ML081900165
| ML081900165 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 06/03/2008 |
| From: | NRC/RGN-II |
| To: | |
| References | |
| 50-390/08-301 | |
| Download: ML081900165 (57) | |
See also: IR 05000390/2008301
Text
Question Number:
51
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DATA SHEET
KIA: 076 K4.03
Knowledge of SWS design feature(s) andlor interlock(s) which provide for the following: Automatic opening
features associated with SWS isolation valves to CCW heat exchang ers.
Tier:
2
Group:
1
RO Imp:
2.9
RO Exam:
SRO Imp: 3.4
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
WBN Bank
Applicable 10CFR55 Section :
41.7
Learning Objective: 3-0 T-SYS067A, Objective 13: Given a loss of power, determine the correct respons e
of the ERCW System including: a. "C" CCS Heat Exchanger outlet valves .
References:
3-0 T-SYS067A, Rev. 10. 1-47W611-67-5, E-OAppend ix A. Rev. 27.
Question:
Reactor trip and safety injection signals have been manually initiated. Which ONE of the following describes
the required positions for the listed ERCW valves in accordance with E-O, Appendix A, Equipment
Verification?
O-FCV-67-144,
O*FCV-67-152,
"CCS Heat Exchanger 'C'
"CCS Heat Exchanger 'C'
Disch to Hdr A"
Alt Disch to Hdr B"
A.
CLOSED
OPEN to Position A
B.
THROTTLED
CLOSED
C
THROTTLED
OPEN to Position A
U.
OPEN
CLOSED
DISTRACTOR ANALYSIS
a.
CORRECT. Under normal conditions O-FCV-67-152 is closed with power on the valve . Upon the
receipt of an SI signal 0-FCV-67-152 strokes to the 35% position automa tically. During performance of
E-O, Appendi x A the operator places the O-FCV-67-152 handswitch for the valve in the Position A. 0-
FCV-67-144 is normally open and is closed by manual operator action during performan ce of E-O,
Appendix A.
b.
Incorrect
Plausible since the valve positions are reversed.
c.
Incorrect. Plausible since the operator may confuse listed valves with others with similar numbers.
d.
Incorrect. Plausible since the operator may confuse listed valves with others with similar numbers .
55 of 81
Question Number:
52
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA: 076 A1.02
Ability to predict and/or monito r changes in parameters (to prevent exceeding design limits) associated with
operating the SWS controls including: Reactor and turbine building closed cooling water temperatures.
Tier:
2
Group:
1
RO Imp:
2.6
RO Exam:
Yes
Yes
Cognitive Level:
Source:
HIGH
NEW
Applicable 10CFR55 Section:
41.5/45.5
Learning Objective: 3-0T-SYS067A, Objective 8: State the ERCW System normal discharge path and
given a failure of the path. discuss the alternate discha rge paths.
References:
AOI-13, LOSS OF ESSENTIAL RAW COOLING WATER , Rev. 35.
Question:
Given the following plant conditio ns:
The plant is operating at 100% power with 1B CCP in service .
The Control Room Operator shutdowns down the C-A ERCW pump in preparation for a test on the
2A 6.9 KV Shutdown Board.
Which ONE of the following describes the impact of shutting down the pump on the listed parameters?
(Assume no other operator action.)
16 CCP Oil
Seal Water Return Heat
Temperature
Exchanger Temperature
A.
Rises
Rises
B.
Rises
Remains constant
C.
Remains constant
Rises
D.
Remains constant
Remains constan t
DISTRACTOR ANALYSIS
a.
Incorrect. The C-A ERCW pump supplies flow to the A header. When the pump is stopped, flow to
both the 1A and 2A ERCW headers is decreased. This causes a decrease in flow through the seal
water heat exchangers. A reduction in flow to the 1A CCS heat exchanger will cause the A ESF
header to heat up as well as the Reactor Building and Miscella neous Equipment Headers to heat up.
The 1A CCP which would normally be in service would show an increase in oil temperature. The Seal
Water Heat Exchanger, supplied off the Miscellaneous Equipment Header would also heat up.
b.
Incorrect. Plausible, because oil temperat ure does rise, howeve r, Seal Water Return Heat Exchanger
temperatu re also rises.
c.
CORRECT. Seal Wate r Return Heat Exchanger temperatu re does rise, and CCP oil tempera ture
remains constan t.
d.
Incorrect. Plausible. since candidate incorrectly recalls cooling water supply to this component.
56 of 81
Question Number:
53
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DATA SHEET
KIA : 078 K1.01
Instrument Air
Knowledge of the physical connections and/or cause-effect relationships between the lAS and the following
systems: Sensor air.
Tier:
2
Group:
1
RO Imp:
2.8
RO Exam:
SRO Imp: 2.7
SRO Exam :
Yes
Yes
Cognitive Lev el :
Source:
HIGH
NEW
Applicable 10CFR55 Section:
41.2 to 41.9/45.7 to 45.8
Learn ing Obj ect ive : 3-0 T-SYS032A, Objecti ve 16: List the events and their correspond ing set points that
take place on decreasing control air pressure .
References :
1-47W611-32-2 rev 4, SOI-32.02 rev 19 Note on page 12.
Question:
Which ONE of the following identifies both of the following?
(1)
The lowest of the listed containment pressures that results in 1-FCV -32-80 , Aux Air to Rx Bldg
Train B, being automa tically isolated, and
(2)
The lowest of the listed air pressures sensed downstream of the valve that allows the valve to
REMAIN OPEN after the valve was opened and the control switch on 1-M-15 placed to A-Auto after
the isolation signal was reset.
ill
ill
A.
2.0 psid
68 psig
B
2.0 psid
78 psig
C.
3.0 psid
68 psig
D.
3.0 psid
78 psig
DISTRACTOR ANALYSIS
a.
Incorrect. Containment pressure is not high enough to cause the isolation and the sensor downstream
of the valve will not allow the valve to remain open with the pressure at 68psig , but plausible because
with the containment pressure above 1.5, a Phase A isolation would have occurred and many paths
would have isolated .
b.
Incorrect. Containment pressure is not high enough to cause the isolatio n, but since it is above 1.5,
then a Phase A isolation would have occurred and many paths would have isolated and the
downstream pressure is high enough to allow, but plausible because with the containment pressure
above 1.5 then a Phase A isolation would have occurred and many paths would have isolated and 78
psig is high enough for the downstream sensor to allow air to open the valve .
c.
Incorrect. Containment pressure is high enough to cause the isolation, but the pressur e sensed
downstream of the valve is not high enough to allow operating air pressu re to maintain the valve open
after the switch is placed in A-Auto, but plausible because with the containment pressure above 3.0 psig
then a Phase B isolation would have occurred causing isolation of the valve.
d.
CORR ECT. With containment pressure greater than 2.8 psig, a phase B isolation will automatica lly
occur, and the sensor downstre am of the isolation valve must detect greater than 75 psig to allow the
valve to remain open after the control switch was placed to the A-Auto position,
57 of 81
Question Number:
54
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA : 103 A2.05
Containment
Ability to (a) predict the impacts of the following malfunctions or operations on the containment system-and
(b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those
malfunctions or operations: Emergency containment entry.
Tier:
2
Group:
1
RO Imp:
2.9
RO Exam:
SRO Imp: 3.9
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
HIGH
NEW
Applicable 10CFR55 Section:
41.5/43.5/45.3/45.13
Learning Objective: 3-0T -SYS088A, Objective 10: Describe the containment and penetration testing
required and the acceptance criteria.
References:
3-0T -SYS088A, Technical Specification 3.6.1., 1-SI-88-24, rev 7, TI-1207A, Rev. O.
Question :
Given the following plant conditions:
Plant is in Mode 4.
Lower containment air lock is broken and inner door is jammed and will not open.
If conditions were to require an emergency entry into lower containment by opening the sub-hatch, which
ONE of the following is REQUIRED to be contacted prior to openin g the sub-hatch, and what action is
required as a result of the sub-hatch being opened?
A.
Shift Manager;
Perform 1-SI-88-24, Containment Divider Barrier Personnel Access Hatches & Equipment Hatches,
within one (1) hour.
B.
Shift Manager;
Perform 1-SI-88-24, Containment Divider Barrier Personnel Access Hatches & Equipment Hatches,
prior to Mode 3 entry.
C.
Work Week Manager;
Perform 1-51-88-24, Containment Divider Barrier Personnel Access Hatches & Equipment Hatches ,
within one (1) hour.
D.
Work Week Manager;
Perform 1-SI-88-24, Containment Divider Barrier Personnel Access Hatches & Equipment Hatches,
prior to Mode 3 entry.
D1STRACTOR ANALYSIS
a.
CORRECT. The sign on the sub-hatch requires the Shift Manager to be notified and the Sl must be
completed within one hour, with the plant in Mode 4.
b.
Incorrect. While the sign on the sub-hatch requires the Shift Manager to be notified, the SI must also be
completed with the plant is in Mode 4. Plausible because the notification is correct and the candidate
could conclude the SI is not required until Mode 3.
c.
Incorrect. Plausible since TI-1207A, Containment Access for Modes 1-4, does mention Work Week
Manager as a consultant source only, not as a required notification.
d.
Incorrect. Plausible since TI-1207 A, Containmen t Access for Modes 1-4, does mention Work Week
Manager as a consultant source only, not as a required notification.
58 of 81
Qu estion Number:
55
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
K/A: 103 M .09
Containment
Ability to manually operate and/or monitor in the control room : Containment vacuum system.
Tier :
2
Group:
1
RO Imp:
3.1
RO Exam:
SRO Imp: 3.7
SRO Exam:
Yes
Yes
Cognitive Level :
Source:
HIGH
NEW
Applicable 10CFR55 Section:
41.7/45 .5 to 45.8
Learning Objective: 3-0T-SYS065A. Objective 5: Describe how the EGTS and Annulus Vac uum Systems
maintain annulus pressu re.
References :
SOI-65.01, Annulus Vacuum System , Rev. 18.
Question:
Which ONE of the following identifies the NORMAL pressure band controlled by the Containment Annulus
Vacuum System and the requi red method of controlling pressure if 1-M-278 Window 232- 8, ANNULUS 6P
LO/DAMPER SWA POV ER is LIT?
A.
8 .
C.
D.
Annulus Pressure Band
-6.0 to -6.2" WC
-6.0 to -6.2" WC
-4.3 to -4.5" WC
-4.3 to -4.5" WC
Method of Controlling Pressure
Dispatch a NAUO to RESET the dampers locally.
Swap the dampers using handswiches on 1-M-27B.
Dispatch a NAUO to RESET the dampers locally.
Swap the dampers using handswiches on 1-M-27B.
DISTRACTOR ANALYSIS
a.
CORRECT. The correct pressure band of -6.0 to -6.2 "wc is provided and the correct response to
Window 232-8 is provided .
b.
Incorrect. Plausible , since the correct pressure band of -6.0 to -6.2 "wc is provided but incorrect
actio ns to restore alignment are stated.
c.
Incorrect.
Plausible since the 4.3 to -4.5 "wc range is associated with the swapover of the dampers
and the correct response to Window 232- 8 is provided.
d.
Incorrect. Incorrect damper sequence, and the local action of the NAUO is described corre ctly.
59 of 81
Question Number:
56
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA: 014 A2.02
Rod Position Indication
Ability to (a) predict the impacts of the following malfunctions or operations on the RPIS; and (b) based on
those on those predictions, use procedures to correct , control, or mitigate the consequences of those
malfunctions or operations: Loss of power to the RPIS.
Tier :
2
Group:
2
RO Imp:
3.1
RO Exam :
SRO Imp: 3.6
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
HIGH
NEW
Applicable 10CFR55 Section:
41.5/43.5/45.3/45.13
Learning Objective: 3-0T-SYS085A, Objective 25: Explain the bases , input, alarms, and operator actions
relative to the rod insertion limits.
References:
3-0 T-SYS085A, Attachment 7 Engineering Evaluation of Westinghouse Information
Regardll1g Computer Enhanced Rod Position Indication (CERPI) Displays.
Question:
Given the following plant conditions:
The unit is at 100% power.
1-M-1B, Window 17-0, 120 AC VITAL PWR BD 1-1 UV/CKT TRIP , is LIT.
Which ONE of the following describes the impact on the Computer Enhanced Rod Position Indication
(CERPI) System and what is required to ensure that Tech Spec Rod Group Alignment Limits are met?
Rod position indicati on
A.
Available
B.
NOT Availab le
C.
Available
D.
NOT Availab le
Ac tions Required for Rod Group alignment
Use the "ALL RODS" function on the operating display.
Flux map is required to confirm rod position.
Flux map is required to confirm rod position.
Use the "ALL RODS" function on the operating display.
DISTRACTOR ANALYSIS
a.
CORRECT. The power loss described in the stem results in only the left side CERPI display going
dark. The right side CERPI display remains powered, and the operator can select "All Rods" on the
displa y to determine positions .
b.
Incorrect. The right hand display remains powered, with capabil ity to monitor all rods. No flux map is
required.
c.
Incorrect. The power loss described in the stem results in only the left side CERPI display going dark.
The right side CERPI display remains powered , and the operator can select "All Rods" on the display
to determine positions. No flux map is required.
d.
Incorrect. The right hand display remains powered , with capability to monitor all rods. No flux map is
required .
60 of 81
Question Number:
57
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA: 028 A1.02
Hydrogen Recom biner and Purge Con trol
Ability to predict and /or monitor cha nges in parame ter (to prevent exceeding design limits) associated with
ope rating the HRPS controls includ ing: Containment pressure .
Tier:
2
Group:
2
RO Imp:3.4
SRO Imp:3.7
RO Exam:
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
NEW
Applicable 10CFR55 Section:
41.5 /45.5
Learning Objective: 3-0 T-SY S083A, Objective 8: Describe the major com ponents and operating princip le
of the Hydrogen Recombiners.
References:
Ref: SOI-83.01 Rev 15, TI-83.01 Rev 1, and TS 3.6.7 Basi s.
Question:
Whi ch ONE identifies BOTH of the follo win g for Hydro gen Recombiner operations?
(1)
The MAXIMUM Hydr ogen Recombiner temperature allowed when operating?
AND
(2)
The HIGHEST of the listed containment hydrogen conce ntrations allowed when placi ng the
recombiner in serv ice ?
A.
B.
C.
D.
(1) Maximum Temperature
(2) H2 Concentration
6%
4%
6%
4%
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible due to 1150 °F is the temperature above which the Hydr ogen Rec ombiners will
remove hydrogen, and greater than 6% is the limit (without taking into account instrument accuracy)
where Hydro gen Recombiners shall NOT be placed in service lAW SOI-83.01 .
b.
Incorrect. Plaus ible due to 1150 °F is the temperature above which the Hydro gen Recombi ners will
remove hydr ogen . The hydrogen concentration is correct.
c.
Incorr ect: The ma ximum temperature value is correct. Plausible due to greater than 6% is the limit
(without taking into account instrument accura cy) where Hydrogen Recombiners shall NOT be placed in
service lAW SOI- 83.01 .
d.
CO RRECT. The maximum temperature value is correct, and 4% is the correct hydrogen concentration ,
per the given reference.
61 of 81
Que stion Number:
58
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DATA SHEET
KIA: 033 G2,2,36
Spent Fuel Pool Cooling
Ability to ana lyze the effec t of maintenance activities , such as degraded power sources, on the status of
limiting cond itions for operations ,
Tier:
2
Group:
2
RO Imp:
3.1
RO Exam :
SRO Imp: 3.2
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
HIGH
NEW
Applicable 1OCFR55 Section:
41.10/43.2/45 ,13
l earning Obje ctive: 3-0 T-SYS078A, Object ive 8: Descr ibe the Spen t Fuel Pit pumps including: c. Power
supplies; Object ive 4: Given plant conditions and parameters, correctly determine the Conditions for
Operation or Technical Requi rements for various components listed in Section 7 of Tech . Specs,
References :
SOI-78.01, Spen t Fuel Pool Cool ing and Cleaning System, Rev. 52.
Que stion:
Following a refue ling outage, the followi ng conditions exist:
Unit 1 is in Mode 3 preparing for Mode 2 entry.
Fuel shuffles are being conducted in the Spen t Fuel Pit.
Spent Fuel Pool Cooling pump A is the only Spent Fuel Pool Cooling pump in service,
The 2A-A Shutdown Board normal feeder breaker is inadvertently open ed during testing.
The DG starts and reenergizes the shutdown board,
Which ONE of the following describes the initial effect on the Spent Fuel Pool Cooling system, and the
required action, if any, per Spen t Fuel Pool Tech Specs?
A.
The Spent Fuel Coo ling Pump strips from the board , and then sequences back on to the shutd own
board .
Spent Fuel Pool Tech Specs requires that movement of irradiated fuel assemblies in the fuel storage
pool be immediately suspended.
B.
The Spent Fuel Cooling Pump strips from the board , and then sequences back on to the shutdown
board,
Spent Fuel Pool Tech Specs does NOT require that movement of irradiated fuel assemblies in the fuel
storage pool be suspended,
C,
The Spen t Fuel Cooling Pump strips from the board, and rema ins off.
Spent Fuel Pool Tech Specs requires that movement of irradiated fuel assemblies in the fuel storage
pool be immediately suspended .
D.
The Spent Fuel Cooling Pump strips from the board, and remains off.
Spent Fuel Pool Tech Specs does NOT require that movement of irradiated fuel assemblies in the fuel
storage pool be suspended.
DISTRACTOR ANALYSIS
a,
Incorrect. While it ;s correct that the pump strips from the board due to the blackout relays, the pump is
not sequenced back on after the diesel generator recovers the board. It is also correct that no action is
required to immediately suspend the movement of radiation fuel. Plausible because the pump could be
confused with other pumps that do sequence back on.
b.
Incorrect. While it is correct that the pump strips from the board due to the blackout relays, the pump is
not sequenced back on after the diesel generator recovers the board and there is no requ irement to
suspend the movement of radiation fuel. Plausible because the pump could be confused with other
pump s that do sequence back on and because the immediate suspension of irradiated fuel movemen t
is required for other conditions associated with the spent fuel piUcooling system,
62 of 81
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
c.
Incorrect. While it is correct that the pump strips from the board due to the blackout relays and not
sequenced back on after the diesel generator recovers the board, there is no action require d to
immediately suspend the movement of radiation fuel. Plausible because the immediate suspensio n of
irradiated fuel movement is required for other conditions associated with the spent fuel pit/coo ling
system.
d.
CORRECT. The SFP pump A normal supply is from 480V Shutdown Board 2A2-A. The pump strips
from the board due to the blackout relays and does not sequence back on when the diesel generator
recovers the board. It is also correct that no action is required to immediately suspend the moveme nt of
radiation fuel.
63 of 81
Question Number:
59
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA: 045 K5.17
Main Turbine Generator
Knowledge of the operational implications of the following concepts as the(y) apply to the MT/B(G) System:
Relationship between moderator temperature coefficient and boron concentration in RCS as T/G load
Increases.
Tier:
2
Group:
2
RO Imp:
2.5
RO Exam:
SRO Imp: 2.7
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
HIGH
SQN BANK
Applicable 10CFR55 Section:
41.5/45.7
Learning Objective: 3-0T-G00400, Objective 6: Explain the average coolant tempe rature (Tavg) program
utilized during power increase or decrease. Objective 7: Explain why reactor would "follow" the turbine up
in power during a load increase.
References:
Nuclear Parameter and Operation s Package (NuPOP) Cycle 9.
Question:
Which ONE of the following identifies (a) how main steam header pressu re responds as turbine load is
raised from 25% to 65%, and (b) which method of maintaining Tavg matched with Tref results in the value
for MTC being the MOST negative as turbine load is raised?
A.
(a)
Main steam header pressure lowers.
(b)
Rods are withdrawn to maintain Tavg on program, with Boron concentration held constant.
B.
(a)
Main steam header pressure rises.
(b)
Rods are withdrawn to maintain Tavg on program, with Boron concentration held constant.
C
(a)
Main stearn header pressure lowers.
(b)
Rod position is held constant, while Boron concentration is lowered to maintain Tavg on
program.
D.
(a)
Main steam header pressure rises.
(b)
Rod position is held constant, while Boron concentration is lowered to mainta in Tavg on
program .
DISTRACTOR ANALYSIS
a.
Incorrect. Main Steam Header pressure lowers as turbine load is raised. Plausible since
candidate could conclude that withdrawing rods makes MTC more negati ve. It actually makes
it less negative . However a competing effect of Tavg rising has a negative effect on MTC.
b.
Incorrect. Main Steam Header pressure lowers as turbine load is raised . Plausible since
candidate could conclude that withdrawing rods makes MTC more negative. It actually makes
it less negative. However, a competing effect of Tavg rising has a negative effect on MTC.
c.
CORRECT. Main Steam Header pressure lowers as turbine load is raised. Reduction of
boron concentration results in more negative MTC. Tavg rising has a negative effect on MTC.
This additiv e negative effect is the most negative of all choices given.
d.
Incorrect. Main Steam Header pressure lowers as turbine load is raised. Reduction of boron
concentration results in more negative MTC. Tavg rising has a negative effect on MTC. This
additive negative effect is the most negative of all choices given.
64 of 81
Question Number:
60
Watts Bar 2008 NRC Init ial License Exam
WRITTEN QUESTION DA TA SHEET
KIA : 055 G2.4.3
Condenser Air Removal
Ability to identify post-accident instrumentation.
Tier :
2
Group:
2
RO Imp:
3.7
RO Exam:
SRO Imp: 3.9
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
NEW
Applicable 10CFR55 Section :
41 .6/45.4
Learning Objective: 3-0 T-SYS090A, Objective 16: Identify where Post Accident Monitors are used & read
out.
References :
SOI-90.05, POST-ACCIDENT RAD MONITORS, Rev 12; 1-47W610-90-5 R40.
Question:
Which ONE of the following identifies monitors associated with Condenser Vacuum Pump discharge which
are Post Accident Monitors (PAM) , in accordance with SOI-90 .05, Post Accident Radiation Monitors?
A.
Both 1-RM-90-1 19 and 1-RM-90-404 .
B.
Neither 1-RM-90-119 nor 1-RM-90-404.
C.
1-RM-90-119 is a PAM, but 1-RM-90-404 is NOT.
D.
1-RM-90-404 is a PAM, but 1-RM-90-119 is NOT.
DISTRACTOR ANALYSIS
a.
Incorrect. SOI-90.05 identifies1-RM-90-404 as a Post Accident Rad Monitor, but 1-RM-90-119 is NOT
identified as a Post Accident Rad Monitor. Both being Post Accident Rad Monito rs is plausible because
1-RM-90-11 9 is used during a SGTR event as an indication of the accident.
b.
Incorrect. SOI-90.05 identifies1-RM-90-404 as a Post Accident Rad Monitor, but 1-RM-90-119 is NOT
identified as a Post Accident Rad Monitor. Neither being Post Accident Rad Monitors is plausible
because neither is listed in the Tech Spec for Accident Monitoring Instrumentation.
c.
Incorrect. SOI-90.05 identifies1-RM-90 -404 as a Post Accident Rad Monitor, but 1-RM-90- 119 is NOT
identified as a Post Accide nt Rad Monitor. Plausible because the 1-RM-90-119 is used during a SGTR
event as an indication of the accident and the candidate may know one of the monitoring is a Post
Accide nt Rad Monitor.
d.
CORRECT. 1-RM-90-404 is identified in SOI-90.05 as a Post Accident Rad Monitor, but 1-RM-90-119
is NOT identified as a Post Accident Rad Monitor.
65 of 81
Question Number :
61
Watts Bar 2008 NRC Initial Licens e Exam
WRITTEN QUESTION DA TA SHEET
K/A : 056 K1.03
Condensa te
Knowledge of the physical connections and/or cause-effect relationships between the Cond ensate System
and the following systems: MFW.
Tier:
2
Group:
2
RO Imp:
2.6
RO Exam:
SRO Imp: 2.6
SRO Exam:
Yes
Yes
Cognitive Level :
Source:
NEW
Applicable 10CFR55 Section:
41.2 to 41.9/45.7 to 45.8
Learning Objective: 3-0T-SYS002A Objective 16: List the conditions which will cause the main feed
pump turbine condense r valves to automatically close .
References:
1-47W611-2-1; AOI-16, Rev. 30.
Question:
Which ONE of the following occurs automa tically if the "B" MFP trips due to thrust bearing wear with the
plant initially at 100% power? (Assume no other equipment failures .)
A.
The motor driven AFW Pumps start.
B.
The Condensate DI pumps trip if feedwat er flow drops to <80%.
C.
The "B" MFPT condenser condensate inlet and outlet valves go closed.
D.
The short cycle valve, 1-FCV-2-35, modulates open to dump excessive condensate flow.
DISTRACTOR ANALYSIS
a.
Incorrect. The AFW pumps start on the loss of both MFPs and other conditions that are not described
in the stem of the question. Loss of one MFP at the stated power level starts the standb y MFP.
Further plausibility is added due to some Westinghouse plants are design ed this way.
b.
Incorrect. Suction pressure <20 psig causes the Condensate DI pumps to trip.
c.
CORRECT. Since MFW flow is greater than 40%, the condensa te inlet and outlet valves for the MFP
turbine condensers go closed.
d.
Incorrect. Plausible since a differential pressure is developed across a flow element (FE-2-35). This
L\\ P is converted to a flow signal which is used to control Short Cycle Valve (FCV-2-35) to ensure a
minimum of 5500 gpm condensate flow.
66 of 81
Question Numb er:
62
Watts Bar 2008 NRC Initial Li cense Exam
WRITTEN QUESTION DA TA SHEET
K/A : 068 K6.10
Liquid Radwas te
Knowledge of the effect of a loss or malfunction on the following will have on the Liquid Radwaste System :
Radiation monitors.
Tier:
2
Group:
2
RO lmp:
SRO Imp:
2.5
2.9
RO Exam :
SRO Exam :
Yes
Yes
Cognitive Level:
Source:
HIGH
NEW
Applicable 10CFR55 Section:
41.7 /45.7
Learning Objective: 3-0T-SYS0 90A. Objective 7: Determine Interlocks and/or cause-effect relatio nships
between the Rad Monitoring Systems (ARM & Process) and the areas they monitor. Include HVAC systems
and area isolations.
References:
1-47W611-77-2 R5. ARI-180-187 Rev 30.
Question :
Which ONE of the following identifies a condition that causes an instrument malfunction alarm on 0-RM-90-
122, WDS Liquid releas e radiation monitor, and the effect the instrument malfu nction alarm has on valve 0-
RCV-77-43. CT BLDN LN RAD RELEASE CNTL?
A.
B.
C
D.
Cause of the alarm
Loss of signal from detector
Loss of signal from detector
Loss of flow through the monitor
Loss of flow through the monitor
Effect on O-RCV-77-43
Auto closes 0-RCV-77-43 if the valve was open .
Prevents 0-RCV-77-43 from opening if the valve's
local control handswitch was placed to OPEN.
Auto closes 0-RCV-77-43 if the valve was open.
Prevents 0-RCV-77-43 from opening if the valve's
local control handswitch was placed to OPEN.
DISTRACTOR ANALYSIS
a.
CORR ECT. As identified in ARI-181-C , a loss of signal from the detector would cause an Instrument
Malfunction alarm and if 0-RCV-77-43 was open it would be automatically closed as identified on 1-
b.
Incorrect. As identified in ARI-181-C. a loss of signal from the detector would cause an Instrument
Malfunction alarm and if 0-RCV-77-43 handswitch was placed to open the valve would open but would
reclose when the switch was released as identified on 1-47W6 11-77-2.
c.
Incorrect. A loss of flow through the detector is not identified in ARI-181-C as a condition that would
cause an Instrument Malfunction alarm but if 0-RCV-77-43 handswi tch was open it would be
automatically closed due to the low flow condition as identified on 1-47W61 1-77-2.
d.
Incorrect. A loss of flow through the detector is not identified in ARI-181-C as a condition that would
cause an Instrument Malfunction alarm and if 0-RCV-77-43 handswitch was placed to open the valve
would open but would reclose when the switch was released as identified on 1-47W611-77-2.
67 of 81
Question Number:
63
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA: 075 K1.02
Circulating Water
Knowledg e of the physical connect ions andl or cause-effect relationships between the circulat ing water
system and the following systems: Liquid radwaste discharge.
Tier:
2
Group:
2
RO Imp :
2.9
SRO Imp: 3.1
RO Exam:
SRO Exam :
Yes
Yes
Cognit ive Level :
Source:
WBN BANK
Applicable 10CFR55 Section:
41.2 to 41.9/45 .7 to 45.8
Learning Objective: 3-0T-SYS0077A, Objective 20: Correctly locate the local and MCR controls for the
Liquid Radwaste System ; 3-0T-SYS027A, Objective 16: Explain the minimu m cooling tower blowdown flow
rate interlock with Radwaste, S/G Slowdown , and Cond Demin discharge valves.
References :
SOI-77 .01, Liquid Waste Disposal, Rev 57.
Quest ion:
A release of the Monitor Tank is in progress through the Liquid Radwaste System. Which ONE of the
following conditions directly results in automatic closure of 0-FCV -77-43, Liquid Radwa ste Release Flow
Control Valve?
A.
High radiation signal on 0-RM-90-225, Condensate Demineralizer Release Liquid Radiation Monitor.
S.
River flow drops to less than 3500 cfs after a 1 minute time delay.
C.
Cooling Tower Slowdown flow drops below 25,000 gpm.
D.
SG Slowdown flow exceeds 150 gpm.
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible. since the candidate may believe that this high radiation condition causes
termination of a release. If aligned for a release from the Condensate Demineralizer, this radiation
Monitor would in fact termina te the release.
b.
Incorrect. Plausible, since river flow dropping to less than 3500 cfs after a 1 minute time delay closes
the diffuser valves, causing the release to be discharged to the 35 acre pond, but will not close.
c.
CORRECT. Per 501-77.01, this condition causes automatic closur e of 0-FCV-77-43.
d.
Incorrect. Plausible, since SG Slowdown flow exceeding 150 gpm extends the time by 1.5 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
for a release due to increased backpressure on the release line.
68 of 81
Question Number:
64
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DATA SHEET
KIA: 079 A4.01
Station Air
Ability to manually operate and/or monitor in the control room : Cross-tie valves with lAS.
Tier:
2
Group:
2
RO Imp:
2.7
SRO Imp: 27
RO Exam:
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
HIGH
WBN BANK
Applicable 10CFR55 Section:
41.7 / 45.5 to 45.8
Learning Objective: 3-0T-SYS032A, Objective 2: Describe Auto Actions for Loss of Control Air per AOI-
10.
References:
ARI-36-42, Heaters, Turb Seal, & Air, Rev 16.
Question:
Given the following plant conditions :
Unit 1 is operati ng at power when control air pressure starts to drop.
Annunciator 42-F, SERVICE AIR PCV-33 -4 CLOSED, alarms .
The CRO responds in accordance with the Annunciator Response Instruction (ARI).
Which ONE of the following identifies the decreasing Control Air system pressu re that cause s this alarm to
occur and whether the Auxiliary Air compressors would have started if the air pressure dropped low enough
to cause the alarm, but then recovered without dropping any lower?
Pressure to Caus e Alarm
Aux Air Compressors
A.
83 psig
Will have started
B
83 psig
Will NOT have started
C.
80 psig
Will have started
D.
80 psig
Will NOT have started
DISTRACTOR ANALYSIS
a.
Incorrect. The isolation pressure is 80 psig not 83 psig. Plausible because 83 psig is the pressure
where the aux air compressors start, which makes the second part of the distractor correc t. This
pressure setpoint could be misapplied to the station service air isolation pressu re.
b.
Incorrect. The isolation pressure is 80 psig not 83 psig. Plausible because the 83 psig setpoint could
be misapp lied to the station service air isolation pressure and for the aux air compressor start, the
candidate recalls the pressu re at which the aux air compressors load (79.5 psig) and applies this
pressure as the starting pressure.
c.
CORR ECT. The service air isolates at 80 psig and the aux air compressors would have started at 83
psig.
d.
Incorrect. The isolation pressure is correct , but the aux air compressors would have started. Plausible
if the candidate recalls the pressure at which the aux air compressors load which is 79.5 psig and
applies this pressu re as the starting pressure .
69 of 81
Quest ion Number :
65
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DATA SHEET
KIA: 086 K3.01
Fire Protection
Knowledge of the effect that a loss or malfunction of the Fire Protection System will have on the following :
Shutdown capabil ity with redundant equipment.
Tier :
2
Group :
2
RO Imp:
2.7
RO Exam :
SRO Imp: 3.2
SRO Exam:
Yes
Yes
Cognitive Level:
Source :
High
Bank , modified
Applicable 10CFR55 Section:
41 .7 /45 .6
Learning Objective: 3-0T-AO I3000 Objecti ve 10: State the two primary limiting safety conditions which
must be maintained following a postulated Appendi x R fire as specified in AOI-30.2.
References:
AOI-30.2, Safe Shutdown, rev 27.
Que st ion:
Given the following plant conditions:
Unit 1 is currently at 100%.
A fire occurs in the Cable Spreading Room.
The crew was unable to start the HPFP pumps.
The incident Commander also reports that due to multiple fire damper failures the fire is spreading
quickly.
The crew has entered AOI-30.2, Fire Safe Shutdown .
In accordance with AOI-30.2, which ONE of the following failures results in a loss of a Control Function
required to place the Plant in Hot Standb y?
REFERENCE PROVIDED
A.
Motor Driven AFW Pumps will not start.
B.
RCS Thermal Barrier Booster Pumps trip.
C.
One Main Steam Isolation Valve fails to close.
D.
Letdown Isolation Valve 1-FCV-62-69 fails closed .
DISTRACTOR ANALYSIS
a.
Incorrect. The TO AFW Pump is required for the Safe Shutdown. Plausible since the candidate may
incorrectly assume that ALL AFW pumps are required .
b.
Incorrect. The trip of the thermal barrier booster pumps does not impact the Safe Shutdown capabilities
of the plant. Plausible if the candidate assumes that forced circulation is a concern .
c.
Incorrect. Per the Safe Shutdown Logic Diagram, the failure of the MSIV does not impact the ability to
place the plant in Hot Standby. Plausible. if the candid ate misapp lied the diagram.
d.
CORRECT. Letdown is required by the Safe Shutdown Logic Diagram to place the plant in Hot
Standby.
NOTE:
This question requires the candidate to use the following refere nce:
AOI-30.2 Section 4.5, Safe Shutdown Logic Diagram
70 of 81
Question Number:
66
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA: G2.1.5
Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations,
etc.
Tier:
3
Group:
RO Imp:
2.9
RO Exam :
SRO Imp: 3.9
SRO Exam :
Yes
Yes
Cognitive Level:
Source :
HIGH
Bank Mod
Applicable 10CFR55 Section:
41 .10/43.5/45 .12
Learning Objective: 3-0T-SPP1000, Objective 6: Describe shift staffing requirements.
References :
OPDP-1 CONDUCT OF OPERATIONS, Section 3.1.3; Tech Spec 5.2.2 Shift Staffing.
Question:
For Mode 1 ope ration . which ONE of the following describes the MINIMUM number of Licensed Operator
positions to man a shift, AND the MAX IMUM time requirement to restore if the minim um shift manning for
Licensed Ope rators is not met per OPDP-1, Conduct of Opera tions and Tech Spec 5.2.2, Unit Staff?
Minimum Shift Manning
Time Requirement
A.
Restore within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
B.
Restore within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
C
Restore within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
D.
Restore within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
DISTRACTOR ANALYSIS
a.
CORRECT. Per OPDP-1 and TS 5.2.2, 2 ROs are required with a 2 Hour time limit per TS 5.2.2.
b
Incorrect. 2 ROs required is correct.
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time limit is incorrect. Plausible since TS does have one
hour time requirements and if minimum shift manning is not met student may concl ude this is impo rtant
enough for a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time requirement.
c.
Incorrect. 1 RO is incorrect. Plausible due to in modes 5 and 6 the minimum requiremen t is only 1 RO.
The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time requirement is correct.
d.
Incorrect. 1 RO is incorrect. Plausible due to in modes 5 and 6 the minimum requirement is only 1 RO.
Second part is also incorrect.
Plausible due to TS does have one hour time requi rements and if
minimum shift manning is not met student may conclude this is important enough for a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time
requirement.
71 of 81
Question Number:
67
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DATA SHEET
KIA: G2.1.28
Know ledge of the purpose and function of major system components and controls.
Tier :
Group:
3
RO Imp:
4.1
SRO Imp: 4.1
RO Exam:
SRO Exam :
Yes
Yes
Cognitive Level:
Source:
WBN Bank
Applicable 10CFR55 Section:
41.7
Learning Objective: 3-0 T-SYS001B, Obje ctive 9: Identify which controller is in service when Tavg is
selected with the unit at powe r.
References:
1-47W611-1-2.
Question:
With the Steam Dump Mode switch in the Tavg mode, what determin es whether the load rejection controller
or the Rx trip controller will be in service?
A.
Loss of Load (C-7) Interlock.
B.
LO-LO Tavg Interlock (550°F).
C.
"A" Train Reactor Trip breaker position.
D.
"B" Train Reactor Trip breaker position .
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible. since the Loss of Load interlock will arm the steam dump s, but does not
determine which controller output will position the dump valves.
b.
Incorrect. Plausible, since the LO-LO Tavg Interlock will close all of the steam dumps if temperature
drops to below 550°F
c.
Incorrect. The A train Reactor Trip breaker is used to ARM the steam dumps on a reactor trip.
d.
CORRECT. The B train reactor trip breaker is used to select the controller in service. With NO reactor
trip, the Load Rejection Controller is selected. Once the B train reactor trip break er opens, the Reactor
Trip controller is selected.
72 of 81
Question Number:
68
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA: G2.1.36
Knowledge of procedures and limitations involved in core alterations.
Tier:
3
Group:
SIG
RO Imp:
3.0
RO Exam:
SRO Imp: 4.1
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
SQN BANK
Applicable 10CFR55 Section:
41.10/43.7
Learning Objective: 3-0T-G00700, Objective 5: Describe the major steps that the operator must take
when unloading fuel per this instructio n.
References:
FHI-7 , Rev. 31.
Question:
Given the following plant conditions:
Unit is in Mode 6.
15 fuel assem blies have been reloaded after a complete core off-load.
Source Range N-131 indicates 10 cps and is selected for audible count rate indication.
Source Range N-132 indicates 5 cps.
In accordance with FHI-7 , "Fuel Handling and Movement", which ONE of the following unanticipated
changes in count rate requires suspension of core alterations?
A.
N131 indicates 25 cps and N132 indicates 8 cps.
B.
N131 indicates 15 cps and N132 indicates 20 cps.
C.
N131 indicates 40 cps and N132 indicates 8 cps.
D.
N131 indicates 20 cps and N132 indicates 15 cps.
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible since N131 has increased by greater than a factor of 2, but N132 has not changed
by a factor of 2.
b.
Incorrect. Plausible since N131 has not increased by greater than a factor of 2, but N132 has changed
by a factor of 4 but has not exceeded the factor of 5 which would require suspens ion of core
alterations .
c.
Incorrect. Plausible, since N131 has increased by a factor of 4, but has not exceeded the factor of 5
which would require suspension of core alterations.
d.
CORRECT
Both source range chann els have doubled, which would require suspension of core
alterations.
73 of 81
Questio n Number:
69
Wat ts Bar 2008 NRC Ini tial Licens e Exam
WRITTEN QUESTION DA TA SHEET
KIA : G2.2.12
Know ledge of surveillance procedures.
Tier:
3
Group:
RO Imp:
3.7
RO Exam :
SRO Imp : 4.1
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
WBN Ban k
Applicable 10CFR55 Sec tion:
41 .10 /45 .13
Learning Objective : 3-0T-SPP0802, Objective 2: Explain the diffe rence between the surveillance due
date and the WBN extens ion date for both Tech Spec and Non Tech Spec surveillances.
References:
SPP-8.2, Rev. 3.
Question :
A Power Range channel has failed requiring the crew to impleme nt AO I-4, NIS Malfu nctions. The US has
entered the appropriate Technical Specifications and states that a flux map will be required per Surve illance
Requi rement 3.2.4.2.
SR 3.2.4.2 directs the operators to verify OPTR is withi n limits using moveable incore detectors once within
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
What is the maximum time for the initia l performance of the flux map and the maximum time for subsequent
perfo rmances?
A.
B.
c.
D.
Initial Performance
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
15 hours
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
15 hours
Subsequent Performances
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
12 hours
15 hour1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> s
15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />
DISTRACTOR ANALYSIS
a.
Incorrect.
Plausible since the first performance must be accomplished within the specified time with
NO extension. An extension is allowed for subsequent performances.
b.
Incorrect.
Plausible since an extension is allowed for SUbsequent performances, but the first
performance must be acco mplished within the specified time with NO extension.
c.
CORRECT. There is no extension allowed for the first perfo rmance, and an extension of 25% of the
allowed time may be granted for subsequent performances.
d.
Incorrect. Plausible since an extension is allowed for subsequent perfo rmances, but the first
performance must be accomplished within the spec ified time with NO extension .
74 of 81
Question Number:
70
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
K/A : G2.2.44
Ability to interpret control room indications to verify the status and operation of a system . and understand
how operator actions and directives affect plant and system conditions.
Tier:
3
Group:
RO Imp:
4.2
RO Exam:
SRO Imp: 4.4
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
HIGH
NEW
Applicable 10CFR55 Section:
41.5/43.5/45.12
Learning Objective : 3-0T-SYS003B, Objective 23: Using plant drawings. determine the effect of a loss of
instrument air/control power on the following valves/components: a.
MDAFWP regulating valve (main and
bypass). b TDAFWP regulating valve, c. AFW pumps.
References :
AOI- 10, LOSS OF CONTROL AIR. Rev. 38; SOI-3.02, Auxiliary Feedwater System . Section
8.5.1.
Question:
Given the following conditions:
Unit 1 shutdown is on progress.
Reactor power is 14% and decreasing .
Intermediate Range NI-36 fails HIGH.
Which ONE of the following identifies how the failure of the NI will effect the reactor trip system and the
effect the failure will have on the Source Range Nls?
Reactor Trip System
A.
Reactor trip will occur at the time of failure.
B,
Reactor trip will occur at the time of failure,
C.
Reactor trip will occur if power reduction is
continued .
D,
Reactor trip will occur if power reduction is
continued,
Effect on Source Range Nls
Source Range Nls will have to be
MANUALLY reinstated.
Source Range Nls will AUTOMATICALLY
reinstate.
Source Range Nls will have to be
MANUALLY reinstat ed.
Source Range Nls will AUTOMATICALLY
reinstate.
DISTRACTOR ANALYSIS
a,
Incorrect. Plausible, since power is greater than 10%. the P-10 blocks are active. This would prevent
an immediate reactor trip at the time of the failure, The SR Nls WOULD have to be manually
reinstated since the P-6 permissive requires both IR channels to be below the setpoint to automatically
reinstat e.
b.
Incorrect. Plausibl e, since power is greater than 10%. the P-10 blocks are active. This would prevent
an immediate reactor trip at the time of the failure, The SR Nls WOULD have to be manually
reinstated since the P-6 permissi ve requires both IR channe ls to be below the setpoint to automatically
reinstate,
c.
CORRE CT. Power is greater than 10%, initially and the P-10 blocks are active. This would prevent an
immediate reactor trip at the time of the failure. If the power reduction continues to a point less than
10% on 3/4 PR channels, P-10 would be unblocked and the 1/2 IR trip would occur. The SR Nls
WOULD have to be manuall y reinstated since the P-6 permissive requires both IR channels to be
below the setpoint to automatically reinstate.
d.
Incorrect. Plausible since power is greater than 10% initially. and the P-10 blocks are active, This
would prevent an immediate reactor trip at the time of the failure, If the power reduction continues to a
75 of 81
Wat ts Bar 2008 NRC Initial Licen se Exam
WRITTEN QUESTION DATA SHEET
point less than 10% on 3/4 PR channels. P-10 would be unblocked and the 1/2 IR trip would occur.
The SR Nls would have to be manually reinstated since the P-6 permiss ive requires both IR channels
to be below the setpoint to automatically reinstate.
76 of 81
Question Number :
71
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA: G2.3.7
Ability to comply with radiation work permit requirements during normal or abnormal conditions.
Tier:
3
Group:
RO Imp:
3.5
RO Exam:
SRO Imp: 3.6
SRO Exam :
Yes
Yes
Cognitive Level:
Source:
NEW
Applicable 10CFR55 Section:
41.12 /45.10
Learning Obj ect ive: 3-0 T-RAD0003, Objecti ve 8: Identify the responsibilities of the followin g concerning
the ALARA program: a. Radiation Protection Manager/Radiation Safety Officer, b. TVA NPG Organization ,
c. Employee.
References :
RCI-153 , Radiation Work Permits (RWPs) Rev 0000; RCI-100 , Control of Radiological Work,
Rev 32.
Question:
An individual enters a Radiological Controlled Area (RCA) covered by a General RWP to perform equipment
inspections.
Which ONE of the following identifies an area within the RCA where a Job Specific RWP is required before
entry is allowed , in accordance with RCI-153. Radiation Work Permits?
A.
Area where whole body dose rates exceeds 100 mrem/hr.
B.
Area posted as Hot Particle Area.
C.
Area with general contamination levels greater than 200 dpm/100cm 2.
D.
Area where total expected dose is greater than 5 mrem.
DISTRACTOR ANALYSIS
a.
Incorrect. Required if >1,000mrem/hr.
b.
CORR ECT. Required for area posted as Hot Particle Area.
c.
Incorrect. Required if> 50,000 dpm/tuucrn".
d
Incorrect. Required if total expected dose exceeds 50 mrem.
77 of 81
Question Number:
72
Wat ts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA: G2.3.14
Knowledge of radiation or contamination hazards that may arise during normal , abnormal, or emergency
conditions or activities.
Tier:
3
Group:
RO Imp:
3.4
SRO Imp: 3.8
RO Exam:
SRO Exam :
Yes
Yes
Cognitive Level:
Source:
NEW
Applicable 10CFR55 Section:
4 1.12 /4 3.4 / 45.10
Learn ing Objective: 3-0 T-SYS031A, Objective 3: Describe the ventilation flow path provided by the
control building ventilation system during normal and emergency operation.
References :
ARI-180-187 rev 30, 1-47W866-4 R39.
Question:
Given the following plant conditions:
Following an accident , both Trains of Control Room Isolation have been initiated .
Several Auxiliary Building Area Radiation Monitors rise to the alarm setpoint.
Which ONE of the following MCR air intake radiation monitors will detect and alert the crew of a radiation
hazard entering the control room and what actions the ARI will direct the crew to perform ?
A.
0-RM-90-125,
Stop MCR Emergency Pressurization Fans.
B.
0-RM-90-125,
Align Emergency Pressurization Fan suction to alternate source.
C.
0-RM-90-205,
Stop MCR Emergency Pressurization Fans.
D.
0-RM-90-205,
Align Emergency Pressurization Fan suction to alternate source.
DISTRACTOR ANALYSIS
a.
Incorrect. 0-RM-90-125 is in an isolated flow path due to the CRt. Plausible because the rad monitor is
the incorrect monitor. It would be in the flow path after the supply was realigned and stopping the fans
would be a way of stopping the intake of radiation.
b.
Incorrect. 0-RM-90-125 is in an isolated flow path due to the CRt. Plausib le because the rad monitor is
the incorrect monitor. It would be in the flow path after the supply was realigned and Annunciator 187-A
directs the action to align the Emergency Pressurization Fan suction to alternate source.
c.
Incorrect. 0-RM-90-205 will cause annunciator 187-A to alarm and the ARI directs the action to ALIGN
Emerge ncy Pressurization Fan suction to alternate source, not to stop the Emergency Pressurization
Fan. Plausible because the rad monitor is the correct monitor and stopping the fans would be a way of
stopping the intake of radiation.
d.
CORR ECT. O-RM-90-205 will cause annunciator 187-A to alarm. 0-RM-90-125 is in an isolated flow
path due to the CRI. Annunciator 187-A directs the action to align the Emergency Pressurization Fan
suction to alternate source.
78 of 81
Question Number:
73
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA: G2.4.13
Knowledge of crew roles and responsibilities during EOP usage.
Tier:
3
Group:
RO Imp:
4.0
RO Exam:
SRO Imp: 4.6
SRO Exam :
Yes
Yes
Cognitive Level:
Source:
SQN BANK
Applicable 10CFR55 Section:
4 1.10/45.12
Learning Objective: 3-0T-TI1204, Objective 13: Apply the rules of usage which relate to performing steps
of an EOP in a specified sequence to determ ine when steps may/must be performed.
References:
TI-12 .04, Rev. 7, Page 31 thru 34.
Question:
The Operator-at-the-Controls (OAC) is responding to an accident. He recognizes that he must take actions
which are NOT contained in the emergency operating procedure in effect and are NOT covered by prudent
operator actions. Which ONE of the following describes the proper action to be taken?
A.
The OAC shall take no action until a procedure is developed or revised.
B.
The OAC shall obtain approval from a licensed SRO prior to taking action.
C.
The OAC should obtain concurrence from another licensed RO prior to taking action.
D.
The OAC should immediately take appropriate actions necessary and inform the SRO when time
permits.
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible, since under normal circumstances, the operator would stop a task in progress and
wait until a new procedure was written before taking any additional actions . The operator must stop
actions long enough to get approval from the SRO. The SRO would have to address the situation using
1OCFR50.54(x) criteria.
b.
CORRECT. The operator must stop actions long enough to get approval from the SRO. The SRO
would have to address the situation using 10CFR50.54(x) criteria.
c.
Incorrect. Plausible if the candidate confuses the actions required by a PEER CHECK with actions in b.
d.
Incorrect. Plausible, since Prudent Operator Actions do allow the RO to take manual compensatory
actions which are within the guidelines of an existing procedu re.
79 of 81
Question Number:
74
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DATA SHEET
KIA: G2.4.37
Knowledge of the lines of authority during implementation of the emergency plan.
Tier:
3
Group:
RO Imp:
3.0
RO Exam:
SRO Imp: 4.1
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
WBN Bank
Applicable 10CFR55 Section :
41 .10 /45 .13
Learning Objective: 3-0T-RAD0003, Objective 6: List the extreme emergency exposure guidelines.
References:
EPIP15, Emergency Exposure Guidelines, Rev 13.
Question:
Given the following:
A Site Area Emergency (SAE) has been declared on Unit 1.
Two hours after the SAE declaration, an individual is to be authorized to receive an Emergency
Exposure radiation dose above the TVA whole body dose limit during the mitigation of the
emergency situation .
In accordance with EPIP-15 . Emergency Exposure Guidelines. whose approval is require d for the individual
to receive the dose?
A.
TSC Radcon Manager.
B.
Onshift Shift Manager.
C.
Site Emergency Director.
D.
Site Vice President.
DISTRACTOR ANALYSIS
a.
Incorrect. Per EPIP-15, the Radiation Protection group is responsible for completing necessary
paperwork and obtaining SED's approval
b.
Incorrect. Plausible, if the Shift Manager was in the role of the SED. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time frame stated in
the stem allows for the TSC to be manned and therefore the SED duties would have been assumed
from the Shift Manager.
c.
CORRECT. The SED is the ONLY individual responsible for authori zing emergency dose limits.
d.
Incorrect. Plausible. since the Site VP may be acting as the SED. However the Site VP by title does
not have responsibility for authorizing emergency dose limits.
80 of 81
Question Number:
75
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA : G2.4.42
Knowledge of emergency response facilities.
Tier:
3
Group:
RO Imp:
2.6
RO Exam:
SRO Imp: 3.8
SRO Exam:
Yes
Yes
Cognitive Level:
Source:
NEW
Applicable 10CFR55 Section:
41.10 /45.11
Learning Objective: 3-0 T-PDC-048C, Objective 20: Use the Satellite Phone to make calls during
emergencies.
References:
SOI- 100.01, rev 22.
Question:
Which ONE of the following identifies where a Portable Satellite Telephone, available for use during an
emergency, is located?
A.
Main Control Room
B.
C.
Joint Informat ion Center
D.
Operations Support Center
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible because the Main Control Room does have Satellite phone capabilities, however
the capability is to selected phones via the Stationary Satellite Telephone (SST) system.
b.
CORRECT.
The Portable Satellite Telephone (PST) is located in a cabinet in the Technical Support
Center.
c.
Incorrect. Plausible because the other 3 locations do have selected phones that can be connected to
the Stationary Satellite Telephone (SST) system, and the JIC does not that capabili ty.
d.
Incorrect. Plausible because the Operations support Center does have Satellite phone capabilities,
however, the capability is to selected phones via the Stationary Satell ite Telephon e (SST) system.
81 of 81
Watts Bar Nuclear Plant
NRC Initial License Written Examination - 2008
Master Examination
Please note: The following 29 pages are the Master Examination copy for
the SRO portion of the examination, including the answer key and
distractor analysis data.
Question Number: Y 7/;;;
Walts Ba r 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA: 00000 7 EA2.05
Reactor Trip - Stabilization - Recovery
Ability to determin e or interpret the following as they apply to a reactor trip: Reactor trip first-out indication.
Tier:
1
RO Imp:
RO Exam:
Cognitive Level :
High
Group:
1
SRO Imp: 3.9
SRO Exam :
Yes
Source:
New
Applicable 10CFR55 Sectio n:
41.7 /45.5 /45.6
Learning Objective: 3-0 T-SYS099A, Objecti ve 13: Describe the causes of "General Warning " on SSPS.
Objecti ve 14: Identify where "General Warning" indications can be found.
Referen ces :
SPP- 3.5, Rev. 19: 1-SI-99-10B , Rev. 42.
Questio n:
Given the following conditions:
Unit 1 is at 100% power. Solid State Protection System (SSPS) Train 'B' Actuation Logic testing is being
perform ed using 1-SI-99-1OB with:
Train 'B' SSPS Mode Selector switch in the 'TEST' positio n.
Train 'B' SSPS Input Error Inhibit switch in the "INHIBIT" position.
A unit trip occurs due to the loss of one of the two 48v DC power suppl ies on TRAIN 'A' SSPS. The
followlllg "First Out" annunciators are lit:
1-XA-55-4C, Turbin e Tr ip First
Window 73C - "RX TRIP BKRS RTA & BYA OPEN"
Window 74C - "RX TRIP BKRS RTB & BYB OPEN"
Window 74B - "MFPT A&B TRIPPED"
1-XA-55-4D Reactor Trip First Out
Window 76B "TUR BINE TRIP "
Which ONE of the following identifies both the sequence of events of the unit trip, and the time allowed to
make the required NRC 50.72 notification?
Sequence of Events
A.
Turbine trip caused the Reactor trip.
B
Turbine trip caused the Reactor trip.
C.
Reactor trip caused the Turbine trip.
D.
Reactor trip caused the Turbine trip.
NRC Notification Required With in
Four Hours
Eight Hours
Four Hours
Eight Hours
DISTRACTOR ANALYSIS
a.
Incorrect. The conditions stated in the stem result in a general warning on both trains of the SSPS
which causes the reactor trip and bypass breake rs to open (but no reactor first out annunc iator will be
lit). Thus the turbine trips as a result of the reactor trip. The four hour notification to NRC for a reactor
trip is correct. Plausible due to the notification time being correct and no other reactor trip first out
annunc iator will be lit except the turbine first out annun ciator.
b.
Incorrect. The conditions stated in the stem result in a general warnin g on both trains of the SSPS
which causes the reactor trip and bypass breakers to open (but no reactor first out annunciator will be
lit), thus the turbine trips as a result of the reactor tripping. The notification to NRC for a reactor trip is
a four hour notification , not an eight hour notification. Plausible due to the notification time being
correct and no other reactor trip first out annunciator lit except the turbine first out annunciator; and that
the notification time could be misapplied since the ESF actuation required time is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and an ESF
actuation of AFW does occur on a reactor trip.
c.
CORRECT. The conditions stated in the stem result in a general warning on both trains of the SSPS
which causes the reactor trip and bypass breakers to open (but no reactor first out annun ciator will be
lit), thus the reactor trip causing the turbine trip is correct. The four hour notification to NRC for a
reactor trip is correct.
1 of 29
Watts Bar 2008 NRC Initia l License Exam
WRITTEN QUESTION DA TA SHEET
d.
Incorrect. The conditions stated in the stem result in a general warning on both trains of the SSPS
which causes the reactor trip and bypass breake rs to open (but no reac tor first out annunciator will be
lit), thus the reactor trip causing the turbine trip is correct. The notification to NRC for a reactor trip is a
four hour notification, not an eight hour notificatio n. Plausible beca use the notification time could be
misapplied since an ESF actuation requires an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification and an ESF actuation of AFW does
occur on the reactor trip.
2 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
Question Number: ,z"77
High
NEW
Cognitive Level :
Source:
Yes
KIA : 000022 G2.1.7
Loss of Reacto r Coolant Makeup
Ability to evaluate plant perform ance and make operational judgments based on operating characteristics,
reactor behavior, and instrume nt interpretation.
Tier :
1
RO Imp:
RO Exam :
Group :
1
SRO Imp: 4.7
SRO Exam :
Applicable 10CFR55 Section:
41.5, 43.5
Learn ing Objective: 3-0 T-SYS062A: Explain the automa tic actuation logic and interlocks associated with
the VCT outlet valves, FCV-62-132 and 133 and the CCP suction valves from the RWST , FCV-62-135 and
136.
Referenc es:
SOI-62.02, Rev. 47; ARI-109-115, rev. 16, SPP-10A , 5 0, rev. 5.
Question :
Given the following plant conditions:
Core burnup is 1200 MWD/MTU.
Indicated reactor power is stable at 100%.
VCT low level alarm annunciates .
Auto makeup has failed.
Actual VCT level had lowered to 5% before the crew completed the appropriate corrective action.
Reactor power has stabilized at approximately 97% power.
If this event had occurred with core burnup at 16600 MWD/MTU, the magnitude of the change in reactor
power would be
(1)
, and the Significance Level of the Reactivity Management Event
classification would be recorded on the PER (Problem Evaluation Report) by
(2)
_
(1)
(2)
A.
less
Reactor Engineering.
B
greater
Reactor Engineering.
C.
less
the Management Review Committee.
D.
greater
the Management Review Committee.
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible if candidate confuses the change in boron worth with time in core life to be less,
instead of more. In this case, candidate incorrectly concludes that the power change response is at a
lower magnitude. Correctly recognizes that Reactor Engineering records the significance level of the
event on the PER.
b.
CORRECT. Since boron worth is greater at EOL, injection of RWST inventor y results in a higher
magnitude of change in reactor power. Per the appropriate references, Reactor Engineering records
the significance level of the reactivity event on the PER.
c.
Incorrect. Plausible if candidate confuses the change in boron worth with time in core life to be less,
instead of more. In this case, candidate incorrectly concludes that the power change response is at a
lower magnitude. Further plausible, since Management Review Committee (MRC) is involved with plant
PERs, but assigning the reactivity significance level is not a function of the MRC.
d.
Incorrect. Plausible, since the candidate correctly recognizes EOL conditions, and applies the
knowledge that boron worth is higher. Further plausible, since Management Review Committee (MRC)
is involved with plant PERs, but assigning the reactivity significance level is not a function of the MRC.
3 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
Question Number:
.2 7~,
K/A : 000025 AA2.01
Loss of RHR System
Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System:
Proper amperage of running LPlldecay heat removal/RHR pump(s).
High
NEW
Cognitive Level:
Source:
Tier:
RO Imp:
RO Exam:
Group:
1
SRO Imp: 2.9
SRO Exam:
Yes
Applicable 10CFR55 Section :
43.5 /45.13
Learning Objective: 3-0 T-AOI1400, Rev. 6, Objective 7: Demonstrate ability/knowledge of AOI, to
correctly: a. Recognize entry conditions, b. Respond to Action steps, c. Respond to Contingencies, d.
Respo nd to Notes and Cautions. 3-0T-G01000, Rev. 5, Objec tive 5: Identify the procedure to which the
opera tor is referred if Residual Heat Removal cooling is lost while in during Reduced Inventory/Mid-Lo op
operations. (SOER 88-3 & SOER 85-4)
References:
GO-10, Rev. 37.
Question:
Given the following:
Unit 1 is in Mode 5 follow ing a refueling outage.
The operating crew is drawing vacuum on the Reactor Coolant System.
The RHR pump begins to show signs of cavitatio n.
Which ONE of the following identifies both how the RHR pump motor amps are affected when the pump is
cavitating, and the mitigating strategy that will be implemented if the cavitation cannot be terminated?
Motor Amps
Mitigating Strategy
A.
Unstable and fluctuating.
Break vacuum per GO-10. Reactor Coolant System Drain
and Fill Operations, then enter AOI-14, Loss of RHR
B.
Unstable and fluctuating.
Immediately enter AOI-14. Loss of RHR Shutdown
Cooling . Break vacuum as directed by the AOI.
C.
Stable but reduced.
Break vacuum per GO-1 0. Reactor Coolant System Drain
and Fill Operations, then enter AOI-14, Loss of RHR
D.
Stable but reduced.
Immediately enter AOI-14 , Loss of RHR Shutdown
Cooling. Break vacuum as directed by the AO I.
DISTRACTOR ANALYSIS
a.
CORRECT. GO-10 describes cavitation and amps being unsteady and directs the break ing of
vacuum prior to implementing AOI-14.
b.
Incorrect. While GO- 10 describes cavitation and amps as stated in the distractor, the GO requires
the vacuum break prior to the transition to AOI-14.
c.
Incorrect. Amps will not be "stable but reduced" if the pump was cavitating; however the mitigating
strateg y is correct.
d.
Incorrect. Amps will not be "stable but reduced" if the pump was cavitating. The GO requires the
vacuum break prior to the transition to AOI-14.
4 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
Quest ion Number : ~1 q
KIA: 000027 G2.4.6
Pressurizer Pressure Control System Malfunction
Knowledge of EOP mitigation strategies.
High
BANK
Cogn itive Level :
Source :
Tier :
RO Imp:
RO Exam:
Group:
SRO Imp: 4.7
SRO Exam:
Yes
Applicabl e 1OCFR55 Sect io n: 41.10 143.5 / 45. 13
Learning Objective: 3-0 T-AOI1800; Objective 5: Explain the operator actions for dropping RCS pressure.
References :
E-O, Rev.27 , Drawing 47W8 13-1, AOI- 18, Rev. 21.
Question:
Given the following conditions :
The plant is operating at 100% power steady state conditions.
All systems are aligned normally.
A failure of the Pressurizer Spray Valve PCV-68-340D causes it to go full OPEN .
The OAC has attemp ted to take manua l control of the Spray Valve, but is unable to close it.
Pressurizer pressure continues to LOWER.
What is the appropriate mitigation strategy and which procedure{s) will be used to implement the strategy?
A.
Enter AOI-18, Malfun ction of Pressurizer Pressure Control System, initiate a Reacto r Trip, trip RCP #1,
then enter E-O, Reactor Trip or Safety Injection.
B.
Refer to ARI-90-B, PZR PRESS LO DEVN BACKU P HTRS ON, isolate Train B Essential Air to
Containment to fail the Pressurizer Spray Valve closed, and then enter AOI-1 8.
C.
Refer to ARI 90-B, PZR PRESS LO-DEVN BACKUP HTRS ON, isolate Train A Essential Air to
Containment to fail the Pressurizer Spray Valve closed, and then enter AOI-18.
S' ~
D.
Enter AOI-18. Malfunction of Pressurizer Pressure Con t ro l~, initiate a Reactor Trip AND SI,
trip RCP #1, then enter E-O, Reactor Trip or Safety Injection.
DISTRACTOR ANALYSIS
a.
Incorrect. Guidance is given in AOI-18 for tripping the reactor if the Pressurizer spray valves cannot be
closed. This requires entry into E-O. Candidate must recognize that a Reacto r trip AND an SI is
required.
b.
Incorrect. Plausible, since isolation of Essential Air to Containment causes the spray valve to fail
closed .
c.
Incorrect. Plausible, since isolation of Essential Air to Containment causes the spray valve to fail
closed.
d.
CORRECT. Guidance is given in AOI-18 for tripping the reactor AND an SI if the Pressurizer spray
valves cannot be closed. This requires entry into E-O.
5 of 29
Question Number: KW
Watts Bar 2008 NRC Initial Licens e Exam
WRITTEN QUESTION DA TA SHEET
KIA: 000058 AA2.01
Loss of DC Power
Ability to determine and interpret the following as they apply to the Loss of DC Power : That a loss of dc
power has occurred; verification that substitute power sources have come on line
Tie r:
Group:
ROlmp:
SRO Im p: 3.1
RO Exam:
SRO Exam:
Yes
Cognitive Level:
Source:
High
New
Applicable 10CFR55 Section:
43.5
Learning Objective: 3-0 T-SYS057P, Objective 6: Explain how the operator can tell if the 125v Vital
Charger or the 125v Vital Battery is supplying power to the 125v Vital Battery Boards.
References:
Tech Spec 3.8.4 Bases, page B 3.8-57, Rev 87
Question:
Given the following:
The plant is operating at 100% power.
The 125 V DC VITAL CHGR III fails and its output breaker opens .
A report is received that there is arcing occurring on Vital Battery III.
The Shift Manager directs that 0-BKR-236-3/109 125V Vital Battery III Breaker, between Vital
Battery III and Vital Battery Board III be opened.
The Unit Supervisor has determined that power to Vital Battery Board III will be restored using Vital
Battery V, in conjunction with the Spare Battery Charger.
Vital Battery V is currently in service to Battery Board V.
Which ONE of the following describ es the expected indication on 1-EI-57-94, Vital Batt BD III AMPS which
will confirm that power has been restored to Battery Board III, AND what is the operability status of Vital
Battery Board III?
1-EI-57 -94 (Batt BO III Amps) Indication
A.
Indicating UPSCALE from zero.
B.
Indicating DOWNSCALE from zero.
C.
Indicating UPSCALE from zero .
D.
Indicating DOWNSCALE from zero.
Operability Status of Vital Batt. BO III
Can NOT be considered operable with
both Vital Battery V and the spare charger
connected concurrently.
Can NOT be considered operable with
both Vital Battery V and the spare charger
connected concurrently.
Operable with the spare charger and Vital
Battery V connected to the Battery Board and with all
applicable surveillances on Vital Battery V satisfied.
Operable with the spare charge r and Vital
Battery V connected to the Battery Board and with all
applicable surveillances on Vital Battery V satisfied.
OISTRACTOR ANALYSIS
a.
Incorrect. Upscale is the incorrect indication if a battery charger is in service providing power to the
loads. The battery board is operable with the spare charger and Battery V connected as identified in
SOI-236.03, therefore using them concurrently is not an operab ility concern .
b.
Incorrect. Downscale is the correct indication since a battery charger is providing power to the loads,
even though a different battery is connect ed to the Vital Board III. The battery board is operable with
the spare charger and Battery V connected as identified in SOI-236.03, therefore using them
concurrently is not an operability concern.
6 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
c.
Incorrect. Upscale is the incorre ct indication if a battery charger is in service providing power to the
loads . The Battery Board is operable with the spare charger and Battery V connected provided all
applicable surveillances on Vital Battery V are satisfied.
d.
CORRECT. Downscale is the correct Indication when a battery charger is providing power to the
loads, even though a different battery is connected to the Vital Board III. The SOl has the spare
charger place d in service prior to Battery V being connected. The battery board is operable with the
spare charger and Battery V (if operable) connected as identified in SOI-236.05. Per Tech. Spec
Basis 3.8.4, page B 3.8-57, the Vital Battery V can be considered operable after it is connected to a
board and all applicable SRs have been verified satisfactorily.
7 of 29
Question Number: ~} I
Watts Bar 2008 NRC Initial License Exa m
WRITTEN QUESTION DATA SHEET
KIA : W/E1 1 G2.2.22
Loss of Emergency Coolant Recirc.
Knowledge of limiting conditions for operations and safety limits.
Tier:
Group:
RO Imp:
RO Exam:
SRO Imp: 4.7
SRO Exam :
Yes
Cognitive Level:
Source:
High
NEW
Ap plicabl e 10CFR55 Section :
41.5. 43.2
Learning Objective: 3-0T-SYS0 63A. Rev. 10, Objective 30: Given the condition/statu s of the Emergency
Core Cooling system/component and the appropriate sections of Tech Specs. determ ine if operability
requirements are met and what actions, if any. are required.
References:
LCO 3.3.2, Action K, including basis.
Question :
With the plant at full power, and during the Shift Turnover for the 1900 shift, the Unit Supervisor is informed
of the following:
1-LS-63-50A (RWST Low Level) was declared inoperable at 1000 that day.
It was placed in the configuration required by Technical Specifications at 1400, and is expected to
remain inoperable until 2300.
A required surveillance instruction on 1-LS-63-51A (RWST Low Level) must be completed by 2330
today to preven t being out of frequenc y due to exceeding the NRC late date.
The surveillance involves having 1-LS-63-51A in the required configuration for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
If the surveillance is completed by 2330, which ONE of the following describes the expected effe ct on the
automatic switchover to containment sump function while 1~6 3-51A is out of service?
~_
A.
Functional. Even though two level switches are TRIPP ED and are inoperable, the remaining
operable level switches are sufficient for switchover to be functional.
B.
Functional. Even though two level switches are BYPASS ED, the remaining level switches are
sufficient for switchover to be functional.
C.
Not functional, because two level switches are TRIPPED and at least three level switches are
required for switchover to be functional.
D.
Not functional , because two level switches are BYPASSED, and at least three level switches are
required for switchover to be functional.
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible, since there are other components in the plant that are placed to trip when
discovered inoperable. However, candidate fails to recognize that when one of these level switches
is inoperable, it is bypassed, not tripped. Further plausibili ty is added by the condit ions given; i.e.,
there TWO switches affected, and the remaining two are sufficient for actua tion capabi lity.
b.
CORRECT. Per LCO 3.3.2, Action K, one inoperable channel of "Automatic Switchover to
Containme nt Sump" must be bypassed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Anothe r channel may also be bypassed for
surveillance testing, leavinq only two channels. However, the remaining two channels are sufficient
for functionality of actuation.
c.
Incorrect. Candidate incorrect ly believes that these conditions require tripping of the two channels,
and uses that incorrect information to conclude that actuation capability is rendered not functional.
Plausible, since candidate may believe at least three channels are required to be operable at all
times, for Mode 1.
d.
Incorrect. Candidate correctly recognizes that these conditions require bypassing the two channels .
However, candidate fails to realize that this does not disable emergency coolan t recirculation
actuation capabilit y. Plausible, since some plant equipment requires at least three of four channels to
be operable; however, the candida te misapplie s that concept for these conditions .
8 of 29
Question Number:
'~2
Watts Bar 2008 NRC Initial License Ex am
WRITTEN QUESTION DA TA SHEET
KIA: 000059 AA2.03
Accidental Liquid RadWaste Rei.
Ability to determine and interpret the following as they apply to the Accidental Liquid Radwaste Release:
Failure modes. their symptoms. and the causes of misleading indications on a radioactive-liquid monitor.
Tier:
1
Gro up :
2
RO Imp:
RO Exam:
SRO Imp: 3.6
SRO Exam :
Yes
Cognitive Level:
Source:
Low
NEW
Applicable 1OCFR55 Section:
43.5.45.13
Learning Objective: 3-0T-SYS07 7A, Objective 19: Discuss how processed water is released .
References :
SOI-77 .01, Rev. 0058; AR1180-187, Rev. 30;
0-00 1-90-1, liquid Radwaste Tank Release, Rev. 0028.
Question :
Given the following :
The unit is at 100% power and all equipment is available.
A planned Cask Decontamination Collector Tank (CDCT) release is in progress when the following
occurs :
- Annunciator 181-A "WDS RELEASE LINE 0-RM-90-122 L1Q RAD HI" alarms.
- The Monitor Tank level is 70% and lowering.
Which ONE of the following identifies whether the release permit was violated and the release permit
requirements to allow the CDCT release to continue, in accordance with SOI-77 .01, Liquid Waste Disposal?
Permit Violated
The release permit would be violated
A.
because the liquid releas ed was not
sampled prior to the release.
The release permit would be violated
B.
because the liquid release d was not
sampled prior to the release.
The release permit was NOT
C
violated because the release was
terminated by the high rad signal.
The release permit was NOT
D.
violated because the release was
terminated by the high rad signal.
Release of COCT
The same release permit can be used following
independent verification of correct valve lineup.
A new release permit must be generated.
The same release perm it can be used following
independent verification of correct valve lineup.
A new release permit must be generated.
OISTRACTOR ANALYSIS
a.
Incorrect. The release permit would be violated because liquid released from the Monitor Tank was not
sampled and the same release permit cannot be used to continue the release . Plausible because the
release permit did sample the CDCT liquid intended to be released. and independent verification of the
valve line-up is something required if the rad monitor is inoperable.
b.
CORRECT. The release permit would be violated because liquid released from the Monitor Tank was
not sampled and a new release permit is required to continue the release due to the release being
terminated by a High Rad signal as identified in SOI-77 .01. The release permit did sample the CDCT
liquid intended to be released. and a new release permit is required in order to continue the release.
c.
Incorrect. The terminat ion of the release by the radiation monitor does not prevent a violation of the
permit. The release permit would be violated because liquid released from the Monitor Tank was not
sampled and the same release permit cannot be used to continue the release. Plausible because the
9 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DATA SHEET
release permit did sample the CDCT liquid intended to be released; the rad monitor did terminate the
release and independent verification of valve line-up is something required if the rad monitor is
d.
Incorrect. The termination of release by the radiation monitor does not prevent a violation of the permit.
The release permit would be violated because liquid released from the Monitor Tank was not sampled.
A new release permit is required in order to continue the release. Plausible because the release permit
did sample the CDCT liquid intended to be released; the rad monitor did terminate the release and a
new release permit is required.
10 of 29
Que stion Number:
Il' 0 .-2.,
..... L" ___
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA : 000068 G2.4.8
Control Room Evac.
Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
Tier:
1
Grou p:
2
RO Imp:
RO Exam:
SRO Imp: 4.5
SRO Exam:
Yes
Cognitive Level :
Source:
High
SQN Bank
MODIFIED
Applicable 1OCFR55 Section:
43.b (5)
Learning Objective: 3-0T-AOI3000, Objective 12: Demonstrate ability/k nowledge of AOI -30.1 and 30.2
by: a. Recognizing entry conditions, b. Responding to required actions of the AOI , c. Responding to
contingencies (RNO), d. Responding to Note s/Cautions.
References:
AOI -30.2, Fire Safe Shutdown, Rev 27; Note at beginning of Section 3.0, page 5.
Question:
Given the following plant conditions :
Unit 1 was at 100% power when a Main Control Room (MC R) evacuation was required.
The crew entered AOI-27, Main Control Room Inaccessibility.
The crew performed ES-0.1, Reactor Trip Response, prior to leaving the MCR.
While perfo rming actions from the ACR a Safe ty Inject ion occurs.
Which ONE of the following will be the status of the MSIVs when the crew establishes control from the Aux
Control Room and describes the correct procedure usage ?
A.
B.
C.
D.
MSIV Status
Open
Open
Closed
Closed
Procedure Usage
AO I-27 will be the controlling procedure beca use it is wrillen with
mitigating actions to respond to a Safe ty Injection.
E-O, Reactor Trip or Safety Injection, would be used because AOI-27
is wrillen assuming no other accident is occurring.
AOI -27 will be the cont rolling procedure because it is wrillen with
mitigating actio ns to respond to a Safety Injection.
E-O, Reactor Trip or Safety Injection, would be used because AOI-27
is wrillen assuming no other accident is occurring.
DISTRACTOR ANALYSIS
a.
Incorrect. AO I-27 directs the closure of the MSIVs prior to leaving MCR and is wrillen to assume no
othe r accident is occurring. Plausible becaus e if the MSIVs were ope n, steam dump valve s would be
available and the AO I used for abando ning the MCR during an Appendix R fire is wrillen for assuming a
spurious SI occu rs. (The Appendix R AOI is the controlling procedure during Appendix R fire).
b.
Incorrect. AO I-27 directs the closure of the MSIVs prior to leaving MCR and is wrillen to assume no
other accident is occurring . If an accident is occu rring then the eme rgency procedure network would be
used . Plausible because if the MS IVs were open, steam dum p valves would be availa ble and the
emergenc y procedure network would be used because the AOI is wrillen assuming no othe r accidents.
c.
Incorrect. AOI- 27 directs the closure of the MSIVs prior to leaving MCR however and the AOI is wrille n
to assume no othe r accident is occurring. Plausible because the MSIVS are directed to be closed prior
to leaving the MCR and the AO I used for abandoning the MCR during an Appendix R fire is wrillen
assuming a spurious SI occurs. (The Appendix R AOI is the controlling procedure during Appendix R
fire).
d.
CORR ECT . AOI-27 directs the MSIVs to be closed prior to leaving the MCR. The discussion section of
til e AOI states the MCR inaccessibility is not considered to occur with or subsequently with anoth er
accident.
11 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DATA SHEET
Question Number: % 81
KIA: W/E13 EA2. 1
Steam Generator Over-pressure
Ability to determine and interpret the following as they apply to the (Steam Gene rator Ove rpressure)
Facility cond itions and selection of appropriate procedures during abno rmal and emergency operations.
Tier:
1
Group:
2
RO Imp:
RO Exam:
SRO Imp: 3.4
SRO Exam :
Yes
Cognit iv e Level :
Source:
High
NEW
Ap plicabl e 10CFR55 Sec tion :
43.5 /45.13
Learning Objective: 3-0T-T11 204, Rev. 1, Objective 25: Describe when a Function Restoration Instruction
can be exited or transitioned out of.
References :
FR-H.2, Steam Generator Overpressure, Rev 5 page 3.
FR-H.3, Steam Generator High Level, Rev 6 page 3.
Question:
Given the following:
The crew is performing FR-H.2, Steam Generator Overpressure, for an overpressure cond ition on
SG #2.
When the step is addressed to check affected SGs NR level it is noted that the SG #2 level is
indicating 94% narrow range.
Which ONE of the following identifies the correct crew actions as a result of the SG level indicating 94% ?
A.
Continue in FR-H.2, but do not initiate any steam release until TSC evaluation is com plete.
B.
Continue in FR-H.2, steam release may conti nue until NR level indicates 100%.
C
Transition to FR-H.3, Steam Generator High Level, but do not initiate any steam release until TSC
evaluation is comp lete.
D.
Transition to FR-H.3, Steam Generator High Level, steam release may continue until NR level
indicates 100%.
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible since the release of steam is prohib ited with a high level (> 93%) condition until
after a TSC evaluation is complete. Transition to FR-H.3 is directed from FR-H .2 at Step 3.
b.
Incorrect. The transit ion to FR-H.3 is required, howe ver candidate may corre ctly conclude that FR-H.3
is lower in priority on the FR-H status tree and not recall the transition. Plausible since even when NR
level indicates 100%, there is still significant volume before the steam gen erator fills with water.
c.
CORRECT. The RNO for the step directs the transition to FR-H.3 and FR-H.3 restr icts the release of
steam until a TSC evaluation is complete.
d.
Incorrect. The RNO for the step directs the transition to FR-H.3 and the release of steam is restricted
until a TSC evaluation is complete if the level exceed s 93%, therefore with the level at 100%, the
release will be restricted. Plausible since even when narrow range SG level indicates 100%, there is
still significant volume before the steam generator fills with water.
12 of 29
Question Number:
j.(ft,s
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KIA : W/E08 G2.4.18
Knowledge of the specific bases for EOPs.
Tie r:
1
Group:
2
RO Imp:
RO Exam:
SRO Imp: 4.0
SRO Exam:
Yes
Cogniti ve Level :
Source:
High
NEW
Applicable 1OCFR55 Section:
41.10 /43 .1 145.13
Learning Objective: 3-0 T-FRP0001, Rev 10, Objective 9: Explain the basis for returning to the instruction
in effect after identifying that RCS pressure 5. 150 psig and RHR is delivering flow when performing step 1 of
FR-P.1.
References:
FR-P.1, Pressurized Thermal Shock, Rev 14; TI-12.04, Users Guide for Abnormal and
Emergency Instructions, Rev.0007.
Quest ion:
Given the following:
Unit 1 experienced a Reactor Trip and Safety Injection .
The crew transitioned to FR-Z.1. High Containment Pressure, from E-1, Loss of Reactor or
Secondary Coolant.
While FR-Z.1 was being performed, the crew transitioned to FR-P.1. Pressurized Thermal
Shock.
The STA reports the containment pressure has dropped and the containment status tree is
GREEN and that no other RED or ORANGE paths exist.
Which ONE of the following identifies the basis of the FR-P.1 step for checking RCS pressure greater than
150 psig, and the proced ure the crew will implement if a transition is made from FR-P.1 during performance
of the step?
A.
To preclude the need to perform FR-P.1 action s, since pressurized thermal shock is not a serious
concern for a large-break LOCA;
Transition is made back to E-1.
B.
To preclude the need to perform FR-P.1 actions, since pressurized thermal shock is not a serious
concern for a large-break LOCA ;
Transition is made back to FR-Z.1.
C.
To avoid delays caused by unnece ssary soak periods required by FR-P.1.
Transition is made back to E-1.
D.
To avoid delays caused by unnecessary soak periods required by FR-P.1.
Transition is made back to FR-Z.1.
DISTRACTOR ANALYSIS
a.
Incorrect. While a pressurized therma l shock is not a serious concern for a large-break LOCA , the
transition will be back to FR-Z.1, not to E-1. Plausible, since basis is correct, and with the FR-Z.1
status green the candidate could conclude the return to E-1 will be required.
b.
CORRECT. A pressurized thermal shock is not a serious concern for a large-break LOCA because
the system cannot repressurize with a large break LOCA. The step RNO will transition back to
instruction in effect, which is FR-Z.1 even though the status tree for it is now green .
c
Incorrect The bases is not to prevent the soak periods, but the transition to E-1 is correct. Plausible
because the procedure does contain modified Sl termination criteria and SI will be terminated in the
procedure which does reduce cooling to the core and the transit ion is correct
d.
Incorrect. The bases is not to prevent the soak periods, and the transition to FR-Z.1 is not correct.
Plausible because the procedure does contain modified SI termination criteria and SI will be
terminated in the procedure which does reduce cooling to the core and with the FR-Z.1 status green
the candidate could conclude the return to FR-Z.1 will be required.
13 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
Question Number: .-11' Eb
KIA : 003 A205
Reactor Coolant Pump
Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; and (b) based on
those predict ions. use procedures to correct. control. or mitigate the consequences of those malfunc tions or
operations: Effects of VCT pressure on RCP sealleakoff flows.
Hig h
NEW
Cognitive Level:
Source:
Tier :
2
RO Imp:
RO Exam :
Group:
1
SRO Imp: 2.8
SRO Exam :
Yes
Applicabl e 1OCFR55 Section :
41.5 /43.5/45.3 /45/13
Learning Objective: 3-0T-EOPOOOO, Objective 15: Explain the purpose for and basis of each step in E-O.
ES-O.O. ES-O.1. ES-0.2. ES-0.3, and ES-O.4.
References:
ES-0.2, Natural Circulation Cooldown. Rev 20; SOI-68.02. Reacto r Coolant Pumps. Rev 33.
Quest ion:
Given the following:
Unit 1 was in Mode 2 with startup in progress when a loss of off-site power occurred.
The decision was made to place the plant in Mode 5.
The crew implemented ES-0.2, Natural Circulation Cooldown, and started cooling the plant down.
Four (4) hours after the cooldown was initiated both trains of offsite power were restored to the
plant.
The crew determines all criteria to restart the RCPs are met except for the #1 seal leakoff flow on
RCP #2 which is lower than the normal operating band.
Which ONE of the following identifies a change that causes an increase in the # 2 RCP sealleakoff flow and
the actions to be taken and procedure to be used if the seal leakoff flow cannot be established within the
normal operating band?
A.
Lower PRT pressure;
Start RCP #1, continue performing ES-0.2 until all RCS temperatu res are less than 200°F. then
transition to GO-6 , Unit Shutdown From Hot Standby To Cold Shutdown.
B.
Lower PRT pressure;
Start the other 3 RCPs and immediately transition from ES-0.2 to GO-6 , Unit Shutdown From Hot
Standb y To Cold Shutdown.
C.
Lower VeT pressure;
Start RCP #1. continue performing ES-O.2 until all RCS temperatures are less than 200°F. then
transition to GO-6. Unit Shutdown From Hot Standby To Cold Shutdown .
D.
Lower VCT pressu re;
Start the other 3 RCPs and immediately transition from ES-0.2 to GO-6 , Unit Shutdown From Hot
Standby To Cold Shutdown.
DISTRACTOR ANALYSIS
a.
Incorrect. Lowering the PRT pressure affects the #1 sealleakoff flow. but only if the leakoff flow path,
which is routed to the VCT. is isolated . Plausible because starting an RCP provides forced circulation,
and some spray flow correct and if ES-0.2 was continued until the end of the procedure. a transition to
GO-6 would be made .
b.
Incorrect.
Lowering the PRT pressure affects the #1 sealleakoff flow. but only if the leakoff flow path.
which is routed to the VCT, is isolated. Plausible because starting the other 3 RCPs and transitioning to
GO-6 after starting the RCPS is correct.
c.
Incorrect. Lowering the VCT pressure would allow increased sealleakoff flow. but ES-0.2 would not be
continued after the RCP was started (a transition would be made to GO-6 when the pump was started .)
Plausible because lowering the VCT pressure is correct and if ES-0.2 was continued until the end of the
procedure. a transition to GO-6 would be made.
d.
CORR ECT. Lowering the VCT pressure would allow increased sealleakoff flow and the other 3 RCPs
are directed to be started if RCP #2 cannot be started.
14 of 29
Question Number:
j~8 -7
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
KiA: 012 G2.2.44
Reactor Protection
Ability to interpret control room indications to verify the status and operation of a system, and understand
how operator actions and directives affect plant and system conditions .
Tier:
2
Group:
1
RO Imp:
RO Exam:
SRO Imp: 4.4
SRO Exam:
Yes
Cognitive Level :
Source:
High
NEW
Applicable 10CFR55 Section:
41.5 /435 /45.12
Learning Objective: 3-0T-A0 12100, Objective 7: Describe significance of loss of de to Protection and
Control systems (SOER 83-5, Rec 9)].
References:
AOI-21 .01, Loss of 125v DC Vital Battery Bd I, Rev 21; 45W600-99-1 R6; LCO 3.3.2.
Question:
Given the following:
Unit 1 is at 100% power.
1-SI-99-10-B, 31 Day Functional Test of SSPS Train B and Reactor Trip Breaker B, is in progress
with Reactor Trip bypa ss breaker (BYB) closed.
Appendix F, Reactor Breaker Replacement, of 1-SI-99-1O-B is in progress with Reactor Trip
Breaker B (RTB) racked out.
A feedwater tran sient occurs due to an electri cal power loss, resulting in the MCR operator
MANUALLY tripping the reactor.
When Tavg, Pressurizer Pressure , and SG levels are stabilized post trip, the following conditions are
observed:
All 4 Diesel Generators running but NOT connected to the shutdown boards .
All Reactor Trip Breaker and Reactor Trip Bypass breaker indicating lights on 1-M-4 are DARK
except for the GREEN light on the Reactor Trip Bypass Breaker B (BYB) which is LIT.
Which ONE of the following identifies both the position of Reactor Trip Breaker A (RTA ) and the correct SRO
decision relative to com pleting 1-SI-99-1O-B following the stabil ization of the plant and the electrical power
restoration?
A.
B.
C
D.
RTA Position
Closed
Closed
Tripped
Tripped
1-SI-99-10-B Status
The surveillance is required to be completed.
The surveillan ce is NOT required to be completed with the plant in
this Mode.
The survei llance is required to be completed .
The surveillance is NOT required to be completed with the plant in
this Mode.
DISTRACTOR ANALYSIS
a.
Incorrect. RTA will be tripped by the UV coil. Plausible, if candidate does not correctly apply the
function of the reac tor trip breaker UV coil trip function but does realize the need complete the SI which
is requir ed even with til e plan t in MODE 3.
b.
Incorrect. RTA will be tripped by the UV coil and the SI is still required with the plant in Mode 3.
Plausible, if candida te does not correctly apply function of the reactor trip breaker UV coil trip function
and determines the SI would not be required to be completed becau se the plant is now in Mode 3.
c.
CORRECT. RTB and BYA will have no indicating lights lit because the breakers are racked out of the
cubicles, BYB will be open and have green light lit; RTA will have no light lit due to a loss of 125v DC
15 of 29
Watts Bar 2008 NRC Initial Licens e Exam
WRITTEN QUESTION DATA SHEET
control power , however, the UV coil on RTA will cause the breaker to open . The SI is requir ed with the
plant in MOD E 3.
d.
Incorrect. The UV coil on RTA will cause the breaker to open , however will be isolated by closin g 1-
FCV-62-90 and 91 in accord ance with the AOI. Plausible if the candidate correctly determin es
correctly the RTA position is correct but incorrectly determines SI wou ld not be required to be
completed because the plant is now in Mode 3.
16 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
Question Number:
4-3'gf)
High
NEW
Cognitive Level:
Source:
K/A : 061 A2.09
Auxiliary/Emergency Feedwater
Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on
those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or
operations: Total loss of feedwater.
Tier:
2
RO Imp:
RO Exam:
Group:
1
Yes
Applicable 1OCFR55 Section:
41.5. 43.5
Learning Objective: 3-0T-SYS003B Auxiliary Feedwater System, Objective 26. Identify the steps to gain
local control of the Turbine-Driven Auxiliary Feedwater pump and SG levels.
References:
FRH.1, rev. 17.
Question:
Given the following:
Following a reactor trip the plant experienced a total loss of feedwater.
The condition caused the crew to initiate RCS bleed and feed.
Subsequently the TDAFW pump was restored to service and the crew is ready to establish AFW
flow to the selected steam generator.
Other plant conditions include:
o
Selected SG Wide Range level is 4%.
o
RCS loop hot leg temperature at 558°F.
(I
Core exit thermocouple temperatures are RISING .
In accordance with FR-H.1, Loss of Secondary Heat Sink, feedwater flow will be re-established to the
selected SG at '"
A.
the minimum detectable flow gpm.
B.
less than 100 gpm until WR level >15%.
C.
a rate that causes wide range level to rise and RCS hot leg to drop.
D.
a maximum rate.
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible, since the feedwa ter flow would be correct for the conditions where a steam
generator was being fed to prevent dryout.
b.
Incorrect. Plausible, since the feedwater flow would be correct for the condition which meets the "hot,
dry" conditions and core exit thermocouples are not rising.
c.
Incorrect. Plausibl e, since the feedwater flow would be correct for the condition where Core Exit
thermocouples were stable or lowering.
d.
CORRECT. The maximum flow rate is only used if the steam generator meets the wide range level is
<15% and RCS temperature is >550°Fand core exit thermocouples are rising.
17 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DATA SHEET
Question Number : .;4[J{
KIA: 064 G2.2.37
Ability to determine operability and/or availability of safety related equipment.
High
NEW
Cognitive Level:
Source:
Yes
Tie r:
2
RO Imp:
RO Exam:
Group:
1
SRO Imp: 4.6
SRO Exam:
Applicable 10CFR55 Section:
41.7,43.5
Learning Objective: 3-0T-SYS082A , Obj. 13: State the Technical Specification requirements associated
with AC Sources & DG Support Systems.
References:
LCO 3.8.4, DC Sources Operating , including Basis.
LCO 3.8.6, Table 3.8 .6-1, inc!. Basis.
Question:
Given the followin g conditions:
The plant is shutdown In Mode 4 with all safety related equipment operable.
During the performance of surve illance O-SI-215- 21-A, DIESEL GENERATOR 1A-A BATT ERY
QUARTER LY INSPECTION, it is reported that the Float Voltage values for three (3) connected
cells are as follows :
Cell 17 = 2.09 v
Cell 34 = 2.06 v
Cell 39 = 2.12 v
ALL other connected cells have Float Voltage values greater than 2.13 v.
ALL other parameters measured during the above surveillance are normal.
Which ONE of the following describes the status of DIG 1A-A battery AND of the DIG 1A-A?
REFERENCE PROVIDED
Status of Battery
Status of D/G 1A-A
A.
Battery is degraded but it can
be consi dered operable for 31 days .
NO Tech Spec or tracking only entry required for DG.
B.
Battery is degraded but it can
be considered operable for 31 days.
Tech Spec tracking only entry required for DG .
C.
Battery is inoperable.
DG is inoperable and Tech Spec tracking only
entry is required.
D.
Battery is inoperable .
DG is inoperable and Tech Spec entry
is required.
18 of 29
DISTRACTOR ANALYSIS
a.
Incorre ct. Plausible, since if cell voltages had been below Categ ory B limits, but above Category C
allowable limits, battery could be considered operable for 31 days.
b.
Incorrect.
Plausible, since if cell voltages had been below Category B limits, but above Category C
allowable
Ii~ i ts , battery could be considered operable for 31 days.
I ("(C\\:12J;*~c
~gGU: Per LCO 3.8.6, since Cell 34 voltage is less than 2.07v, it does not satisfy Category C
requirements, which is Condition B of LCO 3.8.6. The battery is therefore inoperable. It is true the DIG
is also inoperable , however, in Mode 4, only one train of DIGs is required. The one inoperable DIG
status woul d be tracked per Tech Spec tracking only.
/"
.\\
_~
Incorrect. Plausible, since first part of distractor is true; however, candidate fails to recognize and
\\..Crlee
\\\\ .~
apply knowledge that in the current plant mode , the one inoperab le DIG only requires Tech Spec
~
tracking, and NOT
} t i {prOVide only the pages of LCO 3.8.6.
~V\\ Pe6- PO~t ~aJ,'.' QC'I\\r.eA.
Di~ 5>i iK£G4fGtL b l\\t
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
CJ"'
Question Number:
.:t.a-
fLJ
KIA: 076 A2.01
Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on
those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or
operations: Loss of SWS.
Tier:
2
Group:
1
RO Imp:
RO Exam:
SRO Imp: 3.7
SRO Exam:
Yes
Cognitive Level:
Source:
High
NEW
Learning Objective: 3-0 T-AOI1300, Objective 4: Identify the genera l location of a rupture given a Hi
strainer \\P and low flow in the same hdr. 3-0T-T11240, Objective 4: Describe the four "risk" levels used in
the Equipment to Plant Risk Matrix and the significance of each level.
References:
TI-124, Equipment To Plant Risk Matrix. Rev 14.
AOI-13, Loss of Essential Raw Cooling Water, Rev 35.
Question:
Given the following conditions:
Unit 1 is operating at 100% power.
It is determined that the strainer must be isolated.
Which ONE of the following identifies the action directed by AOI-1 3, Loss of Essential Raw Cooling Water,
to mitigate the isolation of the 2B strainer and the PSA risk status the plant will be in due to the strainer
isolation?
AOI-13 actions
A.
Realign cooling water to the A train diesel generators.
B.
Crosstie the 2B header with the 1A header until strainer is repaired.
C.
Realign cooling water to the A train diesel generators.
D.
Crosstie the 2B header with the 1A header until strainer is repaired.
PSA risk
Orange
Orange
Red
Red
D1STRACTOR ANALYSIS
a.
Incorrect. The PSA risk is Red (high) not Orange for the isolation of the ERCW header. Plausible,
since the action listed involves realignment of cooling water. The cooling water supply for the A train
DGs will not be affected as a result of isolating the strainer and no DG cooling water alignmen t would
be required.
b.
Incorrect. The PSA risk is Red (high) not Orange for the isolation of the ERCW header . Plausible,
since crosstie of the headers would be performed in accordance with AOI-1 3.
c.
Incorrect. Plausible, since PSA risk is correct, and the action listed involves realignment of cooling
water. The cooling water supply for the A train DGs will not be affected as a result of isolating the
strainer and no DG cooling water alignment would be required.
d.
CORRECT. The 2B ERCW header will be crosstied with the 1A ERCW header during performance of
AOI-13 and the PSA risk would be RED as identified in TI-124.
19 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
Question Number: ~l
KIA: 011 G2.1.32
Pressurizer Level Control
Ability to explain and appl y system limits and precautions.
High
NEW
Cognitive Level:
Source:
Tier :
2
RO Imp:
RO Exam :
Group:
2
SRO Imp : 4.0
SRO Exam :
Yes
Ap pl icable 10CFR55 Section:
41.10,43.2
Learn ing Objective: 3-0 T-T-S0304, Objective 4: Given plant conditions and parameters correctl y
determine the applicable Limiting Conditions for Operatio ns or Technical Requ irements for the various
components of the RCS.
References:
Question:
Given the following plant conditions:
During a startup, the plant is in Mode 4.
GO-1, Appendix C, Mode 4-to-Mode 3 Review and Approval, has been comple ted up to the last
step, Operations Superintendent Hold Point, for granting appro val to enter Mode 3.
During scaffolding removal activities, a worker contacts the air line to 1-FCV-62-93, Charging Flow
Control Valve, resulting in pulling the air line loose from the valve operator, such that the air system
remains intact, due to crimping of the line on the air supply side of the break.
Which procedures will the Unit Supervisor use to respond to this event and the basis for taking quick action
to limit the effects?
A.
The Unit Supervisor will use AOI-10, Loss of Control Air to direct actions to prevent Pressurizer level
from exceeding 92% in order to maintain the presence of a steam bubble in the pressurizer.
B.
The Unit Supervisor will use AOI-10, Loss of Control Air, and direct actions to prevent the level
decrease to less than 17% to ensure subcooli ng margin can be maintai ned.
C.
The Unit Supervisor will use SOI-62.01, CVCS-Charging and Letdown and direct actions to prevent
the level decrease to less than 17% to ensure subcooling margin can be maintained.
D.
The Unit Supervisor will use SOI-62.01, CVCS-Ch arging and Letdown, to direct actions to prevent
Pressurizer level from exceeding 92% in order to maintain the presence of a steam bubble in the
pressurizer.
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible, if a failed open charging flow control valve condition is not mitigated PZR level
would increase to the point where the desired steam bubble would not be maintained.
b.
Incorrect. Plausible since it contains the correct proced ure. Candidate incorrectly concludes that when
1-FCV-62-93, Charging Flow Control Valve, loses air it fails CLOSED. Using this incorrect conclusion,
it is logical to see why a candidate would then conclude that Pressurizer level will lower. Candidate
correctly recalls that there is a lower limit for Pressurizer level of 17%, and applies it to arrive at the
concern for losing all banks of Pressurizer heaters resulting in a potentia l loss of desired subcooling.
c.
Incorrect. Plausible since it contains a procedure with a title similar to the conditions given in the stem.
Candidate incorrectly concludes that when 1-FCV-6 2-93, Charging Flow Control Valve loses air it fails
CLOSED. Using this incorrect conclusion, it is logical to see why a candidate would then conclude that
Pressurizer level will lower. Candidate correctly recalls that there is a lower limit for Pressurizer level
of 17%, and applies it to arrive at the concern for losing all banks of Pressurizer heaters resulting in a
potential loss of desired subcooling.
d.
CORR ECT. When 1-FCV-62-93 loses operating air, it fails OPEN, resulting in significantly more
charging flow. If this is allowed to continue, Pressurizer level will exceed 92%, which could result PZR
level increasing to the point where the desired steam bubble would not be maintained.. SOI-62 .01
contains detailed action steps for isolating and bypassing 1-FCV-62-93, which will restore charging
flow control to the control room, even with the normal flow control valve failed open .
20 of 29
Watts Bar 2008 NRC Initial Licens e Exam
WRITTEN QUESTION DATA SHEET
Question Number: ~qL.
KIA : 075 A2.03
Circulating Water
Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system ;
and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of
those malfunctions or operations : Safety features and relationship between condenser vacuum, turbine trip,
and steam dump .
Tier:
2
Group:
2
RO Imp:
RO Exam:
SRO Imp: 2.7
SRO Exam:
Yes
Cognitive Level:
Source:
Low
NEW
Applicable 10CFR55 Section:
41.5, 43.5
Learning Objectiv e: 3-0 T-SYS027A, Objective 11: Describe the main conden ser. 3-0T-A0 11700,
Objective 1: Demonst rate knowledge of the Purpose/Goal of AOI, Rev. 9.
Referen ces :
AOI-1 1, Loss of Condenser Vacuum, Rev. 27.
Question:
Given the following:
Unit 1 was operating in Mode 1 at 14% reactor power.
A loss of condenser vacuum occurs resulting in an automatic turbine trip.
The operating crew stabilizes the plant.
Which ONE of the following identifi es both how the condenser circulating water box LlTs will be affected and
the notifications required due to the turbine trip in accordan ce with SPP-3.5, Regulatory Reporting
Requirement s?
Water Box LlT
Notifications required
A.
Rises
Internal TVA notifications only
B.
Rises
NRC and Internal TVA notifications
C.
Lowers
Internal TVA notifications only
D.
Lowers
NRC and Internal TVA notifications
DISTRACTOR ANALYSIS
a.
Incorrect. Steam dumps are prevented from opening due to loss of vacuum . Therefore, no steam is
entering the condenser, and water box LlT will drop. The notifications required by SPP-3.5 are internal
only. Plausible if candida te fails to recall the condenser vacuum interlock with steam dump operation
but correctly identifies the required notifications .
b.
Incorrect. Steam dumps are prevented from opening due to loss of vacuum. Therefore, no steam is
entering the condenser and water box LlT will drop. The notification required by SPP- 3.5 are internal
only, no notification to the NRC is required . Plausible if candidate fails to recall the condenser vacuum
interlock with steam dump opera tion and incorrectly identifies the required notifications.
c.
CORR ECT. With a loss of condenser vacuum, steam dump operation is blocked. Therefore, no
steam is entering the condenser, resulting in a lower LlT across the condenser waterboxes . SPP-3.5
requires internal TVA notification be made due to the turbine trip.
d.
Incorrect. No steam is entering the condenser, resulting in a lower LlT across the condenser
waterboxes. The notification required by SPP-3.5 are internal only, no notification to the NRC is
required . Plausible because the waterbo x LlT response is correct and candidate may incorrectly
identify the required notifications .
21 of 29
Question Number: -+8"1>
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DATA SHEET
K/A : 086 G2.4.9
Fire Protection
Knowledge of low power/shutdown implications in accident (e.q., loss of coolant accident or loss of residual
heat removal) mitigation strategies .
Tier:
2
Group:
2
RO Imp:
RO Exam:
SRO Imp: 4.2
SRO Exam :
Yes
Cognitive Level:
Source:
High
NEW
Applicable 10CFR55 Section:
45.5
Learning Objective: 3-0T-AOI-700, Rev. 10: Explain the 2 modes of maintaining the core cool & stable
during flood mode operation.
References:
AOI-7.01, Rev. 16.
Question:
Given the following:
At0100
At 0200
At 1100
At 1200
Unit 1 entered Mode 3 during shutdown for a refueling outage.
AOI-7.01, Maximum Probable Flood, was implem ented due to extremely heavy
rainfall in the upstream watershed.
The plant is in Mode 4 on RHR with cooldown in progress with the Re S pressure 320
psig and temperature 220
0 F.
River System Operations (RSO) confirms the flood level at the plant is predicted to
crest at EI. 730' within the next 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 identifies both the Flood Stage Preparation level(s) that is/are required to be
completed, and how the cooling of the core will be maintained in accordance with AOI-7 after the
preparations are complete?
A.
Only the procedure for Stage 1 Preparations is required to be completed.
Core cooling will be maintained by the steam generators with water being supplied by high pressure fire
protection with the RHR system removed from service.
B.
Only the procedure for Stage 1 Preparations is required to be completed.
Core cooling will be maintained by the Spent Fuel Pool cooling system crosstied with the RHR system.
C.
Both Stage 1 and Stage II Preparations proced ures are required to be completed.
Core cooling will be maintained by the steam generators with water being supplied by high pressure fire
protection with the RHR system removed from service.
D.
Both Stage 1 and Stage II Preparation s procedures are required to be completed.
Core cooling will be maintained by the Spent Fuel Pool cooling system crosstied with the RHR system.
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible, since method of maintaining core cooling is correct.
b.
Incorrect. Plausible, since Stage 1 preparation is part of the correct action. AOI-7.01, Attachment 2
Step 15 refers the operator to Appendix E, which will cross connect the RHR and SFPC systems if in
open mode cooling configuration.
c.
CORRECT. Per given reference.
d.
Incorrect. Plausible, since the first part is correct. AOI-7.01, Attachment 2, Step 15 refers the operator
to Appendi x E, which will cross connect the RHR and SFPC systems if in open mode cooling
configuration.
22 of 29
Watt s Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
Question Number: ~{;71
KIA : G2.1.23
Ability to perform specific system and integrated plant procedures during all modes of plant operation.
Tier:
3
Group:
RO Imp:
RO Exam:
SRO Imp: 4.4
SRO Exam:
Yes
Cognitive Level:
Source:
Low
NEW
Applicable 10CFR55 Section:
41.10, 43.5
Learning Objective: 3-0 T-SYS079A, Rev. 6, Objective 4: Identify the maximum quantity of fuel that shall
be out of approved storage locations during fuel handling operations.
References:
FHI-7, Fuel Handling and Movement, Rev 0034.
Question:
Which ONE of the following satisfies the requirement of FHI-7 for the maximum number of fuel assemb lies
allowed outside of approved storage?
A.
Two unirradiated assemblies within the fuel-handling area.
B.
Two irradiated assemblies within the spent fuel storage pool bounda ry.
C.
One assembly in the transfer cart, two assemblies in the RCCA fixture and one assembly in the
refueling machine mast over its proper location in the reactor vessel.
D.
One assembly in the transfer cart, two assem blies in the RCCA fixture and one assemb ly in the
refueling machine mast over the reactor side upender.
DISTRACTOR ANALYSIS
a.
Incorrect. Only one unirradiated assembly can be outside of approved storage within the fuel -handli ng
area.
b.
Incorrect. Only one irradiated assembly can be outside of approved storage within the spent fuel
storage pool boundary.
c.
CORRECT. Per FHI-7, the assembly in the mast and over its proper location in the reactor vessel is
allowed beyond the three allowed within the refueling canal; i.e., this answer specifies that three
assemblies are within the refueling canal, two can be in the RCCA change fixture and one in the
transfer cart.
d.
Incorrect. The assembly in the refueling mast over the upender must be includ ed in the 3 assemblies
allowed within the refuel canal which would result in 4 assemblies outside approved storage in the
refuel canal, thus exceeding the three allowed in the refueling canal area.
23 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
Question Number: ~4~
KIA: G2.1.34
Knowledge of primary and secondary plant chemistry limits.
Tier:
3
Group:
RO Imp:
RO Exam:
SRO Imp: 3.5
SRO Exam:
Yes
Cognitive Level:
Source:
Low
NEW
Applicable 1OCFR55 Section:
41.10, 43.5
Learning Objective: 3-0 T-T-S0304, Objective 4: Given plant conditions and parameters correctly
determine the applicable Limiting Conditions for Operations or Technical Requi remen ts for the various
components of the RCS.
References:
CM-3.01, "System Chemistry Specifications", Appendix A, pp. 1, 2, 25, and 26; rev 72
Tech Requirement 3.4.4, Chemistry.
Question:
W iltl the plant at full power, the Chemistry Lab has just informed the Unit Supervisor that the RCS chloride
level is 1650 ppb and that the SG chloride level is 200 ppb. The source of any impuri ty ingress has NOT yet
been identified.
Based on the reported values, (1) what is the MOST restrictive Action Level that must be entered and (2) the
impact on plant operations caused by the Action Level entry?
ill
A.
Action Level 3 for RCS chloride level
B.
Actio n Level 3 for SG chloride level
C.
Action Level 2 for RCS Chloride level
D.
Action Level 2 for SG Chloride level
Initiate actions to take the reactor sub-critical as
quickly as practicable in a controlled manner and
reduce RCS temperature to < 250" F.
Initiate actions to take the reactor sub-critical as
quickly as practicable in a controll ed manner and
reduce RCS temperature to < 250" F.
Restore parameter to below Action Level 1 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
or reduce reactor power to less than 5%.
Restore parameter to below Action Level 1 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
or reduce reactor power to less than 5%.
DISTRACTOR ANALYSIS
a
CORRECT. Action Level 3, which is the most restrictive (requiring plant shutdown ), applies due to
RCS chloride exceeding 1500 ppb.
b.
Incorrect. Candidate incorrectly believes that Action Level 3 applies for SG chloride levels, when it is
actually Action Level 2 (SG chloride exceeds 50 ppb, but does not exceed 250 ppb. Plausible because
the impact on plant operations of a chemistry parameter being aut-of-specification is correct.
c.
Incorrect. Candidate fails to recognize Action Level 3 conditions for RCS chloride. However, the
distractor has plausibility because there is similar impact on plant operations if there were Action Level
2 conditions for RCS chloride.
d.
Incorrect. Plausible, since the Action Level for the given SG chloride value is correct. Further
plausibility is added since there is a time value given to restore the parameter. However, it differs from
the correct time (within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />) by a factor of 4. The requirement to be at less than 5% power is
incorre ct.
24 of 29
Watts Bar 2008 NRC Init ial License Exam
WRITTEN QUESTION DA TA SHEET
Question Number: 7f Cffv
KIA : G2.2.40
Ability to apply Technical Specifications for a system.
Tier:
3
Group:
RO Imp:
RO Exam:
SRO Imp: 4.7
SRO Exam:
Yes
Cognitive Level :
Source:
High
NEW
Applicable 10CFR55 Section: 41 .10/43 .2/43 .5/45 .3
Learning Objective: 3-0T-TS0307, Rev 3, Objecti ve 5: Given plant conditions and param eters, determine
applicable Action Conditions, Required Actions, and Completion Times associated with different Plant
Systems . Objective 2: Determine the Bases for each specification, as applicable, to Plant Systems.
References:
TS 3.7.5 and Bases.
Question:
Given the following
Unit 1 at 100% power.
Diesel Generator (DG) 2A-A is to be rem oved from service for a mainten ance outage with a
planned out of service of 102 hours0.00118 days <br />0.0283 hours <br />1.686508e-4 weeks <br />3.8811e-5 months <br />.
Which ONE of the following identifies Unit 1 equipment that is listed in Tech Spec 3.8.1, AC Sources
Operating, Bases Contingency Actions, as equipment that is to remain in service concurrently during the DG
2A-A outage maintenance to be in compliance with the Tech Spec?
A.
Reactor trip breaker A (RTA).
B.
TDAFW pump.
C.
RHR Pump 18-B.
D.
Any S/G AFW level control valve.
DISTRACTOR ANALYSIS
a.
CORRECT. Tech Spec 3.8.1 Bases Table 3.8.1-2, TS Action or Surveillance Table (SR) Contingency
Actions, Item 4 states "Do not rem ove reactor trip breakers from service concurrently during planned
DG outage maintenance".
b.
Incorrect. The Bases Table item 5 states "Do not rem ove the turbine-driven auxiliary feedwater (AFW)
pump from service concurrently with a Unit 1 DG outage". Plausible if the candidate misapplies the
item to the Unit 2 DG outage.
c.
Incorrect. The Bases Table item 7 states "Do not rem ove the opposite train residual heat remova l
(RHR) pump from service concurrently with a Unit 1 DG outage ". Plausibl e if the candidate misapplies
the item to the Unit 2 DG outage.
d.
Incorrect. The Bases Table item 6 states "Do not remove the auxiliary feedwater level control valves to
the steam generators from service concurrently with a Unit 1 DG outage". Plausible if the candidate
misapplie s the item to the Unit 2 DG outage.
25 of 29
Question Number: ..2'2'1l
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DATA SHEET
KIA: G2.3.6
Ability to approve release permits.
Tier:
3
RO Imp:
RO Exam:
Group:
SRO Imp: 3.8
SRO Exam:
Yes
Cognitive Lev el:
Source:
Low
NEW
Applicable 10CFR55 Section:
41.13 I 43.4 145.10
Learning Objective: 3-0 T-SYS077B, Objec tive 10: Describe the general procedure to make a gaseous
release.
Referen ces :
SOI-77 .02, Waste Gas Disposal System, Rev 0034.
Question:
Which ONE of the following identifies BOTH the minimum decay time requi red to allow the contents of a Gas
Decay tank to decay prior to release and who can waive the minimum time in accordance with SOI-77.02,
Waste Gas Disposal system ?
Decay Time Required
Who Can Waive
A.
60 days
Chemistry Duty Manager
B.
60 days
Radiation Protection Manage r
C.
8 days
Chemistry Duty Manager
D.
8 days
Radiation Protection Manage r
DISTRACTOR ANALYSIS
a.
CORRECT. The procedur e requires a 60 day decay time and does provide for waiving of the time by
the Chemi stry Duty Manager.
b.
Incorrect. The decay time required is 60 days but the Radiation Protect ion Mana ger is not the position
that can waive the requirement if earlier release is required. Plausible because the required time is for
radioactive deca y which could be addres sed by RadCon.
c
Incorrect. The decay time required is 60 days, not 8 days, but the waiving of the requirement by the
Chemistry Duty Manager is allowed by the procedure. Plausible because 8 days is identified in the
ODCM for Gaseous Effluents as being the half life of certain radionuclides that set dose rate limits at
and beyond the Unrestricted Area Bound ary.
d.
Incorrect. The deca y time required is 60 days, not 8 days and the waiving of the requirem ent by the
Radiation Prote ction Manager is not allowed by the procedure. Plausible beca use 8 days is identified
in the ODCM for Gaseous Efflu ents as being the half life of certain radionuclides that set dose rate
limits at and beyond the Unre stricted Area Boundary and because the required time is for radioactive
decay which could be addressed by RadCon.
26 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
Question Number: ~1g
K/A : G2.3.15
Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey
instruments , personnel monitoring equipment, etc.
Tier:
Group:
3
RO Imp:
SRO Imp: 3.1
RO Exam :
SRO Exam:
Yes
Cognitive Level:
Source:
High
NEW
Applicable 10CFR55 Section:
41.12,43.4
Learning Objective: 3-0 T-SYS090A, Rev. 13, Objective 8: Regarding Technica l Specifications and
Technical Requirements for this system: Explain the Limiting Conditions for Operation, Applicability, and
Bases.
References:
SOI-90.04, Section 3.0.B, and Section 5.1.(3], Rev. 6.
LCO 3.3.8, Table 3.3.8-1, including Basis; Drawing 45N600-30-4.
Question:
Given the following plant conditions:
Unit 1 is in Mode 4 with preparations being made for fuel moves.
As part of performing the Channel Operational Test (COT) for 0-RM-90-102, Spent Fuel Pool Pit
Area Monitor, a source check is to be performed.
What is the effect of performing the source check portion of this test, and what is the SRO's responsibility for
Tech Spec/LCO Tracking Sheet entry?
A.
This will result in an automatic actuation of Train A of ABGTS . The SRO will make an entry on the
LCO Tracking Sheet that Train A of ABGTS is inoperable.
B.
RM-90-102 outpu t will be blocked during the source check, the SRO will make an entry on the LCO
Tracking Sheet that Train A of ABGTS is inoperable.
C.
RM-90-102 output will be blocked during the source check, the SRO will make an entry on the LCO
Tracking Sheet that RM-90-102 is inoperable.
D.
This will result in an automatic actuation of Train A of ABGTS . The SRO will make an entry on the
LCO Tracking Sheet that RM-90-102 is inoperable.
DISTRACTOR ANALYSIS
a.
Incorrect. Both parts are incorrect, but plausible, because of the close link between ABGTS (Auxiliary
Building Gas Treatment System ) operation and RM-90-102 (causes actuation of ABGTS). However,
candidate incorrectly believes that the COT causes an automatic actuation of ABGTS.
b.
Incorrect. Candidate correctly understands that the COT requires blocking of the outpu t of RM-90-102,
which prevents it from actuating ABGTS, and concludes therefore, that the associated train of ABGTS
is inoperable. This adds to the plausibility of this distractor, but it is incorrect, since a different LCO
(LCO 3.7.12) governs the operability of the ABGTS train itself.
c.
CORRE CT. This test requires blocking of the output of RM-90-102, which prevents it from auto
actuating the associated train of ABGTS . This is done on purpose, to prevent inadvertent/undesired
actuation of ABGTS . Further , this blocking renders the RM-90-102 inoperable.
d.
Incorrect. Candidate correctly concludes that this test renders RM-90-10 2 inoperable, but for the
wrong reason. It is plausible, since there are other components in the plant, that if running, or
actuated, are considered inoperable. However, candidate incorrectly believes that this test causes an
auto actuation of the associated train of ABGTS.
27 of 29
Question Number:
~<f1
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
K/A: G2.4.23
Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.
Tier: 3
Group:
RO Imp:
SRO Imp:4.4
RO Exam:
SRO Exam:
Yes
Cognitive Level:
Low
Source:
Modified Bank
Applicable 10CFR55 Section:
41.10, 43.5, 45.13
Learning Objective: 3-0T-EOP0300 Objective 5: Given a set of plant conditions, use E-3, ES-3.1, ES-3.2,
and ES-3.3 to correctly diagnose and implement: Action Steps, RNOs, Foldout Pages, Notes and Cautions;
Objective 6. Explain the basis for cooling the RCS to a target incore temp prior to depressuization of the
RCS.
References:
E-3, Steam Generator Tube Rupture, Rev. 22.
Question:
Given the following:
Unit 1 experiences a Safety Injection due to a steam generator tube rupture on SG #1.
All Reactor Coolant Pumps were removed from service due to loss of support systems.
RCS cooldown at maximum rate to target incore temperature is in progress.
The STA reports that a RED path for FR-P.1 , Pressurized Thermal Shock, exists on RCS Loop 1
on the FR-O, Status Trees.
Which ONE of the following identifies when the transition to FR-P.1should be made?
A.
Immediately transition to FR-P.1 from E-3 because FR-P.1 provides actions to limit cooldown and
repressurization of the RCS.
B.
Remain in E-3 until the cooldown is complete and then transition to FR-P.1 only if the RED path still
exists because the cooldown is needed to allow depressurization of the RCS.
C.
Remain in E-3 until the cooldown is complet e and then transition to FR-P.1 even if the RED path no
longer exists because the cooldow n is needed to allow depressurization of the RCS.
D.
Remain in E-3 until the safety injection is terminated, then transition to FR-P.1 only if the RED path still
exists because FR-P.1 provides actions to limit cooldown and repressurization of the RCS.
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible, since the normal rules of usage would require implementation of FR-P.1 as soon
as the condition was confirmed to exist.
b.
Incorrect. Plausible since a stagnant loop condition will exist in the loop associated with the ruptured
SG with the RCPs off. CAUTION prior to the cooldown step in E-3 states that the anticipated red or
orange path on FR-P.1 does not require implem entation until after SI termination has been
accomplished.
c.
Incorrect. Plausible since a stagnant loop condition will exist in the loop associated with the ruptured
SG with the RCPs off. CAUTION prior to the cooldown step in E-3 states that the anticipated red or
orange path on FR-P.1 does not require implementation until after SI termination has been
accomplished.
d.
CORRECT. A stagnant loop condition will exist in the loop associated with the ruptured SG with the
RCPs off. CAUTION prior to the cooldown step in E-3 states that the anticipated red or orange path
on FR-P.1 does not require implementation until after SI termination has been accomplished. E-3
Note prior to Step 43 directs the operator to begin reevaluating for PTS/Cold Overpressure conditions.
An evaluation of PTS is delayed to this point since the active SI would contribute to the erroneous
red/orange path.
28 of 29
Watts Bar 2008 NRC Initial License Exam
WRITTEN QUESTION DA TA SHEET
High
NEW
Cognitive Level :
Source:
Quest ion Number: ~ iDO
KIA: G2.4.29
Knowledge of the emergency plan.
Tier:
3
RO Imp:
RO Exam :
Group:
SRO Imp: 4.4
SRO Exam :
Yes
Applicable 10CFR55 Section:
43.5, 45.11
Learning Objective: 3-0T-PCD-048C, Objective 7: Identify Opera tion's responsibilities for the following
emergency response positions: Site Emergency Director (who is initially the SM).
References:
EPIP-1, 3.3.7, rev. 29, SPP-3.5, Appendi x A, 3.1, Rev. 19.
Question:
Given the following conditions :
The plant is at full power.
A report was received of a tornado being sighted over the Watts Bar Training Center, moving in a
northwest direction. The tornado continue d to move across Highway 68 and then dissipa ted
without touching down onsite.
The MCR crew has just entered AOI-8, Tornado Watch or Warning.
Confirm ation was received that no visible damage had been received to any structures or
equipment on site.
The Shift Manager evaluates the Radiological Emergency Plan (REP) and determines the
conditions for an NOUE were initially met but are now fully resolved.
Which ONE of the following identifies the ODS and NRC notification requirements in accordance with the
REP?
ODS notification
NRC Notification required wi th in :
A.
Report but not declare
15 minutes.
the event.
B.
Report but not declare
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
the event.
C.
Declare and terminate the
15 minutes .
event at the same clock time.
D.
Declare and terminate the
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
event at the same clock time.
DISTRACTOR ANALYSIS
a.
Incorrect. Plausible , since a declaration is not made, and a report is required, but candid ate
incorrectly believes that it needs to be made within fifteen minutes.
b.
CORRECT. Per EPIP-1, for events that are totally resolved prior to declaration, no declaration shall
be made: however, a report to the NRC within one hour is required.
c.
Incorrect. Plausible, since the NRC must be notified, and since candidate may fail to recognize
conditions have totally resolved and therefore no declara tion is to be made .
d.
Incorrect. Plausible, since the notification to the NRC time is correc t. Candidate fails to recogn ize
that these conditions have totally resolved prior to declara tion, and therefore that NO declaration
should be made.
29 of 29