ML081370230
| ML081370230 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/08/2008 |
| From: | NRC/RGN-II |
| To: | Tennessee Valley Authority |
| References | |
| 50-259/08-301 50-259/08-301 | |
| Download: ML081370230 (87) | |
See also: IR 05000259/2008301
Text
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36. RO 268000A2.01 OOI/MEMlT2G2///268000A2.0l//RO/SRO/Il/28/07 RMS
Given the following plant conditions:
- Unit 1, 2 and 3 are in operation at Rated power.
- While processing the floor drain sample tank through the Thermx system the following
conditions are noted:
- Floor Drain Collector Tank Level Rising
- Unit 3 Reactor Building Floor Drain Sump Level High High is received
- Unit 3 Announces on PA System that Unit 3 has scrammed.
Which ONE of the following is the expected response by the RADWASTE operator?
A..; Notify the Unit 3 Unit Supervisor of an EOI-3 Entry condition on Unit 3.
B. Notify the Unit 3 Unit Supervisor that an EOI-3 Entry condition exists, but the affected Unit cannot
be determined from RADWASTE.
C. Notify the Shift Manager that an EOI-3 Entry condition exists, but the affected Unit cannot be
determined from RADWASTE.
D. Control room notification is not appropriate during a scram transient since redundant alarms are
available in the affected control room.
KIA Statement:
268000 Radwaste
A2.01 - Ability to (a) predict the impacts of the following on the RADWASTE ; and (b) based on those
predictions , use procedures to correct, control , or mitigate the consequences of those abnormal
conditions or operations: System rupture
KIA Justification: This question satisfies the KiA statement by requiring the candidate to use specific
plant conditions to determine the response of the RADWASTE system due to a rupture of a plant system
and the procedures used to mitigate that condition.
References:
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the follow ing:
1. Whether Radwaste can determine which unit has an EOI entry condit ion.
2. Based on the answer to Item 1, determine the reporting requirements.
A is correct.
B is incorrect. This is plausible because the Unit 3 US is required to be notified , however the affected
unit CAN be determined from the RADWASTE control room.
C is incorrect. This is plausible if the candidate fails to recognize that the affected unit CAN be
determined from the RADWASTE control room. In addition, the RADWASTE operator is required by
procedure to notify the AFFECTED unit control room.
D is incorrect. This is plausible because each control room has a DRYWELL FLOOR DRAIN SUMP
HIGH LEVEL alarm, but no alarm for the Reactor Building. In addition, control room notification of EOI
entry conditions is MANDATORY, especially during transient conditions.
OPL 171.084
Revision 5
Page 13 of 63
( Instructor Notes
(c) Hi-Hi level starts both
pumps & brings in Hi Level
Alarm.
(2) Alternate relay swaps pump start
circuit.
d. Alarms
(1) Low alarms correspond to
automatic pump stop. Maintaining
adequate sump level is necessary
to ensure airborne contamination is
minimized in the area containing
the sump. Airborne contamination
could become significant if the
discharges into the sump were not
covered by water.
(2) High alarm corresponds to the
level at which one of the sump
pumps should start.
(3) High-High alarm corresponds to Hi-Hi Alarm @
the level at which both sump 66"
pumps should have automatically
started.
(4) The Reactor Building floor drain Obj. V.BA
sumps are monitored for entry into
EOI-3 (Secondary Containment Note: RB equip.
Control). The alarms for these drain sump is
sumps are only located in the not listed as
Radwaste Control Room (panel 25- EOI-3 parameter
17).
(a) There is one common alarm
for all three units for Reactor
Building floor drain sump
level High-High.
(b) There is one common alarm
for all three units for Reactor
Building equipment drain
sump level High-High.
(
OPL 171.084
Revision 5
Page 14 of 63
( Instructor Notes
(c) The radwaste operator can
determine which unit has
the high-high sump level by
the "REACTOR UNIT ONE",
"REACTOR UNIT TWO", or
"REACTOR UNIT THREE"
alarms. The unit with both
alarms in would have the
high-high sump level.
(d) The radwaste operator
must notify the affected unit
of the EOI entry condition as
directed by the ARP.
4. Sump locations that pump into radwaste.
a. Turbine building floor and equipment
drain sumps located north end of
condenser room.
b. Condensate pump pit floor and equipment
drain sumps located west side of each
condensate pump pit.
c. Backwash receiver pit sump located in
receiver pit on north wall , and discharges
to FDCT.
d. Reactor building equipment drain sump
located north-east quadrant basement
reactor building .
e. Reactor building floor drain sumps.
(1) 2 per unit - one pump per sump.
(2) Cross connected by 8 inch line at
overflow and 6 inch line at normal
level.
(3) Located in south-east and south-
west quad .
f. Drywell floor and equ ipment drain sumps
located north end of drywell basement.
OPL 171.084
Revision 5
Page 15 of 63
( Instructor Notes
g. Radwaste equipment and floor drain
sumps located north wall in basement of
Radwaste.
h. Off-Gas sump located northwest end of
Radwaste Basement.
i. Standby gas treatment building sumps
located in standby gas treatment building.
j. Off-Gas Building sump located in
basement of off-gas building.
k. Evaporator Building sump located south NOTE: Unit
wall of the first floor of the evaporator station sump
building. pump normally
discharges to
CCW discharge
conduit but may
be lined up to
pump to
radwaste FOCT
through normally
closed valve at
sump.
5. Sump Pump Rating
a. Motors - 480V AC
b. All sumps have two pumps per sump
except reactor building floor drain sumps
which have one.
37. RO 272000K5.01 OOl/C/A/SYS/HWC/B9/272000K5.0l//RO/SRO/
Given the follow ing plant conditions:
- The HWC System is in the Operator Determined Setpo int mode.
( * Hydrogen flow is set at 14 SCFM.
Which ONE of the following describes the plant respo nse if reactor power is reduced?
A~ MSL radiation levels will rise in opposition to the lowering of reactor power due to a rise in volitile
Ammonia production.
B. MSL radiation levels will lower in response to the lowering of reactor power due to a reduction in
Nitrogen concentration.
C. MSL radiation levels will lower in response to the lowering of reactor power due to a reduction in
Hydrogen concentration.
D. MSL radiation levels will rise in oppos ition to the lowering of reactor power due to a rise in Nitrite and
Nitrate production.
KIA Statement:
272000 Radiation Monitoring
K5.01 - Knowledge of the operational implications of the follow ing concepts as they apply to RADIATION
MONITORING SYSTEM : Hydrogen injection operation's effect on process radiation indications:
Plant-Specific
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use plant
conditions to determ ine the effect on radiation levels due to specific operating conditions of the Hydrogen
Injection system.
References :
Level of Knowledge Justification: This question is rated as CIA due to the requ irement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following :
1. The mechanism by which Hydrogen injection causes MSL radiation levels to rise higher than normal.
2. The effect on hydrogen concentration in the reactor at reduced feedwater flow with Hydrogen Injection
flowrate unchanged.
3. The effect on ammonia production due to a reduction of oxygen concentration in the reactor.
A is correct.
B is incorrect. This is plausible because nitrogen concentration DOES decrease with power level.
However, due to the reduction in oxygen concentration, the reduction in nitrite and nitrate production
allows more nitrogen available to combine with the excess hydrogen and form volitile ammonia .
C is incorrect. This is plausible because of the TYPICAL response of the HWC system when operated
in Automatic mode. In addition, understanding that hydrogen injection flowrate is constant may not lead to
an understanding of that relationship to a reduction in feedwater flow where hydrogen is injected ..
However, in Operator Demand mode the hydrogen concentration increases due to the reduction in
feedwater flow.
D is incorrect. This is plausible because MSL radiation levels DO rise in opposition to lowering power.
However, the reduction in available oxygen causes a reduction in nitrite and nitrate production, which
allows more volitle ammonia formation .
(
OPL 171.220
Revision 4
Page 14 of 71
h. O 2 source
LP turbine
(1) Air in-leakage - oxygen in the air blading to
leaks into the low pressure parts discharge of
of the steam cycle condensate
pumps is less
(2) Some air in-leakage is removed than atmospheric
by the SJAE's but some pressure
dissolves in the condensate
(3) Radiolysis reaction - One way O2 is
produced
2H 2O nand y radiation ) 2H +0
2 2
i. O 2 removal
(1) Radiation induced recombination Reaction is an
of H2 and O2 equil ibrium
reaction
2H 2
° ~~~~ ~iation ~ 2H 2 + 0 2
(2) Carry-over with the steam
j. Hydrogen addition
(1) When an excess of hydrogen is This is the basis
injected to the feed water, the of H2 injection
reaction is driven to the left and
less oxygen (and peroxide) is
produced
(2) The chemical environment
becomes less oxidizing
(3) Elements exposed to the coolant
will assume chemical fo rms using
less oxygen and/or more
(4) Solubility and volatility may be This is how MSL
affected by the change in radiation levels
"oxidation state" of the element will increase
( which will be
discussed later
OPL171 .220
Revision 4
Page 27 of 71
G. Operation Be very careful
when selecting
1. Normally controlled from the HWC Main functions from
Control Panel different screens.
Use self
a. When using the Operator Interface Unit checking.
(OIU) function buttons, be aware that
the same function key will cause
different actions on different screens INPO SER 3-05
(1 ) Hydrogen controller Flow controller
operates in 2
(a) Automatic/Power modes
Determined Setpoint Mode
- changes hydrogen Procedure Use
injection flow in response
to changes in reactor
power. Used for normal
operation of the HWC
System and when
reducing hydrogen
injection related dose rates
to support maintenance,
chemistry or radcon
activities while the plant is
operating
(b) Automatic/Operator Hydrogen flow
Determined Setpoint Mode stays constant,
- changes hydrogen regardless of
injection flow in response power changes,
to the setpoint being until operator
manually entered by the manually enters a
operator. Normally used new setpoint
when initially pressurizing,
purging and placing the
HWC System in service or
if Power Determined
setpoint is unavailable
(2) Oxygen controller - Only mode used
Automatic/Hydrogen Determined for oxygen control
Setpoint
OPL171.220
Revision 4
Page 45 of 71
(
n. Supply Facility Trip - A shutdown signal
is generated when either the hydrogen
or oxygen gas supply facility trips
o. Hydrogen or Feedwater Flow Signal
Failed - A shutdown signal is generated
when the hydrogen flow signal or the
feedwater flow signal is less than 2 rnA
or greater than 22 rnA
I. Radiological Effects of HWC On MSL's Obj. V.B .6
Obj. V.BA
1. The primary source of background MSL Obj. V.D.6
radiation levels during reactor operation is due Obj. V.E .6
to the decay of nitrogen-16 (N16 )
16
a. N has a half-life of 7.1 seconds
b. A 6.13 or 7.12 Mev gamma is emitted 6.13 Mev gamma
16
on N decay is more common
16 16 16
2. Mc:Jor source of nitrogen in a BWR is 0 (11,P) 0 + 11 -7 N +P
1
N reaction
3. When using normal water chemistry methods,
16
a major portion of the N present in the
reactor coolant combines with the free oxygen
to form water soluble nitrites (NOz) and nitrates
(N0 3 )
a. These compounds are circulated
through the reactor coolant systems and
are ultimately removed by the RWCU
System
b. A smaller fraction of the N 16 is carried Predominate
over in the steam in the form of nitrogen contributor to
gas (Nz) and ammonia (NH 3 ) background
radiation levels
16
4. Hz injection alters the N carryover ratio
a. Concentrations of N0 3 , NOz, and NO
( decrease
b. Concentration of NH3 increases Ammonia
OPL171.220
Revision 4
Page 46 of 71
(
(1) Agas
(2) High water solubility
5. The net production of N16 is not influenced by
hydrogen injection
6. The increased dose rates are due to the
increased ease with which N16 gets out of the We can maintain
reactor and into the steam pipes when in the up to 2.7 ppm
NH3 form injection
concentration.
7. The initial U2 run was the first week in Nov.
1999. Up to 90 scfm hydrogen was injected. MSL '8' was
Average MSL radiation level increased highest at 5.2
approximately 5 times normal times normal
8. Addition of noble metals to reactor water
a. Noble metals decompose during reactor Rubidium and
startup or shutdown Iridium
b. During this time it produces a thin layer
of noble metal on wetted surfaces
c. The ECP on these surfaces are reduced
significantly during subsequent
operation
d. This leaves a stoichiometric excess of
e. Now the amount of hydrogen injection
can be reduced which will lower MSL
radiation levels
This NOTE
OPL171.220
Revision 4
Page 53 of 71
(
c. Consequences of Event
No effects were noted. However there is
the potential for rapid recombination in
the Offgas Charcoal beds . Additionally
excessive hydrogen increases the risk
for explosion or fire..
2. High radiation on reduction in power event at
Monticello
a. Event description
On December 13, 1997 with Monticello
at approximately 75 percent power,
workers entered the main condenser
room to repair a leaking root valve and
found the dose rate 2.5 times greater
than expected
Reactor power had been reduced from
100 percent power to 75 percent power
for ALARA purposed and hydrogen .
water chemistry injection rate had been
reduced from the normal 40 scfm to 8
Dose rate encountered was significantly Encountered -
higher that expected 4,800 mrem/hr
Expected - 2,000
Job was stopped to evaluate the mrem/hr
situation and management decided to
have power reduced further
At 60 percent power, dose rates were
about 3,200 mrem/hr and the job was
completed
OPL171.220
Revision 4
Page 54 of 71
(
b. Cause of event
Lack of understanding of the Questioning
radiological effect of reducing reactor Attitude could
power under HWC conditions. As have prevented
reactor power is decreased, less N-16 is this.
produced, and steam line dose rates
decrease. However, power level
changes also change feedwater flow
rate and hydrogen concentration
Reactor power and feedwater hydrogen
concentration both affect steam line
dose rates. At a constant hydrogen
injection rate, as power is decreased,
feedwater flow rate decreases, and
hydrogen concentration increases
An increase in hydrogen concentration
increases the ammonia concentration
although hydrogen injection rates were
reduced, the hydrogen concentration
increase that occurred when feedwater
flow rate decreased with reactor power
was not accounted for and resulted in
higher than expected dose rates
c. Consequences of Event:
Rx power needed to be lowered to 60% Work Planning
vice 75% planned. This resulted in
unplanned lost generation.
Work had to be performed in a
3200mr/hr vice a planned 2000 mr/hr
field. This resulted in unplanned
exposure.
Potential lost opportunity to plan and
perform work that required Rx power to
be lowered to 60 % power vice the
planned reduction to 75% power
38. RO 290003A3 .01 OOIIMEM/T2G2/HVAC/4/290003A3.0l/3.3/3.5/RO/SRO/
Given the following plant conditions:
- High radiation has been detected in the air inlet to the Unit 3 control room.
( * Radiation Monitor RE-90-259B is reading 250 cpm.
Which ONE of the following describes the CREV System response?
A. Neither CREVS unit will auto start at that radiation level.
B. Both CREVS units will auto start with suction from the normal outside air path to elevation 3C.
C." The selected CREVS Unit will auto start; the standby CREVS Unit will begin to auto start , but will
only run if the selected CREVS Unit fails to develop sufficient flow.
D. The selected CREVS Unit will auto start and will continue to run until Control Bay Ventilation is
restarted, then it will automatically stop.
KIA Statement:
290003 Control Room HVAC
A3.01 - Ability to monitor automatic operations of the CONTROL ROOM HVAC including :
Initiation/reconfiguration
KIA Justification: This question satisfies the KiA statement by requiring the candidate to use specific
plant conditions to determine the effect on CREV initiation logic.
References:
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information .
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the following :
1. The initiation setpoint for CREV .
2. The initiation sequence for the selected CREV unit.
3. The method required to secure a CREV unit once started automatically.
A is incorrect. This is plausible because the Tech Spec initiation setpoint is 270 cpm , which is less than
the given radiation level. However, the actual CREV initiation setpointis 221 cpm.
B is incorrect. This is plausible since both CREV units receive a start signal on a valid initiation.
However, the CREV unit NOT selected will experience a 30 second time delay on initiation and will only
complete its start sequence if the selected CREV unit fails to start .
C is correct.
(
D is incorrect. This is plausible because the start sequence is correct. However, once initiated, CREV
must be manually secured. There is no automatic shutdown capability, only trips .
OPL171.067
Revision 13
Page 30 of 71
( INSTRUCTOR NOTES
7. Control Room Emergency Ventilation (CREV) is Tech. Spec. 3.7.3
designed to supply and process the outdoor air oej.v.s.z, V.B.5/
needed for pressurization during isolated V.C.6N.C.7
conditions. There are 2 CREV units rated at (Old CREV Units
3000 cfm each. A CREV unit consists of Motor- abandoned in place
driven fan, (power supply is from 480V RMOV as Auxiliary
Bd 1A for CREV Fan A; RMOV Bd 3B for CREV Pressurization
Fan B), HEPA filter (common), charcoal filter Systems)
assemblies located in the CREVS Equipment TP-4
Room, charcoal heater, and inlet isolation 2-47E2865-4
damper and a backflow check outlet damper.
They are designed to maintain a positive
pressurization to 1/8" w.g. minimum to the
control room.
a. A CREV may be started manually from Red indicating lights
control room Panel 2-9-22 if local control on panel 3-9-21 to
switch is in AUTO position via a 3 provide indication of
position, spring-return to center switch. CREV Fan A and/or
(STOP-AUTO-START). Actuates only the B running on Unit 3.
CREVS unit & associated damper, not the Annunciators are on
isolation dampers. panel 9-6 for all
units.
b. There is also a 2 position maintained
contact, one per train, AUTO-INITIATE/
TEST switch which is used to perform
system level actions for that train
(primarily testing). It provides the same
response as auto start.
c. Local start at local control station in relay
room is done using a 2 position
maintained, one per train, AUTO-TEST
switch. Isolation dampers do not operate
automatically if started from local panel.
d. Automatic start signals are: Obj. V.B .1N.B.2
(1) High radiation of 221 cpm above Obj. V.C.1
background (270 cpm Tech Specs) Obj. V.C.17
in air inlet ducts to control room
from (Radiation monitor RE 90- 1. S. 3.3.7.1
259A Units 1 & 2, Radiation
monitor RE 90-259B Unit 3). Either
monitor starts selected CREV unit.
(2) Reactor zone ventilation systems
radiation high ?.72 MR/hr
OPL171.067
Revision 13
Page 31 of 71
INSTRUCTOR NOTES
(3) Refuel zone ventilation systems The inlet damper is
radiation high ~72 MR/hr normally closed &
fails closed. Damper
(4) Low reactor water level at +2 inches
opening takes -70
above instrument zero
seconds. While in
(5) High primary containment pressure the intermediate
~2.45 psig position both red &
e. On receipt of a start signal, normal outside green lights will be lit
air paths (see below) to elevation 3C are on 2-9-22.
isolated. The selected CREV unit starts The unit heater will
once the inlet damper is full open. This energize 10 sec.
supplies pressurizing air to the Unit 1, 2 after the damper is
and 3 control rooms. One CREV unit can full open to allow the
supply all three control rooms, so the STBY fan to come up to
CREV unit will not normally start. Once speed. High Rad or
started, the CREV unit will continue to run PCIS signal will
until manually secured by first clearing the energize relays in
high radiation signals and the PCIS signals Div I (CR1-A) and
(otherwise equipment cycling will occur) Div II (CR1-B).
Contacts from the
f. Control bay (EL 617) isolation is CR1 relays are used
accomplished by five pneumatic and motor- to energize
operated low leakage dampers which solenoids to isolate
isolate all normal air intakes and exhausts the M.C.R. normal
for EL 617. intake dampers
(1) FCO-31-150B , fresh air make-up (150B,D,E,F, and G)
duct to Units 1 and 2 Control Room
and Relay Room AHU.
(2) FCO 31-150G, 3C elevation relief
vent isolation
(3) FCO-31-150E, exhaust from Unit 1
toilet, locker, and other rooms at
elevation 617.
(4) FCO-31-150D, mounted in fresh air
makeup duct to Unit 3 Control Room
AHU.
(5) FCO-31-150F, exhaust from unit 3
toilet, locker, and other rooms at
elevation 617.
OPL 171.067
Revision 13
Page 32 of 71
( INSTRUCTOR NOTES
g. Manual initiation of the emergency mode Adherence to
of operation can be performed from the procedures
control room by operation of the AUTO- INPO SER 03-05
INITIATEffEST switch (putting switch in Obj.V.BAN.B.5
INITIATEffEST position) at Pnl 2-9-22 .
This operation results in energizing the
CR1 relay for that train/division and
isolation of the control room dampers.
Only one solenoid must be energized to
close these dampers. Therefore, either
test switch will initiate a damper isolation.
h. One switch alone in the INITIATEffEST Normally "A" is
position will NOT result in full functionality selected unit via
of the CREVS units. If that train is not the switch in CREV
selected unit, then operation of that train room . If "A" is
will be delayed by approx. 30 seconds, inoperable, switch
waiting for the selected train . This delay 0-XSW-03 1-7214
will result in the operator waiting to see SYSTEM PRIORITY
the result of his operation of the switch. SELECTOR
The operator will see the amber light lit, SWITCH is placed in
indicating energization of the CR 1 relay TRAIN B position to
and the solenoid for isolation damper start it without time
closing , but will see no activity of the delay.
CREVS unit until the delay timer has
timed out.
i. If the selected train's switch is put in the
INITIATEffEST position that train will
immediately enter its initiation sequence,
with the damper's red light being lit as
well as the green, indicating travel of the
damper toward the open position .
However, should there be any failure of
the selected unit; the standby unit will not
start. This is because the CR 1 relay for
the standby unit was not actuated .
j. Therefore, when manually initiating
emergency operation of the new CREVS
units, it is important to put the AUTO-
INITIATEffEST switches of BOTH trains
to the INITIATEffEST position.
OPL171.067
Revision 13
Page 33 of 71
( INSTRUCTOR NOTES
k. Again, to secure operation, the AUTO-
INITIATEITEST switches must both be
returned to the AUTO position and then the
STOP-AUTO-START switches turned to the
STOP position, to reset the CR1 relays in
both divisions.
I. Trips for the units, which are effective at all
times, are the following:
(1) Fan overload
(2) Unit low flow, less than approx. 2700
cfm -- trip is delayed for 10 seconds
after fan start.
(3) High heater discharge temperature,
approx. 220°F
(4) Low heater delta
temperature(between unit inlet and
heater discharge), indicating that the
heater is not getting the relative
humidity below 70 % -- trip is
delayed for approx. 15 seconds after
the heater is energized.
m. When any of these trip signals are ObjV.B.4/ V.B.5
received, the following will occur:
(1) The heater will be immediately
deenerg ized.
(2) The fan will continue to run and the
damper will remain open for approx.
30 seconds, to dissipate the heat
from the heater. (In the case of fan
overload, the fan will trip
immediately.)
(3) The inlet damper will be
deenergized, and when no longer
fully open, the fan will be
deenergized. The damper requires
approx . 20 seconds to close, while
fan coast down is approx. 60-90
seconds.
OPL171.067
Revision 13
Page 34 of 71
( INSTRUCTOR NOTES
n. In addition to the trips shown above, loss
of power to the inlet damper will trip the
unit. In this case, the heater is
immediately tripped and the fan is
deenergized when the damper is no
longer fully open. This action results in
(slightly) faster tripping of the heater to
avoid heat dissipation problems.
o. Flow switches are provided, one for each PDIS 7316 at Unit 2
division/unit, to start the standby unit if Vent Tower Intake
the selected unit does not start or trips Plenum
off. The selected unit not starting is
sensed by low differential pressure
across the common HEPA filter in the Unit
2 vent tower. Low differential pressure
exists when a fan is not operating; this
signal will normally be present. The
circuit for each unit is such that its
initiation sequence is begun upon either
of the following:
(1) Unit is selected as primary unit and
CR1 relay for that division is
energized.
(2) Other unit is selected as primary
unit, low differential pressure exists
across the common HEPA filter,
and CR1 relay for that division has
been energized for approx. 30
seconds .
p. With this circuit design, when an accident
signal is initially received, the selected
unit will enter its initiation sequence
immediately and the other unit will enter
its initiation sequence approx. 30 seconds
later. Once the selected unit fan has
been started (taking approx. 75 seconds-
- 70 for the damper and 5 for the fan), the
low differential pressure signal will no
longer be present in the standby unit
circuitry and its damper will return to the
fail-close position.
OPL171.067
Revision 13
Page 35 of71
INSTRUCTOR NOTES
q. If the selected unit fails to start properly, it
will itself be turned off by the trips noted
above, and the standby unit will continue
in its initiation sequence. The time delay
for startup of the standby unit will be
selected to ensure that regardless of the
primary unit failure, both fans will not be
running at the same time.
r. If the selected unit starts properly, but
then trips at a later time, the standby unit
will only be missing the low differential
pressure signal to receive its start signal.
The standby unit will start when the
selected system has completed its
shutdown process and the fan has been
deenergized. .
s. To secure from emergency operation, the
high rad signals and the PCIS signals
must first be cleared (otherwise
equipment cycling will occur). These
signals must be cleared on both divisions
to not have the standby unit start up when
the selected unit is secure. The STOP-
AUTO-START switches in the control
room should then be moved to the STOP
position for both units. This will reset I
deenergize the CR1 relays in both
divisions, reopen the control room
isolation dampers and remove the' start
signal from the operating CREVS unit.
The CREVS unit heater will then be
deenergized, with the fan continuing to Obj. V.B.2.
run and the damper held open for approx.
30 seconds, and the damper closing and
the fan turned off as discussed earlier.
(
39. RO 295001AK3 .01 OOIIMEMITIGIIRECIRCI129500IAK3.01//RO/SROINEW
Given the following plant conditions:
c *
Unit 3 is operating at 100% power.
The following alarm is received .
- Recirc loop A out of Service
Which ONE of the following describes the Reactor Water Level response?
A. Lower initially due to shrink, then return to normal.
B. ~ Rise initially due to swell, then return to normal.
C. Lower initially due to shrink and remain lower due to the 1055 of core voids .
D. Rise initially due to swell and remain high due to a lower power level.
KIA Statement:
295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4
AK3.01 - Knowledge of the reasons for the following responses as they apply to PARTIAL OR
COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : Reactor water level response
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the effect on reactor water level due to a partial 1055 of Recirculation flow.
References:
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following :
1. The internal response of the Reactor Vessel due to a Recirc Pump trip .
2. The RPV level instrument response to the conditions determined from Item 1 above.
The RPV response occurs in three parts .
First, the trip of the recirc pump causes a sudden reduction of coolant flow up through the fuel bundles
while power level remains approximately 100% . This causes a sudden increase in core void fraction.
The large voiding in the active fuel region coupled with the reduction in inventory being removed from the
downcomer by the tripped recirc pump results in a rapid rise in RPV level outside the core shroud while
water level inside the core shroud lowers. Since RPV level is measured outside the shroud, indication
rises .
Next, the large void content in the active fuel region responds quickly to insert negative reactivity, causing
a large reduction in reactor power and therefore, steam (void) production. This reduction in void fraction
draws water from the downcomer region outside the shroud into the active fuel region inside the shroud .
Even though reactor power will drop to approximately 65% with a 100% rod pattern, core void fraction at
65% is actually greater than at 100% due to the effect of recirc flow. Therefore, RPV level indication does
not immediately return to its original value .
Finally, the Feedwater Control System responds to the transient by reducing feedwater flow below steam
flow to enable RPV level to slowly return to the original setpoint at a lower reactor power.
A is incorrect. This is plausible because level inside the core shroud initially lowers, then returns to
normal, however RPV level is not measured inside the core shroud .
B s correct.
C is incorrect. This is plausible because level inside the core shroud initially lowers, then returns to
normal. In addition, the reduction of core voids is temporary. Final void fraction is actually higher.
However RPV level is not measured inside the core shroud.
D is incorrect. This is plausible because RPV level initially rises in response to the recirc pump trip.
However, the lower power level is compensated by automatic adjustments of Feedwater Control.
40. RO 29500IG2.1.14 OOl/MEM/TlGl/68 - RECIRC/2/295001G2.1.14//RO/SROI0606S NEW6/24/2007
Given the following plant conditions:
- You are the At-The-Controls (ATC) operator on Unit-1
( * Unit 1 is operating at full power when 1A Recirculation pump tripped .
- The Unit Supervisor has directed you to carry out the actions of 1-AOI-68-1A, Recirc Pump
Trip/Core Flow Decrease OPRMs Operable.
Which ONE of the following describes the required operator action(s) that CANNOT be carried out from
your watch station?
A. IMMEDIATELY take actions to insert control rods to less than 95 .2% loadline AND REFER TO
0-TI-464, Reactivity Control Plan Development and Implementation.
B~ Perform 1-SR-3.4.1(SLO), Reactor Recirculation System Single Loop Operation.
C. CHECK parameters associated with the Recirc Drive and Recirc Pump/Motor 1A(1 B) on ICS and
RECIRC PMP MTR 1A & 1B WINDING & BRG TEMPS, 1-TR-68-71 to determine the cause of trip .
D. REFER TO ICS screens VFDPMPA(VFDPMPB) and VFDAAL(VFDBAL) to help determine the
cause of the recirc pump trip/core flow decrease.
KIA Statement:
295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4
2.1.14 - Conduct of Operations Knowledge of system status criteria which require the notification of plant
personnel
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the required actions which require notification of plant personnel outside of
the control room due to a partial loss of Recirculation flow.
References: 1-AOI-68-1
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information .
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the following :
1. Which actions are required to be performed by 1-AOI-68-1 .
2. Which of the actions determined from Item 1 above CANNOT be carried out by the ATC operator.
A is incorrect. This is plausible because the BOP operator typically inserts control rods while the ATC
operator executes 1-AOI-68-1 and acts as Peer Checker if possible . However, manipulating control rods
is part of the ATCwatch station duties.
B is correct. This duty is carried out by Reactor Engineering .
C is incorrect. This is plausible because the BOP operator or STA typically carry out this action.
However, utilizing ICS screens is available at the ATC watch station and within his required duties.
D is incorrect. This is plausible because the BOP operator or STA typically carry out this action.
However, utilizing ICS screens is available at the ATC watch station and within his required duties.
(
BFN Recirc Pump Trip/Core Flow Decrease 1-AOI-68-1A
Unit 1 OPRMs Operable Rev. 0002
Page 7 of 12
4.2 Subsequent Actions (continued)
NOTE
1) Step 4.2[3] through Step 4.2[18.3] apply to any core flow lowering event.
2) Power To Flow Map is maintained in 0-TI-248, Station Reactor Engineer and on ICS.
[3] IF Region I or II of the Power to Flow Map is entered, THEN
(Otherwise N/A)
IMMEDIATELY take actions to insert control rods to less than
95.2% loadline AND REFER TO 0-TI-464, Reactivity Control
Plan Development and Implementation. 0
[4] RAISE core flow to greater than 45% in accordance with
1-01-68. o
[5] INSERT control rods to exit regions if NOT already exited AND
REFER TO 0-TI-464, Reactivity Control Plan Development and
Implementation. o
NOTE
The remaining subsequent action steps apply to a single Reactor Recirc Pump trip.
[6] CLOSE tripped Recirc Pump discharge valve. o
[7] MAINTAIN operating Recirc pump flow less than 46,600 gpm
in accordance with 1-01-68. n
[8] (NERlC] WHEN plant conditions allow, THEN, (Otherwise N/A)
MAINTAIN operating jet pump loop flow greater than
41 x 106 Ibm/hr (1-FI-68-46 or 1-FI-68-48). [GE SIL 517] o
BFN Recirc Pump Trip/Core Flow Decrease 1-AOI-68-1A
Unit 1 OPRMs Operable Rev. 0002
Page 8 of 12
4.2 Subsequent Actions (continued)
CAUTION
The temperature of the coolant between the dome and the idle Recirc loop should be
maintained within 75°F of each other. If this limit cannot be maintained, a plant cool down
should be initiated. Failure to maintain this limit and NOT cool down could result in hangers
and/or shock suppressers exceeding their maximum travel range. [GE SIL 251,430 and 517]
[9] IF Recirc Pump was tripped due to dual seal failure, THEN
(Otherwise N/A)
[9.1] VERIFY TRIPPED, RECIRC DRIVE 1A(1B) NORMAL
FEEDER, 1-HS-57-17(14). o
[9.2] VERIFY TRIPPED, RECIRC DRIVE 1A(1 B)
ALTERNATE FEEDER, 1-HS-57-15(12). o
[9.3] CLOSE tripped recirc pump suction valve using,
RECIRC PUMP 1A(1B) SUCTION VALVE,
1-HS-68-1 (77). o
[9.4] IF it is evident that 75°F between the dome AND the idle
Recirc loop cannot be maintained, THEN
COMMENCE plant shut down and cool down in
accordance with 1-GOI-100-12A. o
[10] NOTIFY Reactor Engineer to perform Reactor Recirculation
System Single Loop Operation, 1-SR-3.4.1(SLO) AN D to
refer to Station Reactor Engineer, 0-TI-248 and Tech
Specs 3.4.1 as necessary. o
[11] (NERlC] WHEN the Recirc Pump discharge valve has been
closed for at least five minutes (to prevent reverse rotation of
the pump) [GE SIL-517], THEN (N/A if Recirc Pump was isolated in
Step 4.2[9])
OPEN Recirc Pump discharge valve as necessary to maintain
Recirc Loop in thermal equilibrium . 0
BFN Recirc Pump Trip/Core Flow Decrease 1-AOI-68-1A
Unit 1 OPRMs Operable Rev. 0002
( Page 9 of 12
4.2 SUbsequent Actions (continued)
[12] REFER TO the following ICS screens to help determine the
cause of recirc pump trip/core flow decrease.
- VFDPMPA(VFDPMPB) 0
- VFDAAL(VFDBAL) 0
[13] CHECK parameters associated with Recirc Drive and Recirc
Pump/Motor 1A(1B) on ICS and RECIRC PMP MTR 1A & 1B
WINDING & BRG TEMPS, 1-TR-68-71 to determine cause of
trip. 0
[14] PERFORM visual inspection of tripped Reactor Recirc Drive . 0
[15] PERFORM visual inspection of Reactor Recirc Pump Drive
relay boards for relay targets . 0
[16] IF necessary, THEN (Otherwise N/A)
REFER TO 1-01-68 for Reactor Recirc Pump trips. 0
[17] INITIATE actions required to make the necessary repairs. 0
NOTE
Restarting a Recirc Pump while in Region 1 is NOT allowed . Tech Spec 3.4.1.A requires
that the Reactor Mode Switch be immediately placed in SHUTDOWN upon entry into
Region 1
[18] PERFORM the following for Single Loop Operation:
[18.1] REFER TO 1-01-68 for guidance on single loop
operation. o
[18.2] REFER TO Tech Specs 3.4.1. o
[18.3] WHEN available, THEN
RETURN tripped Recirc Pump to service in accordance
with 1-01-68. 0
41. RO 295003AA2.0 1 00 l/MEM/TlG l/O-GOI-IOO-4/IRO 295003AA2.0 l/IRO/SR0/12/18/2007
Given the following plant cond itions:
- All three units were at 100% rated power when 500KV PCB 5234 (Trinity 1 feed to Bus 1
( Section 1) tripped and failed to auto close .
- The signal which caused the PCB trip cannot be reset.
- The Chattanooga Load Coordinator has issued a Switching Order directing BFN to open
Motor Ope rated Disconnect (MOD) 5233 and 5235 to isolate 500KV PCB 5234 for
troubleshooting.
Wh ich ONE of the following describes your response to this Switching Order and the basis for that
response?
A. Ensure the PK block for PCB 5234 is installed to facilitate testing PCB 5234 by TPS personnel
assigned to troubleshoot the breaker trip.
B. Ensure the PK block for PCB 5234 is installed to prevent actuating the breaker failure logic and
tripping the remainder of the PCBs on Bus 1.
C~ Remove the PK block from PCB 5234 to prevent actuating the breaker failure logic and tripping the
remainder of the PCBs on Bus 1.
D. Remove the PK block from PCB 5234 to prevent electrical arching across the MOD contacts while
being opened .
KIA Statement:
295003 Partial or Complete Loss of AC / 6
AA2 .01 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE
LOSS OF AC. POWER : Cause of partial or complete loss of AC. power
KIA Justification: This question satisfies the KIA statement by requiring the cand idate to use specific
plant conditions to determine the potential cause of a partial or complete loss of AC power.
References: 0-GOI-300-4
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information .
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following:
1. The function of the PK block and its relationship to the breaker failure logic.
2. Recognize the CAUTION in O-GOI-300-4 related to PK block removal.
A is incorrect. This is plausible since the PK block must be installed to troubleshoot the breaker,
however it is not re-installed until AFTER the MODs are opened .
B is incorrect. This is plausible since the wording is ALMOST identical to the CAUTION, however the PK
block must be removed .
C is correct.
D is incorrect. This is plausible because electrical arching across the MOD contacts is actually what
causes the trip signal to be generated if the PK block is installed. The electrical arching will occur with or
without the PK block installed. However, it will not trip the 500KV breakers on the bus when it happens
without the PK block installed.
BFN Switchyard Manual 0-GOI-300-4
UnitO Rev. 0065
Page 65 of 85
8.2 Response to a Breaker Trip on 161kV or 500kV Breaker
CAUTION
Breaker reclosure times on opposite ends of the transmission lines leaving BFN are 15 to
17 seconds after a trip. The breakers at BFN should reclose immediately thereafter.
[1] IF a line trips, THEN
WAIT 30 seconds before resetting the disagreement to ensure
adequate time for automatic reclosure . 0
NOTE
1. 161kV breakers have high speed and standard speed reclosure.
2. For PCBs equipped with digital relays, the PCB will lockout from AUTO closure if
the affected line does not reclose from the other end within approximately 1.5
seconds. Only the AUTO closure is prevented, the breaker can be manually
closed with Dispatcher concurrence.
CAUTION
Induced currents in the current transformers of a 500KV PCB during cycling of the
associated MOD's , in conjunction with an existing PCB trip signal, may actuate the breaker
failure logic and trip all PCB's on the associated 500KV bus. Thus the MOD's associated
with a tripped PCB should NOT be operated until the trip has been reset; or, if the trip
cannot be reset, the breaker failure PK block has been removed for the associated tripped
PCB during MOD operation . Contact Dispatcher for instruction or assistance to reset the
tripped relay.
42 . RO 295004AK l. 03 OOIlC/A/TlGII24VDCNB9/295004AKl.03//RO/SROI
Given the following plant conditions:
- A reactor startup is in progress and reactor power is on IRM Range 7.
( * The operator observes the follow ing annunciators/indications:
- SRM Channels A and C fail downscale
- IRM Channels A , C, E, and G HI-Hi INOP
Which ONE of the following power sources, if lost, would cause these failures?
A." +1-24V DC Power Distribution Panel
B. 48V DC Power Distribution Panel
C. 120V AC Instrument and Control Power Distribution Panel
D. 120V AC RPS Power Supply Distribution Panel
KIA Statement:
295004 Partial or Total Loss of DC Pwr I 6
AK1.03 - Knowledge of the operational implications of the following concepts as they apply to PARTIAL
OR COMPLETE LOSS OF D.C . POWER : Electrical bus divisional separation
KIA Justification: This question satisfies the KiA statement by requiring the cand idate to use specific
plant conditions to determine the effect on a division of IRM instruments due to a loss of DC power.
References:
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble ,
sort, and integrate the parts of the question to predict an outcome. This requ ires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following:
1. Which of th listed power supplies input to the IRM system.
2. Which power supply, if lost, would provide only those indications listed.
A is correct.
B is incorrect. This is plausible because 48V DC supplies power to annunciator panels in the control
room including the two annunciators listed in the stem. However, other annunciators would also be
affected by the loss that are not included on the list.
C is incorrect. This is plausible because 120V AC I&C Buses supply power to IRM detectors and drives.
A loss of that power supply would also affect the entire division. However, the indications given in the stm
are not indicative of loads supplied by I&C buses .
D is incorrect. This is plausible because RPS supplies trip units associated with IRMs . However, the
given annunciators are not indicative of the loads supplied by RPS.
Unit 1 1-XA-55-5A Rev . 0011
( Page 42 of 43
SensorlTrip Point:
CHAN B.D .F.H Relay K16 A. Hi-Hi
HI-HI/INOP 1. 116.4 on 125 scale.
S. INOP
1. Hi voltage low.
2. Module unplugged .
(Page 1 of 1) 3. Function switch NOT in operate.
4. Loss of +/- 24 VDC to monitor.
Sensor Control Room Panel 1-9-12.
Location:
Probable A. Flux level at or above setpoint.
Cause: B. One or more inoperable conditions exist.
C. Testing in progress.
D. Malfunction of sensor.
E. Control rod drop accident.
Automatic A. Half-scram if one sensor actuates (except with Rx Mode Switch in RUN).
Action: B. Reactor scram if one sensor per channel actuates (except with Rx Mode
Switch in RUN).
Operator A. STOP any reactivity changes. o
Action: B. VERIFY alarm by multiple indications. o
C. RANGE initiating channel or BYPASS initiating channe/.
REFER TO 1-01-92A. o
D. With SRO permission, RESET Half Scram. REFER TO 1-01-99
E. IF alarm is from a control rod drop, THEN
REFER TO 1-AOI-85-1 . o
F. [NRC/C) IF one or more IRM recorder reading is downscale, THEN
CHECK for loss of +/- 24 VDC power. o
G. NOTIFY Instrument Maintenance that functional tests of any
monitors indicating an INOP condition , including a downscale
reading , are required before the instrument can be considered
operable . [NRC IE item 86-40-03) o
H. NOTIFY Reactor Engineer. o
/. REFER TO Tech Spec Table 3.3.1.1-1, TRM Tables 3.3.4-1
and 3.3.5-1. o
References: 1-45E620-6 1-730E237-6, -10 1 ~730E915-1 0
1-730E915RF-12 1-SIMI-92B
Unit 1 1-XA-55-5A Rev. 0011
( Page 10 of 43
SensorlTrip Point:
SRM Relay K-19 Count rate 5 cps.
DOWNSCALE
(Page 1 of 1)
Sensor Panel 1-9-12 , MCR.
Location:
Probable A. An un-bypassed SRM channel having a count rate :5 3 counts per second.
Cause: B. SI (or SR) in progress.
C. Malfunction of sensor.
Automatic Rod block below range 3 on IRM and Rx Mode Sw. NOT in Run.
Action:
Operator A. VALIDATE SRM downscale. o
Action: B. IF alarm valid, THEN
REFER TO 1-01-92 during startup (Mode 2) operation
or 0-GOI-1 00-3A, -3C during refuel (Mode 5) operation. o
C. NOTIFY Unit Supervisor. o
D. REFER TO Tech. Spec. Sect. 3.3.1.2, Table 3.3.1 .2-1, TRM Tables
3.3.4-1 and 3.3.5-1. o
References: 1-45E620-6-1 1-730E237-8
BFN Loss of I&C Bus A 1-AOI-57-5A
Unit 1 Rev. 0042
( Page 5 of 44
2.0 SYMPTOMS (continued)
H. Loss of Main Steam Relief Valve position indication.
I. Loss of power to RCIC and HPCI Turbine Vibration circuitry and position
indication for testable check valves, (Panel 9-3).
J. Loss of RHRSW and EECW Division I instrumentation, (Panel 9-3, 9-20).
K. Loss of SBGT A flow and differential pressure indication, (Panel 9-25).
L. Loss of SLC A and B amber ready lights and valve position indication,
(Panel 9-5).
M. Loss of LPRM meter lights and APRM alarm lights, (Panel 9-5).
N. Loss of Condensate - Feedwater and Heater Drains instrumentation,
(Panel 9-6).
O. Loss of SRM/IRM detector drive power and position indication, (Panel 9-5).
P. Loss of one-half the blue scram lights and accumulator low pressure-high level
( light indications (Panel 9-5,25-04).
Q. Loss of Control Bay Emergency Ventilation System Division I.
R. Loss of AC Supply to +/- 24V NEUTRON MONITORING BATT CHGR A1-1 NEG
SIDE, 1-CHGD-283-0000A1-1 and +/- 24V NEUTRON MONITORING BATT
CHGR A2-1 POS SIDE, 1-CHGD-283-0000A2-1 . (The Neutron Monitoring
Battery System is rated to carry loads for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. STACK GAS CH1 RAD
MON RTMR, 0-RM-090-0147B will be lost after this time period)
S. Loss of Main Steam Line B, D and Feedwater Line B flow indicators and inputs
to 3 Element Control and Rod Worth Minimizer.
1. "B" Fuel Pool Demin valves :
1. 1-FCV-078-0063, FPC F/D OUTBD ISOL VLV Closes,
2. 1-FCV-078-0068, RX WELL INFL INBD VLV Closes,
3. 1-FCV-078-0066, FPC F/D 1A BYP VLV Opens.
These actions result from loss of power to A and C skimmer surge tank low-low
level switches.
U. I&C BUS A VOLTAGE ABNORMAL (1-XA-55-8C, Window 21).
( V. Short Cycle valves 1-FCV-002-0029A and1- FCV-002-0029B fail open due to
loss of power to 1-FC-2-29.
43. RO 295005AAI .04 00 lIe/A/T! G1///295005AAI .04//RO/SRO/lll28/07 RMS
Given the following plant cond itions:
- Unit-1 is at 100% rated power when the Desk Unit Operator notices that the number 3 MTSV
c' *
position indication is reading 0%.
The number 1, 2, and 4 MTSV position indications all read 100%.
- Maintenance investigation determines that the cause of the MTSV position indication failure is
due to a mechanical failure of the LVDT.
- The Unit-1 Main Turbine receives a trip signal
Which ONE of the following describes the effect on Main Turb ine operation and any required action?
A. Main Turbine operation is unaffected. The RPS logic contact is already open for the #3 MTSV so a
turbine trip will still initiate a scram.
B. Main Turbine operation is affected. The RPS logic contact for the #3 MTSV will not function so a
turbine trip may not initiate a scram.
C. Main Turbine operation is unaffected. The Generator output breaker will still open on a turbine trip
due to a 2-out-of-4 logic arragement.
D~ Main Turbine operation is affected . The Generator output breaker will not open on a turbine trip due
to a 4-out-of-4 logic arragement.
( KIA Statement:
295005 Main Turbine Generator Trip I 3
AA1.04 - Ability to operate andlor monitor the following as they apply to MAIN TURBINE GENERATOR
TRIP: Main generator controls
KIA Justification: This question satisfies the KIA statement by requiring the cand idate to use specific
plant conditions to determ ine the required action following a Main Turb ine Generator trip.
References: OPL 171.228
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome .
0610 NRC Exam
(
REFERENCE PROVIDED: None
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following :
1. Whether the LVDT position indicator feeds the RPS logic, the Turbine Trip logic, or both.
2. Based on the above answer, the affect of a turbine trip with the failure active.
A is incorrect. This is plausible because the position of #3 MTSV is supplied to RPS logic. However, the
position indication supplied to RPS is a limit switch, not the LVDT position. Therefore, RPS logic "sees"
- 3 MTSV as open until the turbine trips.
B is incorrect. This is plausible because the Main Turbine operation is affected. However, the position
indication supplied to RPS is a limit switch, not the LVDT position. Therefore, a turbine trip WILL initiate a
C is incorrect. This is plausible based on the different logic associated with the CIVs and RPS on the
main turbine. However, the logic for MTSV inputs to open the generator output breaker is a 4-out-of-4
logic. Therefore, the generator outpu breaker will not automatically open.
D is correct.
(
OPL171.228
Revision 3
Page 63 of 81
( INSTRUCTOR NOTES
c. Consequences of Event
This caused the Bypass valves to start
opening. Due to the short duration of the
error signal the bypass valves did not
reach full open and subsequently closed.
Operation of the bypass valves would
impact Rx pressure, Rx power and
Generator load.
Corrective Action - EHC logic software
was modified to eliminate the possibility of
this type response to a communications
glitch.
2. At BFN on 1115/2006, the Unit 3 generator PER 95370
breaker failed to trip as expected on a turbine
trip.
a. Description of Event
The metal rod moves
At BFN on 1/1512006, the Unit 3 to alter the magnetic
generator breaker failed to trip as coupling of 2
expected on a turbine trip. The logic for opposing
the generator breaker needs to see all the transformer
stop valves closed and the CIV's closed secondary windings
(either intercept or stop). The LVDT for to make an LVDT
S.V#1 was failed such that the generator provide an output
breaker would not open on a turbine trip. proportional to the
Operator action was taken to manually position of the metal
trip the generator breaker rod.
(
OPL171.228
Revision 3
Page 64 of 81
( INSTRUCTOR NOTES
b. Cause of Event
Work practices
The LVDT transformer coupling rod Monitor all
became disconnected from the valve and parameters during a
fell to a position which gave indication of transient and ensure
- 50% valve position. The affect on the automatic actions
logic for tripping the generator PCB on a have occurred
turbine trip was not recognized.
c. Consequences of Event
Tripping of the generator breaker on a
turbine trip prevents a reverse power
situation where the generator and turbine
could attempt to rotate backwards, causing
equipment damage. The unit operator's
quick recognition and response to the
breaker failure to trip prevented damage.
44 . RO 295006AK3.05 OOl/C/A/TlGIIRPS/l/295006AK3.05//RO/SROI
Given the following plant conditions:
- Power ascens ion is in progress on Unit 3 with the main turbine on line.
( * Control rods are being withdrawn to increase power.
- As reactor power approaches 35%, the STA notes that 2 turbine bypass valves are open.
Which ONE of the follow ing describes the effect on the plant?
Regarding the FSAR Chapter 14 analyses for a turbine trip, the above condi tion _
A. is more conservative than the assumptions used in the FSAR because it lowers the actual power
level at which the RPS reactor scram on turbine trip is enabled.
B.1I is less conservative than the assumptions used in the FSAR because it raises the actual power
level at which the RPS reactor scram on turbine trip is enabled .
C. is less conservative than the assumptions used in the FSAR because it raises the actual power
level for a design basis transient in regard to peak cladding temperature.
D. is more conservative than the assumptions used in the FSAR because it lowers the peak vessel
pressure for a design basis transient in regard to transit ion boiling .
KIA Statement:
295006 SCRAM I 1
AK3 .05 - Knowledge of the reasons for the following responses as they apply to SCRAM : Direct turbine
generator trip: Plant-Specific
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the response to a Main Turbine trip and the basis for that response related
to a reactor scram.
References:
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the follow ing:
1. Wh ich assumptions are of concern regard ing the FSAR analysis for a turb ine trip.
2. What affect the above conditions have on that analys is.
3. Which thermal limit is of concern regarding the analyzed transient.
A is incorrect. This is plausible because the initial power level prior to a scram is assumed in the
analysis. However , the given conditions raise the initial power level which is less conservative.
8 is correct.
C is incorrect. This is plausible because the condition is less conservative based on initil power.
However, the limit of concern is not PCT, but MCPR.
o is incorrect. This is plaus ible because RPV pressure affects transition boiling. However, this is not the
limit of concern during this analysis and the initial conditions are LESS conservative with regard to MCPR.
(
45 . RO 295016AA2.04 001lC /A /TlGlIAOI-IOO-211295016AA2.04//RO/SR0/12/17/2007 RMS
Given the follow ing plant conditions:
- Unit-3 control room was abandoned due to a fire.
( * Control has been established at Panel 25-32 and actions are being carr ied out in accordance with
3-AOI-100-2, Control Room Abandonment.
- A cooldown has begun using MSRVs. Pressure is 850 psig and lowering .
- RHR Loop I is in Suppression Pool Cooling.
In accordance with 3-AOI-100-2, Control Room Abandonment, a Suppression Pool Temperature limit of
_ _OF has been established. The basis for this limit is ?
A. ~ 950F, to prevent exceeding the Technical Specification LCO before reaching Mode 4 (Cold
Shutdown) .
B. ~ 110oF , to prevent exceeding the Heat Capac ity Temperature Limit before the reactor can be
verified to be shutdown.
c. ~ 120oF , to prevent damage to the RCIC turbine from over-heated lube oil wh ich is cooled by the
Suppression Pool water.
D~ ~ 120oF, to prevent exceeding the design basis maximum allowable values for
primary containment temperature or pressure .
KIA Statement:
295016 Control Room Abandonment _
AA2.04 - Ability to determine and/or interpret the follow ing as they apply to CONTROL ROOM
ABANDONMENT : Suppression pool temperature
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the limitation and basis for Suppression Pool Temperature during a Control
Room Abandonment.
References: 3-AOI-100-2, Tech Spec Bases 3.6, EOIPM O-V-B
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble ,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its mean ing to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following :
1. The SP Temperature limit established by 3-AOI-100-2 .
2. The basis for the limit.
A is incorrect. This is plausible because ~950F is the normal operating limit imposed by Technical
Specification. However, this limit is not expected to be maintained during Control Room Abandonment.
B is incorrect. This is plausible because the basis for ~ 11OOF is correct. However, this limit is not
expected to be maintained during Control Room Abandonment and the reactor is assumed to be
shutdown .
C is incorrect. This is plausible because elevated SP temperatures can result in over-heating the lube oil
used for lubricating RCIC . This constitues the basis for Caution #6 in the EOls, but does not apply during
Control Room Abandonment.
D is correct.
BFN Control Room Abandonment 3-AOI-100-2
Unit 3 Rev. 0017
( Page 16 of 90
Date _
4.2 Unit 3 Subsequent Actions (continued)
[15.7] ESTABLISH RHR system flow between 7,000 and
10,000 gpm as follows: o
[15.7.1] MONITOR RHR SYS I TOTAL FLOW, 3-FI-74-79 at
Panel 3-25-32. 0
[15.7.2] THROTTLE OPEN 3-HS-074-0059C, RHR
SYSTEM I TEST VLV at 480V RMOV Bd 3A,
Compt.12C, 0
[15.7.3] WHEN RHR SYS I TOTAL FLOW, 3-FI-74-79
indicates between 7,000 and 10,000 gpm, THEN
DIRECT the operator to stop throttling
3-HS-074-0059C. o
[15.7.4] VERIFY CLOSED RHR SYSTEM I MINIMUM
FLOW VALVE, 3-FCV-74-7, at either of the
following: o
- 480V RMOV Bd 3D, Compt. 4E,
3-BKR-074-0007 RHR SYSTEM I MINIMUM
FLOW VLV FCV-74-7 (M010-16A), OR
(Otherwise N/A) 0
- Rx Bldg - SW Quad - E1 541' local control
switch RHR SYSTEM I MINIMUM FLOW
VALVE,3-HS-074-0007B. (Otherwise N/A) 0
[15.8] MONITOR SUPPR POOL TEMPERATURE,
3-TI-64-55B, at Panel 3-25-32 and MAINTAIN
temperature less than 120°F, o
(
Suppression Pool Average Temperature
B 3.6.2.1
(
BASES
ACTIONS D.1, D.2, and D.3 (continued)
Additionally, when suppression pool temperature is > 110°F,
increased monitoring of pool temperature is required to ensure
that it remains :s:; 120°F. The once per 30 minute Completion
Time is adequate, based on operating experience. Given the
high suppression pool average temperature in this Condition,
the monitoring Frequency is increased to twice that of
Condition A. Furthermore, the 30 minute Completion Time is
considered adequate in view of other indications available in
the control room, including alarms, to alert the operator to an
abnormal suppression pool average temperature condition.
E.1 and E.2
If suppression pool average temperature cannot be maintained
at s 120°F, the plant must be brought to a MODE in which the
LCO does not apply. To achieve this status, the reactor
pressure must be reduced to < 200 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and
the plant must be brought to at least MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The allowed Completion Time is reasonable, based on
operating experience, to reach the required plant conditions
from full power conditions in an orderly manner and without
challenging plant systems.
Continued addition of heat to the suppression pool with
suppression pool temperature> 120°F could result in
exceeding the design basis maximum allowable values for
primary containment temperature or pressure. Furthermore, if a
blowdown were to occur when the temperature was> 120°F,
the maximum allowable bulk and local temperatures could be
exceeded very quickly.
(continued)
BFN-UNIT 3 B 3.6-62 Revision 0
Suppression Pool Average Temperature
B 3.6.2.1
(
BASES
LCO b. Average temperature ~ 105°F when any OPERABLE IRM
(continued) channel is > 70/125 divisions of full scale on Range 7 and
testing that adds heat to the suppression pool is being
performed . This required value ensures that the unit has
testing flexibility, and was selected to provide margin below
the 110°F limit at which reactor shutdown is required. When
testing ends, temperature must be restored to ~ 95°F within
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> according to Required Action A.2. Therefore, the
time period that the temperature is > 95°F is short enough
not to cause a significant increase in unit risk.
c. Average temperature s 110°F when all OPERABLE IRM
channels are ~ 70/125 divisions of full scale on Range 7.
This requirement ensures that the unit will be shut down
at > 110°F. The pool is designed to absorb decay heat and
sensible heat but could be heated beyond design limits by
the steam generated if the reactor is not shut down.
Note that 70/125 divisions of full scale on IRM Range 7 is a
convenient measure of when the reactor is producing power
essentially equivalent to 1% RTP. At this power level, heat
input is approximately equal to normal system heat losses.
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause significant heatup of
the suppression pool. In MODES 4 and 5, the probability and
consequences of these events are reduced due to the pressure
and temperature limitations in these MODES. Therefore,
maintaining suppression pool average temperature within limits
is not required in MODE 4 or 5.
(continued)
BFN-UNIT 3 B 3.6-59 Revision 0
- - - - - _ ...*._ - -_ _- .
OPERATOR CAUTIONS EOI PROGRAM MANUAL
SECTiON O-V-B
DISCUSSION: CAUTION #5 and CAUTION #6
CAUTION #5, this warns the operatorof the potentialplant response if injection of cold, unborated water into the
coreis too rapid under conditions where little or no margin to subcriticality may exist. This may result in a large
increase in positive reactivity with a subsequent reactor powerexcursion large enough to substantiallydamagethe
core.
CAUTION#6, the HPCIand RCIC LubeOil Coolers are cooled by routing part of the pump discharge fluid to the
cooler. At elevated temperatures in the suppression pool, the turbine lubeoil may get too hot to provide adequate
lubrication. Only during EOI operationswill the system be needed at such an extreme suppression pool
temperature. Therefore, the EOIs are an appropriate location for this caution.
- REVISION 2 PAGE 15 OF 15 SECTION O-V-B
_. _ __ _-_~
46 . RO 295018AK2.01 OOI/MEM/TlGIIRBCCW/3/295018AK2.01///
Which ONE of the follow ing components would lose cooling upon isolation of the RBCCW non-essential
loop isolation valve (2-FCV-70-48)?
A. Drywell atmospheric coolers
B." Fuel pool cooling heat exchanger
C. Recirculation pump seals
D. Drywell Equipment Drain Sump Heat Exchanger
KIA Statement:
295018 Partial or Total Loss of CCW /8
AK2.01 - Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF COMPONENT
COOLING WATER and the following: System loads
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
system knowledge to determ ine the effect on RBCCW loads due to a partial loss of RBCCW.
References: 1/2/3-AOI-70-1, OPL 171.047
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the following :
1. Which of the loads listed are part of the "Essential Loop".
2. Wh ich of the loads listed are part of the "Non-essential Loop".
A is incorrect. This is an "Essential Loop" load.
B is correct.
C is incorrect. This is an "Essential Loop" load.
D is incorrect. This is an "Essential Loop" load.
(
BFN Loss of Reactor Building Closed 2-AOI-70-1
Unit 2 Cooling Water Rev. 0027
Page 14 of 14
Attachment 1
(Page 1 of 1)
Components Cooled by RBCCW During Normal Plant Operation
SYSTEM COMPONENTS COOLED
Reactor Recirculation Pump Seals
Pump Motor Bearings
Pump Motor Windings
Pump Discharge Sample Cooler
Primary Containment Drywell Atmosphere Cooling Coils
Reactor Water Cleanup Non-Regenerative Heat Exchangers
Pump Seals
Pump Bearings
Fuel Pool Cooling and Cleanup Fuel Pool Heat Exchangers
Equipment Drains Reactor Building Equipment Drain
Sump Heat Exchanger
Drywell Equipment Drain Sump
Heat Exchanger
(
OPL 171.047
Revision 12
Page 10 of 41
d. Proper system flow operation is assured by Done Each Shift
monitoring the system DP (pump discharge
minus pump suction).
2. RBCCW Heat Loads
a. Essential loop loads Obj. V.B.2
- Drywell Blowers(10) Obj. V.D.2
coolers (2)
coolers (2)
- Drywell equipment drain sump heat
exchanger (1)
b. Non-essential loop loads Obj. V.B.3
- Reactor Building equipment drain Obj. V.D .3
sump heat exchanger (1)
- Reactor water cleanup pump seal water
coolers and bearing oil coolers (2)
- RWCU Non-regenerative heat
exchangers (2)
- Fuel pool cooling heat exchangers (2)
- Reactor recirculation pump discharge
sample cooler (1)
3. RBCCW Heat Exchangers
a. These provide the means for heat removal DCN 51195,
from RBCCW by RCW with Emergency replaced HX1A &
Equipment Cooling Water (EECW) as a 1B, HX 1C NOT
backup. replaced.
OPL171.051
b. They are counter-flow type, 50% capacity
each.
- RBCCW flow makes one pass
through the shell side.
- RCW makes one pass through the
tube side.
47. RO 295019AA2.02 00 1/C/A/SYS/CAJI295019AA2.021I/MODIFIED 11/17/07
Given the following plant conditions:
( * The unit is stabilized, and the scram signal is reset.
- All 8 scram solenoid group lights are on.
- Ten minutes later, the following conditions are present:
- RCW pressure low alarm
- CRO charging water pressure high alarm
- Outboard MSIVs closed , Inboard MSIVs open
- SOV vents and drain valves closed
- Scram solenoid air valves open
Which ONE of the following describes the cause for the event?
A':I Loss of Control Air .
B. Loss of both RPS busses.
C. Loss of 9-9 cabinet 5, Unit Non-Preferred.
D. Loss of Orywell Control Air.
KIA Statement:
295019 Partial or Total Loss of Inst. Air 18
AA2.02 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE
LOSS OF INSTRUMENT AIR : Status of safety-related instrument air system loads (see AK2 .1 - AK2.19)
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the effect of a loss of Control Air on safety related loads.
References: 2-AOI-32-2
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following:
1. Which indications given are indicative of the possible causes listed.
A is correct.
B is incorrect. This is plausible because the scram would occur as well as scram valves open and SDV
vents and drains closed, however these indications would NOT be appropriate AFTER the scram was
reset. In fact, the scram could NOT be reset without RPS available.
C is incorrect. This is plausible because the only indications given that would not apply would be
outboard MSIVs closing and SDV vents and drains failure to re-open.
D is incorrect. This is plausible because a loft of Drywell Control Air would cause MSIVs to close,
however the INBOARD valves would close.
BFN Loss of Control Air 2-AOI-32-2
Unit 2 Rev. 0032
Page 5 of 25
1.0 PURPOSE
This Abnormal Operating Instruction provides symptoms, automatic action, operator
actions and expected system responses for loss of control air.
2.0 SYMPTOMS
A. AIR COMPRESSOR ABNORMAL annunciator, (1-XA-55-20B, Window 29) is in
alarm.
B. CONTROL AIR COMP G BKR ENERGIZED (0-XA-55-23B, Window 38) will
reset (extinguish) when panel reset pushbutton is depressed .
C. CONTROL AIR COMP G MOTOR AMPS, 0-EI-32-2901, on Panel 1-9-20
indicates approximately zero amps.
D. Air Compressor G ICS Display shows Compressor G in an unloaded or
shutdown condition .
E. Air Compressor G ICS Display shows lowering Control Air Header Pressure.
F. Control Air Compressor G breaker tripped.
G. SERVICE AIR XTIE VLV OPEN (0-FCV-33-1 Open) annunciator,
0-PA-33-1A11 (3) (Unit 1 and Unit 3) on Panel 1(3)-9-20 is in alarm at
(1(3)-XA-55-20B, Window 30).
H. CONTROL AIR PRESS LOW annunciator, 0-PA-32-88 is in alarm
(2-XA-55-20B, Window 32).
I. SCRAM PILOT AIR HEADER PRESS LOW annunciator, 2-PA-85-38B on
Panel 9-5 is in alarm (2-XA-55-5B, Window 28).
J. MAIN STEAM LINE ISOL VLV POSN HALF SCRAM annunciator is in alarm
(2-XA-55-4A, Window 30).
K. DRYWELL CONTROL AIR PRESSURE LOW 2-PA-32-70 annunciator is in
alarm (2-XA-55-3E, Window 35).
L. CONDENSER A, B OR C VACUUM LOW 2-PA-47-125 annunciator is in alarm
(2-XA-55-7B, Window 17).
M. OG HOLDUP LINE INLET FLOW LOW 2-FA-66-111A annunciator is in alarm (2-XA-55-53, Window 4).
N. HOTWELL A(B)(C) LEVEL ABNORMAL 2-LA-2-3(2-LA-2-6)(2-LA-2-9) is in
( alarm (2-XA-55-6A, Window 5(6)(7)).
BFN Loss of Control Air 2-AOI-32-2
Unit2 Rev. 0032
( Page 6 of 25
2.0 SYMPTOMS (continued)
O. REACTOR WATER LEVEL ABNORMAL 2-LA-3-53 annunciator is in alarm
(2-XA-55-5A, Window 8).
P. REACTOR PRESS HIGH 2-PA-3-53 annunciator is in alarm (2-XA-55-5A,
Window 1).
Q. REACTOR CHANNEL A(B),AUTO SCRAM annunciator in alarm if any scram
setpoint is exceeded (2-XA-55-5B, Window 1(2)).
R. MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW 2-PA-32-31 annunciator
in alarm (2-XA-55-3D, Window 18).
3.0 AUTOMATIC ACTIONS
A. Unit 2 to Unit 3 Control Air Crosstie, 2-PCV-032-3901, will close when Control
Air Header pressure reaches 65 psig and lowering at the valve.
B. U-1 TO U-2 CO NT AIR CROSSTIE, 1-PCV-032-3901, will close to separate
Units 1 and 2 when Control Air Header pressure reaches 65 psig and lowering
at the valve.
C. 2-PCV-84-0654, CAD/CA FLOW SEL, will select nitrogen from CAD tank A to
supply 2-FSV-64-20, 2-FSV-64-21, 2-FSV-64-221 and 2-FSV-64-222 at
I
~75 psig .
D. 2-PCV-84-0033, will select nitrogen from CAD tank A to supply 2-FSV-84-19,
2-FSV-64-29, and 2-FSV-64-32.
E. 2-PCV-84-0034, will select nitrogen from CAD tank B to supply 2-FSV-84-20,
2-FSV-64-31 and 2-FSV-64-34.
I
BFN Loss of Control Air 2*AOI*32*2
Unit2 Rev. 0032
( Page 7 of 25
4.0 OPERATOR ACTIONS
NOTE
[NER/C) Attachment 1 provides expected system responses, critical components that do not
fail in intended positions should be placed in the required positions. [INPO SOER 88-001)
4.1 Immediate Actions
None
4.2 Subsequent Actions
[1] IF a RFP Minimum Flow Valve failed open and flow is required
from the condensate/feedwater system to reactor vessel or to
prevent pump overload , THEN
ISOLATE the associated RFP minimum flow lines in the
appropriate RFPT Room as follows: (N/A any RFP valves not
affected.)
- RFP 2A MIN FLOW SHUTOFF, 2-SHV-003-0508 o
- RFP 2B MIN FLOW SHUTOFF, 2-SHV-003-0517 o
- RFP 2C MIN FLOW SHUTOFF, 2-SHV-003-0526 o
[2] IF CNDS BSTR PUMPS DISCH BYPASS TO COND B,
2-FCV-2-29A
and
CNDS BSTR PUMPS DISCH BYPASS TO COND C,
2-FCV-2-29B fail CLOSED, THEN (Otherwise N/A)
- VERIFY a flow path for condensate system
- STOP the condensate pumps/booster pumps using
2-01-2. o
[3] IF any outboard MSIVs fails closed, THEN:
PLACE associated hand-switch on Panel 2-9-3 to close
position. (Otherwise N/A) o
BFN Loss of Control Air 2-AOI-32-2
Unit2 Rev. 0032
( Page 8 of 25
4.2 Subsequent Actions (continued)
[4] IF RSW STRG TNK ISOLATION VALVE, 0-FCV-025-0032
FAILS CLOSED, THEN
START a high pressure fire pump using 0-01-26 . o
[5] OPEN CAD SYSTEM A N2 SHUTOFF VALVE,
0-FCV-084-0005, at Panel 9-54. o
[6] OPEN CAD SYSTEM 8 N2 SHUTOFF VALVE,
0-FCV-084-0016, at Panel 9-55. o
NOTES
1) All RCW temperature control valves fail open except for 2-TCV-24-808
and 2-TCV-24-858 on 2A and 28 R8CCW heat exchangers and 2-TCV-024-00758 on
the Main Turbine Oil Coolers (4" line) which fail closed.
2) The appropriate computer points may be used for monitoring for the following lube oil
temperatures, or any local temperature monitoring device that may be available, as
necessary
[7] IF RCW pump motor amps indicate that RCW System flow
reduction is required , THEN
REDUCE RCW flows as required: (Otherwise N/A). o
[7.1] CLOSE main turbine lube oil cooler TCV isolation valve
2-SHV-024-0583 or 2-SHV-024-0584, THEN
ESTABLISH lube oil temperature between 80°F and
90°F using TCV 8YPASS VALVE 2-8YV-024-0585 or
2-8YV-024-0586. 0
[7.2] CLOSE the following RFP turbine oil cooler TCV
isolation valves
- A RFP 2-24-624A or 2-24-625A o
- 8 RFP 2-24-6248 or 2-24-6258 o
- C RFP 2-24-624C or 2-24-625C o
BFN Loss of Drywell Control Air 2-AOI-32A-1
Unit2 Rev. 0021
( Page 4 of 9
1.0 PURPOSE
This abnormal operating instruction provides symptoms, automatic actions and
operator actions for the loss of Drywell Control Air System for causes other than
Group 6 Isolation. The loss of Drywell Control Air caused by a Group 6 Isolation is
addressed in 2-AOI-64-2d .
2.0 SYMPTOMS
A. DRYWELL CONTROL AIR PRESS LOW (2-XA-55-3E, Window 35) at
- 87 psig.
B. MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW (2-XA-55-3D,
Window 18) at s 82 psig.
C. Inboard MSIV's close or start to close.
D. Drywell cooler dampers close.
3.0 AUTOMATIC ACTIONS
None
(
48. RO 295021G2.4.50 OO l/C/A/TIGl /74- l//2950212.4.50///7
Given the following plant cond itions :
- Unit 2 is aligned with RHR Loop I in shutdown cooling with Loop /I in standby readiness.
- A leak occu rs wh ich results in the following condit ions:
- RPV level at '0' and slowly lowering
- OWP at 3.0 psig and slowly rising
- RHR pumps 'A' and 'C' tripped
Wh ich ONE of the following describes the minimum actions required to align RHR Loop /I for injection to
the RPV?
A. After FCV-74-47 and FCV-74-48 are closed; reset PCIS; push the RHR SYS /I SO CLG INBO
INJECT ISOL RESET 2-XS-74-132; and open both injection valves .
B." After FCV-74-47 or FCV-74-48 is closed ; push the RHR SYS /I SO CLG INBO INJECT ISOL
RESET 2-XS-74-132.
C. After FCV-74-47 or FCV-74-48 is closed ; reset PCIS; push the RHR SYS /I SO CLG INBO INJECT
ISOL RESET 2-XS-74-132; and open the inboard injection valve .
D. After FCV-74-47 and FCV-74-48 are closed ; start Loop 2 pumps;reset PCIS; and open the inboard
injection valve.
KIA Statement:
295021 Loss of Shutdown Cooling I 4
2.4.50 - Emergency Procedures / Plan Ability to verify system alarm setpoints and operate controls
identified in the alarm response manual
KIA Justification: This question satisfies the KiA statement by requiring the candidate to analyze plant
conditions and determine the required actions during an emergency which have resulted in a loss of
References: 2-AOI-74-1
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemb le,
sort, and integrate the parts of the question to predict an outcome. This requ ires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following :
1. Current RHR Loop II status given the initial conditions .
2. Based on the RHR Loop II status, determine the minimum actions to align Loop II for injection to the
RPV.
A is incorrect. This is plausible because the valve alignment is correct and resetting 2-XS-74-132 is
correct. However, resetting PCIS and re-opening FCV 74-47 & 48 are NOT required .
B is correct.
C is incorrect. This is plausible because the valve alignment is correct and resetting 2-XS-74-132 is
correct. However, resetting PCIS and re-opening FCV 74-47 is NOT required .
D is incorrect. This is plausible because the valve alignment is correct and resetting 2-XS-74-132 is
correct. However, resetting PCIS, re-opening FCV 74-47 and re-starting RHR pumps are NOT required.
(
BFN Loss of Shutdown Cooling 2-AOI-74-1
Unit 2 Rev. 0032
( Page 7 of 31
4.2 Subsequent Actions (continued)
[5] IF Shutdown Cooling isolates on low RPV water level or high
Drywell press (GROUP 2 ISOL) AND RPV water level needs
restoring using LPCI, THEN (Otherwise N/A)
PERFORM the following before reaching -122 inches RPV
water level: o
NOTE
The LPCI inboard injection valve that is aligned per 2-POI-74-2 will already be in the
required accident position with the breakers open and will NOT isolate.
[5.1 ] PERFORM the folloWing on a group 2 isolation: o
[5.1.1] IF 2-POI-74-2 is in effect, THEN
VERIFY CLOSED one of the following valves :
(Otherwise N/A) o
- RHR SHUTDOWN COOLING SUCT OUTBD
ISOL VLV, 2-FCV-74-47. 0
- RHR SHUTDOWN COOLING SUCT INBD
ISOL VLV, 2-FCV-74-48.
o
- VERIFY CLOSED the LPCI inboard injection
valve NOT aligned for 2-POI-74-2, (RHR SYS I
LPCI INBD INJECT VALVE, 2-FCV-74-53 OR
RHR SYS II LPCIINBD INJECT VALVE,
2-FCV-74-67) 0
BFN Loss of Shutdown Cooling 2*AOI*74*1
Unit2 Rev. 0032
Page 8 of 31
4.2 Subsequent Actions (continued)
[5.1.2] IF 2-POI-74-2 is NOT in effect, THEN
VERIFY CLOSED the following valves on a
Group 2 isolation : o
- RHR SHUTDOWN COOLING SUCT OUTBD
ISOL VLV, 2-FCV-74-47. 0
- RHR SHUTDOWN COOLING SUCT INBD
ISOL VLV, 2-FCV-74-48. 0
- RHR SYS I LPCIINBD INJECT VALVE,
2-FCV-74-53. 0
- RHR SYS II LPCIINBD INJECT VALVE,
2-FCV-74-67. 0
[5.2] DEPRESS RHR SYS 1(11) SD CLG INBD INJECT ISOL
RESET, 2-XS-74-126 and 2-XS-74-132 AND VERIFY
2-IL-74-126 and 2-IL-74-132 extinguished. o
(
49. RO 295023AKI.02 OOl/CIA/TlGI179-2N.B.3.B/295023AKI.02111
Fuel loading is in progress on Unit 1 when you notice an unexplained rise in SRM count rate and an
indicated reactor period; you suspect that an inadvertant criticality event is taking place .
Select which ONE of the following actions is an appropriate response to Inadvertant Criticality During
Incore Fuel Movements?
A. If unexpected criticality is observed following control rod withdrawal, manually SCRAM the reactor.
B. If all rods are not inserted/cannot be inserted, verify the fuel grapple is latched onto the fuel
assembly handle and immediately remove the fuel assembly from the reactor core .
C.oI If the reactor cannot be determined to be subcritical, traverse the refueling bridge and fuel assembly
away from the reactor core, preferably to the area of the cattle chute ..
D. Immediately EVACUATE all personnel from the refuel floor.
KIA Statement:
295023 Refueling Acc Cooling Mode I 8
AK1.02 - Knowledge of the operational implications of the following concepts as they apply to
REFUELING ACCIDENTS : Shutdown margin
KIA Justification: This question satisfies the KIA statement by requiring the candidate to analyze
specific plant conditions to determine a reduction in Shutdown Margin has occurred and the actions
required to address that condition .
References: 1-AOI-79-2
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome . This requires mentally using this
knowledge and its meaning to predict the correct outcome .
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the following :
1. The appropriate condition and Immediate Action required by 1-AOI-79-2.
A is incorrect. This is plausible because the condition is correct, but the action to scram is incorrect.
Reinserting the control rod is required .
B is incorrect. This is plausible because the required action is correct, but the condition is NOT correct.
This action is based on unexplained criticality following insertion of a fuel assembly.
C is correct.
D is incorrect. This is plausible because the evacuation of the Refuel Floor MAY be directed, but other
actions to mitigate the problem take precedence until personnel safety is compromised.
BFN Inadvertent Criticality During Incore 1*AOI*79*2 .
Unit 1 Fuel Movements Rev. 0000
( Page 6 of 9
4.0 OPERATOR ACTIONS
4.1 Immediate Actions
[1] IF unexpected criticality is observed following control rod
withdrawal, THEN
REINSERT the control rod. 0
[2] IF all control rods can NOT be fully inserted, THEN
MANUALLY SCRAM the Reactor. 0
[3] IF unexpected criticality is observed following the insertion of a
fuel assembly, THEN
PERFORM the following:
[3.1] VERIFY fuel grapple latched onto the fuel assembly
handle AND IMMEDIATELY REMOVE the fuel
assembly from the Reactor core . 0
[3.2] IF the Reactor can be determined to be subcritical AND
no radiological hazard is apparent, THEN
PLACE the fuel assembly in a spent fuel storage pool
location with the least possible number of surrounding
fuel assemblies and LEAVE the fuel grapple latched to
the fuel assembly handle . 0
[3.3] IF the Reactor can NOT be determined to be subcritical
OR adverse radiological conditions exist, THEN
TRAVERSE the Refueling Bridge and fuel assembly
away from the Reactor core, preferably to the area of the
cattle chute and CONTINUE at Step 4.1 [4]. 0
[4] IF the Reactor can NOT be determined to be subcritical OR
adverse radiological conditions exist, THEN
EVACUATE the refuel floor. 0
50. RO 295024G2 .1.33 OOl/C/A/TlGl/CONT/PRl/BIO/295024G2.1.33/IRO/SRO/
During operation at 100% power a gross failure of both seals on recirculation pump "8" increases drywell
pressure to 2.0 psig.
( Which ONE of the following is the approximate amount and type of RCS leakage?
A." 60 gpm of Unidentified leakage
B. 60 gpm of Identified leakage
C. 30 gpm of Unidentified leakage
D. 30 gpm of Identified leakage
KIA Statement:
295024 High Drywell Pressure / 5
2.1.33 - Conduct of Operations Ability to recognize indications for system operating parameters which are
entry-level conditions for technical specifications
KIA Justification: This question satisfies the KIA statement by requiring the candidate to determine that
entry into Technical Specifications is required based on conditions which have resulted in high drywell
pressure .
References: U2 TSR Sections 1 & 3.4.4, 2-AOI-68-1, OPL 171.007
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the following :
1. Whether the leakage is IDENTIFIED or UNIDENTIFIED leakage .
2. The amount of leakage assicoated with a gross failure of both seals on a single recirc pump.
A is correct.
B is incorrect. This is plausible because the amount of leakage is correct. However, the leakage is not
IDENTIFIED because the leakage is not intentionally captured and directed to a sump and is not
expected.
C is incorrect. This is plausible because the leakage is UNIDENTIFIED and equal to the Tech Spec
value for total leakage, but insufficient for the conditions given.
D is incorrect. This is plausible because the leakage is equal to the Tech Spec value for total leakage,
but insufficient for the conditions given . In addition, the leakage is not IDENTIFIED because the leakage is
not intentionally captured and directed to a sump and is not expected .
3.4.4
(
3.4 REACTOR COOLANT SYSTEM (RCS)
3.4.4 RCS Operational LEAKAGE
LCO 3.4.4 RCS operational LEAKAGE shall be limited to:
a. No pressure boundary LEAKAGE;
b. s 5 gpm unidentified LEAKAGE ; and
c. s 30 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
period; and
d. s 2 gpm increase in unidentified LEAKAGE within the previous
24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in MODE 1.
APPLICABILITY: MODES 1, 2, and 3.
ACTIONS
CONDITION REQUIRED ACTION COMPLETION
TIME
A. Unidentified LEAKAGE A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
not within limit. within limits.
Total LEAKAGE not
within limit.
B. Unidentified LEAKAGE B.1 Reduce LEAKAGE 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
increase not within limit. increase to within limits.
(continued)
BFN-UNIT 2 3.4-9 Amendment No. 253
Definitions
1.1
(
1.1 Definitions (continued)
LEAKAGE LEAKAGE shall be:
a. Identified LEAKAGE
1. LEAKAGE into the drywell, such as that from
pump seals or valve packing, that is captured
and conducted to a sump or collecting tank; or
2. LEAKAGE into the drywell atmosphere from
sources that are both specifically located and
known either not to interfere with the operation
of leakage detection systems or not to be
All LEAKAGE into the drywell that is not identified
LEAKAGE;
c. Total LEAKAGE
Sum of the identified and unidentified LEAKAGE;
LEAKAGE through a nonisolable fault in a Reactor
Coolant System (RCS) component body, pipe wall,
or vessel wall.
LINEAR HEAT The LHGR shall be the heat generation rate per unit
GENERATION RATE length of fuel rod. It is the integral of the heat flux
(LHGR) over the heat transfer area associated with the unit
length.
(continued)
BFN-UNIT 2 1.1-4 Amendment No. 253
Unit 2 2-XA-55-4A Rev. 0032
( Page 22 of 44
SensorlTrip Point:
RECIRC PUMP A 2-FIS-068-0055 0.1-0.2 gpm after second seal.
NO.2 SEAL
LEAKAGE HIGH
2-FA-68-55
(Page 1 of 2)
Sensor Recirculation Pump 2A Drywell
Location:
Probable A. Recirculation Pump 2A No.2 (outer) seal failure.
Cause: B. Sensor malfunction.
Automatic None
Action:
Operator A. COMPARE NO.2 cavity pressure indicator (2-PI-68-63A) to NO.1
Action: cavity pressure indicator (2-PI-68-64A), on Panel 2-9-4 or ICS. NO.2
seal degradation is indicated if the pressure at NO.2 seal is less
than 50% of the pressure at No. 1 seal. o
B. IF seal failure is indicated, THEN
INITIATE seal replacement as soon as possible. Continued
operation is permissible if Drywellieakage is within T.S. limits. o
NOTE
1) Possible indications of dual seal failure include:
- Window 25 on this panel alarming in conjunction with this window.
- Rising drywell pressure and/or temperature.
- Increased leakage into the drywell sump.
- Increased vibration of the recirc pump.
C. IF dual seal failure is indicated, THEN 0
1. SHUTDOWN Recirc Pump 2A by DEPRESSING RECIRC
DRIVE 2A SHUTDOWN, 2-HS-96-19.. 0
2. VERIFY TRIPPED, RECIRC DRIVE 2A NORMAL FEEDER,
2-HS-57-17. 0
3. VERIFY TRIPPED, RECIRC DRIVE 2A ALTERNATE FEEDER,
2-HS-57-15. 0
4. CLOSE RECIRC PUMP 2A SUCTION VALVE, 2-HS-68-1. 0
Continued on Next Page
OPL 171.007
Revision 22
Page 17 of 86
( \
(b) The flow keeps number 1 seal
cavity clean and cool by
flowing out of the seal area,
along the pump shaft, and
into the recirculation system.
(c) * This purge flow reduces the
possibility of seal damage due
to foreign material entering
the seal from an unclean
piping system .
(8) Seal Failures Obj . V.B .8
Obj. V.C .3
(a) Seal failure may be assessed TP-6
by the resulting changes in
flows and pressures.
(b) Failure of the number 1 seal ARPs provide useful
assembly would allow a info/analysis.
higher flow to the number 2 Obj . V.D.2c
seal cavity, forcing the Obj. V.E.3c
number 2 seal to operate at a
higher pressure (i.e., greater
than 500 psig) .
(c) This failure of the number 1
seal will cause leakage
through the controlled seal
leak-off line to rise to
approximately 1.1 gpm. A
flow element in this line
causes a common alarm on
high flow at 0.9 gpm or on low
flow at 0.5 gpm.
(d) Failure of the number 2 seal
assembly would cause its seal
pressureto drop (depending
upon the magnitude of the
failure).
(i) This failure would also
cause a higher leakage
through the seal leak
detection line
downstream from the
number 2 seal.
OPL17 1.007
Revision 22
Page 18 of 86
( (ii) Normally there is no
flow through this line
and flow switches are
set to alarm at 0.1-0 .2
gpm flow.
(e) Failure of both mechanical Would cause
seals would result in a total elevated drywell temp
seal assembly leakage of 60 and pressure, and
gpm as limited by the seal would exceed Tech
breakdown bushings. Spec and EPIP limits
for RCS leakage.
(f) Should the number 1 seal
restricting orifice become
plugged, the RECIRC PUMP
A(B) NO .1 SEAL LEAKAGE
ABN annunciator will alarm on
low flow (less than or equal to
0.5 gpm) . Additionally, a
reduction in number 2 seal
pressure would be seen.
(g) Should the number 2
restricting orifice become
plugged, the RECIRC PUMP
A(B) NO .1 SEAL LEAKAGE
ABN annunciator would also
alarm on low flow; however,
number 2 seal pressure would
rise to near the pressure of
number 1 seal.
(g) Seal Cooling Obj. V.B .9
(a) Cooling for the recirculation Obj . V.C.3
pump seals is required due to Obj. V.B .1gc
the heat generated by the
friction of the sealing surfaces
and the leakage of reactor
water through the seal
assembly.
(b) This cooling is provided by a
combination of supplied
Reactor Building Closed
Cooling Water (RBCCW) and
the leakage of primary coolant
past the seals.
51. RO 295025EK2 .0S OO l/CINTlGl/EHC LOGICI1295025EK2 .0S//RO /SROI
Unit 2 has experienced an inadvertant MSIV closure and subsequent reactor scram . Consequently,
RCIC was placed in level control and is also maintaining reactor pressure 900 to 1000 psig with the
MSIVs still isolated.
( Given these plant conditions, the digital EHC system is in pressure control with the
pressure setpoint set at _ _ psig.
A. Reactor, 970
B. Reactor, 700
C. Header, 970
D..... Header, 700
KIA Statement:
295025 High Reactor Pressure I 3
EK2.08 - Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following :
Reactor/turbine pressure regulating system: Plant-Specific
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the response of the digital EHC system to a transient resulting in a high
reactor pressure.
References: OPL 171.228
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble ,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
( In order to answer this question correctly the candidate must determine the following :
1. Whether the header pressure dropped sufficiently low enough to cause an automatic transfer to
Header Pressure Control.
2. Whether current plant conditions have allowed EHC logic to automatically transfer back to Reactor
Pressure Control.
A is incorrect. This is plausible because this condition is typical for post-scram EHC conditions if the
MSIVs are open.
8 is incorrect. This is plausible because EHC automatically transfers back to Header pressure control if
HEADER pressure returns above 725 psig. However, the MSIVs are still closed so the given readctor
pressure is not being sensed by the header pressure instruments.
C is incorrect. This is plausible because EHC will swap to Header pressure control, but the setpoint will
drop from 970 psig to 700 psig.
D is correct.
OPL171.228
Revision 3
Page 21 of 81
INSTRUCTOR NOTES
2. During turbine start-up and for a brief time Monitor Plant
following synchronization, the bypass valve parameters for
control also maintains the reactor steam expected response
pressure. Once all the bypass valves are closed,
then the turbine control maintains reactor steam
pressure either in Header Pressure or Reactor
Pressure Control depending on which operating
mode is selected.
3. Steam pressure control is selectable from either Obj.V.B.9.a
panel 9-7 or the EHC Workstation by selecting
HEADER PRESSURE CONTROL or REACTOR
PRESSURE CONTROL.
4. Header Pressure Control Input Signal
a. Two redundant pressure transmitters Powered from within
sense header pressure at the main steam the EHC system
throttle just upstream of the main turbine
stop valves.
b. Both signals are monitored for low, high,
difference, and hardware failures.
c. The higher of the two signals when no
failures are detected is selected as the
input.
d. A maximum difference setpoint of 10-PSI
is also established to detect a fault and/or
transmitter drift from either of the inputs.
e. In the event a fault is detected, the Obj.V.B.9.c
channel is prohibited from being used in
the signal processing and the appropriate
BYPASS pushbutton light will illuminate
on 9-7 and on the HMI operator interface.
f. Once the failed signal is corrected,
depressing the BYPASS pushbutton will
reset the BYPASS logic and both input
signals will then be processed.
OPL171.228
Revision 3
Page 22 of 81
( INSTRUCTOR NOTES
g. This mode IS NOT single failure proof -
one of the two pressure sensors failing
upscale can. and generally will be
selected by the logic to control. This will
depressurize the header to the MSIV
isolation setpoint of 852 psig in RUN
Mode.
h. In the unlikely event that both inputs
signals are detected as failed. the control
logic will automatically switch to reactor
pressure control.
i. If header pressure drops below 700-PSI,
and reactor pressure control is the
controlling mode of operation, the control
logic will automatically transfer to header
pressure control. If desired, the operator
may re-select reactor pressure control
after the transfer has been made even
though header pressure is below 700-psi.
The automatic transfer logic will re-
engage if header pressure rises above
725-psi.
5. Reactor Pressure Control Input Signal TP-3
a. Four (4) redundant pressure transmitters
(PT- 204a-d) grouped in pairs with "A"
and "B" constituting one pair and "C" and
"0" the other pair.
b. A pressure-biasing algorithm determines
the lagged high-median value of the four
(4) inputs and biases the remaining three
(3) input signals to that high median
value.
c. The high-median signal is then averaged Four biased signals
with the other three signals and is used are averaged.
as "Actual Rx Pressure" .
52. RO 295026EA2.01 OO l/C/A/T l/G l///295026EA2.0l//RO/SRO/RWM
Given the following plant conditions:
- Unit-2 is in a transient cond ition with current conditions as follows.
( - Suppression pool level: 13.5 feet
- Reactor pressure: 900 psig
- Suppression pool temperature: 105°F
Which ONE of the following describes the required action?
REFERENCE PROVIDED
A':I Operate all available suppression pool cooling .
B. Emergency Depressurize the RPV by opening all six ADS valves .
c. Rapidly depressurize via the Main Turbine bypass valves .
D. Lower Reactor Pressure to stay within the Safe Area of the Heat Capacity Temperature Limit Curve
and maintain cooldown rate below 100 deg. F/hr.
KIA Statement:
295026 Suppression Pool High Water Temp . / 5
EA2.01 - Ability to determ ine and/or interpret the following as they apply to SUPPRESSION POOL HIGH
WATER TEMPERATURE: Suppression pool water temperature
KIA Justification: This question satisfies the KIA statement by requiring the candidate to correctly
identify an adverse cond ition related to Suppression Pool High Temperature and then determ ine the
action required to correct the adverse cond ition.
Reference: 2-EOI-2 Flowchart, EOIPM Section O-V-D Page 85
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to solve a problem . This requires mentally using this
knowledge and its meaning to resolve the problem .
0610 NRC Exam
REFERENCE PROVIDED* HCTL Curve only
Plausibility Analysis:
A is correct.
B is incorrect. No condition has been met requiring Emergency Depressurization at this temperature. It
is plausible if the candidate focuses on the SP level, which is approaching the limit of 11.5 feet for
Emergency Depressurization . If this is the case, other actions on SP/L take priority over ED.
C is incorrect. ED would not be anticipated under these conditions unless the candidate focuses on SP
level which is approaching the limit of 11.5 feet for Emergency Depressurization. If this is the case, other
actions on SP/L take priority over ED.
D is incorrect. It is plausible if the cand idate continues down the SPIT leg of EOI-2 and determines that
exceeding the HCTL is possible. Since EOI-1 is used to lower pressure and cooldown, and no EOI-1 entry
condition has been met, it is unacceptable to assume that exceeding the HCTL is possible .
iii
! I
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II:
~
0 1
e i Ii i
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.......
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Z . ,I I,
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I
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r.= L~ q
C '"
-
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II:
II.
~
-
oIII
N*
- I
E MINATION
REFERENCE
'(,PROVIDED TO
, CANDIDATE
(
CURVE 3
HEAT CAPACITY TEMP LIMIT
260
250
ISAFE WHEN RX PREssl
Its BelOW 80PSfG I
~ RPV Press. 80
240
-
u:
t-
tl.
- &
230
220
r--
--- r--...
-- ~ ........
...w I RPVPrws.300 i"'"'- ~
-I
Q.
a:::
tl.
Q.
210
200
190
I I
RPV pross. 500 ---
r-- ~
~
""'-
I--.... r----.. i'-.
- ) l RPV pross..7'00 --... ..........
ftJ
180 RP\' Pt'8s$. gOO ---. r-.....
--....
~
110 .....1* . \135 '
\ RPI ~eSS. _ .-
--.. I-- """"""
r-...... c-
-..... r---....
160
~
1.0.- ISAAil
150
11.5 12 13 14 15 16 17 18 19
SUPPR PL LVL (FT)
- ACTION REQUUt E D fF ABOve CURVe FOR EXISTING RX PRESS
53. RO 295028EK3.04 OOlle/A/T! G lI480VLS/B5/295028EK3 .04//RO/SRO/
Given the following plant conditions :
- A Loss of Off-site power has occurred in conjunction with a LOCA on Unit-2.
( I * Plant conditions are as follows :
- Reactor Water Level +20 inches, steady
- Average Drywell Temperature 230°F , rising
- Suppression Chamber Pressure 11 psig, rising
- EDGs Tied and loaded to 4 KV Sd Bds
- Reactor pressure Remains> 800 psig
Which ONE of the following describes the final status of Unit 2 Drywell cooling?
A." Drywell coolers are operating with RBCCWavailable.
B. Drywell coolers are operat ing but no RBCCW is available.
c. Drywell coolers must be manually restarted , RBCCW is available.
D. Drywell coolers must be manually restarted , RBCCW is unavailable.
KIA Statement:
295028 High Drywell Temperature / 5
EK3.04 - Knowledge of the reasons for the follow ing responses as they apply to HIGH DRYWELL
TEMPERATURE : Increased drywell cooling
KIA Justification: This question satisfies the KIA statement by requiring the cand idate to use specific
plant conditions to determ ine the status of drywell cooling follow ing a transient wh ich results in high
drywell temperature .
References:
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome .
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
Co In order to answer this question correctly the candidate must determine the following :
1. Whether a CAS 480V Load Shed has been initiated based on the given conditions.
2. The status of RBCCW and DW Blowers based on the answer to Item #1 above.
A is correct.
B is incorrect. This is plausible because the DW blowers would be operating. However, RBCCW does
not receive a trip signal because a 480V Load Shed signal has not yet been iitiated . RPV level and
pressure are too high.
C is incorrect. This is plausible because following a 480V Load Shed, the DW blowers on the accident
unit must be manually started and RBCCW would be available . However, a 480V Load Shed signal has
not yet been iitiated. RPV level and pressure are too high.
D is incorrect. This is plausible because following a 480V Load Shed, the DW blowers on the accident
unit must be manually started . However, RBCCW does not receive a trip signal because a 480V Load
Shed signal has not yet been iitiated .
OPL171.072
Revision 11
Page 7 of 30
( INSTRUCTOR NOTES
x. Lesson Body
A. The 480V Load Shedding Logic System removes selected Obj. V.B.1N.D.1
loads from 480V boards which are powered from the 4kV TP-1, 2
Shutdown Boards
1. The load shedding is initiated by an accident signal Obj. V.B.3f V.D.3
on Unit 1 or 2 with .! diesel generator supplying Obj. V.C.2
one 4kV Shutdown Board as its only source of
power
AND
2. The accident signal is generated in the Core Spray TP-3
System logic Obj. V.B.2.N.D.2
Obj. V.C.1
a. Low-low-low reactor water level (-122"lLevel
1)
OR CASAsignal
b. High drywell pressure (2.45 psig) with low
reactor pressure (450 psig)
c. For load shed signal on U1 or U2, the
accident is for either unit
d. Unit 3 accident signal won't cause Unit 1 or 2
load shed or vice versa
3. The signal representing "diesel generator supplying TP-4
a 4KV shutdown board" is called "DGVA"
4. For DGVA logic to be satisfied, both conditions must
be present:
a. The DG output breaker or the U2 tie breaker
to U3 being closed
b. The normal and alternate feeder breaker must
beopan
5. All Unit 1-2 DGVA contacts are in parallel
6. Any DfG tied to its Shutdown Board with an accident
signal present will initiate U1-21oad shed logic
BFN Loss of Offsite Power (161 and 500 0-AOI-57-1 A
UnitO KV)/Station Blackout Rev. 0071
( Page'7 of 71
3.0 AUTOMATIC ACTIONS (continued)
V. Unit 1/2 480V Load Shed occurs on a loss of offsite power in conjunction with a
LOCA signal:
1. One RBCCW pump auto restarts (after 40 seconds on U1 and U2).
2. Drywell Blowers auto restart on non-accident unit (after 40 seconds).
Drywell Blowers with their respective auto restart inhibit switches in the
INHIBIT position will not auto restart.
3. Drywell coolers are manually restarted on the accident unit. A Drywell
Blower with its auto restart inhibit switch in the INHIBIT position can be
manually restarted after a ten minute time delay.
4. SGT TRAINS A & B trip, but will AUTO RESTART in 40 seconds when an
initiation signal is present.
5. Loss of Control Bay Chilled Water Pumps A & B. (may be restarted after 10
minutes with use of bypass switch).
W. Unit 3 480V load shedding occurs as follows:
1. Division I 480V load shedding will occur when an accident signal is present
and diesel generator voltage is available on the 4160V shutdown board
supplying the 480V shutdown board 3A as follows :
a. RBCCW pump 3A trips
b. Drywell blowers 3A1 & 3A2 trip
c. After a 40 second time delay, with the control switch in Normal After
Start, RBCCW pump 3A restarts
d. After a 40 second time delay, Drywell blowers 3A1 and 3A2 can be
manually restarted
e. Drywell blowers 3A3, 3A4 and 3A5 cannot be restarted until the load
shed signal is corrected
(
54. RO 295030EA 1.06 oor/c,A/T! G 113 .5/3.511295030EA1.06//RO/SR0/11l20107 RMS
I Given the following plant conditions:
( I * A LOCA has caused gross fuel failure on Unit 3.
- The SED/SRO has approved implementation of EOI Appendix 18, Suppression Pool Water
Inventory Removal and Makeup.
- The control room crew has just closed 3-FCV-74-63, RHR RADWASTE SYS FLUSH VALVE
- Suppression Pool level is -3.5 inches and steady.
Which ONE of the following describes the next appropriate action(s)?
A. Open 3-FCV-74-62, RHR MAIN CNDR FLUSH VALVE and direct suppression pool water to the Main
Condenser ONLY.
B. Re-open 3-FCV-74-63, RHR RADWASTE SYS FLUSH VALVE and direct suppression pool water to
Radwaste ONLY.
c. Verify open the 3-FCV-73-40, HPCI CST SUCTION VALVE, and open 3-FCV-73-30, HPCI PUMP
MIN FLOW VALVE
D~ Appendix 18 is complete; the Suppression Pool level is acceptable.
KIA Statement:
295030 Low Suppression Pool Water Level / 5
EA1.06 - Ability to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL
WATER LEVEL: Condensate storage and transfer (make-up to the suppression pool) : Plant-Specific
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the effectiveness of actions to control Suppression Pool level using the
Condensate storage and transfer system.
References: 2-EOI Appendix 18
Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,
sort, and integrate the parts of the question to predict an outcome. This requires mentally using this
knowledge and its meaning to predict the correct outcome.
0610 NRC Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly the candidate must determine the following :
1. Which actions are required based on the given conditions.
NOTE : Each distractor is plausible because they are all actions directed by 2-EOI Appendix 18 to control
Suppression Pool level.
A is incorrect. The given SP level is sufficiently low enough that additional inventory removal is not
necessary. In addition, following a gross fuel failure, rejecting water to the main condenser is also
inappropriate.
8 is incorrect. The given SP level is sufficiently low enough that additional inventory removal is not
necessary. However, following a gross fuel failure, rejecting water to Radwaste is more appropriate than
to the main condenser.
C is incorrect. The given SP level is sufficiently high enough that additional inventory makeup is not
necessary.
D is correct.
3-EOI APPENDIX-1 8
...
Rev. 2
Page 1 of 4
( 3-EOI APPENDIX-18
SUPPRESSION POOL WATER
INVENTORY REMOVAL AND MAKEUP
LOCATION: Unit 3 Control Room
ATTACHMENTS: None
CAUTION
[NRC/C] Suppression Pool water will be highly radioactive
after a LOCA . Chemical Engineering recommendations are used
to determine location to pump contaminated water.
[NRC Inspection Report 89-16]
NOTE : All panel operations performed at Control Room
Panel 3-9-3 unless otherwise stated.
1. IF Suppression Pool Water makeup is required,
THEN CONTINUE in this procedure at Step 5.
2. IF Gross fuel failure is suspected,
THEN OBTAIN SED/SRO permission to pump down Suppression
Pool BEFORE continuing in this procedure .
3. IF Directed by SRO,
THEN REMOVE water from Suppression Pool as follows:
a. DISPATCH personnel to perform the following
(Unit 3 RB, EI 519 ft, Torus Area) :
1) VERIFY OPEN 3-SHV-074-0786A(B), RHR DR PUMP
A(B) DISCH SHUTOFF VALVE.
2) OPEN the following valves:
- 3-SHV-074-0564A(B) , RHR DR PUMP A(B) SEAL WTR SPLY _
3) UNLOCK and OPEN 3-SHV-074-0765A(B) , RHR DR PUMP
A(B) DISCH .
4) NOTIFY Unit Operator that RHR Drain Pump
3A(3B) is lined up to remove water from
Suppression Pool.
5) REMAIN at torus area UNTIL Unit 3 Operator
directs starting of RHR Drain Pump 3A(3B) .