators are expected to enter 2EOP-01-RSP.
Event 8: Degraded RPV Level indicators
- As a result of the steam break, all in-service RPV level indications should
indicate reference leg flashing (level oscillations should be severe enough to
negate their use).
- Following immediate actions required by 2EOP-01-RSP, Operators are expected
to enter 0EOP-01-RXFP. 0EOP-02-PCCP entry criteria (required at 1.7 psig DW
pressure) should not be met at this point.
Event 9: Transition to 0EOP-01-RXFP
- Operating crew is expected to determine that RPV level cannot be determined
based on plant conditions and transition to 0EOP-01-RXFP. (Critical Task is met
if Operating crew transitions to 0EOP-01-RXFP without transition to 2EOP-01-
RVCP (required at 1.7 psig DW pressure), due to not challenging the RPV level
and pressure control mitigation strategies)
Event 10: Only one SRV available for depressurization
- Once step RXFP-14 is directed, Operators will determine that only one SRV can
be opened. All other SRVs are failed closed.
Event 11: Rapid depressurization required to lower RPV pressure to within 100 psig of
Torus pressure
- Operators are expected to utilize step RXFP-18 to rapidly depressurize the RPV.
(Critical Task is met if Operating crew utilizes any available system in Table P-2
to effect a rapid depressurization in accordance with step RXFP-18.)
Once the applicant crew has performed a rapid depressurization in accordance with
step RXFP-18 to lower RPV pressure to within 100 psig of Torus Pressure, the Chief
Examiner can terminate this scenario.
Appendix D Scenario Outline Form ES-D-1
Facility: Brunswick Scenario No.: NRC-4 Rev 0 Op-Test No.: 2017-301
Developer: J. Viera
Technical Review: Validators: _____________________________
__________________________ _____________________________
Facility Representative: _____________________________
__________________________
Initial Conditions: Unit 2 Reactor Power is at 55% RTP with 2PT-40.2.12 currently in progress. Testing
of BPVs 1 through 7 have been completed.
Turnover: Complete 2PT-40.2.12 beginning with Bypass Valve #8, Section 7.11, Step 1. Following PT
completion, perform reactor power reduction IAW 0GP-12 to 50% RTP using control rods.
Event Malf. Event Event
No. No. Type* Description
Perform 2PT-40.2.12, Section 7.11.
1 n/a N (BOP)
(2PT-40.2.12)
BPVs 9 and 10 Stuck Shut
2 1, 2 TS (SRO)
Lower power using control rods
3 n/a R (ATC)
(0GP-12)
Failure of DWED pump to run
4 3 C (ATC)
CW Lube Water Header Leak
5 4 C (BOP)
Failure of CRD Flow Controller in AUTO
6 5 C (ATC)
C (BOP) Loss of Stack Rad Monitor, failure of RB Ventilation isolation
7 6, 7
TS (SRO) (previously used)
8, 9, LOOP (Dual Unit) with EDG 1 Lockout and EDG 3 failure in AUTO
8 M
10 (2EOP-01-RSP, 0AOP-36.1)
Unit 1 requests to cross-tie E1 to E3
9 n/a n/a
(0AOP-36.1)
Operate EDG 3 to not exceed > 3850 kW loading for 2 minutes
10 T (BOP)
(0AOP-36.1)
Failure of EDG 3 and 4 (Unit 2 SBO)
11 11, 12 n/a
(2EOP-01-RSP, 2EOP-01-RVCP, 2EOP-01-SBO)
100o F/hr cooldown rate initiated within 60 minutes of SBO start
12 T (ATC)
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, Critical (T)ask
Event 1: Perform 2PT-40.2.12, Section 7.11
(2PT-40.2.12)
- Ensure an in-progress 2PT-40.2.12 is pre-marked up to Section 7.11 with a
complete copy available to the applicant crew for pre-review.
Bypass Valve Number 8. Upon conclusion of this section, the applicant crew will
proceed to Event 2.
Event 2: BPVs 9 and 10 Stuck Shut
7.13, to test BPVs 9 and 10. During performance, the BOP operator is expected
to discover that both BPVs 9 and 10 are stuck fully shut and will not open upon
RTGB demand.
- Facility consultation required to ensure correct TS application, as the PT
specifies failure of two BPVs will result in an INOPERABLE BPV system (ref
Section 5.0 NOTE on page 6) yet TS Bases 3.7.6 specifies that failure of three
BPVs will result in an INOPERABLE BPV system.
3.7.6, Condition A.
- Once the PT is complete with TSs addressed or upon Chief Examiner direction if
the applicant elects not to pursue TSs at this time, the Chief Examiner can move
to Event 3.
Event 3: Lower power using control rods
(0GP-12)
- Ensure that initial plant configuration is consistent with 0GP-12 for starting power
level (e.g. one RFPT currently idled, one HDP secured, etc).
- Applicant crew coordination required to meet plant turnover required activities.
Ensure pre-filled 0GP-10 sheets are available at crew turnover. Expect the
operating crew to lower power to as-directed 50% reactor power. During power
reduction, applicant manipulation of HD-V57 and CO-FIC-49 may be required.
- If applicant crew contacts shift management following PT performance, ensure a
Simulator Booth cue is available to instruct the applicant crew to complete the
shift turnover direction to lower power.
- Following observation of sufficient power reduction, the Chief Examiner can
move to Event 4.
Event 4: Failure of DWED pump to run
- Following receipt of annunciator 2-A-4 (1-1), the ATC operator is expected to
diagnose a failure of the DWED pump to run. The ATC is then expected to start a
DWED pump IAW the automatic actions of annunciator 2-A-4 (1-1).
- Upon initiation of this event, the Chief Examiner can move directly to Event 5.
Event 5: CW Lube Water Header Leak
- With an initial scenario configuration of the 2A Bearing Lube Water Pump in
service (and an appropriate number of CWOD pumps in service), initiation of this
event will prompt receipt of annunciator 2-UA-24 (6-7), CW DISCH PMP LUBE
WATER PRESSURE LOW. The BOP operator is expected to start the 2B
Bearing Lube Water Pump IAW the APP (no auto start enabled).
- After one minute, the BOP operator will respond to annunciation of 2-UA-24 (5-
7), CW DISCH PMP LUBE WATER FLOW LOW. The BOP operator is expected
to diagnose that a break in the lube water header has occurred with a failure of
the running CWOD pump(s) to trip. The BOP operator is expected to trip all
running CWOD and Bearing Lube Water pumps IAW the APP.
- Once all CWOD and Bearing Lube Water Pumps are tripped, the Chief Examiner
can move to Event 6.
Event 6: Loss of CRD Flow Controller in AUTO
flow) will result in elevated HCU temperatures causing receipt of alarm 2-A-5 (1-
2), CRD HYD TEMP HIGH. When a field operator is dispatched, report that HCU
(xx) is indicating approximately 400oF.
- Upon receipt of this annunciator, the ATC operator is expected to take MANUAL
control of the CRD Flow Controller and re-establish flow IAW 2OP-08 (referred
from 2APP-A-05 (1-2)).
- When the CRD Flow Controller is in MANUAL control, the Chief Examiner can
move to Event 7. It is expected that shortly after flow is re-established
annunciator 2-A-5 (1-2) will clear.
Event 7: Stack Rad Monitor Failure
(previously used)
- Loss of UPS power to the Stack Rad Monitor will initiate a Group 6 isolation.
Group 6 valves will isolate, but RB Ventilation will fail to isolate. The BOP
operator is expected to perform the manual isolation required.
- This is a scenario event previously used on the 2014-301 exam (ref scenario
LOIX-032, Event 4)
- Once RB Ventilation actions are complete with TSs addressed or upon Chief
Examiner direction if the applicant elects not to pursue TSs at this time, the Chief
Examiner can move to Event 8.
Event 8: LOOP (Dual Unit), Loss of EDG 1, EDG 3 failure to start in AUTO
(2EOP-01-RSP, 0AOP-36.1)
- Upon initiation, a LOOP will affect the operating crew. An automatic SCRAM and
closure of MSIVs will result due to the loss of power on Unit 2. It is expected for
the applicant crew to enter both 2EOP-01-RSP and 0AOP-36.1.
AUTO and will start EDG 3 in CR-MANUAL. The ATC operator is expected to be
manually controlling SRVs.
- Once EDG 3 is running in CR-MANUAL, the Chief Examiner can move to Event
9.
Event 9: Unit 1 requests to cross-tie E1 to E3
(0AOP-36.1)
- Upon initiation, the simulator booth will contact the applicant crew to inform them
that EDG 1 has suffered a lockout and is unavailable, additionally, Unit 1 will
request to perform a cross-tie between E1 and E3.
- It is expected that the BOP operator will perform action associated with closure of
breaker AG0 concurrent with Simulator Booth (as Unit 1 operator).
- Following completion of cross-tie, the Simulator Booth (as Unit 1 CRS) will
contact the applicant SRO and inform them that Unit 1 will be starting loads on
E1.
Event 10: Operate EDG 3 to not exceed > 3850 kW loading for 2 minutes
(0AOP-36.1)
- Following the Unit 1 notification above, the simulator booth will simulate starting
the following loads on E1 (i.e. E3). Each load should be timed in 15 second
increments.
o Batt Chgr 1A-1 and 1A-2 (128 kW)
o CS Pump 1A (940 kW)
o CRD Pump 1A (190 kW)
o FP Cooling Pump 1A (50kW)
o NSW Pump 1A (225 kW)
o RB Vent Sup Fan 1A and 1C (150 kW)
o RHR Pump 1C (750 kW)
o RBCCW Pumps 1A and 1C (96 kW)
o CSW Pump 1B (225 kW)
o RHRSW Pump 1C (600 kW)
exceed 3850 kW (including 480V house loads). (Will require validation during
prep week)
- (Critical Task is met when operator action is taken to reduce loading or to instruct
Unit 1 to reduce loading to not exceed > 3850 kW loading for 2 minutes.) Facility
consultation required to finalize bounding criteria for this Critical Task.
- Once operator action is taken to mitigate EDG 3 loading or once two minutes has
elapsed with EDG 3 loading > 3850 kW, the Chief Examiner can proceed to
Event 11.
Event 11: Failure of EDG 3 and 4 (Unit 2 SBO)
(2EOP-01-RSP, 2EOP-01-RVCP, 2EOP-01-SBO)
- Upon entry into this event, the applicant crew will receive indications of a SBO
(EDG 3 and 4 lockouts). It is expected for the SRO to transition from 2EOP-01-
RSP to 2EOP-01-SBO and 2EOP-01-RVCP.
- It is expected for the SRO to direct entry into 0EOP-01-SBO-01, 0EOP-01-SBO-
02, and 0EOP-01-SBO-04. Additionally, it is expected for the crew to carry out
2EOP-01-RVCP pressure control actions.
Event 12: 100oF/hr cooldown rate initiated within 60 minutes of SBO start
reactor pressure to a control band of 150 to 300 psig. Critical Task to be
performed within 60 minutes of entry into SBO per 0OI-37.4, Section 5.19.2.
(Critical Task is met when operator action to lower pressure results in a cooldown
rate of 100oF/hr within 60 minutes from SBO start.) Facility consultation
required to determine upper bounding of cooldown rate for this Critical Task.
the system used) with applicant control of pressure, the Chief Examiner can
terminate this scenario.
DUKE ENERGY PROGRESS
BRUNSWICK NUCLEAR PLANT
LESSON TITLE: Shift CRD pumps IAW OP-08 (alternate path)
LESSON NUMBER:
REVISION NO: 0
PREPARER/DATE
TECHNICAL REVIEWER/DATE
VALIDATOR/DATE
TRAINING SUPERVISION APPROVAL/ DATE
RELATED TASKS:
K/A REFERENCE AND IMPORTANCE RATING:
210001 Control Rod Drive Hydraulic System
A2 Ability to (a) predict the impacts of the following on the CONTROL ROD DRIVE
HYDRAULIC SYSTEM; and (b) based on those predictions, use procedures to correct, control,
or mitigate the consequences of those abnormal conditions or operations:
A2.01 Pump trips
Importance 3.2/3.3
REFERENCES:
1OP-08, Control Rod Drive Hydraulic System Operating Procedure, Rev.98
0AOP-02.0, Control Rod Malfunction/Mispsoition, Rev. 28
1APP-A-05, Annunciator Procedure for Panel A-05
TOOLS AND EQUIPMENT:
None
ADMINISTRATIVE CATEGORY (from NUREG 1123, Rev 2. Supp 1):
1 - Reactivity Control
SAFETY CONSIDERATIONS:
None.
SPECIAL INSTRUCTIONS:
None
SAFETY CONSIDERATIONS:
None
Page 2 of 12 REV. 0
EVALUATOR NOTES: (Do not read to trainee)
1. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021,
Appendix E, or similar to the trainee.
2. This JPM may be administered in the simulator.
Read the following to performer.
TASK CONDITIONS:
You have been directed to shift the operating CRD pumps.
INITIATING CUE:
Shift the CRD pumps from CRD 1A running to CRD pump 1B running.
Page 3 of 12 REV. 0
PERFORMANCE CHECKLIST
TIME START: __________
Step 1 - Obtain a copy of 1OP-08.
Standard: Obtains a copy of 1OP-08.
SAT/UNSAT
Step 2 - Review applicable precautions and limitations.
Standard: Reviews precautions and limitations.
SAT/UNSAT
Step 3 - Verifies from the initial conditions that Section 5 Prerequisites have been completed.
Standard: Completes Section 5 Prerequisites
SAT/UNSAT
Step 4 - Goes to Section 6.3.2 Shifting CRD Pumps .
Standard: Proceeds to Section 6.3.2 of 1OP-08
SAT/UNSAT
Step 5 - (Procedure step 1) Ensure the CRD System in operation per Section 6.1.1.
Standard: Signs off step per initial conditions.
SAT/UNSAT
Step 6 - (Procedure step 2) If Starting CRD 1A N/A.
Standard: Step is N/A
SAT/UNSAT
Page 4 of 12 REV. 0
Step 7 - (Procedure step 3a, b. c)
a. Ensure C11-F013B (CRD Pump 1B Suction Isolation Valve) LOCKED OPEN.
b. Ensure C11-F014B (CRD Pump 1B Discharge Isolation Valve) LOCKED OPEN.
c. Ensure C11-F015B (CRD Pump 1B Recirculation Line Isolation Valve) LOCKED OPEN.
Examiner Cue: When asked Step 3a,b,c has been completed
Standard: Step completed per the cue
SAT/UNSAT
Step 8 - (Caution) Acknowledge Caution prior to step 4
Standard: Caution Acknowledged
SAT/UNSAT
Step 9 - (Procedure step 4,5,6)
Shift C11-FC-R600 (CRD Flow Control) to BAL
Null C11-FC-R600 (CRD Flow Control) using the manual potentiometer.
Shift C11-FC-R600 (CRD Flow Control) to MAN.
Standard: C11-FC-R600 shifted to Manual in accordance with the procedure
SAT/UNSAT
Step 10 - (Procedure step 7) Set CRD flow rate to 35 gpm.
Standard: CRD Flow adjusted to 35 gpm
SAT/UNSAT
Step 11 - (Procedure step 8) Step is N/A
Standard: Mark Step as N/A
SAT/UNSAT
Step 12 - (Procedure step 9a) Start CRD Pump 1B.
Standard: CRD Pump 1B Started
- CRITICAL STEP ** SAT/UNSAT
Step 13 - (Procedure step 9b) Stop CRD Pump 1A.
Standard: CRD Pump 1A Stopped
- CRITICAL STEP ** SAT/UNSAT
Begin Alternate Path: CRD pump 1B shaft shears resulting in no flow to the CRDs
Page 5 of 12 REV. 0
Step 14 -Recognize that No CRD flow exist and enters 0AOP-02.0.
Standard: Enter 0AOP-02.0
SAT/UNSAT
Step 15 - (AOP-2.0 Immediate actions):
- Stop any power changes in progress (there are none)
- IF more than one control rod is drifting, THEN insert a manual scram (none drifting)
- IF less than or equal to 25% power, AND more than one control rod scrams, THEN
insert a manual scram (No Control Rods Have Scramed)
- IF greater than 25% power, AND nine or more control rods scram, THEN insert a
manual scram. (No Control Rods Have Scramed)
Standard: Complete IA of 0AOP-2.0 prior to control rods scraming
- CRITICAL STEP ** SAT/UNSAT
Step 16 - (Supplemental Actions Step 1, 2 and 3) Steps is not applicable at this point
Standard: Reads steps and continues on.
SAT/UNSAT
Step 17 - (Supplemental Actions Step 4) Maintain core thermal parameters within Technical
Specification limits
Standard: All parameters within limits
SAT/UNSAT
Step 18 - (Supplemental Actions Step 5) Contact the Reactor Engineer for further control rod
movement instructions
Standard: Contacts Reactor Engineer
CUE Reactor Engineer acknowledges
SAT/UNSAT
Step 19 - (Supplemental Actions Step 6) Monitor off-gas radiation
Standard: All parameters within limits
CUE Another Operator will Monitor Off Gas
SAT/UNSAT
Step 20 - (Supplemental Actions Step 7) Notify E&C to sample reactor coolant
Standard: Step is N/A there are no indications of fuel failure
SAT/UNSAT
Page 6 of 12 REV. 0
Step 21 - (Supplemental Actions Step 8 a.1, a.2) Close B32-V22 (Seal Injection Vlv) for Recirc
Pump A and Close B32-V30 (Seal Injection Vlv) for Recirc Pump B
Standard: Operator closes B32-V22 and B32-V30 (green light on Red Light off)
- CRITICAL STEP ** SAT/UNSAT
Step 22 - (Supplemental Actions Step 8 a.3) Place C11-FC-R600 (CRD Flow Control) in
MAN
Standard: Controller is already in MAN
SAT/UNSAT
Step 23 - (Supplemental Actions Step 8 a.4) Reduce C11-FC-R600 (CRD Flow Control)
potentiometer to minimum setting
Standard: Operator reduces flow controller to Min setting
- CRITICAL STEP ** SAT/UNSAT
Step 24 - (Supplemental Actions Step 8 a.5, a.6)
- Ensure C11(C12)-PCV-F003 (Drive Press Vlv) is OPEN.
Standard: No operator action is necessary all is already aligned
SAT/UNSAT
Step 25 - (Supplemental Actions Step 8 a.7) Start a CRD pump
Standard: Operator Starts the Standby CRD Pump
- CRITICAL STEP ** SAT/UNSAT
Step 25 - (Supplemental Actions Step 8 a.8) WHEN all accumulator low pressure alarms are clear,
THEN adjust CRD flow rate to between 30 and 60 gpm, using the manual potentiometer of
C11(C12)-FC-R600 (CRD Flow Control)
Standard: Operator Adjusts CRD flow to between 30 and 60 gpm
- CRITICAL STEP ** SAT/UNSAT
TERMINATING CUE: When candidate adjusts CRD flow to between 30 and 60 gpm
Page 7 of 12 REV. 0
- Comments required for any step evaluated as UNSAT.
TIME COMPLETE: __________
Page 8 of 12 REV. 0
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and Adherence (PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer SAT/ UNSAT/ NE
checking)
E. Proper Equipment Use (observe starting limitations, throttle SAT/ UNSAT/ NE
valve closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety SAT/ UNSAT/ NE
equipment, etc.)
G. Electrical Safety And Arc Flash Protection Compliance (refer to SAT/ UNSAT/ NE
SAF-NGGC-2175, as applicable)
H. Security Compliance (controlled area entry and exit, key control, SAT/ UNSAT/ NE
etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, etc.) SAT/ UNSAT/ NE
J. Radiation Protection (ALARA, understanding and use of RWP, SAT/ UNSAT/ NE
frisking, etc.)
COMMENTS:
Page 9 of 12 REV. 0
Validation Time: 10 Minutes (approximate)
Time Taken: ___ Minutes
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: N/A
Setting: In-Plant Simulator Admin X
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes No X
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
Page 10 of 12 REV. 0
Revision Summary:
REVISION REVISION SUMMARY
NUMBER
0 New
Page 11 of 12 REV. 0
TASK CONDITIONS:
You have been directed to shift the operating CRD pumps.
INITIATING CUE:
Shift the CRD pumps from CRD 1A running to CRD pump 1B running.
Page 12 of 12
DUKE ENERGY PROGRESS
BRUNSWICK NUCLEAR PLANT
LESSON TITLE: SBGT System Operations to Reduce Humidity
LESSON NUMBER: LOT-SIM-JP-010-01
REVISION NO: 01
PREPARER DATE
TECHNICAL REVIEWER DATE
LINE REVIEW/VALIDATOR DATE
Facility Representative DATE
SBGT System Operations to Reduce Humidity
Revision Summary:
REV. No. REVISION SUMMARY
1 JPM formatted for new template
LOT-SIM-JP-010-01 Page 2 of 10 Rev. 01
SBGT System Operations to Reduce Humidity
RELATED TASKS:
261003B101, Perform A Control Room Manual Startup Of SBGT Per OP-10
261006B101, Secure The SBGT System Per OP-10
K/A REFERENCE AND IMPORTANCE RATING:
288000 A4.01 3.1/2.9
Ability to manually operate and or monitor in the control room: Start and stop fans.
REFERENCES:
TOOLS AND EQUIPMENT:
None.
SAFETY FUNCTION (from NUREG 1123, Rev 2.):
9 - Radioactivity Release (Plant Ventilation Systems)
REASON FOR REVISION:
Updated to the new format
LOT-SIM-JP-010-01 Page 3 of 10 Rev. 01
SBGT System Operations to Reduce Humidity
SIMULATOR SETUP:
Initial Conditions
Rx. Pwr. Any
Core Age Any
Required Plant Conditions
None
Malfunctions
None
Overrides
None
Special Instructions
None
SAFETY CONSIDERATIONS:
None
LOT-SIM-JP-010-01 Page 4 of 10 Rev. 01
SBGT System Operations to Reduce Humidity
EVALUATOR NOTES: (Do not read to trainee)
1. The applicable procedure section WILL be provided to the trainee.
2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG
1021, Appendix E, or similar to the trainee.
Read the following to the JPM performer
TASK CONDITIONS:
1. The Reactor Building Auxiliary Operator has reported 2A SBGT relative
humidity at 72%.
2. SBGT System is in standby in accordance with Section 5.1.
3. No painting has been done in the last 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> on the Reactor Building 80
west, 50 west or 20 south elevations, AND no welding has been done in
these areas within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
4. The Unit CRSs permission has been granted to operate 2A SBGT to reduce
humidity.
INITIATING CUE:
You are directed by the Unit CRS to operate 2A SBGT System to reduce relative
humidity to <70%. Inform the Unit CRS when relative humidity has been reduced to
<70% and the SBGT System has been secured.
LOT-SIM-JP-010-01 Page 5 of 10 Rev. 01
SBGT System Operations to Reduce Humidity
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
Step 1 - Obtain a current revision of 2OP-10, Section 8.1.
Current Revision of 2OP-10, Section 8.1 obtained and verified if
applicable.
SAT/UNSAT*
Time Start: _____________
Step 2 - Place 2A SBGT control switch to ON and observe the following indications:
a. Red FAN ON light is on
b. Red HTR ON light is on
c. Red PREF light is off.
d. SBGT System Flow increases to greater than or equal to 3000 scfm
on Panel XU-51, as indicated by VA-FI-3150-1.
SBGT A control switch placed to ON.
- CRITICAL STEP ** SAT/UNSAT*
Step 3 - Monitor 2A SBGT System operation in accordance with Section 6.0.
2A SBGT operation monitored in accordance with Section 6.0.
SAT/UNSAT*
PROMPT: After 2A SBGT System has been verified to be operating properly inform
examinee that relative humidity is 65%.
Step 4 - When relative humidity is <70% then shutdown the SBGT System as follows.
a. Place 2A SBGT control switch to PREF and ensure the associated red
PREF lamp is ON.
2A SBGT control switch placed in PREF and the PREF red lamp is
on.
- CRITICAL STEP ** SAT/UNSAT*
Step 4 (continued)
b. Depress 2A SBGT PUSH OFF push button and ensure the associated
green PUSH OFF indicating lamp is on.
2A SBGT PUSH OFF push button is depressed and
LOT-SIM-JP-010-01 Page 6 of 10 Rev. 01
SBGT System Operations to Reduce Humidity
the PUSH OFF light is on.
- CRITICAL STEP ** SAT/UNSAT*
c. Momentarily place 2A SBGT control switch in RESET and release,
allowing it to return to PREF.
2A SBGT control switch taken to reset, released and
spring returned to PREF.
- CRITICAL STEP ** SAT/UNSAT*
d. Ensure the following indicating lights are on:
- Green PUSH OFF light
- Red PREF light
Green PUSH OFF and red PREF lights are on.
SAT/UNSAT*
PROMPT: Inform examinee that another operator will perform Attachment 4.
Step 5 - Inform Unit CRS that 2A SBGT System relative humidity is 70% and the SBGT
System has been secured.
SAT/UNSAT*
TERMINATING CUE: When 2A SBGT System relative humidity is <70% and 2A SBGT
is secured this JPM is complete.
- Comments required for any step evaluated as UNSAT.
Time Completed: _____________
LOT-SIM-JP-010-01 Page 7 of 10 Rev. 01
SBGT System Operations to Reduce Humidity
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and Adherence
(PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer SAT/ UNSAT/ NE
checking)
E. Proper Equipment Use (observe starting limitations, SAT/ UNSAT/ NE
throttle valve closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety SAT/ UNSAT/ NE
equipment, etc.)
G. Electrical Safety And Arc Flash Protection Compliance SAT/ UNSAT/ NE
(refer to SAF-NGGC-2175, as applicable)
H. Security Compliance (controlled area entry and exit, key SAT/ UNSAT/ NE
control, etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, SAT/ UNSAT/ NE
etc.)
J. Radiation Protection (ALARA, understanding and use of SAT/ UNSAT/ NE
RWP, frisking, etc.)
COMMENTS:
LOT-SIM-JP-010-01 Page 8 of 10 Rev. 01
SBGT System Operations to Reduce Humidity
Validation Time: 10 Minutes (approximate).
Time Taken:
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: 2
Setting: In-Plant Simulator X Admin
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes X No
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
(Reference TAP-411 for evaluation guidance)
Comments:
Comments Reviewed With Performer
Evaluator Signature: Date:
LOT-SIM-JP-010-01 Page 9 of 10 Rev. 01
TASK CONDITIONS:
1. The Reactor Building Auxiliary Operator has reported 2A SBGT relative humidity at
72%.
2. SBGT System is in standby in accordance with Section 5.1.
3. No painting has been done in the last 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> on the Reactor Building 80 west, 50
west or 20 south elevations, AND no welding has been done in these areas within 1
hour.
4. The Unit CRSs permission has been granted to operate 2A SBGT to reduce
humidity.
INITIATING CUE:
You are directed by the Unit CRS to operate 2A SBGT System to reduce relative
humidity to <70%. Inform the Unit CRS when relative humidity has been reduced to
<70% and the SBGT System has been secured.
Page 10 of 10
DUKE ENERGY PROGRESS
BRUNSWICK NUCLEAR PLANT
LESSON TITLE: Place FW master controller in Service per GP-02 - master Controller
Fails.
LESSON NUMBER: LOT-SIM-JP-032-C02
REVISION NO: 04
PREPARER DATE
TECHNICAL REVIEWER DATE
LINE REVIEW/VALIDATOR DATE
Facility Representative DATE
Place FW master controller in Service per GP-02 - master Controller Fails.
Revision Summary:
REV. No. REVISION SUMMARY
4 JPM formatted for new template
LOT-SIM-JP-032-C02 Page 2 of 12 Rev. 04
Place FW master controller in Service per GP-02 - master Controller Fails.
RELATED TASKS:
259 001 B4 01 Respond To A Condensate/Feedwater System Failure Per
0AOP-23.0
K/A REFERENCE AND IMPORTANCE RATING:
295002 AA1.04 4.0/4.0
Ability to operate and/or monitor the following as they apply to LOW REACTOR
WATER LEVEL: Reactor Water Level Control
REFERENCES:
0AOP-23.0
TOOLS AND EQUIPMENT:
None.
SAFETY FUNCTION (from NUREG 1123, Rev 2.):
2 Reactor Water Inventory Control
REASON FOR REVISION:
Updated to the new format
LOT-SIM-JP-032-C02 Page 3 of 12 Rev. 04
Place FW master controller in Service per GP-02 - master Controller Fails.
SIMULATOR SETUP:
Initial Conditions
IC 06
Rx. Pwr. 6%
Core Age BOC
Required Plant Conditions
Plant startup in progress ready to place master FW controller in automatic.
RFPT A and B speed controllers display pump demand. Master Controller displays
Level Setpoint. SULCV Controller displays Valve demand.
Triggers
T: Master Level Controller A/M Switch (K2811A) to OFF, AUTO activated by
!K2812B1Y (SULCV placed to manual).
Malfunction
None
Overrides
K2811A to OFF activated by Trigger 1
Special Instructions
None
SAFETY CONSIDERATIONS:
None
LOT-SIM-JP-032-C02 Page 4 of 12 Rev. 04
Place FW master controller in Service per GP-02 - master Controller Fails.
EVALUATOR NOTES: (Do not read to trainee)
1. The applicable procedure section WILL be provided to the performer.
2. Provide examinee with a copy of GP-02 marked completed to step 5.3.55.
3. If this is the first JPM of the JPM set, read the JPM briefing contained in NUREG
1021, Appendix E, or similar to the performer.
Read the following to the JPM performer
TASK CONDITIONS:
1. Unit Two startup is in progress per GP-02.
2. Reactor Feed Pump discharge pressure is greater than 900 psig.
3. GP-02 is completed up to step 5.3.55.
INITIATING CUE:
You are directed to place the FW Master Controller in Automatic per GP-02, step 5.3.55
and inform the Unit SCO when the step is completed.
LOT-SIM-JP-032-C02 Page 5 of 12 Rev. 04
Place FW master controller in Service per GP-02 - master Controller Fails.
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
Step 1 - Obtain marked up copy of 0GP-02.
0GP-02 obtained.
SAT/UNSAT*
TIME START
___________
Step 2 - Ensure MSTR RFPT SP/RX LVL CTL, C32-SIC-R600 in M (Manual).
Master level controller C32-SIC-R600 is in Manual.
- CRITICAL STEP**SAT/UNSAT*
Step 3 - Ensure FEEDWATER CONTROL MODE SELECT in 1 ELEM.
Feedwater Control is selected for single element.
SAT/UNSAT*
Step 4 - DEPRESS SEL pushbutton on RFPT A SP CTL, C32-SIC-R601A until A BIAS
is indicated and ENSURE bias is set at 0%.
Reactor Feed Pump A Bias is 0%.
SAT/UNSAT*
Step 5 - DEPRESS SEL pushbutton on RFPT A SP CTL, C32-SIC-R601A until PUMP
A DEM is displayed.
Pump A Demand displayed on RFPT A Speed controller.
SAT/UNSAT*
LOT-SIM-JP-032-C02 Page 6 of 12 Rev. 04
Place FW master controller in Service per GP-02 - master Controller Fails.
Step 6 - DEPRESS SEL pushbutton on MSTR RFPT SP/RX LVL CTL, C32-SIC-R600
until MASTR DEM is displayed.
Master Demand is displayed on the Master Controller.
SAT/UNSAT*
Step 7 - SET MASTR DEM to equal PMP A DEM value displayed on RFPT A SP CTL,
C32-SIC-R601A, using the raise and lower pushbuttons on MSTR RFPT
SP/RX LVL CTL, C32-SIC-R600.
Master Controller demand set to equal RFPT A demand.
- CRITICAL STEP ** SAT/UNSAT*
Step 8 - DEPRESS A/M pushbutton on RPFT A SP CTL, C32-SIC-R601A, AND
CHECK the indicator on control station changes to A (automatic) AND PMP
DEM signal remains unchanged.
RFPT A Speed Controller is in automatic.
- CRITICAL STEP ** SAT/UNSAT*
Step 9 - DEPRESS SEL pushbutton on the out-of-service RFPT B SP CTL, C32-SIC-
R601B, until LVL ERROR is indicated AND CHECK LVL ERROR is
approximately 0 inches.
Level Error selected on RFPT B Controller.
SAT/UNSAT*
Step 10 - DEPRESS A/M pushbutton on MSTR RFPT SP/RX LVL CTL, C32-SIC-R600,
AND CHECK the indicator on the control station changes to A (automatic).
Master Controller is placed to automatic.
- CRITICAL STEP ** SAT/UNSAT*
Step 11 - ENSURE PMP A DEM and VALVE DEM signals remain unchanged.
Pump Demand remains unchanged.
SAT/UNSAT*
LOT-SIM-JP-032-C02 Page 7 of 12 Rev. 04
Place FW master controller in Service per GP-02 - master Controller Fails.
NOTE: Monitor the SULCV Controller on panel mimic. When the SULCV Controller is
shifted to Manual, initiate Trigger to activate override on Master Controller.
Step 12 - DEPRESS A/M pushbutton on SULCV, FW-LIC-3269, AND CHECK the
indicator on the control station changes to M (manual).
SULCV Controller is placed to manual
- CRITICAL STEP ** SAT/UNSAT*
NOTE: Master controller output (demand) will slowly change. The examinee will
probably not immediately recognize the slowly changing output and begin to
slowly open the SULCV.
If the examinee recognizes the Master controller failure before beginning to
open the SULCV, the next step in this JPM is not applicable.
Step 13 - SLOWLY OPEN SULCV, using raise pushbutton on FIC-LIC-3269, until
VALVE DEM is 100% and CHECK reactor water level is being maintained
between 182 and 192 inches.
SULCV is slowly opened using raise pushbutton.
SAT/UNSAT*
NOTE: There are a variety of methods to take manual control of Feedwater (i.e. Master
manual, RFP Speed controller manual, Woodward manual).
It is critical that the examinee demonstrates the ability to maintain reactor level
>166 and <206. Since the examinee is unable to complete the task given in
the initiating cue, the examiner should stop the JPM when the terminating cue
is satisfied.
Step 14 - Take manual control of Reactor Level.
Maintains RPV level between +166 and +206 inches.
LOT-SIM-JP-032-C02 Page 8 of 12 Rev. 04
Place FW master controller in Service per GP-02 - master Controller Fails.
- CRITICAL STEP ** SAT/UNSAT*
Step 15 - Inform SCO feedwater control is not properly maintaining level in auto and
manual control has been taken.
SCO informed.
SAT/UNSAT*
TERMINATING CUE: When the examinee demonstrates the ability to maintain RPV
level between +166 and +206, the JPM may be terminated.
TIME COMPLETED _____________
NOTE: Comments required for any step evaluated as UNSAT.
LOT-SIM-JP-032-C02 Page 9 of 12 Rev. 04
Place FW master controller in Service per GP-02 - master Controller Fails.
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and Adherence
(PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer SAT/ UNSAT/ NE
checking)
E. Proper Equipment Use (observe starting limitations, SAT/ UNSAT/ NE
throttle valve closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety SAT/ UNSAT/ NE
equipment, etc.)
G. Electrical Safety And Arc Flash Protection Compliance SAT/ UNSAT/ NE
(refer to SAF-NGGC-2175, as applicable)
H. Security Compliance (controlled area entry and exit, key SAT/ UNSAT/ NE
control, etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, SAT/ UNSAT/ NE
etc.)
J. Radiation Protection (ALARA, understanding and use of SAT/ UNSAT/ NE
RWP, frisking, etc.)
COMMENTS:
LOT-SIM-JP-032-C02 Page 10 of 12 Rev. 04
Place FW master controller in Service per GP-02 - master Controller Fails.
Validation Time: 10 Minutes (approximate).
Time Taken:
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: 2
Setting: In-Plant Simulator X Admin
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes X No
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
(Reference TAP-411 for evaluation guidance)
Comments:
Comments Reviewed With Performer
Evaluator Signature: Date:
LOT-SIM-JP-032-C02 Page 11 of 12 Rev. 04
TASK CONDITIONS:
1. Unit Two startup is in progress per GP-02.
2. Reactor Feed Pump discharge pressure is greater than 900 psig.
3. GP-02 is completed up to step 5.3.55.
INITIATING CUE:
You are directed to place the FW Master Controller in Automatic per GP-02, step 5.3.55 and
inform the Unit SCO when the step is completed.
Page 12 of 12
DUKE ENERGY PROGRESS
BRUNSWICK NUCLEAR PLANT
LESSON TITLE: Place RHR In Suppression Pool Cooling Per AOP-36.2.
LESSON NUMBER: LOT-SIM-JP-303-A10
REVISION NO: 01
PREPARER DATE
TECHNICAL REVIEWER DATE
LINE REVIEW/VALIDATOR DATE
Facility Representative DATE
Place RHR In Suppression Pool Cooling Per AOP-36.2.
Revision Summary:
REV. No. REVISION SUMMARY
1 JPM formatted for new template
LOT-SIM-JP-303-A10 Page 2 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
RELATED TASKS:
205014B101, Start Up RHR In Suppression Pool Cooling Mode Per OP-17
262004B401, Respond To A Total Loss Of All AC Electrical Distribution (Station
Blackout) Per AOP-36.2
K/A REFERENCE AND IMPORTANCE RATING:
219000 A4.02 3.7/3.5
Ability to manually operate and/or monitor in the Control Room: Valve lineup.
REFERENCES:
TOOLS AND EQUIPMENT:
None.
SAFETY FUNCTION (from NUREG 1123, Rev 2.):
5 - Containment Integrity (RHR/LPCI: Torus/Suppression pool Cooling)
REASON FOR REVISION:
Updated to the new format
LOT-SIM-JP-303-A10 Page 3 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
SIMULATOR SETUP:
Initial Conditions
IC-11 BOC
Rx Pwr 100%
Core Age BOCRequired Plant Conditions
Triggers
See LOT-AOP-128
Malfunction
See LOT-AOP-128
Overrides
See LOT-AOP-128
Special Instructions
Load Scenario File for LOT-AOP-128.
Run scenario to point of starting Suppression Pool Cooling (Cross-tie E2-E4, E5-
E6 and E7-E8, Start Drywell Cooling on Non Blacked Out Unit)
SAFETY CONSIDERATIONS:
None
LOT-SIM-JP-303-A10 Page 4 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
EVALUATOR NOTES: (Do not read to trainee)
1. The applicable procedure section WILL be provided to the trainee.
2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG
1021, Appendix E, or similar to the trainee.
Read the following to the JPM performer
TASK CONDITIONS:
1. Units 1 and 2 have lost Off-Site Power.
2. DG4 is the only available Diesel Generator.
3. Both Units are performing AOP-36.2.
4. Unit 2 is being maintained in Hot Shutdown above 500 psig.
5. Emergency Buses E2 to E4, E5 to E6 and E7 to E8 have been cross-tied per
AOP-36.2, Section 3.2.12.
6. Drywell Cooling has been started on Unit 2.
INITIATING CUE:
You are directed to start Suppression Pool Cooling on Unit 2 in accordance with
AOP-36.2, Section 3.2.16 using RHR Loop B, and inform the SCO when
Suppression Pool Cooling has been established.
LOT-SIM-JP-303-A10 Page 5 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
Step 1 - Obtain a current revision of AOP-36.2, Section 3.2.16.
Current Revision of AOP-36.2, Section 3.2.16 obtained.
SAT/UNSAT*
NOTE: MCC 2PB is not stripped per cross-tie actions if DG4 is available.
Step 2 - Ensure energized MCC 2PB.
MCC 2PB is energized.
SAT/UNSAT*
Step 3 - Ensure closed CSW Pump 2B discharge valves.
CSW Pump 2B discharge valves are closed.
SAT/UNSAT*
Step 4 - Ensure CSW Pump 2B mode selector switch is in MAN.
CSW Pump 2B mode selector switch is in MAN.
SAT/UNSAT*
Step 5 - Ensure CSW Pump 2B discharge valve selector switch in NUC HDR.
CSW Pump 2B discharge valve selector switch is in NUC HDR.
- CRITICAL STEP ** SAT/UNSAT*
LOT-SIM-JP-303-A10 Page 6 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
NOTE: CSW Pump 2B Fuses are not removed per cross-tie actions if DG4 is
available.
Step 6 - Start CSW Pump 2B
CSW Pump 2B is running.
- CRITICAL STEP ** SAT/UNSAT*
Step 7 - Ensure CSW Pump 2B discharge valve opens.
CSW Pump 2B Nuclear Header discharge valve is open.
SAT/UNSAT*
NOTE: MCC 2XB is de-energized per cross-tie actions.
PROMPT: When requested as AO to close 480V Sub E8 Compt AO2 (MCC 2XB),
delete malfunction EE_030M [11], and report MCC 2XB is energized.
Step 8 - Request an AO energize 480V Sub E8, Compt AO2 MCC 2XB.
MCC 2XB is energized.
- CRITICAL STEP ** SAT/UNSAT*
Step 9 - Ensure open Nuclear SW Supply Valve SW-V105.
SW-V105 is open.
- CRITICAL STEP ** SAT/UNSAT*
Step 10 - Ensure open HX 2B SW Discharge E11-F002A.
E11-F002A is open.
SAT/UNSAT*
LOT-SIM-JP-303-A10 Page 7 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
PROMPT: When requested to place E11-CS-5607B (F068B Interlock Bypass switch) to
Bypass, modify Remote Function RS_IARHBYPB, BYPASS, and report
action complete.
Step 11 - Request E11-CS-5607B be placed to Bypass.
E11-CS-5607B is in Bypass
- CRITICAL STEP ** SAT/UNSAT*
NOTE: A maximum flow of 2000 gpm can be achieved without a RHR SW Booster
Pump.
Step 12 - Open Heat Exchanger 2B SW Discharge valve E11-F068B to obtain desired
flow rate.
E11-F068B opened to achieve RHR SW flow 1000 gpm.
- CRITICAL STEP ** SAT/UNSAT*
Step 13 - Ensure open Nuclear SW To Vital Header SW-V117.
SW-V117 is full open.
SAT/UNSAT*
Step 14 - Ensure closed Well Water To Vital Header SW-V141.
SW-V141 is closed.
SAT/UNSAT*
Step 15 - Close HX 2B Inlet Valve E11-F047B.
E11-F047B is closed.
- CRITICAL STEP ** SAT/UNSAT*
LOT-SIM-JP-303-A10 Page 8 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
Step 16 - Close HX 2B Bypass Valve E11-F048B
E11-F048B is full closed.
- CRITICAL STEP ** SAT/UNSAT*
Step 17 - Ensure open HX 2B Outlet Valve E11-F003B.
E11-F003B is open.
SAT/UNSAT*
NOTE: MCC 2XB-2 is de-energized per cross-tie actions.
PROMPT: When requested as AO to close 480V Sub E6 Compt AV8 (MCC 2XB-2),
delete malfunction EE_030M [32], and report MCC 2XB-2 is energized.
Step 18 - Request an AO energize 480V Sub E6, Compt AV8 MCC 2XB-2.
- CRITICAL STEP ** SAT/UNSAT*
Step 19 - Open Loop B Torus Discharge Isol Valve E11-F028B.
E11-F028B is open.
- CRITICAL STEP ** SAT/UNSAT*
PROMPT: When requested as AO to de-energize MCC 2XB-2, insert malfunction
EE_030M [32], and report MCC 2XB-2 is de-energized.
Step 20 - Request AO de-energize MCC 2XB-2.
SAT/UNSAT*
LOT-SIM-JP-303-A10 Page 9 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
NOTE: RHR Pump 2B Fuses are not removed per cross-tie actions if DG4 is
available.
Step 21 - Start RHR Pump 2B
RHR Pump 2B is running.
- CRITICAL STEP ** SAT/UNSAT*
Step 22 - Slowly Throttle open HX 2B Bypass Valve E11-F048B.
E11-F048B is throttled open.
- CRITICAL STEP ** SAT/UNSAT*
Step 23 - Slowly throttle open Torus Cooling Isolation Valve E11-F024B to increase
flow to 10,000 gpm.
E11-F024B throttled open to achieve a flow of 10,000 gpm.
- CRITICAL STEP ** SAT/UNSAT*
PROMPT: When requested as AO to close MCC 2XB, Compt DN4 (E11-F103B),
modify Remote Function RH_ZBRH93BT, ON, and report action complete.
Step 24 - Request AO place breaker for RHR HX Vent Valve E11-F103B at MCC 2XB,
Compt DN4 to On.
Breaker for E11-F103B is On.
SAT/UNSAT*
Step 25 - Open RHR HX 2B Inboard vent Valve, E11-F104B and Outboard Vent Valve
E11-F103B
LOT-SIM-JP-303-A10 Page 10 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
E11-F104B and E11-F103B are open.
SAT/UNSAT*
Step 26 - Open HX 2B Inlet Valve, E11-F047B
HX 2B Inlet Valve E11-F047B is full open
- CRITICAL STEP ** SAT/UNSAT*
PROMPT: When E11-F047B is full open, inform examinee that 5 minutes have passed.
Step 27 - After venting for 5 minutes, close RHR HX 2B Inboard vent Valve, E11-
F104B and Outboard Vent Valve E11-F103B.
E11-F104B and E11-F103B are closed.
SAT/UNSAT*
PROMPT: When requested as AO to open MCC 2XB, Compt DN4 (E11-F103B), modify
Remote Function RH_ZBRH93BT, OFF, and report action complete.
Step 28 - Request AO place breaker for RHR HX Vent Valve E11-F103B at MCC 2XB,
Compt DN4 to OFF.
Breaker for E11-F103B is OFF.
SAT/UNSAT*
Step 29 - Close HX 2B Bypass Valve E11-F048B.
E11-F048B is full closed.
- CRITICAL STEP ** SAT/UNSAT*
PROMPT: If requested as SCO to notify E&RC for increased activity, inform examinee
that E&RC has been notified.
Step 30 - Direct E&RC to monitor Service water discharge for increased activity.
LOT-SIM-JP-303-A10 Page 11 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
E&RC notified to monitor Service Water discharge
SAT/UNSAT*
PROMPT: When requested as AO to open 480V Sub E8 Compt AO2 (MCC 2XB), insert
malfunction EE_030M [11], and report MCC 2XB is de-energized.
Step 31 - Request AO to De-energize Sub E8, Compt AO2 (MCC 2XB).
MCC 2XB is de-energized.
SAT/UNSAT*
Step 32 - Inform Unit SCO that RHR Loop B has been placed in suppression pool
cooling.
Unit SCO informed RHR Loop B is operating in suppression pool cooling.
SAT/UNSAT*
TERMINATING CUE: When AOP-36.2, Section 3.2.16 is complete, this JPM may be
terminated.
- Comments required for any step evaluated as UNSAT.
LOT-SIM-JP-303-A10 Page 12 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
LOT-SIM-JP-303-A10 Page 13 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and Adherence
(PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer SAT/ UNSAT/ NE
checking)
E. Proper Equipment Use (observe starting limitations, SAT/ UNSAT/ NE
throttle valve closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety SAT/ UNSAT/ NE
equipment, etc.)
G. Electrical Safety And Arc Flash Protection Compliance SAT/ UNSAT/ NE
(refer to SAF-NGGC-2175, as applicable)
H. Security Compliance (controlled area entry and exit, key SAT/ UNSAT/ NE
control, etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, SAT/ UNSAT/ NE
etc.)
J. Radiation Protection (ALARA, understanding and use of SAT/ UNSAT/ NE
RWP, frisking, etc.)
COMMENTS:
LOT-SIM-JP-303-A10 Page 14 of 16 Rev. 0
Place RHR In Suppression Pool Cooling Per AOP-36.2.
Validation Time: 20 Minutes (approximate).
Time Taken:
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: 2
Setting: In-Plant Simulator X Admin
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes X No
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
(Reference TAP-411 for evaluation guidance)
Comments:
Comments Reviewed With Performer
Evaluator Signature: Date:
LOT-SIM-JP-303-A10 Page 15 of 16 Rev. 0
TASK CONDITIONS:
1. Units 1 and 2 have lost Off-Site Power.
2. DG4 is the only available Diesel Generator.
3. Both Units are performing AOP-36.2.
4. Unit 2 is being maintained in Hot Shutdown above 500 psig.
5. Emergency Buses E2 to E4, E5 to E6 and E7 to E8 have been cross-tied per AOP-36.2,
Section 3.2.12.
6. Drywell Cooling has been started on Unit 2.
INITIATING CUE:
You are directed to start Suppression Pool Cooling on Unit 2 in accordance with AOP-36.2,
Section 3.2.16 using RHR Loop B, and inform the SCO when Suppression Pool Cooling
has been established.
Page 16 of 16
DUKE ENERGY
BRUNSWICK TRAINING SECTION
LESSON TITLE: Restarting RCIC after AUTO initiation and Turbine Trip using the
Hard Card - Controller Failure
LESSON NUMBER: LOT-SIM-JP-016-01
REVISION NO: 1
Lou Sosler 9/9/2013
PREPARER DATE
Bob Bolin 11/1/2013
TECHNICAL REVIEWER DATE
Joel Gordon 9/11/2013
LINE REVIEWER/VALIDATOR DATE
Robert Scott 11/1/2013
TRAINING SUPERVISION APPROVAL DATE
LOT-SIM-JP-016-01 Page 1 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
RELATED TASKS:
217 200 B4 01
Restart The RCIC System After Auto Initiation And Turbine Trip Per OP-16.
K/A REFERENCE AND IMPORTANCE RATING:
217000 K4.06 3.5,3.5
A1.01 3.7, 3.7
A1.02 3.3, 3,3
REFERENCES:
OP-16, Section 8.7, Restarting the RCIC System After Auto Initiation and Turbine trip
OP-16, Att. 7, Restarting RCIC After Auto Initiation and Turbine Trip (OP-16 Section 8.7),
Hard Card S/928
TOOLS AND EQUIPMENT:
N/A
SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1):
4 - Heat Removal From the Core
LOT-SIM-JP-016-01 Page 2 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
SETUP INSTRUCTIONS
SIMULATOR SETUP:
A. Initial Conditions:
IC-11
Reactor Power: 100%
Required Plant Conditions
Reactor water level approximately 150 inches following < LL2 (105).
RFPs and HPCI tripped not available.
B. Triggers
K1H12PAU, RCIC Flow Controller Auto Mode, Push, On
C. Malfunctions
Tirgger 1, ES052F, RCIC FIC EPU Failure
C. Overrides
D. Remote Function
None required
E. Special Instructions
1. Manually trip the RCIC Turbine.
2. Place HPCI Aux Oil Pump to LOCKOUT.
3. Scram the Reactor and reduce level by tripping both RFPT's.
4. Complete initial EOP operator actions.
5. After level drops below LL2 (+105"), and is beginning to recover, place
simulator in FREEZE.
LOT-SIM-JP-016-01 Page 3 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
SAFETY CONSIDERATIONS:
1. Notify SM/CRS of JPM performance prior to commencing In-plant JPM.
2. Determine actual radiological conditions and potentially contaminated areas to achieve ALARA.
3. Ensure all electrical safety requirements are observed.
4. Review Work Practices section prior to conduct of the JPM.
5. DO NOT OPERATE any plant equipment during performance of this JPM.
EVALUATOR NOTES: (Do not read to performer)
1. The applicable procedure section WILL be provided to the trainee.
2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1021,
Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).
3. This JPM may be performed in the Simulator on Unit 2.
4. Critical Step Basis
a) Prevents Task Completion
b) May Result in Equipment Damage
c) Affects Public Health and Safety
d) Could Result in Personal Injury
Read the following to the JPM performer.
TASK CONDITIONS:
1. Both Reactor Feed Pumps have tripped.
2. RCIC AUTO started and has been manually tripped.
3. SBGT is running.
INITIATING CUE:
The Unit SCO has directed you to RESET RCIC and inject water to the Reactor Vessel at
500 gpm to raise level to 170-200 inches, using the Hard Card, and inform the Unit SCO
when the required actions are complete.
LOT-SIM-JP-016-01 Page 4 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
TIME START:
Step 1 - Obtain RCIC Hard Card S/928
RCIC Hard Card S/928, RESTARTING RCIC AFTER AUTO INITIATION AND
TURBINE TRIP (2OP-16, Section 8.7) obtained.
SAT/UNSAT
Step 2 - ENSURE E51-V8 (VALVE POSITION) AND E51-V8 (MOTOR OPERATOR) ARE
CLOSED.
E51-V8 (Valve/Motor Operator) verified CLOSED.
SAT/UNSAT
Step 3 - PLACE RCIC FLOW CONTROL IN MANUAL (M) AND ADJUST OUTPUT TO 0%.
E51-FIC-R601, RCIC Flow Control placed in MANUAL with 0% output.
SAT/UNSAT
Step 4 - JOG OPEN E51-V8 UNTIL THE TURBINE SPEED IS CONTROLLED BY THE
GOVERNOR
E51-V8 is jogged open and turbine speed is increasing.
- CRITICAL STEP** SAT/UNSAT
Step 5 - FULLY OPEN E51-V8
E51-V8 is full open.
- CRITICAL STEP** SAT/UNSAT
Step 6 - SLOWLY RAISE TURBINE SPEED UNTIL FLOW RATE OF AT LEAST 120 GPM
Turbine speed raised to obtain flow rate of at least 120 gpm.
SAT/UNSAT
LOT-SIM-JP-016-01 Page 5 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
Step 7 - ENSURE E51-F019 IS CLOSED WITH FLOW GREATER THAN 80 GPM
E51-F019 verified closed.
SAT/UNSAT
NOTE: When the flow controller is placed in AUTO, the Flow Indicating Controller (FIC) CPU
will fail, as indicated by the red FAIL light on the controller. The capacitor in the FIC
will continually discharge to cause flow to lower, unless the output controller demand
is raised to recharge the capacitor.
Step 8 - WHEN SYSTEM CONDITIONS ARE STABLE, THEN ADJUST SETPOINT, AND
TRANSFER RCIC FLOW CONTROL TO AUTO (A).
E51-FIC- R601 placed in AUTO.
- CRITICAL STEP** SAT/UNSAT
NOTE: Performer should diagnose failure of RCIC flow controller and raise the controller
output demand as necessary to recharge the FIC capacitor and raise flow.
Step 10 - Diagnose failure of RCIC flow controller, E51-R601, and raise controller output to
inject at the desired flow rate.
Raise controller output to inject at a flow rate of up to 500 gpm, to raise Reactor
water level toward 170 - 200 inches.
- CRITICAL STEP** SAT/UNSAT
Step 11 - ENSURE THE FOLLOWING:
BAROMETRIC CNDSR VACUUM PUMP HAS STARTED
SBGT STARTED (2OP-10) - part of initial conditions
Conditions verified or verbalized as part of initial conditions.
SAT/UNSAT
TERMINATING CUE: When RCIC has been reset and manually started and is injecting to the
RPV at approximately 500 gpm, this JPM is complete.
TIME COMPLETED:
LOT-SIM-JP-016-01 Page 6 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
NOTE: Comments required for any step evaluated as UNSAT.
COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY.
LOT-SIM-JP-016-01 Page 7 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute / Two Minute Rule
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and Adherence
(PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer checking) SAT/ UNSAT/ NE
E. Proper Equipment Use (observe starting limitations, throttle valve SAT/ UNSAT/ NE
closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety equipment, SAT/ UNSAT/ NE
etc.)
G. Electrical Safety And Arc Flash Protection Compliance (refer to SAT/ UNSAT/ NE
SAF-NGGC-2175, as applicable)
H. Security Compliance (controlled area entry and exit, key control, SAT/ UNSAT/ NE
etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, etc.) SAT/ UNSAT/ NE
J. Radiation Protection (ALARA, understanding and use of RWP, SAT/ UNSAT/ NE
frisking, etc.)
LOT-SIM-JP-016-01 Page 8 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
REVISION SUMMARY
1 Revised to new JPM Template, Revision 3.
Changed RCIC Flow Controller failure to HPU failure.
Updated to current Hard Card.
LOT-SIM-JP-016-01 Page 9 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
Validation Time: 10 Minutes (approximate)
Time Taken: Minutes
APPLICABLE METHOD OF TESTING
Performance: Simulate: Actual: X Unit: 2
Setting: In-Plant Simulator: X Admin:
Time Critical: Yes No X Time Limit:
(Ensure reference section on previous page identifies the regulation
or procedure that mandates this time limit requirement)
Alternate Path: Yes X No
EVALUATION
Performer:
JPM Results: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
LOT-SIM-JP-016-01 Page 10 of 11 Rev. 1
TASK CONDITIONS:
1. Both Reactor Feed Pumps have tripped.
2. RCIC AUTO started and has been manually tripped.
3. SBGT is running.
INITIATING CUE:
The Unit SCO has directed you to RESET RCIC and inject water to the Reactor Vessel at
500 gpm to raise level to 170-200 inches, using the Hard Card, and inform the Unit SCO
when the required actions are complete.
Page 11 of 11
DUKE ENERGY
BRUNSWICK TRAINING SECTION
LESSON TITLE: Restarting RCIC after AUTO initiation and Turbine Trip using the
Hard Card - Controller Failure
LESSON NUMBER: LOT-SIM-JP-016-01
REVISION NO: 1
Lou Sosler 9/9/2013
PREPARER DATE
Bob Bolin 11/1/2013
TECHNICAL REVIEWER DATE
Joel Gordon 9/11/2013
LINE REVIEWER/VALIDATOR DATE
Robert Scott 11/1/2013
TRAINING SUPERVISION APPROVAL DATE
LOT-SIM-JP-016-01 Page 1 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
RELATED TASKS:
217 200 B4 01
Restart The RCIC System After Auto Initiation And Turbine Trip Per OP-16.
K/A REFERENCE AND IMPORTANCE RATING:
217000 K4.06 3.5,3.5
A1.01 3.7, 3.7
A1.02 3.3, 3,3
REFERENCES:
OP-16, Section 8.7, Restarting the RCIC System After Auto Initiation and Turbine trip
OP-16, Att. 7, Restarting RCIC After Auto Initiation and Turbine Trip (OP-16 Section 8.7),
Hard Card S/928
TOOLS AND EQUIPMENT:
N/A
SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1):
4 - Heat Removal From the Core
LOT-SIM-JP-016-01 Page 2 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
SETUP INSTRUCTIONS
SIMULATOR SETUP:
A. Initial Conditions:
IC-11
Reactor Power: 100%
Required Plant Conditions
Reactor water level approximately 150 inches following < LL2 (105).
RFPs and HPCI tripped not available.
B. Triggers
K1H12PAU, RCIC Flow Controller Auto Mode, Push, On
C. Malfunctions
Tirgger 1, ES052F, RCIC FIC EPU Failure
C. Overrides
D. Remote Function
None required
E. Special Instructions
1. Manually trip the RCIC Turbine.
2. Place HPCI Aux Oil Pump to LOCKOUT.
3. Scram the Reactor and reduce level by tripping both RFPT's.
4. Complete initial EOP operator actions.
5. After level drops below LL2 (+105"), and is beginning to recover, place
simulator in FREEZE.
LOT-SIM-JP-016-01 Page 3 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
SAFETY CONSIDERATIONS:
1. Notify SM/CRS of JPM performance prior to commencing In-plant JPM.
2. Determine actual radiological conditions and potentially contaminated areas to achieve ALARA.
3. Ensure all electrical safety requirements are observed.
4. Review Work Practices section prior to conduct of the JPM.
5. DO NOT OPERATE any plant equipment during performance of this JPM.
EVALUATOR NOTES: (Do not read to performer)
1. The applicable procedure section WILL be provided to the trainee.
2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1021,
Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).
3. This JPM may be performed in the Simulator on Unit 2.
4. Critical Step Basis
a) Prevents Task Completion
b) May Result in Equipment Damage
c) Affects Public Health and Safety
d) Could Result in Personal Injury
Read the following to the JPM performer.
TASK CONDITIONS:
1. Both Reactor Feed Pumps have tripped.
2. RCIC AUTO started and has been manually tripped.
3. SBGT is running.
INITIATING CUE:
The Unit SCO has directed you to RESET RCIC and inject water to the Reactor Vessel at
500 gpm to raise level to 170-200 inches, using the Hard Card, and inform the Unit SCO
when the required actions are complete.
LOT-SIM-JP-016-01 Page 4 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
TIME START:
Step 1 - Obtain RCIC Hard Card S/928
RCIC Hard Card S/928, RESTARTING RCIC AFTER AUTO INITIATION AND
TURBINE TRIP (2OP-16, Section 8.7) obtained.
SAT/UNSAT
Step 2 - ENSURE E51-V8 (VALVE POSITION) AND E51-V8 (MOTOR OPERATOR) ARE
CLOSED.
E51-V8 (Valve/Motor Operator) verified CLOSED.
SAT/UNSAT
Step 3 - PLACE RCIC FLOW CONTROL IN MANUAL (M) AND ADJUST OUTPUT TO 0%.
E51-FIC-R601, RCIC Flow Control placed in MANUAL with 0% output.
SAT/UNSAT
Step 4 - JOG OPEN E51-V8 UNTIL THE TURBINE SPEED IS CONTROLLED BY THE
GOVERNOR
E51-V8 is jogged open and turbine speed is increasing.
- CRITICAL STEP** SAT/UNSAT
Step 5 - FULLY OPEN E51-V8
E51-V8 is full open.
- CRITICAL STEP** SAT/UNSAT
Step 6 - SLOWLY RAISE TURBINE SPEED UNTIL FLOW RATE OF AT LEAST 120 GPM
Turbine speed raised to obtain flow rate of at least 120 gpm.
SAT/UNSAT
LOT-SIM-JP-016-01 Page 5 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
Step 7 - ENSURE E51-F019 IS CLOSED WITH FLOW GREATER THAN 80 GPM
E51-F019 verified closed.
SAT/UNSAT
NOTE: When the flow controller is placed in AUTO, the Flow Indicating Controller (FIC) CPU
will fail, as indicated by the red FAIL light on the controller. The capacitor in the FIC
will continually discharge to cause flow to lower, unless the output controller demand
is raised to recharge the capacitor.
Step 8 - WHEN SYSTEM CONDITIONS ARE STABLE, THEN ADJUST SETPOINT, AND
TRANSFER RCIC FLOW CONTROL TO AUTO (A).
E51-FIC- R601 placed in AUTO.
- CRITICAL STEP** SAT/UNSAT
NOTE: Performer should diagnose failure of RCIC flow controller and raise the controller
output demand as necessary to recharge the FIC capacitor and raise flow.
Step 10 - Diagnose failure of RCIC flow controller, E51-R601, and raise controller output to
inject at the desired flow rate.
Raise controller output to inject at a flow rate of up to 500 gpm, to raise Reactor
water level toward 170 - 200 inches.
- CRITICAL STEP** SAT/UNSAT
Step 11 - ENSURE THE FOLLOWING:
BAROMETRIC CNDSR VACUUM PUMP HAS STARTED
SBGT STARTED (2OP-10) - part of initial conditions
Conditions verified or verbalized as part of initial conditions.
SAT/UNSAT
TERMINATING CUE: When RCIC has been reset and manually started and is injecting to the
RPV at approximately 500 gpm, this JPM is complete.
TIME COMPLETED:
LOT-SIM-JP-016-01 Page 6 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
NOTE: Comments required for any step evaluated as UNSAT.
COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY.
LOT-SIM-JP-016-01 Page 7 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute / Two Minute Rule
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and Adherence
(PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer checking) SAT/ UNSAT/ NE
E. Proper Equipment Use (observe starting limitations, throttle valve SAT/ UNSAT/ NE
closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety equipment, SAT/ UNSAT/ NE
etc.)
G. Electrical Safety And Arc Flash Protection Compliance (refer to SAT/ UNSAT/ NE
SAF-NGGC-2175, as applicable)
H. Security Compliance (controlled area entry and exit, key control, SAT/ UNSAT/ NE
etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, etc.) SAT/ UNSAT/ NE
J. Radiation Protection (ALARA, understanding and use of RWP, SAT/ UNSAT/ NE
frisking, etc.)
LOT-SIM-JP-016-01 Page 8 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
REVISION SUMMARY
1 Revised to new JPM Template, Revision 3.
Changed RCIC Flow Controller failure to HPU failure.
Updated to current Hard Card.
LOT-SIM-JP-016-01 Page 9 of 11 Rev. 1
Restarting RCIC after AUTO initiation and Turbine Trip using the Hard Card - Controller Failure
Validation Time: 10 Minutes (approximate)
Time Taken: Minutes
APPLICABLE METHOD OF TESTING
Performance: Simulate: Actual: X Unit: 2
Setting: In-Plant Simulator: X Admin:
Time Critical: Yes No X Time Limit:
(Ensure reference section on previous page identifies the regulation
or procedure that mandates this time limit requirement)
Alternate Path: Yes X No
EVALUATION
Performer:
JPM Results: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
LOT-SIM-JP-016-01 Page 10 of 11 Rev. 1
TASK CONDITIONS:
1. Both Reactor Feed Pumps have tripped.
2. RCIC AUTO started and has been manually tripped.
3. SBGT is running.
INITIATING CUE:
The Unit SCO has directed you to RESET RCIC and inject water to the Reactor Vessel at
500 gpm to raise level to 170-200 inches, using the Hard Card, and inform the Unit SCO
when the required actions are complete.
Page 11 of 11
PROGRESS ENERGY CAROLINAS
BRUNSWICK TRAINING SECTION
SIMULATOR
LOT-SIM-JP-07.1-03
TITLE: Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
REVISION NO: 2
RECOMMENDED BY:K A Bowdon 7/23/12
Instructor/Developer DATE
CONCURRENCE BY:Andy Padleckas 9/25/12
Line Reviewer DATE
APPROVED BY: Robert Scott 9/25/2012
Superintendent/Supervisor Training DATE
Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
SIMULATOR SETUP
IC-02
Triggers
None
Malfunctions
RD042F, Insert Withdraw Rod Block Failure
Overrides
None
Remote Functions
None
Special Instructions
Clear scram buffer at RWM computer display (back panel - place keylock switch to Inop,
press Etc until scram data option appears, press scram data, then delete data)
Ensure RWM Computer Display and Operator Display in Operate (Operate Mode displayed
Activate malfunction RD042F, Insert Withdraw Rod Block Failure.
LOT-SIM-JP-07.1-03 Page 2 REV. 2
Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
SAFETY CONSIDERATIONS:
None
EVALUATOR NOTES: (Do not read to trainee)
1. The applicable procedure section WILL be provided to the trainee.
2. Provide examinee with copy of 0PT-01.6.2 complete up to Section 7.4, RWM Functional
Test. Ensure a copy of GP-10, sequence A2 and GP-01 is available.
3. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021,
Appendix E, or similar to the trainee.
Read the following to trainee.
TASK CONDITIONS:
1. Unit Two startup is planned following a forced outage.
2. GP-01, Prestartup Checklist is being performed and is complete up to the step to perform
0PT-01.6.2, Rod Worth Minimizer System Operability Test.
3. 0PT-01.6.2, Rod Worth Minimizer Operability Test has been completed to Section 7.4,
RWM Functional Test.
4. GP-10, sequence A2 has been selected in RWM for startup.
INITIATING CUE:
You are directed by the Unit CRS to perform the RWM Functional Test per 0PT-01.6.2,
Section 7. Inform the Unit CRS when the RWM Functional Test is complete.
LOT-SIM-JP-07.1-03 Page 3 REV. 2
Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
NOTE: The marked up copy of 0PT-01.6.2 should be provided to the examinee per the
evaluator notes.
Step 1 - Obtain a current revision of 0PT-01.6.2.
Current Revision of 0GP-01 and 0PT-01.6.2.
SAT/UNSAT*
START TIME:
Step 2 - Ensure rod select power ON and confirm RWM Operator Display indicates Insert and
Withdraw blocks.
Rod select power is On.
- CRITICAL STEP ** SAT/UNSAT*
Step 3 - Select a Step 01 rod per GP-10 and withdraw the rod to position 04.
Step 01 rod selected and withdrawn to position 04.
- CRITICAL STEP ** SAT/UNSAT*
LOT-SIM-JP-07.1-03 Page 4 REV. 2
Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
Step 4 - Confirm no Insert or Withdraw blocks are indicated on RWM.
Confirmed no Insert or Withdraw blocks are indicated on RWM.
SAT/UNSAT*
Step 5 - Confirm Withdraw permissive light is on.
Confirmed Withdraw permissive light is on.
SAT/UNSAT*
Step 6 - Confirm selected rod is indicated at Position 04 on the RWM Operator Display.
Confirmed selected rod is indicated at Position 04 on the RWM Operator Display.
SAT/UNSAT*
Step 7 - Select a rod from any GP-10 step other than Step 01.
A rod from any GP-10 step other than Step 01 is selected.
- CRITICAL STEP ** SAT/UNSAT*
NOTE: The withdraw permissive light will be lit even though RWM displays a Withdraw Block.
The examinee may determine RWM is not operating properly at this point.
LOT-SIM-JP-07.1-03 Page 5 REV. 2
Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
PROMPT: If informed as CRS that RWM is not functioning at this point, direct examinee to
continue actions of the PT until a step in the PT is unsatisfactory.
Step 8 - Confirm a Select Error (SE) and a Withdraw Block (WB) are indicated on the RWM
Operator Display.
Confirmed a Select Error (SE) and a Withdraw Block (WB) are indicated on the
RWM Operator Display.
SAT/UNSAT*
Step 9 - Select the Step 01 rod at Position 04 and insert to position 00.
The Step 01 rod at Position 04 is inserted to position 00.
- CRITICAL STEP ** SAT/UNSAT*
Step 10 - Place the RWM Operator Display keylock switch on RTGB Panel H12-P603 in the
Test position.
RWM Operator Display keylock switch on RTGB Panel H12-P603 is in the Test
position.
- CRITICAL STEP ** SAT/UNSAT*
Step 11 - Press the Etc softkey on the Operator Display until the Rod Test selection appears.
Rod Test selection appears.
- CRITICAL STEP ** SAT/UNSAT*
LOT-SIM-JP-07.1-03 Page 6 REV. 2
Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
Step 12 - Press the Rod Test softkey on the Operator Display.
Rod Test softkey on the Operator Display is pressed.
- CRITICAL STEP ** SAT/UNSAT*
Step 13 - Select a Step 13 rod per GP-10 and withdraw it to Position 04.
A Step 13 rod per GP-10 is withdrawn to Position 04.
- CRITICAL STEP ** SAT/UNSAT*
Step 14 - Place the RWM Operator Display keylock switch in Operate.
RWM Operator Display keylock switch is in Operate.
- CRITICAL STEP ** SAT/UNSAT*
NOTE: The withdraw permissive light will be lit even though RWM displays a Withdraw Block.
The examinee may determine RWM is not operating properly at this point.
PROMPT: If informed as CRS that RWM is not functioning at this point, direct examinee to
continue actions of the PT until a step in the PT is unsatisfactory.
Step 15 - Confirm withdraw block (WB) is indicated on the Operator Display.
Confirmed withdraw block (WB) is indicated on the Operator Display.
SAT/UNSAT*
LOT-SIM-JP-07.1-03 Page 7 REV. 2
Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
Step 16 - Confirm withdraw error is indicated on the Operator Display identifying the Step 13
rod at Position 04.
Confirmed withdraw error is indicated on the Operator Display identifying the
Step 13 rod at Position 04.
SAT/UNSAT*
Step 17 - Confirm latched step is Step 01.
Confirmed latched step is Step 01.
SAT/UNSAT*
Step 18 - Confirm Operate mode is indicated on the Operator Display.
Confirmed Operate mode is indicated on the Operator Display.
SAT/UNSAT*
NOTE: The withdraw permissive light will be lit after the rod in the next step is selected.
RWM display shows withdraw block and select error, but allows the rod to be
withdrawn due to the malfunction in the setup.
Step 19 - Select any other control rod.
A different control rod is selected.
- CRITICAL STEP ** SAT/UNSAT*
LOT-SIM-JP-07.1-03 Page 8 REV. 2
Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
Step 20 - Confirm Rod Block RWM/RMCS Sys Trouble (A-05 5-2) alarms.
Confirmed Rod Block RWM/RMCS Sys Trouble (A-05 5-2) alarms.
SAT/UNSAT*
Step 21 - Confirm select error (SE) is indicated on the Operator Display.
Confirmed select error (SE) is indicated on the Operator Display.
SAT/UNSAT*
NOTE: The operator may determine the rod can be withdrawn by observing withdraw
permissive light being lit without actually attempting to withdraw the rod.
Step 22 - Confirm the selected control rod cannot be withdrawn.
Determines the selected rod can be withdrawn and that the test is unsatisfactory.
- CRITICAL STEP ** SAT/UNSAT*
PROMPT: When informed test is unsatisfactory, as Unit CRS direct examinee to ensure all
control rods are fully inserted.
Step 23 - Inform Unit SCO that test is found to be unsatisfactory.
Unit SCO informed that test is found to be unsatisfactory.
SAT/UNSAT*
LOT-SIM-JP-07.1-03 Page 9 REV. 2
Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
NOTE: If rods are not inserted using RMCS, they will insert when the mode switch is placed
to Shutdown.
Step 24 - Insert control rod withdrawn (from Step 22 of this JPM).
Control rod is at Position 00.
- CRITICAL STEP ** SAT/UNSAT*
Step 25 - Select the Step 13 rod at Position 04 and insert it to Position 00.
Control rod is at Position 00.
- CRITICAL STEP ** SAT/UNSAT*
PROMPT: When all rods are inserted, inform examinee that maintenance has been contacted
for RWM repairs and the JPM is complete.
COMPLETION TIME:
TERMINATING CUE: When the RWM functional test is determined to be unsatisfactory, and
all rods are fully inserted, this JPM is complete.
- Comments required for any step evaluated as UNSAT.
LOT-SIM-JP-07.1-03 Page 10 REV. 2
Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
RELATED TASKS:
201001B201, Conduct Rod Worth Minimizer Operability Test Per PT-01.6.2
K/A REFERENCE AND IMPORTANCE RATING:
201006 A3.04 3.5/3.4
Ability to monitor automatic operations of the Rod Worth Minimizer System (RWM) including
control rod movement blocks
REFERENCES:
0PT-01.6.2
TOOLS AND EQUIPMENT:
None
SAFETY FUNCTION (from NUREG 1123):
7 - Instrumentation
REASON FOR REVISION:
Updated to current revision of 0GP-01
Modified to perform actions of PT-01.6.2 only (GP-01 no longer directs placing mode switch
to shutdown if PT is unsatisfactory)
LOT-SIM-JP-07.1-03 Page 11 REV. 2
Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
HUMAN PERFORMANCE WORK PRACTICES:
A. Proper Communications (content, repeat backs, 3 step SAT / UNSAT/ NE
communications, etc)
B. Proper Procedure Use (place-keeping, compliance, etc.) SAT / UNSAT/ NE
C. Use of STAR (Stop, Think, Act, Review) SAT / UNSAT/ NE
D. Peer Checking (performer may request the evaluator to act as
SAT / UNSAT/ NE
peer).
SAT / UNSAT/ NE
E. Human Performance error prevention tool use.
F. Proper Equipment Use (starting limitations, throttle valve closures, SAT / UNSAT/ NE
etc.)
G. Safety Compliance (use of safety equipment, knowledge of safety SAT / UNSAT/ NE
equipment, etc.)
H. Security Compliance (controlled area entry and exit, key control, SAT / UNSAT/ NE
etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, etc.) SAT / UNSAT/ NE
J. Radiation Protection (ALARA, understanding and use of RWP, SAT / UNSAT/ NE
frisking, etc.)
- Comments required for any step evaluated as UNSAT.
LOT-SIM-JP-07.1-03 Page 12 REV. 2
Rod Worth Minimizer Functional Test - Failure To Enforce Blocks.
Validation Time: 15 Minutes (approximate).
Time Taken: _________
APPLICABLE METHOD OF TESTING
Performance: Simulate ___ Actual X Unit: 2
Setting: In-Plant Simulator X Admin
Time Critical: Yes No X Time Limit
Alternate Path: Yes X No
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
(Reference TAP-411 or NUREG-1021 ES-303 as applicable for guidance)
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
LOT-SIM-JP-07.1-03 Page 13 REV. 2
TASK CONDITIONS:
1. Unit Two startup is planned following a forced outage.
2. GP-01, Prestartup Checklist is being performed and is complete up to the step to perform
0PT-01.6.2, Rod Worth Minimizer System Operability Test.
3. 0PT-01.6.2, Rod Worth Minimizer Operability Test has been completed to Section 7.4,
RWM Functional Test.
4. GP-10, sequence A2 has been selected in RWM for startup.
INITIATING CUE:
You are directed by the Unit CRS to perform the RWM Functional Test per 0PT-01.6.2,
Section 7. Inform the Unit CRS when the RWM Functional Test is complete.
LOT-SIM-JP-07.1-03 Page 14 REV. 2
CAROLINA POWER & LIGHT COMPANY
BRUNSWICK TRAINING SECTION
SIMULATOR
LOT-SIM-JP-014-A02
LESSON TITLE: Reduce RPV Water Level Using RWCU To Radwaste.
REVISION NO: 0
RECOMMENDED BY: K A Bowdon 2/24/03
Instructor/Developer DATE
CONCURRENCE BY: Michael S Williams 7/31/03
Line Superintendent/Supervisor DATE
APPROVED BY: M A Pearson 8/5/03
Superintendent/Supervisor Training DATE
Reduce RPV Water Level Using RWCU To Radwaste.
SIMULATOR SETUP
IC-03 BOC
Rx Pwr 0%
Core Age BOC
Triggers
None
Malfunctions
None
Overrides
None
Remote
None
Special Instructions
Secure RWCU reject flow by closing G31-F033 and G31-F034.
Raise RPV level to approximately 195. Ensure alarm A7 2-2 is sealed in.
Ensure both RWCU Filter Demins are in service.
LOT-SIM-JP-014-A02 Page 2 of 9 REV. 0
Reduce RPV Water Level Using RWCU To Radwaste.
SAFETY CONSIDERATIONS:
None.
EVALUATOR NOTES: (Do not read to trainee)
1. The applicable procedure section WILL NOT be provided to the trainee.
2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021,
Appendix E, or similar to the trainee.
3. There is no direct procedural guidance for this JPM. It is considered within the operators
capability to control RPV water level using CRD and RWCU as directed by General
Operating Procedures.
Read the following to trainee.
TASK CONDITIONS:
1. Unit Two is in Mode 2, performing a reactor startup per GP-02.
2. RPV level band is 182 to 192 as indicated on C32-LI-N004A/B/C.
3. CRD is operating for control rod withdrawal.
4. RWCU is in operation. RWCU was aligned for reject to the main condenser, but that
flow path has been secured for maintenance to repair a small leak in the header
downstream of G31-F034, Reject To Condenser. G31-F034 has been placed under
clearance in the closed position.
5. Reactor Level Hi/Lo has alarmed since securing the RWCU reject flow path.
INITIATING CUE:
You are directed by the Unit SCO establish a RWCU reject to Radwaste in accordance with
GP-02, Step 5.3.4, and lower RPV water level to <192 as indicated on narrow range
instruments C32-LI-N004A/B/C. You are to inform the Unit SCO when the Reactor Hi/Lo
Level alarm is clear.
LOT-SIM-JP-014-A02 Page 3 of 9 REV. 0
Reduce RPV Water Level Using RWCU To Radwaste.
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
NOTE: The following actions are directed by GP-02. There are no specific instructions in
OP-14 (Reactor Water Cleanup Operating Procedure) that apply for current plant
conditions. The operator is expected to be able to control RPV level using CRD and
RWCU reject during execution of General Operating Procedures.
PROMPT: When notified of reject operation to Radwaste, as Radwaste CO, inform examinee
that you have aligned the reject flow path to the Waste Collector Tank. Report
Waste Collector Tank level is 20% and sufficient capacity exists for approximately
20,000 gallons of reject flow.
Step 1 - Obtain current revision of GP-02.
Current revision of GP-02 is obtained.
SAT/UNSAT*
Step 2 - Notify Radwaste of intention to reject to Radwaste, and coordinate to ensure
Radwaste can accept the reject flow.
Radwaste notified of reject to Radwaste.
SAT/UNSAT*
Step 3 - Open Reject To Radwaste Vlv, G31-F035.
Reject To Radwaste Vlv, G31-F035 is open.
- CRITICAL STEP ** SAT/UNSAT*
LOT-SIM-JP-014-A02 Page 4 of 9 REV. 0
Reduce RPV Water Level Using RWCU To Radwaste.
Step 4 - Throttle open RWCU Reject Flow Control Vlv, G31-F033 to achieve a lowering RPV
water level.
G31-F033 is opened to achieve a lowering reactor water trend.
- CRITICAL STEP ** SAT/UNSAT*
PROMPT: If asked, as Radwaste CO, report Water Collector Tank level rising.
NOTE: The operator may (but is not required to) reduce CRD flow rate to the RPV to aid in
reducing RPV level to the desired band. If the operator reduces CRD flow rate, CRD
parameters should be maintained per OP-08, Section 6.0 (CRD Dive Water Pressure
restored 260-275 psig by throttling closed Dive Pressure Control Valve).
NOTE: The operator should monitor Regen HX Outlet, Point 2, and/or Filter Inlet, Point 3, on
G31-TI-R607, to ensure Non-Regen heat exchangers outlet temperature remains
below 130°F, but with RPV temperature <212°F, this limitation will not be exceeded.
Step 5 - Monitor Regen HX Outlet, Point 2, and/or Filter Inlet, Point 3, on G31-TI-R607, to
ensure Non-Regen heat exchangers outlet temperature remains below 130°F.
Operator monitors Regen HX Outlet, Point 2, and/or Filter Inlet, Point 3, on G31-
TI-R607, to ensure Non-Regen heat exchangers outlet temperature remains
below 130°F.
SAT/UNSAT*
LOT-SIM-JP-014-A02 Page 5 of 9 REV. 0
Reduce RPV Water Level Using RWCU To Radwaste.
Step 6 - Lower RPV water level at or below 192 as indicated by C32-LI-N004A/B/C.
RPV water level indicates at or below 192 as indicated by C32-LI-N004A/B/C
and Reactor Level Hi/Lo alarm is clear.
- CRITICAL STEP ** SAT/UNSAT*
Step 7 - Inform Unit SCO that RPV water level is in the required band.
Unit SCO informed RPV water level is in the required band.
SAT/UNSAT*
TERMINATING CUE: When RPV water level indicates 192 on narrow range indicators
N004A/B/C, this JPM is complete.
- Comments required for any step evaluated as UNSAT.
LOT-SIM-JP-014-A02 Page 6 of 9 REV. 0
Reduce RPV Water Level Using RWCU To Radwaste.
RELATED TASKS:
204002B101, Place The RWCU System In Service With The Reactor Not In Cold Shutdown
Per OP-14.
K/A REFERENCE AND IMPORTANCE RATING:
204000, A4.08 3.4/3.4
REFERENCES:
GP-02
TOOLS AND EQUIPMENT:
None.
SAFETY FUNCTION (from NUREG 1123, Rev 2.):
2 - Inventory Control (Reactor Water Cleanup)
REASON FOR REVISION:
New JPM developed by NRC for 2003 license exam.
LOT-SIM-JP-014-A02 Page 7 of 9 REV. 0
Reduce RPV Water Level Using RWCU To Radwaste.
Time Required for Completion: 15 Minutes (approximate).
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual 4 Unit: 2
Setting: Control Room Simulator 4 ( Not applicable to In-Plant JPMs )
Time Critical: Yes No 4 Time Limit N/A
Alternate Path: Yes No 4
EVALUATION
Trainee: SSN:
JPM: Pass Fail
Remedial Training Required: Yes No
Did Trainee Verify Procedure as Authorized Copy?: Yes No
(Each Student should verify one JPM per evaluation set.)
Comments:
Comments reviewed with Student
Evaluator Signature: Date:
LOT-SIM-JP-014-A02 Page 8 of 9 REV. 0
TASK CONDITIONS:
1. Unit Two is in Mode 2, performing a reactor startup per GP-02.
2. RPV level band is 182 to 192 as indicated on C32-LI-N004A/B/C.
3. CRD is operating for control rod withdrawal.
4. RWCU is in operation. RWCU was aligned for reject to the main condenser, but that
flow path has been secured for maintenance to repair a leak in the header downstream
of G31-F034, Reject To Condenser. G31-F034 has been placed under clearance in the
closed position.
5. Reactor Level Hi/Lo has alarmed since securing the RWCU reject flow path.
INITIATING CUE:
You are directed by the Unit SCO establish a RWCU reject to Radwaste in accordance with
GP-02, Step 5.3.4, and lower RPV water level to <192 as indicated on narrow range
instruments C32-LI-N004A/B/C. You are to inform the Unit SCO when the Reactor Hi/Lo
Level alarm is clear.
DUKE ENERGY PROGRESS
BRUNSWICK NUCLEAR PLANT
LESSON TITLE: 6.3.14 Transfer of Recirc VFD-CONT-UPS-A(B) from Inverter
Operation to Maintenance Bypass (JPM I)
LESSON NUMBER: New In Plant
REVISION NO: 0
PREPARER DATE
TECHNICAL REVIEWER DATE
LINE REVIEW/VALIDATOR DATE
Facility Representative DATE
Transfer of Recirc VFD-CONT-UPS- A from Inverter Operation to Maintenance Bypass
Revision Summary:
REV. No. REVISION SUMMARY
0 New JPM
New In Plant Rev. 0
Transfer of Recirc VFD-CONT-UPS- A from Inverter Operation to Maintenance Bypass
RELATED TASKS:
K/A REFERENCE AND IMPORTANCE RATING:
226001 K3.04 3.1/3.3
Uninterruptible power supply.
REFERENCES:
TOOLS AND EQUIPMENT:
None.
SAFETY FUNCTION (from NUREG 1123, Rev 2.):
6 - Electrical
REASON FOR REVISION:
New JPM
New In Plant Rev. 0
Transfer of Recirc VFD-CONT-UPS- A from Inverter Operation to Maintenance Bypass
SIMULATOR SETUP:
Initial Conditions
Any Core time in core life.
Triggers
Facility to Determine
Malfunction
None
Overrides
As Required
Special Instructions
Facility to determine
SAFETY CONSIDERATIONS:
None
New In Plant Rev. 0
Transfer of Recirc VFD-CONT-UPS- A from Inverter Operation to Maintenance Bypass
EVALUATOR NOTES: (Do not read to trainee)
1. The applicable procedure section WILL be provided to the trainee.
2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG
1021, Appendix E, or similar to the trainee.
Read the following to the JPM performer
TASK CONDITIONS:
1-VFD-CONT-UPS-A is currently operating in accordance with 1OP-02, Section
6.1.4.
In preparation for a maintenance activity, the 1-VFD-CONT-UPS-A inverter must be
realigned with UPS loads powered through the Maintenance Bypass Switch.
INITIATING CUE:
You are to perform 1OP-02, Section 6.3.14, Transfer of Recirc VFD-CONT-UPS-
A(B) from Inverter Operation to Maintenance Bypass and inform the Main Control
Room when complete.
New In Plant Rev. 0
Transfer of Recirc VFD-CONT-UPS- A from Inverter Operation to Maintenance Bypass
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
Step 1 - Obtain a current revision of 1OP-2, Section 6.3.14.
Current Revision of 1OP-2, Section 6.3.14 obtained.
SAT/UNSAT*
Step 2 - Ensure Recirc VFD-CONT-UPS-A(B) is operating in accordance
with Section 6.1.4.
Condition is met per the initial conditions.
SAT/UNSAT*
Step 3 - Confirm plant conditions require removing VFD-CONT-UPS-A(B)
inverter from service and powering UPS loads through the
Maintenance Bypass switch.
Confirmed in initial conditions
SAT/UNSAT*
Step 4 - Record VFD-CONT-UPS-A(B) _____.
Record VFD-CONT-UPS-A.
SAT/UNSAT*
New In Plant Rev. 0
Transfer of Recirc VFD-CONT-UPS- A from Inverter Operation to Maintenance Bypass
NOTE: This section is applicable to only the VFD-CONT-UPS-A(B).
Step 5 - Confirm the Maintenance Bypass Available light, located on the
UPS back panel, is ON.
Maintenance Bypass Available light is lit.
SAT/UNSAT*
Step 6 - Place the Maint Bypass switch, in UTILITY
Switch is rotated to the UTILITY position.
- CRITICAL STEP ** SAT/UNSAT*
Step 7 - Confirm NO Recirc VFD or RFCS control alarms due to the UPS transfer.
Contacts the MCR and verifies no VFD control alarms are present.
SAT/UNSAT*
Step 8 - Place the UPS Output breaker, located on the UPS back panel, in OFF.
Output breaker is placed in the off position.
- CRITICAL STEP ** SAT/UNSAT*
Step 9 - Place the UPS Input breaker, located on the UPS back panel, in OFF.
INPUT breaker is placed in the off position.
- CRITICAL STEP ** SAT/UNSAT*
Step 10 - Depress the UPS Off/Bypass button, located on the UPS front panel, twice to
shut down the UPS..
The UPS Off/Bypass Button is depressed 2 times
- CRITICAL STEP ** SAT/UNSAT*
New In Plant Rev. 0
Transfer of Recirc VFD-CONT-UPS- A from Inverter Operation to Maintenance Bypass
Step 11 - Confirm 1-A-06, 1-5 (1-A-07, 1-6), Recirc VFD PDC UPS A(B) Fault alarm is
ON.
Contacts MCR and verifies alarm status
SAT/UNSAT*
TERMINATING CUE: When - 1OP-02, Section 6.3.14 is complete, this JPM may be
terminated.
- Comments required for any step evaluated as UNSAT.
New In Plant Rev. 0
Transfer of Recirc VFD-CONT-UPS- A from Inverter Operation to Maintenance Bypass
Validation Time: 20 Minutes (approximate).
Time Taken:
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: 2
Setting: In-Plant Simulator X Admin
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes X No
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
(Reference TAP-411 for evaluation guidance)
Comments:
Comments Reviewed With Performer
Evaluator Signature: Date:
New In Plant Rev. 0
TASK CONDITIONS:
1-VFD-CONT-UPS-A is currently operating in accordance with 1OP-02, Section 6.1.4.
In preparation for a maintenance activity, the 1-VFD-CONT-UPS-A inverter must be
realigned with UPS loads powered through the Maintenance Bypass Switch.
INITIATING CUE:
You are to perform 1OP-02, Section 6.3.14, Transfer of Recirc VFD-CONT-UPS-A(B) from
Inverter Operation to Maintenance Bypass and inform the Main Control Room when
complete.
DUKE ENERGY
BRUNSWICK TRAINING SECTION
LESSON TITLE: Restoring Seal Purge Flow with Pump Running
- Seal Leakage Abnormal
LESSON NUMBER: AOT-OJT-JP-002-A03
REVISION NO: 06
Chris Michaels 9/5/16
PREPARER DATE
Matt Wooldridge 9/8/16
TECHNICAL REVIEWER DATE
Matt Wooldridge 9/8/16
LINE REVIEWER/VALIDATOR DATE
Bryan Wooten 9/9/16
LINE REVIEWER/VALIDATOR DATE
Jim Barry 9/12/16
TRAINING SUPERVISION APPROVAL/ DATE
Restoring Seal Purge Flow with Pump Running - Seal Leakage Abnormal
REVISION SUMMARY
6 Minor changes and format updates.
Updated procedure sections
Enhanced explanations in Critical Step Bases Table.
Revalidated for 2016 Annual Exam.
Page numbers changed to hidden text on Task Cue sheet.
Reviewed for Time Critical or Time Sensitive Operator Actions
5 New JPM format.
AOT-OJT-JP-002-A03 Page 2 of 11 REV. 06
Restoring Seal Purge Flow with Pump Running - Seal Leakage Abnormal
RELATED TASKS:
202602B104 Restore Reactor Recirculation Pump Seal Purge Flow With
Pump Running Per OP-02.
K/A REFERENCE AND IMPORTANCE RATING:
202001 A1.09 ( 3.3/3.3) Ability to predict and/or monitor changes in parameters
associated with operating the Recirculation System
controls including: Recirculation pump seal pressures
REFERENCES:
1(2)OP-02, Reactor Recirculation System Operating Procedure
TOOLS AND EQUIPMENT:
None
SAFETY FUNCTION (from NUREG 1123, Rev 2.):
Safety Function 1, Reactivity Control (Recirculation System)
TIME CRITICAL OPERATOR ACTIONS:
None for this JPM
AOT-OJT-JP-002-A03 Page 3 of 11 REV. 06
Restoring Seal Purge Flow with Pump Running - Seal Leakage Abnormal
SAFETY CONSIDERATIONS:
None
EVALUATOR NOTES: (Do not read to performer)
1. The applicable procedure section WILL be provided to the trainee.
2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1021,
Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).
3. This JPM may be performed on either Unit.
4. Critical Step Basis
a) Prevents Task Completion
b) May Result in Equipment Damage
c) Affects Public Health and Safety
d) Could Result in Personal Injury
Read the following to the performer.
TASK CONDITIONS:
1. Both Reactor Recirc Pumps are in service.
2. Seal purge flow has been interrupted to the 1A, 1B, 2A, 2B (EVALUATOR
TO SELECT ONE) Reactor Recirculation Pump Seals.
3. Indications are that Upper Seal Leakage is abnormal.
4. All applicable prerequisites have been met.
5. Additional qualified personnel are available to assist.
6. The CRD System is in operation.
7. The Recirc Pump Seal Injection Flow Controller is operable.
8. Seal No. 1 inlet temperature is 175°F.
INITIATING CUE:
You are directed by the Reactor Operator to restore seal purge flow to the
1A, 1B, 2A, 2B (EVALUATOR TO SELECT ONE) Reactor Recirculation Pump
Seals and inform the control room when those actions
are complete.
AOT-OJT-JP-002-A03 Page 4 of 11 REV. 06
Restoring Seal Purge Flow with Pump Running - Seal Leakage Abnormal
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
Step 1 - Obtain a current revision of 1 (2)OP-02. Performer should select Section 6.3.8
since pump seal leakage is abnormal. Section 6.3.7 is performed when seal
leakage is Normal.
Current Revision of 1 (2)OP-02, Section 6.3.8 obtained and verified, if
applicable.
SAT/UNSAT*
TIME START
PROMPT: As requested, notify the trainee that IV steps will be performed by
another operator. If needed, provide cue that IV performed at applicable
steps.
NOTE: Because temperature is >170º F, Step 3 should NOT be performed;
student should go to Step 4. With seal temperature > 170°F but less
than 200°F, seal purge pressures are controlled by slowly throttling flow
controller (which is operable) in this JPM.
Step 2 - Close Seal injection flow controller upstream isolation valve, B32-V45(V47).
B32-V45(V47) closed.
- CRITICAL STEP ** SAT/UNSAT*
PROMPT: If asked, inform the student that Independent Verifications, IVs have
been performed for appropriate steps.
Step 3 - Ensure Seal Injection Flow Controller Bypass Valve B32-V43 (V44) is closed.
B32-V43(V44) is closed.
SAT/UNSAT*
AOT-OJT-JP-002-A03 Page 5 of 11 REV. 06
Restoring Seal Purge Flow with Pump Running - Seal Leakage Abnormal
PROMPT: If asked, inform student that B32-V22(30) is OPEN and that CRD is in
service.
NOTE: Recirculation Pump seal temperatures are monitored by recorder B32-
R601 on Control Room Back Panel H12-P614:
- Point 12(24) for Pump A(B) Seal No. 1 Inlet (T2).
PROMPT: If asked, inform student that temperature reading is dropping 0.2º F
every 15 seconds.
Step 4 - WHILE monitoring Recirculation Pump seal temperatures, SLOWLY throttle
open seal injection flow controller upstream isolation valve, B32-V45(V47), so
that T-2
does NOT decrease more than 5º F per minute or 50º F per hour.
B32-V45(V47) throttled open slowly.
- CRITICAL STEP ** SAT/UNSAT*
PROMPT: After several minutes inform student that all seal temperatures are
normal.
Step 5 - When seal temperatures are normal, fully open B32-V45(47).
B32-V45(47) fully open.
- CRITICAL STEP ** SAT/UNSAT*
PROMPT: If asked, indicate on local indicator B32-FI-1839 (1842) that the flow
indicator reads 4 gpm.
Step 6 - Ensure seal purge flow is 3 to 5 gpm as indicated on local indicator
B32-FI-1839 (1842).
Flow indicates 3 to 5 gpm on B32-FI-1839 (1842).
SAT/UNSAT*
AOT-OJT-JP-002-A03 Page 6 of 11 REV. 06
Restoring Seal Purge Flow with Pump Running - Seal Leakage Abnormal
PROMPT: If asked, inform the student that the affected Recirc Seal No. 1 inlet
temperature are below 200°F
Step 7 - Inform Control Room that seal purge flow has been restored.
Control Room informed seal purge is restored.
SAT/UNSAT*
TERMINATING CUE: When the B32-V45(47) is fully open, this JPM is complete.
Time Completed: ___________
NOTE: Comments required for any step evaluated as UNSAT.
COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY.
AOT-OJT-JP-002-A03 Page 7 of 11 REV. 06
Restoring Seal Purge Flow with Pump Running - Seal Leakage Abnormal
Critical Step Bases
Step Critical / Not Critical Basis / Reason
1 Not Critical Administrative
2 Critical If valve not closed, would result in rapid seal
pressurization and/or thermal shock, resulting in
degradation or failure.
3 Not Critical Flowpath not affected
4-5 Critical If not performed per the procedure, could result in seal
degradation and failure.
6-7 Not Critical Ensure and Recognize steps.
AOT-OJT-JP-002-A03 Page 8 of 11 REV. 06
Restoring Seal Purge Flow with Pump Running - Seal Leakage Abnormal
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and Adherence
(PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer SAT/ UNSAT/ NE
checking)
E. Proper Equipment Use (observe starting limitations, throttle SAT/ UNSAT/ NE
valve closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety SAT/ UNSAT/ NE
equipment, etc.)
G. Electrical Safety And Arc Flash Protection Compliance (refer SAT/ UNSAT/ NE
to SAF-NGGC-2175, as applicable)
H. Security Compliance (controlled area entry and exit, key SAT/ UNSAT/ NE
control, etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, SAT/ UNSAT/ NE
etc.)
J. Radiation Protection (ALARA, understanding and use of SAT/ UNSAT/ NE
RWP, frisking, etc.)
Comments:
AOT-OJT-JP-002-A03 Page 9 of 11 REV. 06
Restoring Seal Purge Flow with Pump Running - Seal Leakage Abnormal
Time Required for Completion: 15 Minutes (approximate).
Time Taken _______
APPLICABLE METHOD OF TESTING
Performance: Simulate X Actual Unit: 1/2
Setting: Control Room Simulator ___ In Plant X
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes X No ____
EVALUATION
Trainee:
JPM: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
AOT-OJT-JP-002-A03 Page 10 of 11 REV. 06
Read the following to the performer.
TASK CONDITIONS:
1. Both Reactor Recirc Pumps are in service.
2. Seal purge flow has been interrupted to the 1A, 1B, 2A, 2B (EVALUATOR
TO SELECT ONE) Reactor Recirculation Pump Seals.
3. Indications are that Upper Seal Leakage is abnormal.
9. All applicable prerequisites have been met.
10. Additional qualified personnel are available to assist.
11. The CRD System is in operation.
12. The Recirc Pump Seal Injection Flow Controller is operable.
13. Seal No. 1 inlet temperature is 175°F.
INITIATING CUE:
You are directed by the Reactor Operator to restore seal purge flow to the
1A, 1B, 2A, 2B (EVALUATOR TO SELECT ONE) Reactor Recirculation Pump
Seals and inform the control room when those actions
are complete.
Pie
rce
08/
15/
DUKE ENERGY PROGRESS 14
BRUNSWICK NUCLEAR PLANT F
a
i
l
i
t
y
R
e
p
LESSON TITLE: Setting Service Air Dryer Sweep Value to Zero
r
e
LESSON NUMBER: AOT-OJT-JP-302-K01 s
e
n
t
REVISION NO: 0 a
t
i
Robert Bolin 07/18/14 v
PREPARER DATE e
D
Lou Sosler 07/18/14 A
TECHNICAL REVIEWER DATE T
E
Derek
Pickett 07/18/14
LINE REVIEW/VALIDATOR DATE
Bruce
Leitch 07/18/14
LINE REVIEW/VALIDATOR DATE
Jerry
Setting Service Air Dryer Sweep Value to Zero
RELATED TASKS:
200504B504
Perform Emergency Actions Associated with a Pneumatic (Air/Nitrogen) System Failure
per AOP-20.
K/A REFERENCE AND IMPORTANCE RATING:
Gen 2.1.30 4.4 / 4.0
Ability to locate and operate components, including local controls.
300000 A2.01 2.9/2.8
Ability to (a) predict the impacts of air dryer and filter malfunctions on the Instrument Air
System and (b) based on those predictions, use procedures to correct, control, or mitigate
the consequences of those abnormal operations.
REFERENCES:
0AOP-20.0, Rev 42
TOOLS AND EQUIPMENT:
None
SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1):
8 - Plant Service Systems (300000 - Instrument Air)
AOT-OJT-JP-300-K01 Page 2 of 12 Rev. 0
Setting Service Air Dryer Sweep Value to Zero
SAFETY CONSIDERATIONS:
1. Use caution in the vicinity of operating equipment.
2. Hard hat, safety glasses and hearing protection are required in the plant.
3. Ensure good ALARA practices while in the plant.
EVALUATOR NOTES: (Do not read to performer)
1. The applicable procedure section WILL be provided to the trainee.
2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021,
Appendix E, or similar to the trainee.
3. Critical Step Basis(s)
a. Prevents Task Completion
b. May Result in Equipment Damage
c. Affects Public Health and Safety
d. Could Result in Personal Injury
Read the following to the JPM performer.
TASK CONDITIONS:
1. A loss of pneumatics has occurred.
2. 0AOP-20.0 is being executed.
3. The temporary air compressor has been placed in service.
INITIATING CUE:
You are directed by the Unit Two SCO to perform 0AOP-20.0, Pneumatic (Air/Nitrogen)
System Failures, Attachment 3, Setting Service Air Dryer(s) Maximum Sweep Value to Zero,
on 2A Air Dryer and inform the CRS when the required actions are complete.
AOT-OJT-JP-300-K01 Page 3 of 12 Rev. 0
Setting Service Air Dryer Sweep Value to Zero
PERFORMANCE CHECKLIST
Step 1 - Perform Take A Minute to review task and conditions at the job site.
As a minimum Identifies correct location and any safety hazards in the area.
SAT/UNSAT
TIME START ___________
Step 2 - Identify applicable Dryer Control System display cabinet.
Circles 2-SA-2A-AIR-DRY-PNL on step 1 of the Attachment 3.
SAT/UNSAT*
NOTE: The cabinet is not locked, the use of anything (i.e. screwdriver, a key, etc) will turn the
key mechanism to allow the cabinet door to open.
NOTE: DIP switch #9 is located inside the Dryer Control System display cabinet. DIP switch
- 9 is the second switch from the bottom, located between the door and circuit board.
PROMPT: When the performer is going to open the cabinet, give them the pictures, as
needed, of the inside of the cabinet to demonstrate operation of the Dip Switch.
Step 3 - Place DIP SWITCH #9 to ON.
DIP switch is in the on position when the switch is switched away from the circuit
board.
- CRITICAL STEP ** SAT/UNSAT*
Step 4 - Depress MODE (F1) key until CONFIGURE page is displayed.
The MODE (F1) key is depressed until the CONFIGURE page is displayed on
the Dryer Control System display.
SAT/UNSAT*
Step 5 - Depress PAGE (F2) key until CONFIGURE PAGE 2 of 2 is displayed.
The PAGE (F2) key is depressed until the CONFIGURE PAGE 2 of 2 is
displayed on the Dryer Control System display.
SAT/UNSAT*
AOT-OJT-JP-300-K01 Page 4 of 12 Rev. 0
Setting Service Air Dryer Sweep Value to Zero
Step 6 - Depress ENTER (F5).
The ENTER (F5) key is depressed.
SAT/UNSAT*
NOTE: Values are adjusted by using the +(F3) and -(F4) keys. The nominal value for the
setting is 90.
Step 7 - Adjust MAXIMUM SWEEP VALUE to 0(zero) with the +(F3) and -(F4) keys.
Adjusts the MAXIMUM SWEEP VALUE to zero (from 90) using the -(F4) key.
- CRITICAL STEP ** SAT/UNSAT*
Step 8 - Place DIP SWITCH #9 to OFF.
DIP switch is in the on position when the switch is switched away from the circuit
board.
- CRITICAL STEP ** SAT/UNSAT*
Step 9 - Inform control room actions for 0AOP-20.0, Attachment 3 for Service Air Dryer 2A
service air dryer maximum sweep value to zero are complete.
Acknowledge receipt of the information from the performer using proper 3-way
communications.
SAT/UNSAT*
TERMINATING CUE: 0AOP-20.0, Attachment 3, Setting Service Air Dryer(s) Maximum
Sweep Value to Zero, actions are completed, this JPM is complete.
TIME COMPLETE: _____________
NOTE: Comments required for any step evaluated as UNSAT
AOT-OJT-JP-300-K01 Page 5 of 12 Rev. 0
Setting Service Air Dryer Sweep Value to Zero
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute / Two Minute Rule
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and Adherence
(PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer checking) SAT/ UNSAT/ NE
E. Proper Equipment Use (observe starting limitations, throttle valve SAT/ UNSAT/ NE
closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety equipment, SAT/ UNSAT/ NE
etc.)
G. Electrical Safety And Arc Flash Protection Compliance (refer to SAT/ UNSAT/ NE
SAF-NGGC-2175, as applicable)
H. Security Compliance (controlled area entry and exit, key control, SAT/ UNSAT/ NE
etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, etc.) SAT/ UNSAT/ NE
J. Radiation Protection (ALARA, understanding and use of RWP, SAT/ UNSAT/ NE
frisking, etc.)
Comments:
AOT-OJT-JP-300-K01 Page 6 of 12 Rev. 0
Setting Service Air Dryer Sweep Value to Zero
REVISION SUMMARY:
REVISION REVISION SUMMARY
NUMBER
0 New JPM.
AOT-OJT-JP-300-K01 Page 7 of 12 Rev. 0
Setting Service Air Dryer Sweep Value to Zero
Validation Time: 12 Minutes (approximate).
Time Taken: _________
APPLICABLE METHOD OF TESTING
Performance: Simulate X Actual Unit: 2
Setting: In-Plant X Simulator Admin
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes No X
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
AOT-OJT-JP-300-K01 Page 8 of 12 Rev. 0
Page 9 of 12
Page 10 of 12
Page 11 of 12
TASK CONDITIONS:
1. A loss of pneumatics has occurred.
2. 0AOP-20.0 is being executed.
3. The temporary air compressor has been placed in service.
INITIATING CUE:
You are directed by the Unit Two SCO to perform 0AOP-20.0, Pneumatic (Air/Nitrogen)
System Failures, Attachment 3, Setting Service Air Dryer(s) Maximum Sweep Value to Zero,
on 2A Air Dryer and inform the CRS when the required actions are complete.
Page 12 of 12
DUKE ENERGY
BRUNSWICK TRAINING SECTION
LESSON TITLE: DC Ground Calculation
LESSON NUMBER: AOT-ADM-JP-051-05
REVISION NO: 04
Chris Michaels 9/6/16
PREPARER DATE
Josh Ashcroft 9/9/16
TECHNICAL REVIEWER DATE
Josh Ashcroft 9/9/16
LINE REVIEWER/VALIDATOR DATE
Bryan Wooten 9/9/16
LINE REVIEWER/VALIDATOR DATE
Jim Barry 9/12/16
TRAINING SUPERVISION APPROVAL/ DATE
AOT-ADM-JP-051-05 Page 1 of 10 Rev. 04
DC Ground Calculation
REVISION SUMMARY
Minor changes and format updates.
4 Enhanced explanations in Critical Step Bases Table.
Revalidated for 2016 Annual Exam.
Page numbers changed to hidden text on Task Cue sheet.
Replaced format to current format with Duke Energy Progress Logo.
3
Task Conditions were changed to state that 2OP-51 is complete
through step 8.1.2.11 instead of step 8.1.2.8. This change was made
due to step 8.1.2.11 being the last field action.
Updated Work Practices section to match requirements of TAP-301,
Development of Job Performance Measures.
2 Formatting and numbering changes to match the JPM template, rev. 2.
Step where performer demonstrates they can obtain procedure and
verify current revision is no longer required to be observed and
evaluated. Procedure use is now evaluated in the Work Practices
section.
Deletes reference to old Touch STAR and reformats Work Practices
section.
Revises Settings to In-Plant, Simulator, and Admin per the JPM
Template.
Ensured Time Start and Time Stop points are clearly identified.
Added Validation Time and Time Taken to top of Evaluation form.
Changed reference to SCO to CRS.
Added Basis for critical task and annotated in JPM
Deleted KAs that did not match JPM
Changed safety function to electrical from generic
Made step 2 critical and reworded steps 2 and 3
AOT-ADM-JP-051-05 Page 2 of 10 Rev. 04
DC Ground Calculation
RELATED TASKS:
263613B104 Perform DC Ground Isolation For Bus P, N, And P/N Per OP-51.
K/A REFERENCE AND IMPORTANCE RATING:
263000 A4.04 3.0/3.2
Ability to manually operate and/or monitor in the control room - ground detection
circuit.
LIST OF REFERENCES
2OP-51 DC Electrical System Operating Procedure
Section 6.3, DC Ground Isolation for Bus P, N & PN, AND
Attachment 4 Data Sheet for Battery Ground Detection
TOOLS AND EQUIPMENT:
Calculator
SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1):
6 - Electrical
TIME CRITICAL OPERATOR ACTIONS:
None for this JPM
AOT-ADM-JP-051-05 Page 3 of 10 Rev. 04
DC Ground Calculation
SAFETY CONSIDERATIONS:
None
EVALUATOR NOTES: (Do not read to performer)
1. The applicable procedure section WILL be provided to the trainee.
2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1021,
Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).
3. This JPM may be performed on Unit 2.
4. Critical Step Basis
a) Prevents Task Completion
b) May Result in Equipment Damage
c) Affects Public Health and Safety
d) Could Result in Personal Injury
Read the following to the JPM performer
TASK CONDITIONS:
1. The Control Room has received annunciation of a ground on 125/250 VDC
Battery Switchboard 2B.
2. All Battery Chargers are in Float.
3. Another operator has determined the following:
P Bus reading is 1.8 mA.
PN Bus reading is 0.08 mA.
N Bus reading is 2.2 mA.
4. Field operator reports 2OP-51 Section 6.3.1 complete through Step 14.
INITIATING CUE:
You are directed by the Control Operator to complete DC Ground Isolation For Bus
P, N, and PN per 2OP-51 and perform the calculation to determine the total
resistance to ground for 125/250 VDC Battery Switchboard 2B, then report your
findings to the Control Room Supervisor.
AOT-ADM-JP-051-05 Page 4 of 10 Rev. 04
DC Ground Calculation
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
Step 1 - Obtain current revision of 2OP-51 Section 6.3.1 and Attachment 4 and verify
copy if applicable.
Current revision of 2OP-51 Section 6.3.1and Attachment 4 obtained and
verified if applicable.
SAT/UNSAT
TIME START: _________
PROMPT: If asked, inform examinee that battery chargers 2B1 and 2B2 are both in
FLOAT.
Step 2 - DETERMINE battery charger status and RECORDS the voltage on Attachment
4.
135 VDC recorded on attachment for both 2B-1 and 2B-2.
- CRITICAL STEP ** SAT/UNSAT
Step 3 - DETERMINE the value of the battery ground.
Calculates 135 volts + 135 volts 50 K ohms = 17.5 K ohms
1.8 mA + 2.2 mA
Acceptable correct range is 16.5K to 18.5K ohms
- CRITICAL STEP ** SAT/UNSAT
Step 4 - Completes Attachment 4 documentation, including as Performed by.
Documents and Initials Attachment 4.
SAT/UNSAT
AOT-ADM-JP-051-05 Page 5 of 10 Rev. 04
DC Ground Calculation
Step 5 - Informs Control Room Supervisor resistance to ground is 17.5 K ohms.
CRS notified.
SAT/UNSAT
TERMINATING CUE: When calculation is performed and the CRS is informed of the
results the JPM is complete.
Time Completed: ___________
NOTE: Comments required for any step evaluated as UNSAT.
COLLECT AND CONTROL ALL JPM EXAM MATERIALS FOR EXAM SECURITY.
AOT-ADM-JP-051-05 Page 6 of 10 Rev. 04
DC Ground Calculation
Critical Step Bases
Step Critical / Not Critical Basis / Reason
1 Not Critical Administrative
2-3 Critical Calculations must be accurately performed based on the
data provided. Improper calculation will result in incorrect
ground calculation per Attachment 4.
4-5 Not Critical Informing CRS of results.
AOT-ADM-JP-051-05 Page 7 of 10 Rev. 04
DC Ground Calculation
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and Adherence
(PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer SAT/ UNSAT/ NE
checking)
E. Proper Equipment Use (observe starting limitations, throttle SAT/ UNSAT/ NE
valve closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety SAT/ UNSAT/ NE
equipment, etc.)
G. Electrical Safety And Arc Flash Protection Compliance (refer SAT/ UNSAT/ NE
to AD-HS-ALL-0110, as applicable)
H. Security Compliance (controlled area entry and exit, key SAT/ UNSAT/ NE
control, etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, SAT/ UNSAT/ NE
etc.)
J. Radiation Protection (ALARA, understanding and use of SAT/ UNSAT/ NE
RWP, frisking, etc.)
Comments:
AOT-ADM-JP-051-05 Page 8 of 10 Rev. 04
DC Ground Calculation
Validation Time: 7 Minutes (approximate).
Time Taken: ____ Minutes
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: 2
Setting: Classroom X Simulator In-Plant:
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes No X
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
AOT-ADM-JP-051-05 Page 9 of 10 Rev. 04
Read the following to the JPM performer
TASK CONDITIONS:
1. The Control Room has received annunciation of a ground on 125/250 VDC Battery
Switchboard 2B.
2. All Battery Chargers are in Float.
3. Another operator has determined the following:
P Bus reading is 1.8 mA.
PN Bus reading is 0.08 mA.
N Bus reading is 2.2 mA.
4. Field operator reports 2OP-51 Section 6.3.1 complete through Step 14.
INITIATING CUE:
You are directed by the Control Operator to complete DC Ground Isolation For Bus P,
N, and PN per 2OP-51 and perform the calculation to determine the total resistance to
ground for 125/250 VDC Battery Switchboard 2B, then report your findings to the
Control Room Supervisor.
DUKE ENERGY PROGRESS
BRUNSWICK NUCLEAR PLANT
LESSON TITLE: Evaluate Jet Pump Performance Per 0PT-13.1
LESSON NUMBER: LOT-ADM-JP-002-02
REVISION NO: 1
Original accidentally shredded
PREPARER/DATE
TECHNICAL REVIEWER/DATE
VALIDATOR/DATE
Eddie Rau 2/17/2015
TRAINING SUPERVISION APPROVAL/ DATE
Evaluate Jet Pump Performance Per 0PT-13.1
RELATED TASKS:
202001B201, Perform Reactor Recirculation Jet Pump Operability Test Per PT-13.1
K/A REFERENCE AND IMPORTANCE RATING:
GEN 2.1.33 3.4/4.0
REFERENCES:
0PT-13.1
TOOLS AND EQUIPMENT:
Calculator
ADMINISTRATIVE CATEGORY (from NUREG 1123, Rev 2. Supp 1):
Generic 2.1 - Conduct of Operations
LOT-ADM-JP-002-02 Page 2 of 14 Rev. 1
Evaluate Jet Pump Performance Per 0PT-13.1
SAFETY CONSIDERATIONS
None
EVALUATOR NOTES: (Do not read to trainee)
1. The applicable procedure section WILL be provided to the trainee.
2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021,
Appendix E, or similar to the trainee.
Read the following to trainee.
TASK CONDITIONS:
1. Unit Two is operating at 100% power with all recirculation loop flow and differential
pressure indicators operable. Recirc/Jet Pump readings are as indicated in the provided
attachment.
INITIATING CUE:
You are directed by the Unit CRS to perform 0PT-13.1, Reactor Recirculation Jet Pump
Operability, using the data provided. Determine if the surveillance is completed SAT or
UNSAT.
Results of 0PT-13.1.
SURVEILLANCE
SAT UNSAT
LOT-OJT-JP-002-A01 Page 3 of 14 REV. 1
Evaluate Jet Pump Performance Per 0PT-13.1
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
Step 1 - Student performs applicable CORE 4 practice(s) prior to commencing task.
CORE 4 trait used, if applicable.
SAT/UNSAT*
Time Start: ___________
NOTE: The following steps require evaluation of curves located in 2OP-02, Attachments 4
through 17.
Step 2 - Obtain Recirculation Pump A Calculated Speed, Panel on P603.
Recirculation Pump A Calculated Speed determined to be 88.76%.
- CRITICAL STEP ** SAT/UNSAT*
Step 3 - Obtain Recirculation Pump B Calculated Speed, Panel on P603.
Recirculation Pump B Calculated Speed determined to be 90.43%.
- CRITICAL STEP ** SAT/UNSAT*
NOTE: The readings in steps 4 and 5 may vary slightly from student to student. Reading should
be within 1000 gpm of listed values.
Step 4 - Obtain Recirculation Pump A flow on B32-FR-R614.
Recirculation Pump A flow on B32-FR-R614 determined to be 45 x 1000 gpm.
- CRITICAL STEP ** SAT/UNSAT*
Step 5 - Obtain Recirculation Pump B flow on B32-FR-R614.
Recirculation Pump B flow on B32-FR-R614 determined to be 46 x 1000 gpm.
- CRITICAL STEP ** SAT/UNSAT*
NOTE: The readings in steps 6 and 7 may vary slightly from student to student. Reading
should be within 1x106 lbm/hr of listed values.
LOT-OJT-JP-002-A01 Page 4 of 14 REV. 1
Evaluate Jet Pump Performance Per 0PT-13.1
Step 6 - Obtain Jet Pump Loop A flow from Flow Indicator B21-FI-R611A.
Determines Jet Pump Loop A flow is 39x106 lbm/hr
- CRITICAL STEP ** SAT/UNSAT*
Step 7 - Obtain Jet Pump Loop B flow from Flow Indicator B21-FI-R611B.
Determines Jet Pump Loop B flow is 39x106 lbm/hr
- CRITICAL STEP ** SAT/UNSAT*
Step 8 - Record the current revision number of 2OP-02, Reactor Recirculation System
Operating Procedure.
Current revision of 2OP-02 recorded.
SAT/UNSAT*
NOTE: The readings in step 9, 10 & 11 may vary slightly from student to student. Reading
should be within 1% of listed values.
Step 9 - Record individual jet pump % psid.
Records the following values:
JP 1 @ 31% JP 11 @ 37%
JP 2 @ 31% JP 12 @ 36%
JP 3 @ 40% JP 13 @ 35%
JP 4 @ 33% JP 14 @ 36%
JP 5 @ 40% JP 15 @ 40%
JP 6 @ 35% JP 16 @ 41%
JP 7 @ 32% JP 17 @ 37%
JP 8 @ 32% JP 18 @ 34%
JP 9 @ 34% JP 19 @ 34%
JP 10 @ 32% JP 20 @ 37%
SAT/UNSAT*
LOT-OJT-JP-002-A01 Page 5 of 14 REV. 1
Evaluate Jet Pump Performance Per 0PT-13.1
Step 10 - Determine Loop A average % psid.
Loop A average % psid is 34.0% psid (acceptable range 33.0% - 35.0%).
- CRITICAL STEP ** SAT/UNSAT*
Step 11 - Determine Loop B average % psid.
Loop B average % psid is 36.7% psid (acceptable range 35.7% - 37.7%).
- CRITICAL STEP ** SAT/UNSAT*
Step 12 - Compare the % psid for each jet pump and calculated average jet pump% psid for
each loop with the established curves.
Determines Jet Pump #3 is UNSAT
- CRITICAL STEP ** SAT/UNSAT*
Step 13 - Notify CRS that PT-13.1, Jet Pump Performance is completed UNSAT due to Jet
Pump #3 being out of range.
Determines Jet Pump #3 reading is UNSAT
SAT/UNSAT*
TERMINATING CUE: When examinee has evaluated jet pump operability per 0PT-13.1 this
JPM is complete.
- Comments required for any step evaluated as UNSAT.
Time Complete: ___________
LOT-OJT-JP-002-A01 Page 6 of 14 REV. 1
Evaluate Jet Pump Performance Per 0PT-13.1
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute
3. Correct Component Verification (CCV), Validate Assumptions
4. Procedure and Work Instruction Use and Adherence (PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer checking) SAT/ UNSAT/ NE
E. Proper Equipment Use (observe starting limitations, throttle valve SAT/ UNSAT/ NE
closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety equipment, SAT/ UNSAT/ NE
etc.)
G. Electrical Safety And Arc Flash Protection Compliance (refer to SAT/ UNSAT/ NE
SAF-NGGC-2175, as applicable)
H. Security Compliance (controlled area entry and exit, key control, SAT/ UNSAT/ NE
etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, etc.) SAT/ UNSAT/ NE
J. Radiation Protection (ALARA, understanding and use of RWP, SAT/ UNSAT/ NE
frisking, etc.)
COMMENTS:
LOT-OJT-JP-002-A01 Page 7 of 14 REV. 1
Evaluate Jet Pump Performance Per 0PT-13.1
Validation Time: 20 Minutes (approximate)
Time Taken: ___ Minutes
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: N/A
Setting: In-Plant Simulator Admin X
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes No X
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
LOT-OJT-JP-002-A01 Page 8 of 14 REV. 1
Revision Summary:
REVISION REVISION SUMMARY
NUMBER
Formatting changes to match the JPM template, revision 3.
Validated using 0PT-13.1 Reactor Recirculation Jet Pump Operability
revision 44 and 2OP-02 Reactor Recirculation System Operating Procedure
revision 161.
1 Added attachment that includes photos to enable being performed in
classroom.
Added Duke Energy Progress image to cover page.
Updated work practices to current plant expectations.
Removed Step 1 which was obtaining the procedure (it is given to them).
Page 9 of 14
Page 10 of 14
Page 11 of 14
Page 12 of 14
Page 13 of 14
TASK CONDITIONS:
1. Unit Two is operating at 100% power with all recirculation loop flow and differential
pressure indicators operable. Recirc/Jet Pump readings are as indicated in the provided
attachment.
INITIATING CUE:
You are directed by the Unit CRS to perform 0PT-13.1, Reactor Recirculation Jet Pump
Operability, using the data provided. Determine if the surveillance is completed SAT or
UNSAT.
Inform the Unit CRS of the results.
SURVEILLANCE
SAT UNSAT
Page 14 of 14
CAROLINA POWER & LIGHT COMPANY
BRUNSWICK TRAINING SECTION
ADMIN
LOT-ADM-JP-201-E01
LESSON TITLE: Evaluate Proposed Temporary Change.
REVISION NO: 0
RECOMMENDED BY: K A Bowdon 03/13/03
Instructor/Developer DATE
CONCURRENCE BY: Michael S Williams 7/31/03
Line Superintendent/Supervisor DATE
APPROVED BY: M A Pearson 08/05/03
Superintendent/Supervisor Training DATE
Evaluate Proposed Temporary Change.
Special Instructions
Obtain copies of prints D-02549, Sheet 1B, 2-FP-05924, Sheet 2, LL-09243, Sheet 15 and
LL09244, Sheet 15
NOTE: Electronic file of drawing changes available under K:\Training\JPM\Admin JPMS
(Power Point File). This file can be used to denote drawing changes without actually
marking copies of plant drawings.
On D-02549, change V35 and V36 to show closed, and denote valves closed per EC TC
On 2-FP-9524, Sheet 2, draw cloud around G41-N005 connections 1NC and 1C, denote
leads lifted per EC TC 03-001.
On 2-FP-9524, Sheet 2, draw cloud around G41-N005 connections 2NC and 2C, denote
leads lifted per EC TC 03-001.
On LL-09243, Sheet 15, draw jumper from terminals 2-H02/CC5 to 2-H02/AA11. Denote
jumper installed per TC EC 03-001.
On LL-09244, Sheet 15, draw jumper from terminals 2-H02/CC17 to 2-H02/BB14. Denote
jumper installed per TC EC 03-001.
Mark up EGR-NGGC-005, Form 2 to reflect temporary changes.
LOT-ADM-JP-201-E01 Page 2 of 9 REV. 0
Evaluate Proposed Temporary Change.
SAFETY CONSIDERATIONS:
None.
EVALUATOR NOTES: (Do not read to trainee)
1. The applicable procedure section WILL NOT be provided to the trainee.
2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021,
Appendix E, or similar to the trainee.
3. This JPM may be administered in any setting.
Read the following to trainee.
TASK CONDITIONS:
1. Unit Two (2) Fuel Pool Cooling Skimmer Surge Tank low level alarm switch 2-G41-LSL-
N005 has failed. The shelf condition of this level switch is closed. This failed level
switch has resulted in Annunciator A-4 6-7 (Fuel Pool Cooling Alarm) being sealed in on
Panel P603.
2. The failed level switch is obsolete. A temporary change has been developed in
accordance with EGR-NGGC-0005 to abandon this level switch in place until a suitable
replacement can be located and installed.
3. This temporary change is to valve out the failed level switch and lift leads and/or jumper
the level switch contacts. This temporary change will defeat the level switch input to
Annunciator A-4 6-7 (a multiple input Annunciator). This temporary change will also
defeat low level indicating lights located on local panels 2-G41P001 and 2-G41-P002.
The low level trip function for the Fuel Pool Cooling Pumps will not be affected.
4. A Temporary Change Log form (EGR-NGGC-0005, Form 2) has been prepared
indicating a description of this temporary change and the Priority 0 drawings affected.
The Priority 0 drawings referenced have been marked up to reflect the temporary
change.
LOT-ADM-JP-201-E01 Page 3 of 9 REV. 0
Evaluate Proposed Temporary Change.
INITIATING CUE:
You are directed by the WCC SRO to review the marked up Priority 0 drawings identified
on the Temporary Change Log Form 2 and ensure the changes reflected on the identified
Priority 0 drawings meet the intent of the temporary change, and inform the WCC SRO of
the results of your evaluation.
LOT-ADM-JP-201-E01 Page 4 of 9 REV. 0
Evaluate Proposed Temporary Change.
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
Step 1 - Review Temporary Change Log, Form 2 and evaluate drawings listed as Priority 0 on
the Log.
Reviews Temporary Change Log, Form 2 and evaluate drawings listed as Priority
0 on the Log.
SAT/UNSAT*
Step 2 - Determine jumper installed from terminals 2-H02/CC5 to 2-H02/AA11 on drawing LL-
09243, Sheet is for level switch N006, not N005 and should not be installed.
Determines jumper should not be installed from terminals 2-H02/CC5 to 2-
H02/AA11 on LL-09243, Sheet 15.
- CRITICAL STEP ** SAT/UNSAT*
Step 3 - Determine jumper installed from terminals 2-H02/CC17 to 2-H02/BB23 on drawing
LL-09244, Sheet is for level switch N006, not N005 and should not be installed.
Determines jumper should not be installed from terminals 2-H02/CC17 to 2-
H02/BB23 on LL-09244, Sheet 15.
- CRITICAL STEP ** SAT/UNSAT*
LOT-ADM-JP-201-E01 Page 5 of 9 REV. 0
Evaluate Proposed Temporary Change.
Step 4 - Determine other drawing changes meet the intent of the TC.
Determines other drawing changes meet the intent of the TC.
- CRITICAL STEP ** SAT/UNSAT*
Step 5 - Inform WCC SRO of results.
SAT/UNSAT*
TERMINATING CUE: This box will tell the evaluator when the JPM is complete based on
student performance.
- Comments required for any step evaluated as UNSAT.
LOT-ADM-JP-201-E01 Page 6 of 9 REV. 0
Evaluate Proposed Temporary Change.
RELATED TASKS:
299020B301, Develop A Clearance Per OPS-NGGC-1301
K/A REFERENCE AND IMPORTANCE RATING:
GEN 2.2.11 2.5/3.4
REFERENCES:
Prints D-02549, Sheet 1B, 2-FP-05924, Sheet 2, LL-09243, Sheet 15 and LL09244, Sheet
15
TOOLS AND EQUIPMENT:
None.
ADMINISTRATIVE CATEGORY (from NUREG 1123)
Equipment Control
REASON FOR REVISION:
New JPM developed for 2003 Initial License Exam.
LOT-ADM-JP-201-E01 Page 7 of 9 REV. 0
Evaluate Proposed Temporary Change.
Time Required for Completion: 30 Minutes (approximate).
APPLICABLE METHOD OF TESTING
Performance: Simulate 4 Actual Unit: 2
Setting: Control Room Simulator ( Not applicable to In-Plant JPMs )
Time Critical: Yes No 4 Time Limit N/A
Alternate Path: Yes No 4
EVALUATION
Trainee:
JPM: Pass Fail
Remedial Training Required: Yes No
Did Trainee Verify Procedure as Authorized Copy?: Yes No
(Each Student should verify one JPM per evaluation set.)
Comments:
Comments reviewed with Student
Evaluator Signature: Date:
LOT-ADM-JP-201-E01 Page 8 of 9 REV. 0
TASK CONDITIONS:
1. Unit Two (2) Fuel Pool Cooling Skimmer Surge Tank low level alarm switch 2-G41-LSL-
N005 has failed. The shelf condition of this level switch is closed. This failed level
switch has resulted in annunciator A-4 6-7 (Fuel Pool Cooling Alarm) being sealed in on
Panel P603.
2. The failed level switch is obsolete. A temporary change has been developed in
accordance with EGR-NGGC-0005 to abandon this level switch in place until a suitable
replacement can be located and installed.
3. This temporary change is to valve out the failed level switch and lift leads and/or jumper
the level switch contacts. This temporary change will defeat the level switch input to
annunciator A-4 6-7 (a multiple input annunciator). This temporary change will also
defeat low level indicating lights located on local panels 2-G41P001 and 2-G41-P002.
The low level trip function for the Fuel Pool Cooling Pumps will not be affected.
4. A Temporary Change Log form (EGR-NGGC-0005, Form 2) has been prepared
indicating a description of this temporary change and the Priority 0 drawings affected.
The Priority 0 drawings referenced have been marked up to reflect the temporary
change.
INITIATING CUE:
You are directed by the WCC SRO to review the marked up Priority 0 drawings identified
on the Temporary Change Log Form 2 and ensure the changes reflected on the identified
Priority 0 drawings meet the intent of the temporary change, and inform the WCC SRO of
the results of your evaluation.
DUKE ENERGY
BRUNSWICK TRAINING SECTION
LESSON TITLE: Determine TEDE While Working in a High Airborne Area
LESSON NUMBER: LOT-ADM-XXXXX
REVISION NO: 0
PREPARER / DATE
TECHNICAL REVIEWER / DATE
VALIDATOR / DATE
LINE SUPERVISOR / DATE
TRAINING SUPERVISION APPROVAL / DATE
RELATED TASKS:
None
K/A REFERENCE AND IMPORTANCE RATING:
Generic 2.3.4 3.2/3.7
Knowledge of Radiation Exposure Limits under normal or emergency conditions
REFERENCES:
PD-RP-ALL-0001, Rev. 4, Radiation Worker Responsibilities
TOOLS AND EQUIPMENT:
Calculator
SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1):
A.3 Radiation Control
SETUP INSTRUCTIONS
None
SAFETY CONSIDERATIONS:
1. None
EVALUATOR NOTES: (Do not read to performer)
1. The applicable procedure section WILL be provided to the trainee.
2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1021,
Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).
3. Critical Step Basis
a) Prevents Task Completion
b) May Result in Equipment Damage
c) Affects Public Health and Safety
d) Could Result in Personal Injury
Read the following to the JPM performer.
TASK CONDITIONS:
I will explain the initial conditions, which steps to simulate, discuss or perform, and
provide initiating cues. When you complete the task successfully, the objective for this
Job Performance Measure will be satisfied.
Initial Conditions:
- The plant is shut down for refueling
- While doing contract work the operator received a combined 1,325 mRem TEDE this
calendar year.
- In the same calendar year, the operator has received another 495 mRem TEDE at the
Harris plant.
- The estimated dose rate in the work area is 465 mRem/hr.
- An airborne contamination concern exists.
- It is estimated that it will take 25 minutes to complete the inspection if the operator uses
a respirator.
- If the operator does NOT wear a respirator, the inspection will take 12 minutes, but
Radiation Protection projects that the internal exposure will be 10 DAC-hrs.
Initiating Cue:
1.a. Calculate the resultant total effective dose equivalent for with a respirator for this job and
the accumulated dose for the year.
1.b. Calculate the resultant total effective dose equivalent for without a respirator for this job
and the accumulated dose for the year.
2. Using the lowest dose determined in number 1, determine if the individual can perform
the task without exceeding Duke Energys Annual Administrative Dose Limit.
Show your calculations on the next page
Task Standard:
Determination made that NOT wearing a respirator will result in a lower
TEDE and that the individual can perform the task without exceeding Duke
Energys Annual Administrative Dose Limit.
Required Materials:
Laptop with General Reference PD-RP-ALL-0001, Radiation Worker
Responsibilities, Rev 3 or hard copy
General References:
PD-RP-ALL-0001, Radiation Worker Responsibilities, Rev. 4
Handouts:
JPM Cue Sheet
Time Critical Task:
No
Validation Time:
XXXXX
Critical Step Justification
Determining external exposure while wearing a respirator is required to
Step 2
complete the calculations and answer.
Determining internal exposure while NOT wearing a respirator is
Step 4
required to complete the calculations and answer.
Determining external exposure while NOT wearing a respirator is
Step 5
required to complete the calculations and answer.
Determining individuals total exposure for the year if the work is allowed
Step 8
without a respirator is required to complete the calculations and answer.
Determining individuals total exposure for the year if the work is allowed
Step 9
with a respirator is required to complete the calculations and answer.
Determining if the individual can perform the work without exceeding
Step 10 Duke Energys Annual Administrative Dose Limit of 2000 mRem is
required to complete the calculations and answer.
PERFORMANCE CHECKLIST
Start Time: __________.
NOTE: Steps in this JPM may be performed in any order.
Performance Step: 1 Determines internal exposure while wearing a respirator
Standard: Determines internal exposure to be ZERO while wearing a
respirator
Comment:
Performance Step: 2 Determines external exposure while wearing a respirator
Standard: Determines external exposure to be 193.8 mRem TEDE while
wearing a respirator
(465 mRem / hr x 25 min = 193.8 mRem)
(NOTE: Could round to 194)
Comment:
Performance Step: 3 Determines TOTAL exposure while wearing a respirator
Standard: Determines total exposure to be 193.8 mRem while wearing a
respirator
(0 mRem internal + 193.8 mRem external = 193.8 mRem total)
(NOTE: Could round to 194)
Comment:
Indicates a Critical Step
Performance Step: 4 Determines internal exposure while NOT wearing a respirator
Standard: Determines internal exposure to be 25 mRem while not wearing
a respirator
(2.5 mRem / hr x 10 DAC-hr = 25 mRem)
(NO tolerance)
Comment:
Performance Step: 5 Determines external exposure while NOT wearing a respirator
Standard: Determines external exposure to be 93.0 mRem TEDE while not
wearing a respirator
(465 mRem / hr x 12 min = 93 mRem)
(NO tolerance)
Comment:
Performance Step: 6 Determines TOTAL exposure while NOT wearing a respirator
Standard: Determines total exposure to be 118 mRem while not wearing a
respirator
(25 mRem internal + 93 mRem external = 118 mRem total)
(NO tolerance)
Comment:
Performance Step: 7 Determines individuals total exposure for the year
Standard: Determines individuals total exposure for the year to be 1820
mRem
(1325 mRem + 495 mRem = 1820 mRem)
Comment:
Indicates a Critical Step
Performance Step: 8 Determines individuals total exposure for the year if the work is
allowed without a respirator
Standard: Determines individuals total exposure for the year if the work is
performed without a respirator to be 1938 mRem
(1820 mRem + 118 mRem = 1938 mRem)
(NO tolerance)
Comment:
Performance Step: 9 Determines individuals total exposure for the year if the work is
allowed with a respirator
Standard: Determines individuals total exposure for the year if the work is
performed with a respirator to be 2013.8 mRem
(1820 mRem + 193.8 mRem = 2013.8 mRem)
Note: If calculated wearing a respirator the total exposure for the
year will be 2013.8 mRem and work cannot be performed without
an extension. The directions were to use the lowest dose and
this represents UNSAT performance.
(NOTE: Could have rounded 193.8 to 194 and answer would
be 2014 mRem)
Comment:
Performance Step: 10 Determines if the individual can perform the work without
exceeding Duke Energys Annual Administrative Dose Limit of
2000 mRem without an administrative dose limit extension
Standard: Determines the individual CAN perform the work without wearing
a respirator will NOT exceed Duke Energys Annual
Administrative Dose Limit of 2000 mRem
(1820 mRem + 118mRem = 1938 mRem)
Comment:
When all calculations have been completed and the
Evaluator Cue: determination that work can proceed, this JPM is complete.
Stop Time: _________
Indicates a Critical Step
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and
Adherence (PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer SAT/ UNSAT/ NE
checking)
E. Proper Equipment Use (observe starting limitations, SAT/ UNSAT/ NE
throttle valve closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety SAT/ UNSAT/ NE
equipment, etc.)
G. Electrical Safety And Arc Flash Protection Compliance SAT/ UNSAT/ NE
(refer to AD-HS-ALL-0110, as applicable)
H. Security Compliance (controlled area entry and exit, key SAT/ UNSAT/ NE
control, etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, SAT/ UNSAT/ NE
etc.)
J. Radiation Protection (ALARA, understanding and use of SAT/ UNSAT/ NE
RWP, frisking, etc.)
Comments:
Validation Time: XX Minutes (approximate).
Time Taken: ____ Minutes
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: 1
Setting: In-Plant Simulator Admin X
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes No X
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
KEY
1.a Calculation for resultant total effective dose equivalent with a respirator.
Determines internal exposure to be ZERO while wearing a respirator
Determines external exposure to be 193.8 mRem TEDE while wearing a respirator
(465 mRem / hr x 25 min = 193.8 mRem)
(NOTE: Could round to 194)
Determines TOTAL exposure while wearing a respirator
(0 mRem internal + 193.8 mRem external = 193.8 mRem total)
(NOTE: Could round to 194)
Determines individuals total exposure for the year to be 1820 mRem
(1325 mRem + 495 mRem = 1820 mRem)
Determines individuals total exposure for the year if the work is allowed to be 2013.8 mRem
(1820 mRem + 193.8 = 2013.8 mRem)
(NOTE: Could round to 2014 mRem)
Note: If calculated wearing a respirator the total exposure for the year will be 2013.8 mRem and
work CANNOT be performed without an extension. The directions were to use the lowest dose
and this represents UNSAT performance.
1.b Calculation for resultant total effective dose equivalent without a respirator.
Determines internal exposure to be 25 mRem while not wearing a respirator
(2.5 mRem / hr x 10 DAC-hr = 25 mRem)
(NO tolerance)
Determines external exposure to be 118 mRem TEDE while not wearing a respirator
(495 mRem / hr x 12 min = 93 mRem)
Determines total exposure to be 118 mRem while not wearing a respirator
(25 mRem internal + 93 mRem external = 118 mRem total)
Determines individuals total exposure for the year if the work is allowed to be 1938 mRem
(1820 mRem + 118 = 1938 mRem)
The individual CAN perform the task without exceeding Duke Energys Annual Admin Dose
Limit if the task is performed WITHOUT a respirator.
KEY
2. Using the lowest dose determined from the above calculations (1a or 1b):
CAN the individual perform the task without exceeding Duke Energys Annual Administrative
Dose Limit?
YES (without a respirator total dose would be 1,938 mRem which is < 2,000 mRem)
TASK CONDITIONS:
I will explain the initial conditions, which steps to simulate, discuss or perform, and provide
initiating cues. When you complete the task successfully, the objective for this Job
Performance Measure will be satisfied.
Initial Conditions:
- The plant is shut down for refueling
- While doing contract work the operator received a combined 1,325 mRem TEDE this calendar
year.
- In the same calendar year, the operator has received another 495 mRem TEDE at the Harris
plant.
- The estimated dose rate in the work area is 465 mRem/hr.
- An airborne contamination concern exists.
- It is estimated that it will take 25 minutes to complete the inspection if the operator uses a
respirator.
- If the operator does NOT wear a respirator, the inspection will take 12 minutes, but Radiation
Protection projects that the internal exposure will be 10 DAC-hrs.
Initiating Cue:
1.a. Calculate the resultant total effective dose equivalent for with a respirator for this job and the
accumulated dose for the year.
1.b. Calculate the resultant total effective dose equivalent for without a respirator for this job and the
accumulated dose for the year.
2. Using the lowest dose determined in number 1, determine if the individual can perform the task
without exceeding Duke Energys Annual Administrative Dose Limit.
Show your calculations on the next page
Name:
Date:
1.a Calculation for resultant total effective dose equivalent with a respirator.
1.b Calculation for resultant total effective dose equivalent without a respirator.
2. Using the lowest dose determined from the above calculations (1a or 1b):
CAN the individual perform the task without exceeding Duke Energys Annual Administrative
Dose Limit?
DUKE ENERGY PROGRESS
BRUNSWICK NUCLEAR PLANT
LESSON TITLE: Determine Reportability Requirements
LESSON NUMBER:
REVISION NO: 0
PREPARER/DATE
TECHNICAL REVIEWER/DATE
VALIDATOR/DATE
TRAINING SUPERVISION APPROVAL/ DATE
RELATED TASKS:
K/A REFERENCE AND IMPORTANCE RATING:
2.1.25 (3.9/4.2)
REFERENCES:
0OI-01.07, Notifications, Rev. 38
TOOLS AND EQUIPMENT:
None
ADMINISTRATIVE CATEGORY (from NUREG 1123, Rev 2. Supp 1):
Conduct of Operations
SAFETY CONSIDERATIONS:
None.
SPECIAL INSTRUCTIONS:
None
EVALUATOR NOTES: (Do not read to trainee)
1. For Licensed personnel, provide memory stick with POM documents and access to
non-network computer.
2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021,
Appendix E, or similar to the trainee.
3. This JPM may be administered in the simulator, control room, or classroom setting.
Page 2 of 8 REV. 0
Read the following to performer.
TASK CONDITIONS:
- The Unit 1 is in Mode 3.
- RCIC is out of service and isolated.
- Shutdown Cooling A was being aligned for operation.
- While aligning Shutdown Cooling A, an error caused flow to be diverted to the Suppression Pool
via the minimum flow valve.
Level 3.
- Reactor water level dropped to 15 inches Wide Range before the inventory reduction was
terminated by the successful RHR isolation.
- CRD has been used to raise water level back to the normal band.
INITIATING CUE:
- You are the Shift Manager.
- Determine the MOST limiting reportability of this incident and complete Attachments 1 and
3 of 0OI-01.07, Notifications, if applicable.
- Another SRO will complete any other administrative requirements such as initiating LCOs,
CRs, Work Requests, etc.
Page 3 of 8 REV. 0
PERFORMANCE CHECKLIST
START TIME:
1. Obtain a copy of 0OI-01.07, Notifications
SAT/UNSAT
2. Utilize attachment 1 and determine that an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification is required due to
item 3.3 as the event did result in a valid actuation of one of the systems listed.
Did the event or condition result in valid actuation of any of the systems listed below except
when the actuation resulted from and is part of a pre-planned sequence during testing or
reactor operation? (Note 1) [10 CFR 50.72(b)(3)(iv)(A)]
These systems are:
Reactor protection system (RPS) including: reactor scram and reactor trip.
[10 CFR 50.72(b)(3)(iv)(B)(1)]
YES
- CRITICAL STEP ** SAT/UNSAT*
3. Fill out attachment 3 of 0OI-01.07, page 1 is only one required
Applicant must check block under the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification per 10 CFR 50.72 (b)(3) for Specified
System Actuation.
- CRITICAL STEP ** SAT/UNSAT*
TERMINATING CUE: When applicant indicates that the reportability has been determined.
- Comments required for any step evaluated as UNSAT.
TIME COMPLETE: __________
Page 4 of 8 REV. 0
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and Adherence (PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer SAT/ UNSAT/ NE
checking)
E. Proper Equipment Use (observe starting limitations, throttle SAT/ UNSAT/ NE
valve closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety SAT/ UNSAT/ NE
equipment, etc.)
G. Electrical Safety And Arc Flash Protection Compliance (refer to SAT/ UNSAT/ NE
SAF-NGGC-2175, as applicable)
H. Security Compliance (controlled area entry and exit, key control, SAT/ UNSAT/ NE
etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, etc.) SAT/ UNSAT/ NE
J. Radiation Protection (ALARA, understanding and use of RWP, SAT/ UNSAT/ NE
frisking, etc.)
COMMENTS:
Page 5 of 8 REV. 0
Validation Time: Minutes (approximate)
Time Taken: ___ Minutes
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: N/A
Setting: In-Plant Simulator Admin X
Time Critical: Yes X No Time Limit 15
Alternate Path: Yes No X
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
Page 6 of 8 REV. 0
Revision Summary:
REVISION REVISION SUMMARY
NUMBER
0 New JPM for 2017 ILOT NRC Exam
Page 7 of 8 REV. 0
TASK CONDITIONS:
1. 2OP-05, Standby Liquid Control System Operating Procedure Section 8.6, Manual
Volume determination has been completed.
2. The distance from the surface of the liquid to the top of the tank is 71 inches.
3. The B-10 Atomic Enrichment is 51%.
4. Standby Liquid Control tank solution temperature (2-C41-TIC-R002) is 66°F.
5. The SLC Tank concentration is 8.7%.
INITIATING CUE:
You are directed by the CRS to verify SLC normal system operating parameters and inform
the Control Room Supervisor of the results.
Page 8 of 8
DUKE ENERGY PROGRESS
BRUNSWICK NUCLEAR PLANT
LESSON TITLE: Evaluate Jet Pump Performance Per 0PT-13.1
LESSON NUMBER: LOT-ADM-JP-002-02 SRO Version
REVISION NO: 2
Original accidentally shredded
PREPARER/DATE
TECHNICAL REVIEWER/DATE
VALIDATOR/DATE
Eddie Rau 2/17/2015
TRAINING SUPERVISION APPROVAL/ DATE
Evaluate Jet Pump Performance Per 0PT-13.1
RELATED TASKS:
202001B201, Perform Reactor Recirculation Jet Pump Operability Test Per PT-13.1
K/A REFERENCE AND IMPORTANCE RATING:
GEN 2.1.33 3.4/4.0
REFERENCES:
0PT-13.1
TOOLS AND EQUIPMENT:
Calculator
ADMINISTRATIVE CATEGORY (from NUREG 1123, Rev 2. Supp 1):
Generic 2.1 - Conduct of Operations
LOT-ADM-JP-002-02 Page 2 of 14 Rev. 1
Evaluate Jet Pump Performance Per 0PT-13.1
SAFETY CONSIDERATIONS
None
EVALUATOR NOTES: (Do not read to trainee)
1. The applicable procedure section WILL be provided to the trainee.
2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021,
Appendix E, or similar to the trainee.
Read the following to trainee.
TASK CONDITIONS:
1. Unit Two is operating at 100% power with all recirculation loop flow and differential
pressure indicators operable. Recirc/Jet Pump readings are as indicated in the provided
attachment.
INITIATING CUE:
You are directed by the Unit CRS to perform 0PT-13.1, Reactor Recirculation Jet Pump
Operability, using the data provided. Evaluate the results of the surveillance
LOT-OJT-JP-002-A01 Page 3 of 14 REV. 1
Evaluate Jet Pump Performance Per 0PT-13.1
PERFORMANCE CHECKLIST
NOTE: Sequence is assumed unless denoted in the Comments.
Step 1 - Student performs applicable CORE 4 practice(s) prior to commencing task.
CORE 4 trait used, if applicable.
SAT/UNSAT*
Time Start: ___________
NOTE: The following steps require evaluation of curves located in 2OP-02, Attachments 4
through 17.
Step 2 - Obtain Recirculation Pump A Calculated Speed, Panel on P603.
Recirculation Pump A Calculated Speed determined to be 88.76%.
- CRITICAL STEP ** SAT/UNSAT*
Step 3 - Obtain Recirculation Pump B Calculated Speed, Panel on P603.
Recirculation Pump B Calculated Speed determined to be 90.43%.
- CRITICAL STEP ** SAT/UNSAT*
NOTE: The readings in steps 4 and 5 may vary slightly from student to student. Reading should
be within 1000 gpm of listed values.
Step 4 - Obtain Recirculation Pump A flow on B32-FR-R614.
Recirculation Pump A flow on B32-FR-R614 determined to be 45 x 1000 gpm.
- CRITICAL STEP ** SAT/UNSAT*
Step 5 - Obtain Recirculation Pump B flow on B32-FR-R614.
Recirculation Pump B flow on B32-FR-R614 determined to be 46 x 1000 gpm.
- CRITICAL STEP ** SAT/UNSAT*
NOTE: The readings in steps 6 and 7 may vary slightly from student to student. Reading
should be within 1x106 lbm/hr of listed values.
LOT-OJT-JP-002-A01 Page 4 of 14 REV. 1
Evaluate Jet Pump Performance Per 0PT-13.1
Step 6 - Obtain Jet Pump Loop A flow from Flow Indicator B21-FI-R611A.
Determines Jet Pump Loop A flow is 39x106 lbm/hr
- CRITICAL STEP ** SAT/UNSAT*
Step 7 - Obtain Jet Pump Loop B flow from Flow Indicator B21-FI-R611B.
Determines Jet Pump Loop B flow is 39x106 lbm/hr
- CRITICAL STEP ** SAT/UNSAT*
Step 8 - Record the current revision number of 2OP-02, Reactor Recirculation System
Operating Procedure.
Current revision of 2OP-02 recorded.
SAT/UNSAT*
NOTE: The readings in step 9, 10 & 11 may vary slightly from student to student. Reading
should be within 1% of listed values.
Step 9 - Record individual jet pump % psid.
Records the following values:
JP 1 @ 31% JP 11 @ 37%
JP 2 @ 31% JP 12 @ 36%
JP 3 @ 40% JP 13 @ 35%
JP 4 @ 33% JP 14 @ 36%
JP 5 @ 40% JP 15 @ 40%
JP 6 @ 35% JP 16 @ 41%
JP 7 @ 32% JP 17 @ 37%
JP 8 @ 32% JP 18 @ 34%
JP 9 @ 34% JP 19 @ 34%
JP 10 @ 32% JP 20 @ 37%
SAT/UNSAT*
LOT-OJT-JP-002-A01 Page 5 of 14 REV. 1
Evaluate Jet Pump Performance Per 0PT-13.1
Step 10 - Determine Loop A average % psid.
Loop A average % psid is 34.0% psid (acceptable range 33.0% - 35.0%).
- CRITICAL STEP ** SAT/UNSAT*
Step 11 - Determine Loop B average % psid.
Loop B average % psid is 36.7% psid (acceptable range 35.7% - 37.7%).
- CRITICAL STEP ** SAT/UNSAT*
Step 12 - Compare the % psid for each jet pump and calculated average jet pump% psid for
each loop with the established curves.
Determines Jet Pump #3 is UNSAT
- CRITICAL STEP ** SAT/UNSAT*
Step 13 - Determine that PT-13.1, Jet Pump Performance is completed UNSAT due to Jet
Pump #3 being out of range.
Determines Jet Pump #3 reading is UNSAT
SAT/UNSAT*
Step 14 - Determine that TS 3.4.2 is applicable and the unit is required to be in mode 3 in 12
hours..
Determines TS requiremnets
- CRITICAL STEP ** SAT/UNSAT*
TERMINATING CUE: When examinee has evaluated jet pump operability per 0PT-13.1 this
JPM is complete.
- Comments required for any step evaluated as UNSAT.
Time Complete: ___________
LOT-OJT-JP-002-A01 Page 6 of 14 REV. 1
Evaluate Jet Pump Performance Per 0PT-13.1
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute
3. Correct Component Verification (CCV), Validate Assumptions
4. Procedure and Work Instruction Use and Adherence (PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer checking) SAT/ UNSAT/ NE
E. Proper Equipment Use (observe starting limitations, throttle valve SAT/ UNSAT/ NE
closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety equipment, SAT/ UNSAT/ NE
etc.)
G. Electrical Safety And Arc Flash Protection Compliance (refer to SAT/ UNSAT/ NE
SAF-NGGC-2175, as applicable)
H. Security Compliance (controlled area entry and exit, key control, SAT/ UNSAT/ NE
etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, etc.) SAT/ UNSAT/ NE
J. Radiation Protection (ALARA, understanding and use of RWP, SAT/ UNSAT/ NE
frisking, etc.)
COMMENTS:
LOT-OJT-JP-002-A01 Page 7 of 14 REV. 1
Evaluate Jet Pump Performance Per 0PT-13.1
Validation Time: 20 Minutes (approximate)
Time Taken: ___ Minutes
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: N/A
Setting: In-Plant Simulator Admin X
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes No X
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
LOT-OJT-JP-002-A01 Page 8 of 14 REV. 1
Revision Summary:
REVISION REVISION SUMMARY
NUMBER
Formatting changes to match the JPM template, revision 3.
Validated using 0PT-13.1 Reactor Recirculation Jet Pump Operability
revision 44 and 2OP-02 Reactor Recirculation System Operating Procedure
revision 161.
1 Added attachment that includes photos to enable being performed in
classroom.
Added Duke Energy Progress image to cover page.
Updated work practices to current plant expectations.
Removed Step 1 which was obtaining the procedure (it is given to them).
Page 9 of 14
Page 10 of 14
Page 11 of 14
Page 12 of 14
Page 13 of 14
TASK CONDITIONS:
1. Unit Two is operating at 100% power with all recirculation loop flow and differential
pressure indicators operable. Recirc/Jet Pump readings are as indicated in the provided
attachment.
INITIATING CUE:
You are directed by the Unit CRS to perform 0PT-13.1, Reactor Recirculation Jet Pump
Operability, using the data provided. Determine if the surveillance is completed SAT or
UNSAT.
Inform the Unit CRS of the results.
SURVEILLANCE
SAT UNSAT
Page 14 of 14
DUKE ENERGY PROGRESS
BRUNSWICK NUCLEAR PLANT
LESSON TITLE: Determine Post Maintenance Testing Requirements
LESSON NUMBER:
REVISION NO: 0
PREPARER/DATE
TECHNICAL REVIEWER/DATE
VALIDATOR/DATE
TRAINING SUPERVISION APPROVAL/ DATE
RELATED TASKS:
K/A REFERENCE AND IMPORTANCE RATING:
Generic 2.2.21 Knowledge of pre- and post-maintenance operability requirements.
2.9/4.1
REFERENCES:
0PLP-20, Rev. 48, POST-MAINTENANCE TESTING PROGRAM
TOOLS AND EQUIPMENT:
None
ADMINISTRATIVE CATEGORY (from NUREG 1123, Rev 2. Supp 1):
3 - Equipment Control
SAFETY CONSIDERATIONS:
None.
SPECIAL INSTRUCTIONS:
None
SAFETY CONSIDERATIONS:
None
Page 2 of 9 REV. 0
EVALUATOR NOTES: (Do not read to trainee)
1. For Licensed personnel, provide memory stick with POM documents and access to
non-network computer.
2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021,
Appendix E, or similar to the trainee.
3. This JPM may be administered in the simulator, control room, or classroom setting.
Read the following to performer.
TASK CONDITIONS:
1-SW-V382, RBCCW Heat Exchanger Service Water Outlet Flow Control Valve has
been removed from the system rebuilt in the valve shop and reinstalled. The post
maintenance testing listed in the package indicated that the following post maintenance
tests were performed:
- During the tests, visually observe the valve to verify proper operation, positive
isolation, and that no packing or external leakage exists.
- Perform a position verification check.
You have been tasked with determining if the Post Maintenance Testing is adequate
prior to returning the system to service and closing out the work order.
INITIATING CUE:
You are directed to determine the Post Maintenance Testing Requirements for 1-SW-V382,
RBCCW Heat Exchanger Service Water Outlet Flow Control Valve maintenance.
Determine if the post maintenance testing performed completes the requirements to return
the valve to service.
List the post maintenance testing requirements.
Page 3 of 9 REV. 0
PERFORMANCE CHECKLIST
NOTE: Sequence is not essential to completing the task.
TIME START: __________
Step 1 - Obtain a copy of 0PLP-20.
Standard: Obtains a copy of 0PLP-20.
SAT/UNSAT
Step 2 - Identifies that 1-SW-V-382 is a manual valve and goes to attachment 19.
Standard: References attachment 19
SAT/UNSAT
Step 3 - Identifies from Attachment 19 that 1-SW-V382 was replaced and condition A is
applicable.
Standard: Identifies that valve was replaced
SAT/UNSAT
Step 4 - Identifies that Post maintenance Tests 1, 2, 3 and 5 are required.
Standard: Identifies that the following tests are required:
- During the tests, visually observe the valve to verify proper operation, positive
isolation, and that no packing or external leakage exists.
- Exercise (stroke) check.
- For locked in position throttle valves, perform a flow verification check.
- Seat leakage test (if limits are specified). (not Critical as limits were not
specified in cue)
- CRITICAL STEP ** SAT/UNSAT*
Page 4 of 9 REV. 0
Step 5 - Determine if the post maintenance testing that was performed meets the
requirements of Attachment 19 of 0PLP-20.
Standard: Identifies that testing was inadequate for the maintenance that was
performed
- CRITICAL STEP ** SAT/UNSAT*
Step 6 - List the required Post maintenance tests in 0PLP-20.
Standard: Lists the following testing requirements:
- Exercise (stroke) check.
- For locked in position throttle valves, perform a flow verification check.
- Seat leakage test (if limits are specified). (not Critical as limits were not
specified in cue)
TERMINATING CUE: When candidate identifies that testing requirements have been
determined
- Comments required for any step evaluated as UNSAT.
TIME COMPLETE: __________
Page 5 of 9 REV. 0
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and Adherence (PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer SAT/ UNSAT/ NE
checking)
E. Proper Equipment Use (observe starting limitations, throttle SAT/ UNSAT/ NE
valve closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety SAT/ UNSAT/ NE
equipment, etc.)
G. Electrical Safety And Arc Flash Protection Compliance (refer to SAT/ UNSAT/ NE
SAF-NGGC-2175, as applicable)
H. Security Compliance (controlled area entry and exit, key control, SAT/ UNSAT/ NE
etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, etc.) SAT/ UNSAT/ NE
J. Radiation Protection (ALARA, understanding and use of RWP, SAT/ UNSAT/ NE
frisking, etc.)
COMMENTS:
Page 6 of 9 REV. 0
Validation Time: 10 Minutes (approximate)
Time Taken: ___ Minutes
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: N/A
Setting: In-Plant Simulator Admin X
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes No X
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
Page 7 of 9 REV. 0
Revision Summary:
REVISION REVISION SUMMARY
NUMBER
0 New
Page 8 of 9 REV. 0
TASK CONDITIONS:
1. 2OP-05, Standby Liquid Control System Operating Procedure Section 8.6, Manual
Volume determination has been completed.
2. The distance from the surface of the liquid to the top of the tank is 71 inches.
3. The B-10 Atomic Enrichment is 51%.
4. Standby Liquid Control tank solution temperature (2-C41-TIC-R002) is 66°F.
5. The SLC Tank concentration is 8.7%.
INITIATING CUE:
You are directed by the CRS to verify SLC normal system operating parameters and inform
the Control Room Supervisor of the results.
Page 9 of 9
DUKE ENERGY
BRUNSWICK TRAINING SECTION
LESSON TITLE: Determine TEDE While Working in a High Airborne Area
LESSON NUMBER: LOT-ADM-XXXXX
REVISION NO: 0
PREPARER / DATE
TECHNICAL REVIEWER / DATE
VALIDATOR / DATE
LINE SUPERVISOR / DATE
TRAINING SUPERVISION APPROVAL / DATE
RELATED TASKS:
None
K/A REFERENCE AND IMPORTANCE RATING:
Generic 2.3.4 3.2/3.7
Knowledge of Radiation Exposure Limits under normal or emergency conditions
REFERENCES:
PD-RP-ALL-0001, Rev. 4, Radiation Worker Responsibilities
TOOLS AND EQUIPMENT:
Calculator
SAFETY FUNCTION (from NUREG 1123, Rev. 2, Supp. 1):
A.3 Radiation Control
SETUP INSTRUCTIONS
None
SAFETY CONSIDERATIONS:
1. None
EVALUATOR NOTES: (Do not read to performer)
1. The applicable procedure section WILL be provided to the trainee.
2. Prior to the first JPM of the JPM set, provide the JPM briefing contained in NUREG-1021,
Appendix E, or similar briefing (for non-regulated exams) to the trainee(s).
3. Critical Step Basis
a) Prevents Task Completion
b) May Result in Equipment Damage
c) Affects Public Health and Safety
d) Could Result in Personal Injury
Read the following to the JPM performer.
TASK CONDITIONS:
I will explain the initial conditions, which steps to simulate, discuss or perform, and
provide initiating cues. When you complete the task successfully, the objective for this
Job Performance Measure will be satisfied.
Initial Conditions:
- The plant is shut down for refueling
- While doing contract work the operator received a combined 1,325 mRem TEDE this
calendar year.
- In the same calendar year, the operator has received another 495 mRem TEDE at the
Harris plant.
- The estimated dose rate in the work area is 465 mRem/hr.
- An airborne contamination concern exists.
- It is estimated that it will take 25 minutes to complete the inspection if the operator uses
a respirator.
- If the operator does NOT wear a respirator, the inspection will take 12 minutes, but
Radiation Protection projects that the internal exposure will be 10 DAC-hrs.
Initiating Cue:
1.a. Calculate the resultant total effective dose equivalent for with a respirator for this job and
the accumulated dose for the year.
1.b. Calculate the resultant total effective dose equivalent for without a respirator for this job
and the accumulated dose for the year.
2. Using the lowest dose determined in number 1, determine if the individual can perform
the task without exceeding Duke Energys Annual Administrative Dose Limit.
Show your calculations on the next page
Task Standard:
Determination made that NOT wearing a respirator will result in a lower
TEDE and that the individual can perform the task without exceeding Duke
Energys Annual Administrative Dose Limit.
Required Materials:
Laptop with General Reference PD-RP-ALL-0001, Radiation Worker
Responsibilities, Rev 3 or hard copy
General References:
PD-RP-ALL-0001, Radiation Worker Responsibilities, Rev. 4
Handouts:
JPM Cue Sheet
Time Critical Task:
No
Validation Time:
XXXXX
Critical Step Justification
Determining external exposure while wearing a respirator is required to
Step 2
complete the calculations and answer.
Determining internal exposure while NOT wearing a respirator is
Step 4
required to complete the calculations and answer.
Determining external exposure while NOT wearing a respirator is
Step 5
required to complete the calculations and answer.
Determining individuals total exposure for the year if the work is allowed
Step 8
without a respirator is required to complete the calculations and answer.
Determining individuals total exposure for the year if the work is allowed
Step 9
with a respirator is required to complete the calculations and answer.
Determining if the individual can perform the work without exceeding
Step 10 Duke Energys Annual Administrative Dose Limit of 2000 mRem is
required to complete the calculations and answer.
PERFORMANCE CHECKLIST
Start Time: __________.
NOTE: Steps in this JPM may be performed in any order.
Performance Step: 1 Determines internal exposure while wearing a respirator
Standard: Determines internal exposure to be ZERO while wearing a
respirator
Comment:
Performance Step: 2 Determines external exposure while wearing a respirator
Standard: Determines external exposure to be 193.8 mRem TEDE while
wearing a respirator
(465 mRem / hr x 25 min = 193.8 mRem)
(NOTE: Could round to 194)
Comment:
Performance Step: 3 Determines TOTAL exposure while wearing a respirator
Standard: Determines total exposure to be 193.8 mRem while wearing a
respirator
(0 mRem internal + 193.8 mRem external = 193.8 mRem total)
(NOTE: Could round to 194)
Comment:
Indicates a Critical Step
Performance Step: 4 Determines internal exposure while NOT wearing a respirator
Standard: Determines internal exposure to be 25 mRem while not wearing
a respirator
(2.5 mRem / hr x 10 DAC-hr = 25 mRem)
(NO tolerance)
Comment:
Performance Step: 5 Determines external exposure while NOT wearing a respirator
Standard: Determines external exposure to be 93.0 mRem TEDE while not
wearing a respirator
(465 mRem / hr x 12 min = 93 mRem)
(NO tolerance)
Comment:
Performance Step: 6 Determines TOTAL exposure while NOT wearing a respirator
Standard: Determines total exposure to be 118 mRem while not wearing a
respirator
(25 mRem internal + 93 mRem external = 118 mRem total)
(NO tolerance)
Comment:
Performance Step: 7 Determines individuals total exposure for the year
Standard: Determines individuals total exposure for the year to be 1820
mRem
(1325 mRem + 495 mRem = 1820 mRem)
Comment:
Indicates a Critical Step
Performance Step: 8 Determines individuals total exposure for the year if the work is
allowed without a respirator
Standard: Determines individuals total exposure for the year if the work is
performed without a respirator to be 1938 mRem
(1820 mRem + 118 mRem = 1938 mRem)
(NO tolerance)
Comment:
Performance Step: 9 Determines individuals total exposure for the year if the work is
allowed with a respirator
Standard: Determines individuals total exposure for the year if the work is
performed with a respirator to be 2013.8 mRem
(1820 mRem + 193.8 mRem = 2013.8 mRem)
Note: If calculated wearing a respirator the total exposure for the
year will be 2013.8 mRem and work cannot be performed without
an extension. The directions were to use the lowest dose and
this represents UNSAT performance.
(NOTE: Could have rounded 193.8 to 194 and answer would
be 2014 mRem)
Comment:
Performance Step: 10 Determines if the individual can perform the work without
exceeding Duke Energys Annual Administrative Dose Limit of
2000 mRem without an administrative dose limit extension
Standard: Determines the individual CAN perform the work without wearing
a respirator will NOT exceed Duke Energys Annual
Administrative Dose Limit of 2000 mRem
(1820 mRem + 118mRem = 1938 mRem)
Comment:
When all calculations have been completed and the
Evaluator Cue: determination that work can proceed, this JPM is complete.
Stop Time: _________
Indicates a Critical Step
WORK PRACTICES:
Comments are required for any step evaluated as UNSAT
A. CORE 4 SAT/ UNSAT/ NE
1. Task Preview / Pre-Job Briefs
2. Take-A-Minute
3. Correct Component Verification (CCV), Validate
Assumptions
4. Procedure and Work Instruction Use and
Adherence (PU&A)
B. Communications (proper content, repeat backs, 3 step SAT/ UNSAT/ NE
communications, etc.)
C. STAR (Use of Stop, Think, Act, Review) SAT/ UNSAT/ NE
D. Peer Checking (if performer requests or discusses peer SAT/ UNSAT/ NE
checking)
E. Proper Equipment Use (observe starting limitations, SAT/ UNSAT/ NE
throttle valve closures, etc.)
F. Safety Compliance (use of PPE, knowledge of safety SAT/ UNSAT/ NE
equipment, etc.)
G. Electrical Safety And Arc Flash Protection Compliance SAT/ UNSAT/ NE
(refer to AD-HS-ALL-0110, as applicable)
H. Security Compliance (controlled area entry and exit, key SAT/ UNSAT/ NE
control, etc.)
I. Proper Tool Use (proper tool for the job, calibration dates, SAT/ UNSAT/ NE
etc.)
J. Radiation Protection (ALARA, understanding and use of SAT/ UNSAT/ NE
RWP, frisking, etc.)
Comments:
Validation Time: XX Minutes (approximate).
Time Taken: ____ Minutes
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: 1
Setting: In-Plant Simulator Admin X
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes No X
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
KEY
1.a Calculation for resultant total effective dose equivalent with a respirator.
Determines internal exposure to be ZERO while wearing a respirator
Determines external exposure to be 193.8 mRem TEDE while wearing a respirator
(465 mRem / hr x 25 min = 193.8 mRem)
(NOTE: Could round to 194)
Determines TOTAL exposure while wearing a respirator
(0 mRem internal + 193.8 mRem external = 193.8 mRem total)
(NOTE: Could round to 194)
Determines individuals total exposure for the year to be 1820 mRem
(1325 mRem + 495 mRem = 1820 mRem)
Determines individuals total exposure for the year if the work is allowed to be 2013.8 mRem
(1820 mRem + 193.8 = 2013.8 mRem)
(NOTE: Could round to 2014 mRem)
Note: If calculated wearing a respirator the total exposure for the year will be 2013.8 mRem and
work CANNOT be performed without an extension. The directions were to use the lowest dose
and this represents UNSAT performance.
1.b Calculation for resultant total effective dose equivalent without a respirator.
Determines internal exposure to be 25 mRem while not wearing a respirator
(2.5 mRem / hr x 10 DAC-hr = 25 mRem)
(NO tolerance)
Determines external exposure to be 118 mRem TEDE while not wearing a respirator
(495 mRem / hr x 12 min = 93 mRem)
Determines total exposure to be 118 mRem while not wearing a respirator
(25 mRem internal + 93 mRem external = 118 mRem total)
Determines individuals total exposure for the year if the work is allowed to be 1938 mRem
(1820 mRem + 118 = 1938 mRem)
The individual CAN perform the task without exceeding Duke Energys Annual Admin Dose
Limit if the task is performed WITHOUT a respirator.
KEY
2. Using the lowest dose determined from the above calculations (1a or 1b):
CAN the individual perform the task without exceeding Duke Energys Annual Administrative
Dose Limit?
YES (without a respirator total dose would be 1,938 mRem which is < 2,000 mRem)
TASK CONDITIONS:
I will explain the initial conditions, which steps to simulate, discuss or perform, and provide
initiating cues. When you complete the task successfully, the objective for this Job
Performance Measure will be satisfied.
Initial Conditions:
- The plant is shut down for refueling
- While doing contract work the operator received a combined 1,325 mRem TEDE this calendar
year.
- In the same calendar year, the operator has received another 495 mRem TEDE at the Harris
plant.
- The estimated dose rate in the work area is 465 mRem/hr.
- An airborne contamination concern exists.
- It is estimated that it will take 25 minutes to complete the inspection if the operator uses a
respirator.
- If the operator does NOT wear a respirator, the inspection will take 12 minutes, but Radiation
Protection projects that the internal exposure will be 10 DAC-hrs.
Initiating Cue:
1.a. Calculate the resultant total effective dose equivalent for with a respirator for this job and the
accumulated dose for the year.
1.b. Calculate the resultant total effective dose equivalent for without a respirator for this job and the
accumulated dose for the year.
2. Using the lowest dose determined in number 1, determine if the individual can perform the task
without exceeding Duke Energys Annual Administrative Dose Limit.
Show your calculations on the next page
Name:
Date:
1.a Calculation for resultant total effective dose equivalent with a respirator.
1.b Calculation for resultant total effective dose equivalent without a respirator.
2. Using the lowest dose determined from the above calculations (1a or 1b):
CAN the individual perform the task without exceeding Duke Energys Annual Administrative
Dose Limit?
DUKE ENERGY PROGRESS
BRUNSWICK NUCLEAR PLANT
LESSON TITLE: Protective Action Recommendation (EP)
LESSON NUMBER:
REVISION NO: 0
PREPARER/DATE
TECHNICAL REVIEWER/DATE
VALIDATOR/DATE
TRAINING SUPERVISION APPROVAL/ DATE
RELATED TASKS:
K/A REFERENCE AND IMPORTANCE RATING:
Generic 2.4.44 Knowledge of emergency plan protective action recommendations.
REFERENCES:
0PEP-02.6.21, 0PEP-02.6.28, 0PEP-02.1
TOOLS AND EQUIPMENT:
None
ADMINISTRATIVE CATEGORY (from NUREG 1123, Rev 2. Supp 1):
4 - Emergency Plan
SAFETY CONSIDERATIONS:
None.
SPECIAL INSTRUCTIONS:
None
SAFETY CONSIDERATIONS:
None
REV. 0
EVALUATOR NOTES: (Do not read to trainee)
1. For Licensed personnel, provide memory stick with POM documents and access to non-network
computer.
2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021, Appendix E, or
similar to the trainee.
3. This JPM may be administered in the simulator, control room, or classroom setting.
Read the following to performer.
Ensure all Applicants have access to Plant references prior to proceeding.
TASK CONDITIONS:
It is now 1300 on Thursday, July 27th 2017.
You are the Site Emergency Coordinator during a plant emergency.
The Emergency Operations Facility is not yet fully staffed.
You have just declared a General Emergency based on the following conditions:
- Unisolable steam line break with a failure of the MSIVs to isolate
- Emergency Depressurization has been performed
- As of this time, an HP Dose Assessment has indicated > 5000 mrem thyroid
CEDE at the Site Area Boundary with the release originating from the Unit 2
Turbine Building
- As of this time, the Control Room Emergency Coordinator has the appropriate
Offsite Agencies on the line. They are aware of the emergency and are awaiting
formal reporting.
- As of this time, the following meteorological conditions exist:
o Wind Direction: 150o
o Wind Speed: 10 mph
o No precipitation forecast
o Stability Class: C
INITIATING CUE:
You are to complete the Emergency Notification Form required following the given
conditions above.
THIS IS A TIME CRITICAL JPM
REV. 0
PERFORMANCE CHECKLIST
NOTE: Sequence is not essential to completing the task.
TIME START: __________
Step 1 - 0PEP-02.6.21, Attachment 2, ENF Line #1
Marks DRILL
- CRITICAL STEP ** SAT/UNSAT
Step 2 - 0PEP-02.6.21, Attachment 2, ENF Line #2
Marks BRUNSWICK
- CRITICAL STEP ** SAT/UNSAT
Step 3 - 0PEP-02.6.21, Attachment 2, ENF Line #3
Marks GENERAL EMERGENCY
- CRITICAL STEP ** SAT/UNSAT
Step 4 - 0PEP-02.6.21, Attachment 2, ENF Line #4 (Ref 0PEP-02.1)
Marks Declaration Date 07/27/2017 1300
Marks Description Dose Assessment using actual meterology indicates doses >
1000 mrem TEDE or 5000 mrem thyroid CDE at or beyond the SITE BOUNDARY
- CRITICAL STEP ** SAT/UNSAT
Step 5 - 0PEP-02.6.21, Attachment 2, ENF Line #5
Marks Release IS OCCURRING
- CRITICAL STEP ** SAT/UNSAT*
REV. 0
Step 6 - 0PEP-02.6.21, Attachment 2, ENF Line #6 (Ref 0PEP-02.6.28, Attachment 2)
Marks Evacuate F, G, H, J
Marks Shelter A, B, E
Marks Consider the use of KI
- CRITICAL STEP ** SAT/UNSAT*
Step 7 - 0PEP-02.6.21, Attachment 2, ENF Lines #7 through 11
Use of these lines is not required for Initial Notifications. Post JPM
questioning is required if any of these items are completed.
SAT/UNSAT*
Step 8 - 0PEP-02.6.21, Attachment 2, ENF Line #12
This line may contain any applicant provided information.
N/A
Step 9 - 0PEP-02.6.21, Attachment 2, ENF Line #13
Marks Applicant signature in Approved By line
Marks SEC (Site Area Coordinator) as Title
Marks Date/Time (referenced off given Date/Time)
TIME designated must be within 15 minutes of JPM start for successful completion
- CRITICAL STEP ** SAT/UNSAT*
Step 10 - 0PEP-02.6.21, Attachment 2, ENF Lines #14 and 15
These lines may be left blank.
SAT/UNSAT*
- Comments required for any step evaluated as UNSAT.
TIME COMPLETE: __________
REV. 0
Validation Time: 10 Minutes (approximate)
Time Taken: ___ Minutes
APPLICABLE METHOD OF TESTING
Performance: Simulate Actual X Unit: N/A
Setting: In-Plant Simulator Admin X
Time Critical: Yes No X Time Limit N/A
Alternate Path: Yes No X
EVALUATION
Performer:
JPM: Pass Fail
Remedial Training Required: Yes No
Comments:
Comments reviewed with Performer
Evaluator Signature: Date:
REV. 0
Revision Summary:
REVISION REVISION SUMMARY
NUMBER
0 New
REV. 0
TASK CONDITIONS:
It is now 1300 on Thursday, July 27th 2017.
You are the Site Emergency Coordinator during a plant emergency.
The Emergency Operations Facility is not yet fully staffed.
You have just declared a General Emergency based on the following conditions:
- Unisolable steam line break with a failure of the MSIVs to isolate
- Emergency Depressurization has been performed
- As of this time, an HP Dose Assessment has indicated > 5000 mrem
thyroid CEDE at the Site Area Boundary with the release originating from
the Unit 2 Turbine Building
- As of this time, the Control Room Emergency Coordinator has the
appropriate Offsite Agencies on the line. They are aware of the
emergency and are awaiting formal reporting.
- As of this time, the following meteorological conditions exist:
o Wind Direction: 150o
o Wind Speed: 10 mph
o No precipitation forecast
o Stability Class: C
INITIATING CUE:
You are to complete the Emergency Notification Form required following the given
conditions above.
THIS IS A TIME CRITICAL JPM
REV. 0