ML042450258
| ML042450258 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 04/05/2004 |
| From: | Ernstes M Operator Licensing and Human Performance Branch |
| To: | Scalice J Tennessee Valley Authority |
| References | |
| 50-390/04-301 50-390/04-301 | |
| Download: ML042450258 (43) | |
See also: IR 05000390/2004301
Text
Draft Submittal
WATTS BAR JULY 2004 EXAM
5 0 -39012
0 04-3 0 1
JULY 23, & JULY 26-30,2004
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8;
WATTS BAR NUCLEAR PLANT
SENIOR REACTOR
OPE RATOR
NRC
ITTEN EXA
JULY 23,2004
REFERENC COPY
007 C 2.1.34 001
Given the following plant conditions:
-
-
-
Plant trip occurred from I
OQ% power.
The plant is stable at no-load temperatures and pressures.
Chemistry has sampled the primary and secondary plants.
--.
Which ONE of the following sets of chemistry parameters lists the minimum
values that could result in exceeding offsite dose limits in the event a design
'::,:~sbi:a~~id~nti:occurred
following the plant trip?
Secondary
RCS Dose
SDecific Activity
Equivalent 1-132
A.
.265pcilgm
.'I
Opcilgm
B.
2.65pcilgm
7 .OFci/gm
2.65pcVg m
265pciIgm
The correct answer is D.
A. Incorrect -All parameters are off by one decimal place. the examinee may
have a misconception or confuse values with transient values used in the
Technical Requirement for RCS chemistry parameters.
B Correct - These are the correct values listed in Tech Specs far the given
.- 0
parameters.
C. Incorrect - All parameters are off by one decimal place and reversed.
D. Incorrect - parameters are reversed
REFERENCES:
Technical Specifications 3.4.16 and 3.7.14
1QCFR55 41.7
Reactor Trip: Ability to maintain primary and secondary plant chemistry within
allowable limits
SRO - 76
Reference:
TIS 3.04
KIA value:
2.9
Level:
1
TierGrp:
3!3
KIA Kumber: ti 2.1.34
Last Used:
Source:
NEW
SRO Only:
YES
015 c; 2.4.45 001
'Given the following plant conditions:
- A large LOCA has occurred.
-
The operating crew has implemented FR-C.1, "Response to Inadequate
Core Cooling".
-
Core exit thermocouples are -1225OF.
-
RCP # I and #3 are started.
- Annunciator 964,
RCP VIBRATION MON BETBECTED, goes into alarm
and local verification indicates #I
RCP vibration is at 21 mils.
-
Annunciator 99-C, RCP LWR BEARING TEMP HI, goes into alarm and
the operator verifies that #3 RCP lower bearing temperature is 238OF.
Which ONE of the following would be the correct response to this condition?
A. Stop both running RCPs
BY Leave both RCPs running.
C. Start any available RCP, then stop #I RCP.
D. Start any available RCP, then stop both running RCPs.
The correct answer is B.
A. Incorrect no pumps should be stopped after being started in FR-C.1 in
response in rising incore temperatures > 1200OF. Examinee may think
normal RCP trips should be considered.
6. Correct- no pumps should be stopped after being started in FR-C.1 in
..
response in rising incore temperatures > 1200W.
C. Incorrect - no pumps should be stopped after being started in FK-(2.1 in
response in rising incore temperatures > 720OOF. Normal operating
conditions are not required to run RCPs under these conditions.
D. Bncorrect - no pumps should be stopped after being started in FR-C.1 in
response in rising incore temperatures > 1200OF. Normal operating
conditions are not required to run RCPs under these conditions.
REFERENCES:
FR-C.? 3-QT-FRCOOO
lOCFR55.41.7 t 45.5 to 45.8
015 RCP Malfunctions: G 2.4.45, ability to prioritize and interpret the significance of
each annunciator or alarm.
RO-N/A
SRO-77
Reference:
AOI-24
K!A value:
3.6
Level:
2
TicrErp:
111
W A Number: 015 Ci 2.4.45
Last l.!sed:
Source:
KEW
SRO Only:
YES
038 EA2.06 001
Unit 1 has experienced a Reactor Trip and Safety Injection due to a steam
generator tube rupture (SGTR) on SIG #I.
E-3, "Steam Generator Tube
Rupture", is in effect and the crew is about to commence a cooldown at
maximum rate.
The following conditions exist:
~L
- SIG #I
Level is 65% Narrow Range and RISING.
-
RCS Pave is 557 degrees F and STABLE.
-
PZR Pressure is 1950 psig.
-
CONDENSER VACUUM LOW annunciator is LIT.
-
Main Condenser Vacuum is I 5 in. Hg and STABLE.
- #I
and #2 Condenser Circulating Water pumps have tripped.
- #3 and #4 Condenser Circulating Water pumps are running.
action is necessary to commence cooldown per E-3, "Steam
b 8
uhEbP
$,bL~y*i Generator Tube Rupture":
A. Take Steam Dumps to the Steam Pressure Mode and manually open Steam
/
r
Dumps to commence cooldown.
,
ALB. Take Steam Dumps to the Steam Pressure Mode, take both Steam Dump
Control Selector Switches momentarily to Bypass Interlock, and then
manually open Steam Dumps to commence cooldown.
',!&!!
/'J
. ~.
CY Commence cooldown by manually opening the intact SlGs Atmospheric
'
D. Take both Steam Dump Control Seiector Switches momentarily to Bypass
Interlock, and then manually open Steam Bumps to commence cooldown.
Reliefs to 100% demand.
The correct answer is C
A. Incorrect - Since Tave is 557OF student may believe that taking SDs to steam
pressure mode will allow use. This would normally he the method of choice.
Condenser vacuum will not permit SD use.
8. lncorrect - Since Tave is 557OF student may believe that taking SDs to steam
pressure mode will allow use. This would be the method of choice.
C. Correct - The RNO in E-3 step 28 directs the use of PORV since the
condenser is not available to accept SBs.
D. Incorrect - Student may believe that the interlock for condenser vacuum may
be bypassed and select this option.
REFERENCES:
Lesson plan 3-OT-EQP0300
E-3 'SIG Tube Rupture'
10CFR.55.43.5/45.13
-
Ability to determine or interpret the following as they apply to a SGBR: Viable
alternatives for placing plant in a safe condition when condenser is not available.
SRQ-78
Reference:
3-0T-EOP0300
KIA value:
4.4
Level:
2
TieriGrp:
Ill
KjA Number: 038 EATOX
Last Used:
Source:
RANK
SRC) Only:
YES
058 AA2.01 001
Given the following plant conditions:
-
Plant is operating at 180%.
- Annunciator 18-A, 125 BC VITAL CHGRIBATT II ABNORMAL goes into
alarm.
-
1-8-57-92, VIT BATT BD II AMPS, reading -75 amps above zero.
-
Operating crew ensures the plant is stable and maintaining 100% power.
Which ONE of the following indicates the current status of the power supplies to
125V Vital Battery Board II?
A. 125VVital Battery Charger IB is operating NORMALLY; 125Vvital Battery II
operating NORMALLY.
25V Vital Battery Charger 11 is FAILED; 125V vital Battery I! operating
NQRMALLY.
6. 125V Vital Battery Charger II is operating NORMALLY; 125V vital Battery II
FAILED.
D. 125V Vital Battery Charger !I is FAILED; 125V vital Battery II FAILED.
The correct answer is B.
Correct - amps on this indicator normally indicate slightly less than zero as
the charger is inservice providing power to the battery board while charging
the battery. With the alarm in and indicator indicating above zero the
examinee should conclude that the charger is failed and the battery is
supplying power to the battery board.
B. Incorrect - amps on this indicator normally indicate slightly less than zero as
the charger is inservice providing power to the battery board while charging
the battery. With the alarm in and indicator indicating above zero the
examinee should conclude that the charger is failed and the battery is
supplying power to the battery board.
6. Incorrect - amps on this indicator normally indicate slightly less than zero as
the charger is inservice providing power to the battery board while charging
the battery. With the alarm in and indicator indicating above zero the
examinee should conclude that the charger is failed and the battery is
supplying power to the battery board. If the fuse had blown the amps would
be reading -zero.
B. Incorrect - amps on this indicator normally indicate slightly iess than zero as
the charger is inservice providing power to the battery board while charging
the battery. With the alarm in and indicator indicating above zero the
examinee should conclude that the charger is failed and the battery is
supplying power to the battery board. With both components failed amps
indicator would read zero and the unit would trip due to secondary transients.
REFERENCES:
801-21 .QI; 3-OT-SYS057P
i
1QCFR55 43.5 145.1 3
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 online.
SRO - 79
Reference:
3-O7'-SYSO57P
KIA value:
4.1
Level:
2
Tieri'Cirp:
313
KiA Number: 058 AA2.01
Last Uscd:
Source:
NEW
SRO ot>iy:
yrs
062 G 2.4.4 001
Given the following plant conditions:
-
-
-
Unit operating at 100% power.
ERCW system in normal alignment.
ERCW headers 1A and 28 indicating LOW flow.
The following annunciators are LIT:
- "ERCW Header A Supply Pressure Low".
-
"IPS Valve and Strainer Room A Sump Level Hi".
-
"ERCW PMP A-A Discharge Pressure Low".
- "ERCW PMP D-A Discharge Pressure LQW".
Which ONE of the following describes what has occurred in the ERCW
sy stein?
A. 'A header pumps have tripped.
3.
Both 'W header strainers need to be backwashed.
C? '1A header has ruptured upstream of the '18' strainer
B. '1A header has ruptured between the IPS and Auxiliary Bldg
The correct answer is 6.
A. Incorrect - All alarms except the sump high level would lead to this choice
-_
B. Incorrect - Strainer alarm would be lit for a clogged strainer. No sump
alarm with high pressure.
C. Correct - All alarms stated would be lit for this accident. Pressure indicator
is located just upstream of the strainer.
D. Incorrect - Strainer Dp alarm would be lit all other conditions match except
sump.
REFERENCES:
Lesson plan 3-OT-AOI 1300
AOI-13 LOSS of ERCW
IOCFR55.41.10/43.2/45.6
Ability to recognize abnormal indications for system operating parameters which
are entry-level conditions for emergency and abnormal operating procedures.
SRO-80
WA changed to G 2.4.4 from AA2.01 based on feedback from Ron Aiello
Reference:
3-0'r-AOI 1300
K;A value:
3.6
Level:
2
Tier/Grp:
l!i
W A Number. 062 G 2.4.4
Last Used:
Source:
BANI<
SRO Only:
YES
065 AA2.01 001
Given the following plant conditions:
-
-
-
-
Unit 1 has just entered Mode 5 with RHR Shutdown Cooling in Service
RCS temperature is 195T lowering slowly.
'D' C&SS air compressor is tagged out for maintenance.
The following alarms are received simultaneously in the MCR:
-
4245, C&SS AIR COMP SEQUENCER UNDERVOLTAGE
-
41-F, CONTROL AIR PRESS LO.
Which ONE of the following describes the effect of this malfunction, and the
correct operator actions required.
A!' RHR Heat Exchanger Flow Control Valves FCV-74-16 & 28 will fail OPEN
and cause RCS temperature to DROP. Manually load and unload running
C&SS air compressor,
B. RHR Heat Exchanger Flow Control Valves FCV-74-16 & 28 will fail CLOSED
and cause RCS temperature to RISE. Manually load and unload running
C&SS air compressor.
C. RHR Heat Exchanger Flow Control Valves FCV-74-26 8 2% will fail OPEN
and cause RCS temperature to DROP. Ensure all unloading solenoids
manual valves OPEN.
D. RHR Heat Exchanger Flow Control Valves FCV-74-16 & 28 will fail CLOSED
'.%.-..
and cause RCS temperature to RISE. Ensure all unloading solenoids
manual valves OPEN.
The correct answer is A
A. Correct
~ Flow control valves fail open on loss of nan-essential air. AWI
directs operators to manually load and unload compressors since loss of
sequencer power has unloaded all remaining C&SS compressors.
B. Incorrect - Student may be distracted by plant conditions, thinking that a
mode change would occur if a slight heat-up was the result of valves failing
closed.
C. Incorrect - Solenoid valves in the open position would keep the compressors
D. Incorrect - Student may be distracted by plant conditions, thinking that a
unloaded.
mode change would occur if a slight heat-up was the result of valves failing
closed. Solenoid valves in the open position would keep the compressors
unloaded.
REFERENCES:
Lesson Plan 3-OT-8011000
801-10 'Loss of Control Air'
ARI 41 FM2E
--
IOCFR55.43.5/45.13
Ability to determine and interpret the following as they apply to the Loss Qf instrument
Air: Cause and effect of low-pressure instrument air alarm.
SRO-81
Reference:
3-OT-A011000
KjA value:
3.2
Lwei:
2
TierGrp:
lil
KIA Number: 065 AA2.01
Last [!sed:
Source:
NEW
SROOniy:
YES
E12 EA2.1 001
Given the following piant conditions:
-
The unit has sustained a main steam line break affecting all SIGs.
cur
'.--
{p:&
-
-
The operator has reduced AFW flow to minimum detectable flow to
minimize the RCS cooldown.
Local actions were successful in closing #3 SIG MSIV.
Safety Injection termination criteria is NQT met.
The following conditions exist:
-
SIG
Level
Pressure
1
32% WR slowly DRQPPING
320 psig and DROPPING
2
31% WR slowly DROPPING
370 psig and DROPPING
3
4
33% WR slowly BROQPING
320 psig and DROPPING
35% WR slowly RISING
380 psig and RISING
Which ON of the following describes the required action and the reason for the
action?
A? Transition to E-2, "Faulted SI6 Isolation", because there is an intact SIG
available.
-.- ..
5. Transition to E-3, "Steam Generator Tube Rupture", because there is an
unexplained RISE in SIG level.
C. Transition to FR-H.1, "Loss of Secondary Heat Sink", because there is a
RED condition on the Heat Sink Status Tree.
D. Continue with ECA-2 1, "Uncontrolled Depressurization of All Steam
Generators", because Safety Injection termination is NOT complete.
The correct answer is A.
A. Correct - rising pressure in #3 S/G indicates it has been isolated and is
intact and per the CAUTION at the beginning of ECA3.1, when any intact
SiG is isolated recovery actions should continue with E-2.
3. incorrect - one SIG level is rising however it is due to apparent isolation of
the SI6 allowing pressure and level to recover.
6.
incorrect - AFW was reduced by operator action. FR-H.l should NOT be
entered per CAUTION prior to step 3 in ECA-2.1.
D. incorrect - SI termination had not yet began, transition to E-2 is still allowable
-
per the CAUTION prior to step 1 of ECA-2.2.
REFERENCE:
lNPO Bank - Beaver Valley 12/01/2002
Lesson Plan 3-OT-ECA0201; ECAZ.1, "Uncontrolled Depressurization of All
113 CFR 55 43.5 145.13
Ability to determine and interpret the following as they apply to the (Uncontrolled
Depressurization of all Steam Generators): Facility conditions and selection of
appropriate procedures during abnormal and emergency operations.
-_
SRO - 82
Reference:
ECA-2.1
KIA value:
4.0
Level:
2
Tiw'Grp:
3
KIA Number: El2 EA2.1
Last Used:
12/01/2002
Source:
BANK
SRO Only:
YES
'_
003 G2. I . 14 001
Given the following plant conditions:
-
-
-
-
Unit is operating at 100% power.
Bank D control rod H-12 drops.
Reactor power has been reduced to less than 75% in accordance with
,408-2, "Malfunction of the Reactor Control System".
Repairs have been made to the dropped rod.
Which ONE of the following identifies individuals that must be notified of the
dropped control rod ANB agree on a retrieval rate?
A? Reactor Engineering and STA
B. Reactor Engineering and Operations Duty Manager.
6. Plant Manager and STA
D. Operations Duty Manager and Plant Manager.
The correct answer is A.
A. Correct, Both the Reactor Engineer and STA are required to be notified and
agree on a retrieval rate per AOI-2
B. Incorrect, Operations Duty Manager is required to be notified per 801-2 but
does not have to agree on a retrieval rate.
._ .....
C. Incorrect, Plant Manager is required to be notified per AOf-2 but does not
have to agree on a retrieval rate.
B. Incorrect, Operations Duty Manager and Plant Manager are required to be
notified per ,401-2 but do not have to agree on a retrieval rate.
REFERENCES:
Lesson Plan 3-QT-AQI0200
AOI-2
KA 0036.2.1.14 - Knowledge of system status criteria which require the notification of
plant personnel: Dropped Control Rod
KO - NIA §BO - 83
Reference:
AOI-2
U A value:
3.3
Level:
1
...~.
'TieriGrp:
1 12
KiAiiumber: 003 G2.1.14
Last Used:
Source:
NEW
SKOOnly:
YES
033 2.1.22 001
Unit 1 is at 4% power with a startup in progress. All systems are operating
normaliy and the rod control system is in MANUAL when the following
indications are received:
I..-__-.
~ NIS Intermediate Range Channel I (1-611-135) indicating offscale LOW
- Trip Status Light XX-55-5, Window 3, IR > P6, goes out.
Based on these indications, which ONE of the following actions will satisfy Tech
Spec requirements?
A. Maintain reactor power level below 5% until the IRM is restored to
OPERABLE status.
B. Maintain reactor power level below 10% until the IRM is restored to
OPERABLE status.
C. Raise reactor power to greater than 5% THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
I3 Raise reactor power to greater than 10% THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The correct answer is 6).
A. Incorrect - Student may believe that a Mode change is not permitted unless
both IRM are in service.
B. Incorrect - Student may not want to exceed P-10 as a specified condition of
'*.,
K O 3.0.4.
C. Incorrect - TIS allows continued operation if power is raised to
10% within
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Student may confuse the limit as the power level of the mode
change.
D. Correct - TIS allows continued operation if power is raised to > 10% within 2
hours.
KEFERENCES:
Lesson plan 3-OP-TIS0302
TIS 3.3.1
AOI-4
10CFW55.43.5t45.13
Ability to determine Mode of Operation.
SRQ-84
Reference:
3-OT-Ti'S0301
U A viilue:
3.3
Level:
2
Tieri'Grp:
112
.
KIA Number: 035 2.1.22
Last Used:
Source:
RANK
SRO Only:
YES
067 (22.4.30 001
Given the following plant conditions:
-
-
During a surveillance test, a fuel oil fire occured on la-A Diesel
The incident commander reported that Fire Operations extinguished the
fire and some damage is visible to ventilation ductwork located in the
room.
During the event the Shift Manager declared an Alert classification per the
Radiological Emergency Plan (REP).
'..
./'
Generator.
-
Which ONE of the following lists thqcmi.nWnatm,:$ime for making notification to the
NRC regarding this incident?
A. 5 Minutes.
3.
15 Minutes
C? 1 Hour,
D. 4Hours.
The correct answer is C
A. Incorrect - SED is required to notify the ODS of a REP declaration within 5
minutes. NRC notification is required within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
'k .*
B. Incorrect - Site Emergency Director is required to make the REP declaration
within 15 minutes. MRC notification is required within I hour.
C. Correct - declaration of any REP emergency class requires a ?-hour
notification for the NRC Operations Center.
D. Incorrect - declaration of any REP emergency class requires a 1-hour
notification for the NRC Operations Center. 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reports are also
immediate reports to the NRC.
REFERENCES:
SPP-3.5, "Regulatory Reporting Requirements"; Radiological Emergency Plan
qOCFK55 43.5 145.1 I
Plant fire on site: Knowledge of which events related to system
operationslstatus should be reported to outside agencies.
RO - N/A
SRO - 85
Reference:
SPP-3.5
K:A valut:
3.6
Level:
1
TierKirp:
1 :2
KiA Number: 067 (32.4.30
Last Used:
Source:
NEW
SROOnly:
YES
049 AA2.01 00 I
Given the following plant conditions:
-
Unit 1 is in mode 1.
-
1-SI-30-701, "Containment isolation Valve Local Leak Rate Test Purge
Air" is in progress.
- The report to the Unit SRO shows that penetration X-6, 1-FCV-30-50
Inboard Purge CIV has failed the local leak rate test. Leakage is in
excess of the administrative limit.
The failure of the Purge CIV has elevated overall containment leakage to
>.25% as verified by 1-SI-0-700, "Primary Containment Total Leak
Rate".
---
Which ONE of the following is the
condition? (use references provided)
A? Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, return the containment to an operable status
B. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, return the valve to an operable status or isolate the
required for this
C. Within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, be in mode 3 and mode 5 within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
D. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, return the valve to an operable status or isolate the
.-.-.
The correct answer is A.
A. Correct - T/S action 3.6.1 requires total containment leakage to be less than
25% or the containment is considered inoperable.
3. Incorrect - LCO 3.6.3.8 requires isolation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for a penetration with
two isolation valves. This does not pertain to purge valves.
C. Incorrect - LCQ 3.6.3.F and 3.6.IB both require this shutdown if the
penetration can not be isolated.
D. Incorrect - LC03.6.3.E allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to isolate a purge path if leakage is in
excess of limits.
REFERENCES:
Lesson plan 3-OT-TiS0306
TIS 3.6.3
lOCFR55.43.5l45.13
.
.*
Ability to determine and interpret the following as they apply to the boss of containment
Integrity: Loss of Containment Integrity.
§BO-86
Reference:
3-0'I'-'I'/S0306
KIA value:
1.3
I .eve1 :
1
TierGrp:
1 !2
KiA Number: 069 AA2.01
Last Used:
Source:
NEW
SROOnly:
YES
074 EA2.01 001
Given the following plant conditions:
- A major plant transient has occurred.
- The operating crew is monitoring inadequate core cooling entry
requirements,
- RCS pressure on the RVLIS - ICCM display is reading 2400 psig.
-
Both Pressurizer PORVs are closed.
~
The controlling PZR pressure instruments are reading 150
psig lower than the RVklS ICCM display.
\\--.'
-
Si and Phase 'B' have actuated.
As Core Exit Thermocouple temperatures RISE, which ONE of the following
describes the impact on the Subcooling Margin Monitor (ICCM)?
Train 'A'
subcooling indication will be
A. Accurate and slowly RISING.
B. Inaccurate and slowly RISING.
C? Accurate and slowly LOWERING
B. Inaccurate and slowly LOWERING
The correct answer is C.
-
A. Incorrect - WR RCS pressure instrument is part of the Post Accident
monitoring system and should be accurate. As temperature rises, subcooling
with lower.
5. Correct - WR RCS pressure instrument is part of the Post Accident
monitoring system and should be accurate. As temperature rises, subcooling
with lower.
C. Incorrect - WR RCS pressure instrument is part of the Post Accident
monitoring system and should be accurate. As temperature rises, subcooling
with lower.
B. Incorrect - WR RCS pressure instrument is part of the Post Accident
monitoring system and should be accurate. As temperature rises, subcooling
with lower.
REFERENCES:
Lesson plan 3-OT-SYS068F
lOCFR55 43 5145.13
Ability to determine and interpret the following as they apply to an Inadequate
Core Cooling: Subcooling Margin.
--,
SRO-87
Reference:
3-CYr-SYSU68F
L'A value.
4.9
Level:
2
TieriGrp:
1'1 'ti2
KiA Number: 074 EA2.UI
1,ast Used:
Source:
NEW
SRO Only:
YES
-.
004 n2.07 1101
Given the following plant conditions:
-
-
-
-
1-LT-68-335 failed LOW.
The plant is at 100% power.
All Control Systems are in AUTOMATIC.
PZR level control is selected to LT-68-339 and 335
Which ONE of the following describes the corrective actions required by
procedure to mitigate this failure and why?
A? Isolate charging in accordance with
Level Control System", because th
isolate.
, "Malfunction of Pressurizer
ent failure caused letdown to
B. Place charging valve
in MANUAL and restore level to
program in
of Pressurizer bevel
caused 1 -FCV-62-93,
Charging Flow Control, to go OPEN.
C. Isolate charging in accordance wit
of Reactor Makeup
Control " because the instrument
8. Place charging valve controller, 1-HiC-62-93 in MANUAL and restore level to
program in accordance witG?%J%4alfunction of Reactor Makeup Control",
because the instrument failure-caused I-FCV-62-93, Charging Flow PZR
Level Control, to go OPEN.
The correct answer is A
A. Correct - letdown will isolate due to the instrument channel failing low,
therefore charging will be isolated in order to minimize thermal transients on
the regenerative heat exchanger. AOI-20, "Malfunction of Pressurizer Level
Control System" provides response to this failure.
\\-.:
B. incorrect - charging flow controller would not cause the charging flow control
valve to open since another channel provides it's control. Charging must be
isolated in any case since letdown isolated due to the instrument failure.
AOI-20, "Malfunction of Pressurizer Level Control System" provides response
to this failure.
C. Incorrect - letdown will isolate due to the instrument channel failing low,
therefore charging will be isolated in order to minimize thermal transients on
the regenerative heat exchanger. However, AOI-3 is not the correct
procedure and the examinee may confuse the two AOls.
D. Incorrect - charging flow controller would not cause the charging flow control
valve to open since another channel provides it's control. Charging must be
isolated in any case since letdown isolated due to the instrument failure.
However, AOI-3 is not the correct procedure and the examinee may confuse
the two AQls.
REFERENCES:
Lesson Plan 3-0T-A0l02000
AQI-20
lOCFR55 41.5 145.5 1 45.3 145.5
Ability to (a) predict the impacts of the following malfunction or operations on the
CVCS; and (b) based on those predictions, use procedures to correct. control, or
mitigate the consequences of those malfunctions or operations: Isolation of
ietdownlmakeup.
Rcference:
SYS068C
lciA value:
3.7
Lewl:
2
Tieri'Grp:
3
KjA Number: 004 A2.07
h s t Used:
Source:
NEW
SKO Only:
YES
06 1 A2.05 00 1
Given the following plant conditions:
as been implemented.
te with operators locally to
R will need to c
-
Instrument air h
A. Pump speed varied with Trip and Throttle valve
B. adjustments onL Vs (Level Control Valves) at the Valve.
..~--.- 3
..-:::::.;:<:
. . ~ J
--.. -... 3
....
-. .
. .. - ...
ents &XV.
ressure regulators at backup nitrogen station.
C.' "~
8. Adj;as_tments
The correct answer is C,
LCVs from the control room after backup nitrogen aligned
A. Incorrect - AQI-30.2 does not provide direction to control T&T valve. This
would be effective in controlling level if used.
effective in controlling level if used.
',..-
~
5. Incorrect - AOI-30.2 does not use local valve adjustments. This would be
C. Correct - Nitrogen backup is in place for Appendix R considerations and is
the method directed by 801-30.2 for control of TDAFW subsystem.
B. incorrect - Nitrogen backup is available for use during Appendix R events
however when it is aligned control is by local means only. Examinee may
think that after nitrogen is aligned control is restored to the control room just
as though instrument air is available
REFERENCES:
Lesson plan 3-OT-A013000
1 OCFR55.43.5i45.13
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:
Automatic control malfunction.
SRQ-69
.
d
.id-
'.---
Reference:
R-C'~-AO13000
KIA value:
3.4
Level:
2
Tier!Grp:
2 1
KIA Number: 061 h2.05
Last LJsed:
Source:
NEW
SROOnIy:
YES
'.---
073 (> 1.4 4 001
With the plant at 100% power the following annunciators are LIT:
-
7 9 4 , CCS HX A 1-RM-90-124 LIB RAD HI.
237-C, RCP THRM BAR REP FLOW LO.
-
23743, RCP THRM BAR RET HDR TEMP HI.
-
Which ONE of the following identifies the correct procedure that should be
entered to mitigate this event?
A. AOI-6, "Small Reactor Coolant System Leak".
B? AOi-15, "Loss of Component Cooling Water".
C. ,401-24, "Reactor Coolant Pump Malfunction"
D. 801-31, "Abnormal Release of Radioactive Material".
The correct answer is B.
A. Incorrect - AOI-6 would be entered for a small RCS leak however
interpretation of the alarms provided should indicate to the examinee that the
leak auto isolated the thermal barrier. AOI-75 ensures this action.
5. Correct- AOI-15 directs action to mitigate leaks into the CCS including the
C. Incorrect - AOI-24 normally addresses problems with RIcPs. Examinee may
thermal barrier.
.
incorrectly transition to this procedure due to the problem with the RCP
thermal barrier
B. Incorrect - AOI-31 addresses release fo radioactive material. Examinee may
incorrectly transition to this procedure due to the radiation monitor alarm.
Reference:
3-OT-AOl1 500; AOI-I 5
ARI 179-A
ARI 237-5, D
1 OCFR55 41.20 / 43.2 I45.13
Process Radiation Monitoring System: Ability to recognize abnormal indications
for system operating parameters which are entry-level conditions for emergency
and abnormal operating procedures.
KO - N/A
SRO - 90
Kefirence:
AOI- 15
K;A value:
4.3
'h-.."
6.evel:
I
TierIGi-p:
3
KIA Kiimtxr: 073 G 2.4.4
Last Used:
Source:
NEW
srw oniY: YES
076 G2.?.?? 001
Given the following plant conditions:
-
-
ERCW Supply Header Temperatures:
Unit is operating at 100% power.
-
lA-84.5OF
-
2A-85.5'F
-
lB-85.6'F
-
28-84.9OF
Which ONE of the following identifies the entry requirements of Technical
Specification 3.7.9, Ultimate Heat Sink?
A. Since none of the temperatures are above the TS limit: no entry is required
B. Since only one temperature per train is above the TS limit; no entry is required
C. Since two of the four temperatures are above the TS limit; entry is required.
DY Since the average temperature is above the TS limit; entry is required
The correct answer is Q.
A. Incorrect, The temperature limit is 85 degrees, requires the applicant to know
the entry requirement of the LCO.
B. Incorrect, applicant may incorrectly assume the UHS limit is trained since the
LCO for ERCW Tech Spec 3.7.8 is trained.
C. Incorrect, applicant may incorrectly assume the U H S limit is similar to
instrumentation Tech Spec for number of channels required to be operable
D. Correct, The average of the four ERCW supply headers is used to determine
---.'
the UHS temperature. (84.5 + 85.5 + 85.6 + 84.9= 340.5 14 = 85.125) The
Tech Spec limit for UWS is 85 OF.
REFERENCES:
Lesson Plan 3-OT-SYS067, 3-OT-TIS0307
TIS 3.7.9
I-Si-0-2B-01
lQCFR5543.2145.13
KA 076 G2.2.22 - Service Water: Knowledge of limiting conditions for
operations and safety limits.
RO - N1A
SRQ - 91
KIA iLumDer: 076 G2.?.?2
Last [!sed:
Source:
N E W
SRO Only:
YES
011 AL041101
Given the following plant conditions:
-
-
-
Unit is operating at 100% power
1-HIC-62-93, Charging Flow PZR Level Control, is in automatic
1A-A CCP trips on overcurrent
Which ONE of the following identifies the effect this will have on 1-FCV-62-93
and the correct procedural direction?
A. Valve will OPEN and PZR level will RISE. Immediately start 15-E3 CCP per
AOi-20, Malfunction of the PZR Level Control System.
B! Valve will OPEN but PZR level will DROP. Realign charging and letdown per
AOI-20: Malfunction of the PZR Level Control System.
C. Vaive will CLOSE but PZR level will RISE. Immediately start 1B-B CCP per
SO\\-62.01, CVCS, Charging and Letdown.
D. Valve will CLOSE and PZR level will DROP. Realign charging and letdown
per Sol-62.02, CVCS, Charging and Letdown.
The correct answer is B
A. Incorrect, 1-FCV-62-93 will open but PZR level will drop vs. rise. Examinees
may think that if immediate actions are taken to isolate letdown that PZR
level will rise. The immediate start of1 B-B CCP is not allowed by 801-20.
Examinee may mistake this with a precaution in SOi-62.01 that allow
immediate start of the available pump, but all criteria are not met to allow
this.
-.---.
B. Correct, 1-FCV-62-93 will open and PZR level will drop. After the operator
takes immediate actions to isolate letdown the SRO should enter ,401-20 to
start the 1 B-3 CCP and reestablish charging and letdownl.
C. Incorrect, 1-FCV-62-93 will open but PZR level will drop vs. rise. Examinees
may think that if immediate actions are taken to isolate letdown that PZR
level will rise causing valve to close. The immediate start oflB-B CCP is
allowed by SOI-62.01, but all criteria is not met to allow immediate start.
B. Incorrect, 1-FCV-62-93 will not close but PZR level will drop. Although
Sol-62.01 does contain steps to reestablish charging and letdown the SWO
should enter AOI-20.
REFERENCES:
Lesson Plan 3-QT-A012000
AOI-20
SOI-62.01
'~. .--"
.'
10 CFR5541.5143.5145.3145.13
KA 01 lA2.04 - Ability to (a) predict the impacts of the following malfunctions or
operations on the PZR LCS; and (b) based on those predictions, use procedures
to correct, control, or mitigate the consequences of those malfunctions: Loss of
two or three charging pumps.
RO - N1A
SRQ - 92
Reference:
AOI-20
KIA value:
3.7
Level:
2
TieriGrp:
2'2
KIA Noinber: 01 i A2.04
Last [!sed:
Source:
NEW
SRO Only:
YES
072 ti2.4.48 001
Given the following plant conditions:
-
Unit is operating at 100% power.
-
There is NO movement of irradiated fuel in progress.
-
'I-RM-90-102, Spent Fuel Pit Area Radiation Monitor, Green, Amber and
Red lights are NOT LIT.
-
The following annunciators are LIT:
-
-
84-D, SFP Q-RM-90-102/103 lNST MALF.
184-B. SFP O-RM-S0-I02/103 RAD HI.
Which ONE of the following describes the the entry requirements into Technical
Specification 3.3.8, ABGTS Actuation Instrumentation, and the directions that
should be given to the operator for the Auxiliary Building Gas Treatment system?
(use attached reference)
A. ~ n t r y
into LCO is required; arect operator to ensure automatic ABGTS
B. Entry into LCQ is rquiredvtpf,k$utomatic ABGTS actuation does NOT occur.
Cf Entry into LCQ is NOT requiredtkhect operator to ensure automatic ABGTS
actuation occurs.
actuation occurs.
B. Entry into LCO is NOT required
utomatic ABGTS actuation does NOT
occur.
-.
The correct answer is C.
A. Incorrect, Entry into the LC0 is only required during movement of irradiated
B. Incorrect, Entry into the LCO is only required during movement of irradiated
fuel, ABGTS does start but applicant may be confused since a power failure
to the rate meter will cause alarm lights on the monitor to be out.
fuel, ABGTS does start
C. Correct, Entry into the LCO is not required and ABGPS does start
B. Incorrect, Entry into the LCQ is not required but ABGTS does start, applicant
may be confused since a power failure to the rate meter will cause alarm
lights on the monitor to be out.
REFERENCES:
Lesson Plan 3-OT-SYS0908
ARI-184-B AND 284-D
k-
10 CFR 55 43.5 / 45.12
KA 872G2.4.48 Area Radiation Monitoring System: Ability to interpret control room
indications to verify the status and operation of the Area Radiation Monitoring system.
and understand how operator actions and directives affect plant and system conditions:
80 - N/A SRO - 93
Reference:
3-07-SYSOOOA
KIA d u e :
3.8
Level:
2
TierIGrp:
212
K!A Number: 071 G2:1.48
Last Used:
Source:
NEW
SRO Only:
YES
G2.I.13 001
Nuclear Security has determined that a specific credible insider threat exists
Which ONE of the following describes the compensation measure taken for
the operating crew, and the individual responsible to authorize implementation
as directed by SPP-1.3 "Plant Access and Security".
A. Operations personnel will obtain an armed security escort for entry into vital
areas, as authorized by the Shift Manager.
B. Operations personnel will obtain an armed security escort for entry into vital
areas, as authorized by the Site Security Manager.
-
(P? Operations personnel will implement a two-person line of sight rule for entry
into vital areas, as authorized by the Shift Manager.
B. Operations personnel will implement a two-person line of sight rule for entry
into vitat areas, as authorized by the Site Security Manager.
The correct answer is C,
A. Incorrect - Security escort is not required for this event. A more serious
threat to the plant may require this comp measure, and Students may
confuse the requirements. Shift manager authorization is correct.
B. Incorrect - Security escort is not required for this event. A more serious
threat to the plant may require this comp measure, and Students may
confuse the requirements. Site Security Manager authorization is
incorrect. Students may choose Site Security Manager due to this high
level event.
---.
C. Correct - Phis is as directed by SPP-1.3
D. Incorrect - Site Security Manager authorization is incorrect. Students may
choose Site Security Manager due to this high level event.
REFERENCES:
SPP-1.3: 'Plant Access and Security' p. 25
1 OCFR55.41.10/43.5/45.9/45.10
Knowledge of facility requirements for controlling vital / controlled access
SRO-94
Reference:
SPP-1.3
KIA value:
2.9
LXVCl:
I
TierKirp:
3
KIA Number. G 2.1.13
Last Used:
Source:
NEW
SRO Only:
YES
G 2 I 34 001
Chemistry reports that reactor coolant pH is outside it's allowable limit. The shift
chemist recommends action be taken to reduce the Lithium concentration to
restore reactor coolant pH to within limits.
Which ONE of the following identifies the action that should be taken to reduce
lithium concentration and how that action affects reactor coolant pH7
.-
A. Maximize letdown flow to the CVCS mixed beds; LOWER pH
B. Maximize letdown flow to the CVCS mixed beds; RAISE pH
Cf Align letdown flow to the CVCS cation beds; LOWEK pH.
D. Align letdown flow to the CVCS cation beds; RAISE pH
The correct answer is C
A. Incorrect- letdown flow may be maximized in certain instances, such as
during failed fuel events, to remove fission products. Examinee may
confuse the use of mixed beds with cation beds. Not effective in removing
Lithium to adjust pH.
B. Incorrect - letdown flow may be maximized in certain instances, such as
during failed fuel events, to remove fission products. Examinee may
confuse the use of mixed beds vdith cation beds. Not effective in removing
Lithium to adjust pH.
C. Correct - letdown can be aligned to the cation beds to remove Lithium which
reduces the pH of reactor coolant.
D. Incorrect - letdown can be aligned to the cation beds to remove Lithium and
reduce pH of reactor coolant. Examinee may confuse whether Lithium
removal reduces or raises reactor coolant pH.
REFERENCES:
3-QT-SYSO62A
10CFR55 41.10 143.5 145.12
Ability to maintain primary and secondary plant chemistry within allowable limits.
SRO - 95
Reference:
SYS062A
ti;A value:
2.9
Level:
I
TierGrp:
1
K!A Number:
G 2.1.34
I.ast Used:
Source:
NEW
SROOnly:
YES
G 2.2. I 4 002
Which ONE of the following situations requires tracking via a Configuration
Status Sheet?
A? A pump breaker is opened in accordance with Shift Manager direction
--
~
due to inadvertent oil discharge.
B. Repositioning a normally locked open valve in accordance with a system
clearance.
C. Opening a 125V DC control power breaker for a safety related pump in
accordance with a system operating instruction.
D. Opening a normally closed valve in accordance with a troubleshooting work
document.
The correct answer is A.
A. Correct - unanticipated problems that require configuration changes must
be tracked. The configuration Status Sheet would be included with the
work order used for repairs.
B. Incorrect - repositioning a valve in accordance with the system clearance
procedure is allowable and the clearance process would provide for
tracking component configuration.
C. Incorrect - opening a breaker in accordance with the system operating
procedure is allowable and the procedure would provide for tracking
component configuration.
D. Incorrect - component configuration changes are allowable during trouble
shooting activities using a work document that contains general
configuration control instruction.
REFERENCES:
3-OT-SPPIOOI
1 OCFR55.43.3145.13
Knowledge of the process for making configuration changes
RQ-NA
SRO-96
Keferenc.e:
3-OT-SPP1001
KIA \\wlue:
3.0
Level:
2
TierIGrp:
-
K/A Number: G 2.2 I 4
Last Used:
Source:
KEW
SKO Only:
YES
.-.
.
(i 2.3.3 001
Given the following plant conditions:
-
-
-
-
Core off-load is in progress.
A fuel assembly has just been removed from the core and
is in the mast of the refueling machine.
Failure of the reactor cavity seal occurs.
AOI-29, "Dropped or Damaged Fuel or Refueling Cavity Seal
Failure" was implemented.
In which ONE of the following locations should the fuel assembly be placed?
A. In the RCCA change fixture.
B./ In any analyzed core location
C. In the cavity side upender and lowered to horizontal position.
D. Transport to the SFP side upender and leave in horizontal position,
The correct answer is B.
A. Incorrect - RCCA change fixture may normally be used as an alternate
storage location however not in the event
B. Correct - 801-29, "Dropped or Damaged Fuel or Refueling Cavity Seal
Failure", directs fuel handlers to place any fuel assembly carried by the
refueling machine in any analyzed core location.
C. Incorrect - if placed in the upender and lowered the assembly would be
underwater but inadequate level would be over the fuel for shielding.
D. Incorrect - assemblies in the transfer cart may be transported to the SFP side
upender and maintained in horizontal position. Examinee may think this is
allowed for any assembly contained in the refueling machine.
x.. .
REF ERE N C E S :
3-QT-SYS079A; 801-29, "Dropped or Damaged Fuel or Refueling Cavity Seal
Failure
10CFR55 43.4 l45.10
Knowledge of SRO responsibilities for auxiliary systems that are outside the
control room (e.g., waste disposal and handling systems).
80 - N/A
SRQ - 97
Rctbrcncc:
AOI-29
KIA villue:
2.9
L.evel:
2
TieriCrp:
3
- u i
K I A Kliniber: G 2.3.3
I.ast IJsed:
Source:
NEW
SRO Only:
YES
__I
G 2.1.8 001
Given the following plant conditions:
-
-
RE-90-1 18, Waste Gas Effluent Monitor and 1-RE-90-400, shield Building
Exhaust Monitors are both out of service.
Chemistry is requesting a Waste Gas Decay Tank release.
Which ONE of the foliowing statements describes the condition(s) which will
allow the WGDT (Waste Gas Decay Tank) release? (use references provided)
A. Both monitors must be returned to service prior to WGDT release
B. WGDT release can not be performed until RE-90-1 78 is returned to service.
C. WGDT release can not be performed until 1-RE-90-400 is returned to
service.
W WGDP release is permitted even though both monitors remain out of service.
The correct answer is D.
A. Incorrect - Student may select because ODCM lists both monitiors as required
for a gaseous release.
B. Incorrect - Student may select because ODCM lists this monitor as required at
all times when release in progress.
k,...
C. Incorrect - Student may select because ODCM lists this monitor as required
at all times.
D. Correct - ODCM will allow release with both monitors 00s if all Comp measures
taken.
REF ERE N C E S :
Lesson Plan 3-OT-ODCM
10CFR55 43.4145.1 0
Knowledge of the process for performing a planned gaseous radioactive release.
SRO - 98
Reference:
&'A value:
3.2
Level:
2
TizrKrp:
3
G.4 Number: (i 23.S
Last Used:
NEW
Source:
NEW
SRO Only:
YES
'a-
G2.4.41 001
Given the following plant conditions:
-
-
-
-
-
-
While handing fuel in the spent fuel pool, a fuel element is dropped.
Gas bubbles are rising from the element.
All SFP radiation monitor alarms are LIT in the control room.
1-RE-90-1, SFP area radiation monitor is reading 2758 mR/hr
0-RE 98-1016, Auxiliary Building Rad Monitor has been is indicating
2.3E+06 cpm for the last 20 minutes.
Field surveys indicate 110 mremlhr at the EAB (Exclusion Area
Boundary).
Which ONE of the following is the proper Emergency Plan declaration for this
event? (use references provided)
A. Site Area in accordance with category 7.1.
5f Alert in accordance with category 7.1
C. Alert in accordance with category 7.3.
B. Alert in accordance with category 7.4
The correct answer is B.
A. Correcf - Field Survey results > I80 mremlhr at the SP (Site Perimeter) satisfy #2 of
-_
the Site Area classification of section 7.1.
B. Incorrect - Alert section of 7.1 is satisfied but is not the highest classification
C. Incorrect - Alert section of 7.3 is satisfied but is not the highest classification
D. Incorrect - Alert section of 7.4 is satisfied but is not the highest classification.
REFERENCES:
Lesson Plan 3-OT-PDC-048C
10 CFR 55.43.5/45.11
Knowledge of the emergency action level thresholds and classifications.
SRO-99
Keierence:
~-OYT-PC'I)-O~SC
K/A valuc:
4.i
1 .evel:
2
TierIGrp:
3
KiA Numhet' G 2.4.41
Last Used:
NEW
Source:
NEW
SRO Only:
G 2.4.45 001
Given the following plant conditions:
-
Unit was at 100% power.
-
An inadvertent Safety Injection occurred.
-
Operators have just transilioned to ES-1 .I,
"SI Termination".
-
PZR level is 25% and slowly RISING.
-
PZR pressure is 2050 psig and RISING
-
The following annunciators are LIP:
-
88-B, PRT LEVEL W l L Q
-
88-C, PRP PRESS HI
-
88-D, PRT TEMP HI
Which ONE of the following is the cause uf the alarms and the correct procedure
implementation?
A. Pressurizer PQRVs, I-PCV-68-340 and 334, have failed open; transition tu
E-I, "Loss of Reactor or Secondary Coolant".
B. Pressurizer PORVs, 1-PCV-68-340 and 334, have lifted; continue in
ES-I .I,
"SI Termination"
C. RCP #I Seal Leakoff Relief Valve, I-RLF-62-636, has failed open; transition to
El, "Loss of Reactor or Secondary Coolant".
B.J RCP # I Seal Leakoff Relief Valve, 1-RLF-62-636, has lifted; continue in
\\.. %"'
The correct answer is D.
A. Incorrect - diagnosis of given conditions should indicated that PZR PORVs
should be closed since PZR pressure is rising.
B. Incorrect - diagnosis of given conditions should indicated that PZR PORVs
should be closed since PZR pressure is rising. There is no conditions that
indicates PQRVs have opened.
C. Incorrect - RCP # I Seal Leakoff Relief Valve, I-RLF-62-6313 normally opens
following a Safety Injection to provide RCP leakoff flow path.
D. Correct - this relief will open when the seal return line cntmt isolation valves
close due to a Phase A and continue RCP seal leak-off to the PRT.
REF ERE N C E S :
ES-I .I
- E-1; 3-OT-SYS068B;
IOCFR 55 43.5 145.3 145.12
Ability to interpret the significance of each annunciator or alarm
--
RO - N/A
SRO - 700
Reference:
SYSO6XU
KIA value:
3.6
Level:
2
TieriGrp:
3
KIA Number:
(j 2.4.45
Lsst Llsetl:
Source:
NEW
SRO Only:
YES
,.-
"..