ML081370280
| ML081370280 | |
| 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: ML081370280 (107) | |
See also: IR 05000259/2008301
Text
REFERENCE PROVIDED: Unit-2 TRM Section 3.3.1
Plausibility Analysis:
(
In order to answer this question correctly, the candidate must determine the following:
1. Recognize that PIS 1-91A and 1-81A have failed in the conservative direction for the given plant
conditibns. (i.e; >30% power)
2. Based on Item 1 above, recognize that the TCVlSV Closure Scram capability is maintained as
long as power remains above 3QOk.
3. Recognize that an INFORMATION ONLY LCO is mandated to ensure these failures are addressed
when power is reduced below 30%.
4. Recognize that the Applicability NOTE for TRM 3.3.1 applies for the given plant conditions.
A is incorrect. The Applicability NOTE for TRM 3.3.1 applies for the given plant conditions, therefore
the requirement to place the channels in trip or restore them to OPERABLE status is NOT required. This
is plausible because the TCVlSV Closure Scram capability IS maintained and the actions in the
distractor closely resemble the required actions in the LCO.
B is correct.
C is incorrect. TCV/SV Closure Scram capability is maintained. This is plausible because the actions in
the distractor match the required actions in the LCO if the TCVlSV Closure Scram capability was NOT
maintained.
o is incorrect. TCVlSV Closure Scram capability is maintained. This is plausible because the actions in
the distractor match the required actions in the LCO if the TCVlSV Closure Scram capability was NOT
maintained.
(
Reactor Protection System Instrumentation
TR 3.3.1
TR 3.3
INSTRUMENTATION
(
TR 3.3.1
Reactor Protection System (RPS) Instrumentation
There shall be two OPERABLE or tripped trip systems with a
minimum of two OPERABLE instrument channels per trip system
for the Turbine First Stage Pressure Permissive. The pressure
switch allowable values shall be <154 psig.
APPLICABILITY:
~ 30% RTP (Turbine First Stage Pressure ~ 154 psig)
NOTE---------------------------
Required Actions shall be taken only if the permissive fails in such a manner to prevent
the affected RPS logic from performing its intended function. Otherwise, no action is
required.
ACTIONS
A.
CONDITION
One or more required
channels are
REQUIRED ACTION
NOTE----
Inoperable Turbine First Stage
Pressure Permissive channel(s)
or subsystem(s) may also affect
Technical Specifications LCO
3.3.1.1 and LCO 3.3.4.1.
COMPLETION TIME
A.1
Trip the inoperable
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
channel(s) or entire trip
system(s).
A.2.1
Initiate insertion of
OPERABLE rods.
AND
Immediately
A.2.2
Complete insertion of all
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
OPERABLE rods.
(continued)
l
BFN-UNIT 2
3.3-1
TRM Revision 0, 12
August 17, 1999
Reactor Protection System Instrumentation
TR 3.3.1
(
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION TIME
A.
(continued)
A.3
Reduce power to less
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
than 30 percent of rated.
NOTE,----------------
A channel may be placed in an INOPERABLE status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required
surveillance without placing the trip system in the tripped condition provided at least one
OPERABLE channel in the same trip system is monitoring that parameter.
TECHNICAL SURVEILLANCE REQUIREMENTS
TSR 3.3.1.1
TSR 3.3.1.2
BFN-UNIT 2
SURVEILLANCE
Perform CHANNEL FUNCTIONAL TEST.
Perform CHANNEL CALIBRATION.
3.3-2
FREQUENCY
92 days
24 months
TRM Revision 3, 12
August 17, 1999
TR 3.3
INSTRUMENTATION
(
TR 3.3.1
BASES
RPS Instrumentation
B 3.3.1
Reactor Protection System (RPS) Instrumentation
BACKGROUND
Fast closure of the TCVs results in the loss of a heat sink that
produces reactor pressure, neutron flux, and heat flux transients
that must be limited. Therefore, a reactor scram is initiated on TCV
fast closure in anticipation of the transients that would result from
the closure of these valves. The Turbine Control Valve Fast
Closure, Trip Oil Pressure -
Low Function is the primary scram
signal for the generator load rejection event without bypass valve
capability analyzed in the FSAR section 14.5. For this event, the
reactor scram reduces the amount of energy required to be
absorbed and, along with the ACTIONS of the EOC-RPT System,
ensures that the MCPR SL is not exceeded.
Turbine Control Valve Fast Closure, Trip Oil Pressure-Low
signals are initiated by the electrohydraulic control (EHC) fluid
pressure at each control valve. One pressure switch is associated
with each control valve, and the signal from each switch is
assigned to a separate RPS logic channel. This Function must be
enabled at THERMAL POWER ~ 30% RTP. This is normally
accomplished automatically by pressure transmitters sensing
turbine first stage pressure; therefore, opening the turbine bypass
valves may affect this function.
The Turbine Control Valve Fast Closure, Trip Oil Pressure-Low
Allowable Value is selected high enough to detect imminent TCV
fast closure.
Four channels of Turbine Control Valve Fast Closure, Trip Oil
Pressure -
Low Function with two channels in each trip system
arranged in a one-out-of-two logic are required to be OPERABLE
to ensure that no single instrument failure will preclude a scram
from this Function on a valid signal. This Function is required,
consistent with the analysis assumptions, whenever THERMAL
POWER is ~ 30% RTP. This Function is not required when
THERMAL POWER is < 30% RTP, since the Reactor Vessel
Steam Dome Pressure-High and the Average Power Range
Monitor Fixed Neutron Flux-High Functions are adequate to
maintain the necessary safety margins.
BFN-UNIT 2
B 3.3-1
~viSion o
(
BASES
BACKGROUND
(continued)
RPS Instrumentation
B3.3.1
Closure of the TSVs results in the loss of a heat sink that produces
reactor pressure, neutron flux, and heat flux transients that must be
limited. Therefore, a reactor scram is initiated at the start of TSV
closure in anticipation of the transients that would result from the
closure of these valves. The Turbine Stop Valve -
Closure
Function is the primary scram signal for the turbine trip event
analyzed in the FSAR section 14.5. For this event, the reactor
scram reduces the amount of energy required to be absorbed and,
along with the ACTIONS of the End of Cycle Recirculation Pump
Trip (EOC-RPT) System, ensures that the MCPR SL is not
exceeded.
Turbine Stop Valve-Closure signals are initiated from position
switches located on each of the four TSVs. Two independent
position switches are associated with each stop valve. One of the
two switches provides input to RPS trip system A; the other, to RPS
trip system B. Thus, each RPS trip system receives an input from
four Turbine Stop Valve-Closure channels, each consisting of
one position switch. The logic for the Turbine Stop Valve-
Closure Function is such that three or more TSVs must be closed
to produce a scram. This Function must be enabled at THERMAL
POWER ~ 30% ~TP. 1: .
's normally accomplished automati~
by.J~ressureJ!!nsmitters sensing turbine I
"Stage press~
f
t~.e-m
ct~fQre ,_oJ?enilliDtLe-tOrbln~ bypa~~!~ may a~~
un
Ion.
The Turbine Stop Valve-Closure Allowable Value is selected to
be high enough to detect imminent TSV closure, thereby reducing
the severity of the SUbsequent pressure transient.
Eight channels of Turbine Stop Valve-Closure Function, with four
channels in each trip system, are required to be OPERABLE to
ensure that no single instrument failure will preclude a scram from
this Function if any three TSVs should close. This Function is
required, consistent with analysis assumptions, whenever
THERMAL POWER is ~ 30% RTP. This Function is not required
when THERMAL POWER is < 30% RTP since the Reactor Vessel
Steam Dome Pressure-High and the Average Power Range
Monitor Fixed Neutron Flux- High Functions are adequate to
maintain the necessary safety margins.
BFN-UNIT 2
B 3.3-2
TRM Revision 0
(
RPS Instrumentation
B 3.3.1
BASES
APPLICABLE
This function must inhibit the automatic bypassing of turbine
SAFETY ANALYSIS
control valve fast closure or turbine trip scram and turbine stop
valve closure scram whenever turbine first stage pressure is
greater than or equal to 154 psig.
This in combination with the TSV's 10% closure limit switches and
the TCVs low trip oil pressure are required to prevent exceeding
APPLICABI L1TY
TRM LCO 3.3.1 requires that the 30% RTP sensed by the turbine
first stage pressure either be OPERABLE or tripped. These
pressure switches function to bypass the Scram for the TSVs and
TCVs. The switches sense greater than 30% RTP when in the
conservative condition (Le., will function to cause a Scram on
closure of the TCV's and TSVs). For the Turbine First Stage
Pressure Permission, an instrument channel consists of the
pressure switch assigned to that channel. Pressure switches
PI5-1-81A and PIS-1-91A are assigned to trip system Band
pressur~)witchesPIS-1-81Band PIS-1-91 B are assigned to trip
system A.
If these switches are tripped, the failure is in the conservative
direction.
'--...
~
r----
This Scram is only needed above 30% RTP as indicated by
~ 154 psig Turbine First Stage Pressure since adequate margin to
MCPR SL is assured for any type of turbine trip, with or without
bypass valves.
BFN-UNIT2
B 3.3-3
vision 0
BASES
ACTIONS
TECHNICAL
SURVEILLANCE
REQUIREMENTS
RPS Instrumentation
B 3.3.1
A note is provided to indicate that Required Actions are to be taken
only if the permissive fails in such a manner to prevent the affected
RPS logic from performing its intended function. Otherwise, no
action is required since the RPS function is maintained.
One hour to trip the inoperable channel(s) or trip system(s) is
reasonable and consistent with conservative operation with
degraded safety functions. Since these pressure switches do not
cause a Scram by themselves, but rather in combination with the
TSV's or TCVs, the switches may be tripped without causing a half
A.2.1 and A.2.2
Four hours to insert all OPERABLE control rods provides the same
end function as a Scram and is within conservative operating
considerations given the degraded safety function to maintain
Dropping below 30% RTP also provides for conservative operation
since any turbine trip below that power does not cause the MCPR
SL to be exceeded
TSR 3.3.1.1 and TSR 3.3.1.2
Functional test consists of the injection of a simulated signal into
the electronic trip circuitry in place of the sensor signal to verify
OPERABILITY of the trip and alarm functions.
Calibration consists of the adjustment of the primary sensor and
associated components so that they correspond within acceptable
range and accuracy to known values of the parameter which the
channel monitors, including adjustment of the electronic trip
circuitry, so that its output relay changes state at or more
conservatively than the analog equivalent of the trip level setting.
Surveillance requirement times are based on equipment reliability
and engineering judgment and conservatively set to provide
adequate assurance of safety function performance.
BFN-UNIT2
B 3.3-4
TRM Revision 0
BASES
REFERENCES
RPS Instrumentation
B 3.3.1
1.
BFN Technical Specifications (version prior to standardized
version)
2.
BFN-UNIT2
B 3.3-5
TRM Revision 0
}
E
MINATION
REFERENCE
( PROVIDED TO
CANDIDATE
Reactor Protection System Instrumentation
TR 3.3.1
TR 3.3
INSTRUMENTATION
(
TR 3.3.1
Reactor Protection System (RPS) Instrumentation
There shall be two OPERABLE or tripped trip systems with a
minimum of two OPERABLE instrument channels per trip system
for the Turbine First Stage Pressure Permissive. The pressure
switch allowable values shall be <154 psig.
APPLICABILITY:
~ 30% RTP (Turbine First Stage Pressure ~ 154 psig)
NOTE-------------------------------------------------------
Required Actions shall be taken only if the permissive fails in such a manner to prevent
the affected RPS logic from performing its intended function. Otherwise, no action is
required.
ACTIONS
A.
CONDITION
One or more required
channels are
REQUIRED ACTION
NOTE----
Inoperable Turbine First Stage
Pressure Permissive channel(s)
or subsystem(s) may also affect
Technical Specifications LCO
3.3.1.1 and LCO 3.3.4.1.
COMPLETION TIME
A.1
Trip the inoperable
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
channel(s) or entire trip
system(s).
A.2.1
Initiate insertion of
OPERABLE rods.
Immediately
A.2.2
Complete insertion of all
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
OPERABLE rods.
(continued)
BFN-UNIT 2
3.3-1
TRM Revision Q, 12
August 17, 1999
Reactor Protection System Instrumentation
TR 3.3.1
(
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION TIME
A.
(continued)
A.3
Reduce power to less
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
than 30 percent of rated.
- ---NOTE---
A channel may be placed in an INOPERABLE status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required
surveillance without placing the trip system in the tripped condition provided at least one
OPERABLE channel in the same trip system is monitoring that parameter.
._- -
TECHNICAL SURVEILLANCE REQUIREMENTS
TSR 3.3.1.1
TSR 3.3.1.2
BFN-UNIT2
SURVEILLANCE
Perform CHANNEL FUNCTIONAL TEST.
Perform CHANNEL CALIBRATION.
3.3-2
FREQUENCY
92 days
24 months
TRM Revision 8, 12
August 17, 1999
(
(
16. SRO 268000A2 .01 00 lIe/A/TIG2/0PL171.084//268000A2.0l//SRO ONLY/12/18/2007 RMS
Given the following plant conditions:
BFN is in the process of discharging the Waste Sample Tank to the river in accordance with an
approved Discharge Permit.
With the discharge in progress, the Radwaste operator calls the control room and reports that
0-RR-90-130 (Radwaste Effluent Radiation Monitor) is now reading below the initial background
radiation level prior to commencing the discharge.
Which ONE of the following describes the required action(s), if any, and the potential cause of this
indication?
A.
The discharge may continue. This is an expected indication during a discharge of low activity water
processed by the Thermex System.
B. The discharge may continue.
Discharging water surrounding the detector is acting to shield it from
background radiation .
C. Have Radwaste terminate the discharge. Suspect 0-RR-90-130 has not been properly calibrated
prior to commencing the discharge.
D~ Have Radwaste terminate the discharge. Suspect 0-RR-90-130 was not suspended in water when
background radiation levels were initially recorded.
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 demonstrate
knowledge of the consequences of a rupture or leak involving a Radwaste system or component. This
question is based on Operating Experience at BFN when a leaking drain valve caused 0-RR-90-130 to be
exposed to higher than normal background radiation levels prior to commencing a discharge.
References:
BFPER 971713 (level B) 10/28/97 on Radwaste effluent Radiation Monitor
Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall
or recognize discrete bits of information.
SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (4) Radiation
hazards that may arise during normal and abnormal situations, including maintenance activities and
various contamination conditions.
0610 NRC SRO Exam
(
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly, the candidate must determine the following :
1. Recall the event which occurred at BFN and the causal factors.
2. Determine the appropriate action if a similar event were to occur while they were performing the
function of Unit Supervisor.
A is incorrect. A discharge of low activity water will typically not provide any change in radiation levels
detected by O-RR-90-130. A reduction in radiation levels is not expected. This becomes plausible if the
candidate is not aware that O-RR-90-130 is immersed in water prior to and after a discharge.
B is incorrect. This becomes plausible if the candidate is not aware that O-RR-90-130 is immersed in
water prior to and after a discharge.
C is incorrect. Detector O-RR-90-130 is required to be calibrated prior to each discharge but verification
that the detector is properly immersed in water prior to taking background readings is NOT required. Even
so, improper calibration would not manifest itself by a reduction in radiation levels after the discharge was
begun. This is plausible because the appropriate action was taken to secure the discharge and this was,
in fact, the suspected cause during the actual event.
D is correct.
(
(c)
BFPER 971713 (level B) 10/28/97 on Radwaste effluent
Radiation Monitor.
i.
0-RR-90-130 indicated lower than background during
discharge release on 10/27/97 where fluctuations from
-1500 cps to 900 cps occurred.
ii.
Subsequent discharge releases on 10/27/97 did not
repeat the same phenomenon. However, RWoperators
reported similar activity during release on 10/21/97.
OPL 171.084
Revision 5
Page 37 of 1
(d)
Action taken
i.
Instrument Mechanics verified calibration of both O-RM-
90-130 and 0-RR-90-130. As found was correct.
ii.
Subsequent troubleshooting revealed a leaking drain
valve (0-DRV-077-0879) due to crud buildup on the valve
seat allowing the detector chamber to drain down.
(i)
Water normally contained inside the chamber provides
shielding to the detector from the background radiation
levels.
(ii)
During the period the chamber was empty, the detector
saw a higher than usual background level. When a
relatively "low" activity release occurred, the newly
provided shielding from the water being released was
actually lower in activity than the background level
previously seen by the detector.
iii.
0-DRV-077-0879 was changed out.
Which ONE of the following describes the required operator action?
17. SRO 271000G2.4.36 00 lIelA/TIG2IEPIP- l//271000G2.4.361ISRO ONLY/12/3107 RMS
I
I
A transient has occurred on Unit-1 resulting in the following annunciators in alarm
STACK GAS RADIATION HI (1-RA-90-147B)
-
STACK GAS RADIATION HIGH-HIGH (1-RA-90-147A)
OG PRETREATMENT RADIATION HIGH (1-RA-90-157A)
-
RX BLDG,TURB BLDG, RF ZONE EXH RADIATION HIGH (1-RA-90-250A)
I
Given the following plant conditions:
I
(
Notify Radcon and
_
REFERENCE PROVIDED
C~ have Chemistry take a coolant sample to verify fuel damage. EPIP entry is required due to EAL
1.4-U.
B. a coolant sample is NOT required to verify fuel damage . No EPIP entry is required until radcon
surveys are completed at the site boundry.
A.
have Chemistry take a coolant sample to verify fuel damage. No EPIP entry is required until sample
results are completed.
i
I
I
I
I
II
D. a coolant sample is NOT required to verify fuel damage. EPIP entry is required due to EAL 1.4-U.
L
_
KIA Statement:
271000 Off-gas
2.4.36 - Emergency Procedures 1Plan Knowledge of chemistry 1health physics tasks during emergency
operations
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine that Chemistry and Radcon support is required.
References:
ARPs for listed annunciators
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.
SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (4) Radiation
hazards that may arise during normal and abnormal situations, including maintenance activities and
various contamination conditions.
0610 NRC SRO Exam
(
(
(
REFERENCE PROVIDED: EPIP-1
Plausibility Analysis:
In order to answer this question correctly, the candidate must determine the following:
1. The commmon denominator in all the ARPs for the given alarms is to have Radcon and Chemistry
personnel respond to the event.
2. Recognize that EPIP-1 requires a radiochemical analysis to verify fuel damage PRIOR to making
any classification based on fuel failure.
3. Recognize that a valid Off-Gas pretreatment radiation high alarm requires an Unusual Event
declaration in accordance with EAL 1.4U.
A is incorrect. EPIP entry is required based on multiple radiation alarms eliminating the potential for a
single failed annunciator. Therefore, the OG Pretreatment Radiation High alarm is valid and EAL 1.4U is
required. This is plausible because a radiochemical analysis is required to verify fuel damage.
8 is incorrect. EPIP entry is required based on multiple radiation alarms eliminating the potential for a
single failed annunciator. Therefore, the OG Pretreatment Radiation High alarm is valid and EAL 1.4U is
required. In addition, a radiochemical analysis IS required to verify fuel damage. This is plausible because
several EALs within EPIP-1 require site boundry surveys to verify EAL entry cond itions.
C is correct.
D is incorrect. A radiochemical analys is IS required to verify fuel damage. This is plausible because the
OG Pretreatment Radiation High alarm is valid and EAL 1.4U is required .
BROWNS FERRY
EMERGENCY CLASSIFICATION PROCEDURE
EVENT CLASSIFICATION MATRIX
EPIP-1
MSL I OFFGAS
LOSS OF DECAY HEAT
RADIATION
REMOVAL
Description
1.4-U I
I
I
I
Valid MAIN STEAM LINE RADIATION HIGH-HIGH
alarm, RA-90-135C
Valid OG PRETREATMENT RADIATION HIGH
alarm, RA-90-157A.
OPERATING CONDITION:
Mode 1 or 2 or 3
I
Description
I
I
I
c:zc:
C/)c:>>r-
m<mz-I
I
I
I
I
1.5-A I
I
I
I
Reactor moderator temperature can NOT be
maintained below 2120 F whenever Technical
Specifications require Mode 4 conditions or during
operations in Mode 5.
>>r-m
- U
-I
OPERATING CONDITION:
Mode 4 or 5
I
I
I
I
1.5-S I CURVE I
I
I US
Suppression Pool temperature, level and RPV
C/)
pressure can NOT be maintained in the safe area
=i
of Curve 1.5-S.
m
,
m
~m
- U
G)
m
OPERATING CONDITION:
Zo
Mode 1 or 2 or 3
-<
I
I
I
I
I
I
I
I
G)
mzm
~r-
ms::m
- 0
G)
mzo-<
PAGE 22 OF 201
REVISION 42
BROWNS FERRY
EMERGENCY CLASSIFICATION PROCEDURE
TECHNICAL BASIS
EPlp*1
(
EAL:
MSL/OFFGAS RADIATION
1.4-U
UNUSUAL EVENT
Valid MAIN STEAM LINE RADIATION HIGH-HIGH alarm, RA-90-135C
Valid OG PRETREATMENT RADIATION HIGH alarm, RA-90-157A.
OPERATING CONDITION: Mode 1 or 2 or 3
BASIS:
REFERENCES:
NOTES:
CURVESITABLES:
Main Steam Line radiation high high or offgas radiation high is indicative of fuel
cladding leakage.
The Main Steam Line radiation high high alarm setpoint is normally set at 3 times
normal full power background. 3 times normal full power background is in excess of
any spikes expected from operational transients that do not result in cladding failure.
This alarm setpoint is substantially above that which would be indicative of fuel
cladding damage above Technical Specification allowable limits; however, the
presence
of
a
valid
alarm warrants
declaration
of
an
Unusual
Event and
consideration of other symptoms and event classifications for possible upgrade of the
event based on fission product barrier loss.
The offgas pretreatment radiation high alarm setpoint is set at a value that is
indicative of the ODCM allowable limits for radiation release.
Either of these conditions is considered a potential degradation in the level of safety
of the plant and a potential precursor of a more serious problem.
Escalation to the Alert is based on either Reactor coolant samples exceeding
300 IJCi/gm or Drywell radiation levels indicative of loss of the fuel cladding barrier.
Reg Guide 1.101 Rev. 3, (NUMARC-SU4 example-1)
(
PAGE 102 OF 201
REVISION 42
BROWNS FERRY
EMERGENCY CLASSIFICATION PROCEDURE
TECHNICAL BASIS
EPIP-1
(
EAL:
OTHER
8.4-A
ALERT
Events are in process or have occurred which involve an actual or potential
substantial degradation in the level of safety of the plant or a security event that
involves probable life threatening risk to site personnel or damage to site equipment
because of HOSTILE ACTION.
Any releases are expected to be limited to small
fractions of the EPA ProtectiveAction Guideline exposure levels.
Any loss or potential loss of fuel cladding or RCS pressure boundary.
OPERATING CONDITION: ALL
BASIS:
REFERENCES:
This event classification is intended to address conditions not explicitly addressed
elsewhere but that warrant declaration of an emergency because conditions exist
which are believed by the Site Emergency Director to fall under the Alert
classification.
BFN EALs were developed primarily utilizing the symptom based
grouping methodology. This approach is consistent with the BFN EOI methodology.
It is important to note here that the consideration of fission product barriers has been
incorporated within this symptom based approach. Barrier-based EALs refer to the
level of challenge to principal barriers used to assure containment of radioactive
material. For radioactive materials that are contained within the reactor core, these
barriers
are: fuel
cladding,
reactor
coolant
system
pressure
boundary, and
containment.
The level of challenge to these barriers encompasses the extent of
damage (loss or potential loss) and the number of barriers currently under challenge.
Site Emergency Directors should be continuously aware of all challenges to these
barriers and the number of barriers loss or potentially loss.
Also Site Emergency
Directors should consider that when the loss or potential loss thresholds is imminent
(Le., 1 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) use judgment and classify as if the thresholds are exceeded.
The threshold for fission product barrier loss is considered to be consistent with the
following:
Fuel clad - A Reactor coolant sample that yields a result of 300 j../Cilgm lodine-131
equivalent is indicative of cladding failure (Refer to 1.3-A).
RCS barrier - Reactor coolant leakage of at least 50 GPM from the primary system
(Refer to 2.4-A).
Reg Guide 1.101 Rev. 3, (NUMARC HA6, FA)
NRC Bulletin 2005-02, July 18, 2005 - Attachment 2 (Emergency Classification Level
changes)
NEI White Paper, "Enhancements to Emergency Preparedness Programs for Hostile
Action", May 2005 (Revised November 18, 2005)
PAGE 196 OF 201
REVISION42
OTHER
8.4-5
(
BROWNS FERRY
EMERGENCY CLASSIFICATION PROCEDURE
TECHNICAL BASIS
EPIP-1
EAL:
SITE AREA EMERGENCY
Events are in process or have occurred which involve actual or likely major failures of
plant functions needed for protection of the public or HOSTILE ACTION that results
in intentional damage or malicious acts (1) toward site personnel or equipment that
could lead to the likely failure thereof or, (2) prevent effective access to equipment
needed for protection of the public.
Any releases are not expected to result in
exposure levels which exceed EPA Protective Action Guideline exposure levels
beyond the site boundary.
Any loss or potential loss of both fuel cladding and RCS pressure boundary.
Potential loss of either fuel cladding or RCS pressure boundary and loss of any
additional barrier.
OPERATING CONDITION: ALL
BASIS:
This event classification is intended to address unanticipated conditions not explicitly
addressed elsewhere but that warrant declaration of an emergency because
conditions exist which are believed by the Site Emergency Director to warrant Site
Area Emergency classification.
BFN EALs were developed primarily utilizing the
symptom based grouping methodology.
This approach is consistent with the BFN
EOI methodology.
It is important to note here that the consideration of fission
product barriers has been incorporated within this symptom based approach.
Barrier-based EALs refer to the level of challenge to principal barriers used to assure
containment of radioactive material.
For radioactive materials that are contained
within the reactor core, these barriers are: fuel cladding, reactor coolant system
pressure boundary, and containment.
The level of challenge to these barriers
encompasses the extent of damage (loss or potential loss) and the number of
barriers currently under challenge. Site Emergency Directors should be continuously
aware of all challenges to these barriers and the number of barriers loss or potentially
loss. Also Site Emergency Directors should consider that when the loss or potential
loss thresholds is imminent (l.e., 1 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) use judgment and classify as if the
thresholds are exceeded.
Loss or potential loss of any two fission product barriers must be considered along
with inability to monitor fission product barriers during extreme conditions.
The
threshold for fission product barrier loss is considered to be consistent with the
following:
Fuel clad - A Reactor coolant sample that yields a result of 300 IJCi/gm lodine-131
equivalent is indicative of cladding failure (Refer to 1.3-A).
RCS barrier - Reactor coolant leakage of at least 50 GPM from the primary system
(Refer to 2.4-A).
Primary Containment barrier - Refer to 2.5-U.
(
PAGE 198 OF 201
REVISION 42
(
Unit 1
STACK GAS
RADIATION
HIGH
1-RA-90-147B
MA
~
(Page 1 of 1)
XA-55-3A
SensorlTrip Point:
0-RM-90-147B
0-RM-90-148B
0-RM-90-306
1-ARP-9-3A
Rev. 0036
Page 20 of 50
ill
11 ,948 CPS
11,948 CPS
5.57 X 10-2 mCi/cc
Sensor
Location:
O-RE-090-0147 and
O-RE-090-0148
EI 599'6", Pnl 25-39
inside stack
Probable
Cause:
Automatic
Action:
A. Source check.
B. Resin trap failure (RWCU or Cond Demin).
C. Possible fuel element failure.
D. Sensor malfunction.
None
Operator
Action:
A. CHECK alarm condition on the following:
1. WIDE RANGE GASEOUS EFFLUENT RADIATION MONITOR,
0-RM-90-306 on Panel 2-9-2.
2.
STACK GAS RADIATION, 0-RR-90-147 on Panel 1-9-2.
o
o
B. CHECK following radiation recorders on Panel 1-9-2 and associated
radiation monitors on Panel 1-9-10:
1. OFFGAS RADIATION, 1-RR-90-266
0
2.
OG POST-TREATMENT CH BRAD MaN RTMR,
1-RM-90-265A.
0
3.
OG POST-TREATMENT CH A RAD MaN RTMR,
1-RM-90-266A.
0
C. VERIFY dilution fan running and damper open by checking red light
illuminated above STACK DILUTION AIR FAN A (B), 1-HS-66-29A
(31A) on Panel 1-9-8.
0
D. VERIFY Charcoal Adsorbers in service.
0
E. NOTIFY RADCON and Shift Manager.
0
F. REQUEST Chemistry perform radiochemical analysis to determine
wu~e.
0
References:
1-47E620-3
0-SIMI-90B
0-47E610-90-4 & 20
c
Unit 1
STACK GAS
RADIATION
HIGH-HIGH
1-RA-90-147A
XA-55-3A
SensorlTrip Point:
0-RM-90-147B
0-RM-90-148B
Hi-Hi alarm from drawer
1-ARP-9-3A
Rev. 0036
Page 11 of 50
HI-HI
23,896 CPS
23,896 CPS
MA
~
0-RM-90-306
(Page 1 of 1)
As listed in 2-AOI-90-2
Sensor
Location:
Probable
Cause:
Automatic
Action:
EI 599'6", Panel 25-39. Inside stack.
A. Source check.
B. Resin trap failure (RWCU or Cond Demin).
C. Possible fuel element failure.
D. Sensor malfunction.
E. Off-gas flow abnormal.
None
Operator
Action:
A. CHECK alarm condition on the following:
1. WIDE RANGE GASEOUS EFFLUENT RADIATION MONITOR,
0-RM-90-306 on Panel 2-9-2.
2. STACK GAS RADIATION, 0-RR-90-147 on Panel 1-9-2.
o
o
B. CHECK following radiation recorders on Panel 1-9-2 and associated
radiation monitors on Panel 1-9-10:
1. OFFGAS RADIATION, 1-RR-90-266.
0
2.
OG POST-TREATMENT, CH BRAD MON RTMR,
1-RM-90-265A.
0
3. OG POST-TREATMENT CH A RAD MON RTMR,
1-RM-90-266A.
0
C. VERIFY dilution fan running and damper open by checking red light
illuminated above STACK DILUTION AIR FAN A (B), 1-HS-66-29A
(31A) on Panel 1-9-8.
0
D. VERIFY Charcoal Adsorbers in service.
0
E. NOTIFY Shift Manager and RADCON.
0
F. REQUEST Chemistry perform radiochemical analysis to determine
wu~e.
0
G. REFER TO ODCM.
0
H. REFER TO EPIP-1.
0
(
References:
1-45E620-3
0-47E610-90-4 & 20
1-729E814-2
0-SIMI-90B
(
Unit 1
OG PRETREATMENT
RADIATION
HIGH
1-RA-90-157A
MA
1'5
(Page 1 of 1)
XA-55-3A
Sensorffrip Point:
1-RM-090-0157
Hi alarm from drawer.
1-ARP-9-3A
Rev. 0036
Page 10 of 50
HI
5000 MR/HR
Sensor
Location:
Probable
Cause:
Automatic
Action:
Turb. Bldg, EI. 565', B-T3, 1-RE-090-0157, OG Pretreatment Sample Chamber.
A. High radiation in the pretreatment Off-Gas System .
B. Resin trap failure (RWCU or Condo Demin)
C. Possible fuel element failure.
None
Operator
Action:
A. VERIFY high radiation by checking recorder 1-RR-90-266
(Panel 1-9-2) and ratemeters 1-RM-90-157 (Panel 1-9-10).
B. NOTIFY RADCON.
C. CHECK off-gas flow normal.
D. CHECK Main Steam Line Radiation Recorder 1-RR-90-135
(Panel 1-9-2).
E. CHECK STACK GAS RADIATION, 0-RR-90-147.
F. REQUEST Chemistry perform radiochemical analysis to determine
source.
G. MONITOR off-gas release rate for ODCM compliance. Power
reduction may be required.
H. IF ODCM Limits are exceeded, THEN
REFER TO EPIP-1.
o
o
o
o
o
o
o
o
References:
0-47W600-77
729E814 Series
1-47E610-90-1
1-45E620-3
Unit 1
RX BLDG,TURB BLDG,
RF ZONE EXH
RADIATION HIGH
1-RA-90-250A
MA
~
(Page 1 of 1)
XA-55-3A
SensorlTrip Point:
1-RM-90-250
Gas
1-ARP-9-3A
Rev. 0036
Page 9 of 50
HIGH ALARM - 6594 CPM
ALERT - 3297 CPM
Sensor
Location:
Probable
Cause:
Automatic
Action:
Operator
Action:
EI 664' Refuel Floor R-4 P-Line
A. Daily source check .
B. High radiation in the Reactor Building, Turbine Building, Refuel Zone exhaust
ventilation ducts.
C. Dry Cask storage activities in progress.
None
A. CHECK 1-RM-90-250 on Panel 1-9-2 (O-MON-90-361) and
MONITOR activity levels on recorder AIR PARTICULATE MONITOR
CONTROLLER 1-MON-90-50 on Panel 1-9-2.
D
B. IF high activity is conformed, THEN
NOTIFY RADCON.
D
C. REQUEST Chemistry perform analysis to determine source.
D
D. IF Dry Cask storage activities are in progress, THEN
NOTIFY CASK Supervisor.
D
E. IF the TSC is NOT manned, THEN
EVACUATE personnel from affected areas .
D
F. IF the TSC is manned, THEN
REQUEST the TSC to evacuate unnecessary personnel from
affected areas.
D
G. MONITOR release rate for ODCM compliance.
D
H. IF ODCM Limits are exceeded, THEN
REFER TO EPIP-1.
D
I.
IF Eberline is operable, THEN
REFER TO 1-01-90, to reset alarms.
D
References:
0-47W600-80
1-SIMI-90B
1-47E610-90-1
TVA Calc NDQ00902005008/EDC63693
r
TENNESSEE VALLEY AUTHORITY
BROWNS FERRY NUCLEAR PLANT
EMERGENCY PLAN IMPLEMENTING PROCEDURE
EPIP-1
EMERGENCY CLASSIFICATION PROCEDURE
REVISION 42
PREPARED BY: RANDY WALDREP
PHONE: 2038
RESPONSIBLE ORGANIZATION: EMERGENCY PREPAREDNESS
APPROVED BY:
TONY ELMS
EFFECTIVE DATE: 04/06/2007
LEVEL OF USE: REFERENCE USE
QUALITY-RELATED
DATE: 04/06/2007
I
II(
18. SRO 288000A2.03 OO l/C/A/T2G2/0I-30//288000A2.03//SRO ONLY11211S/2007 RMS
Given the following Unit 3 conditions:
Unit-3 was at 100% rated power
A Loss of Coolant Accident occurred resulting in the following plant indications:
Reactor water level is +30 inches and steady with RCIC injecting.
Reactor pressure is 750 psig and lowering slowly .
Drywell pressure is 5.0 psig and rising slowly.
Reactor Zone exhaust radiation is 65 mR/hr.
Refuel Zone exhaust radiation is 4 mR/hr.
SGT trains A, 8 and C are running.
Which ONE of the following describes the status of Reactor Zone and Refuel Zone ventilation and any
corrective actions required for these conditions?
A.
Reactor and Refuel Zone ventilation systems are isolated. Perform 3-EOI Appendix 8E and restart
ventilation per 3-01-30A and 3-01-308.
B~ Reactor and Refuel Zone ventilation systems are isolated. Perform 3-EOI Append ix 8E and restart
ventilation per 3-EOI Appendix 8F.
C. Reactor Zone ventilation system is isolated. Refuel Zone ventilation is unaffected. Perform 3-EOI
Appendix 8E and restart ventilation per 3-01-308.
D. Reactor Zone ventilation system is isolated. Refuel Zone ventilation is unaffected. Perform 3-EOI
Appendix 8E and restart ventilation per 3-EOI Appendix 8F
KJA Statement:
288000 Plant Ventilation
A2.03 - Ability to (a) predict the impacts of the following on the PLANT VENTILATION SYSTEMS ; and (b)
based on those predictions, use procedures to correct, control, or mitigate the consequences of those
abnormal conditions or operations: Loss of coolant accident: Plant-Specific
KJA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determ ine ventilation system status and take the appropriate corrective actions.
References: 3-01-30A and 8. 3-EOI Append ix 8E & F
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.
SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment
of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency
situations.
0610 NRC SRO Exam
(
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly, the candidate must determine the following :
1. High drywell pressure causes an isolation of BOTH Reactor Zone and Refuel Zone ventilation .
2. 3-EOI Appendix BE is required to bypass a PCIS Group 6 isolation to allow restarting ventilation .
3. 3-EOI Appendix BF is required to restart ventilation following implementation of 3-EOI Appendix BE.
4. It is not appropriate to use 3-01-30A or 30B to restart ventilation once 3-EOI Appendix BE has been
implemented.
A is incorrect. 3-EOI Appendix BF is required to restart ventilation following implementation of 3-EOI
Appendix BE. It is not appropriate to use 3-01-30A or 30B. This is plausible because 3-01-30A and 30B
are functionally correct, but fail to address SGT trains in operation .
B is correct.
C is incorrect. High drywell pressure causes an isolation of BOTH Reactor Zone and Refuel Zone
ventilation. This is plausible due to confusion between differing isolation signals for each zone. For
example, refuel zone high radiation does not impact the reactor zone ventlation system. In addition, it is
not appropriate to use 3-01-30B to restart ventilation once 3-EOI Appendix BE has been implemented.
D is incorrect. High drywell pressure causes an isolation of BOTH Reactor Zone and Refuel Zone
ventilation. This is plausible due to confusion between differing isolation signals for each zone. For
example, refuel zone high radiation does not impact the reactor zone ventlation system.
Reactor Zone Ventilation System
3-01-30B
Unit 3
Rev. 0018
Page 7 of 34
(
3.0
PRECAUTIONS AND LIMITATIONS
A.
The Reactor Zone Ventilation System should be operated in a manner such that
no area will exceed 0.5 inches negative water pressure.
B.
A Work Order (WO) should be submitted to have Reactor Zone inlet filters
replaced when the differential pressure across the filters rises to 0.6 inch water.
C.
Refueling Zone supply and exhaust fans should be operating prior to startup of
Reactor Zone supply and exhaust fans to establish negative refuel zone
pressure prior to start of reactor zone fans.
D.
During cold weather operation, the Air Wash and Air Wash Pump B (Air Wash
Pump) and associated valves should be lined up in accordance with
0-GOI-200-1.
E.
Whenever primary containment is required , the Reactor Zone Ventilation
System should be restored to service as soon as possible following a Reactor
Building isolation to prevent a possible main steam line tunnel high temperature
Group I isolation.
F.
When the Reactor Zone Ventilation System is out of service and primary
containment is required, reference should be made to Technical Specification,
3.3.6.1 and 3-AOI-30B-1, Reactor Zone Ventilation Failure. Shift Manager
should be notified immediately.
G.
The Reactor Zone Supply and Exhaust Fans will auto trip due to any of the
following :
1.
Reactor Zone exhaust duct high radiation.
2.
Drywell high pressure.
3.
Reactor low water level.
4.
Reactor Zone high or low pressure.
H.
Reactor Zone supply and exhaust fans are alternated every six weeks.
Refuel Zone Ventilation System
3-01-30A
Unit 3
Rev. 0025
Page 7 of 33
3.0
PRECAUTIONS AND LIMITATIONS
A.
The Refueling Zone Ventilation System is required to be operated in a manner
such that the area pressure does not exceed 0.5 inches negative water
pressure.
B.
A WO should be submitted to have Refueling Zone inlet filters replaced when
the inlet DP rises to 0.6 inches H20.
C.
Refueling Zone supply and exhaust fans should be operating prior to startup of
Reactor Zone supply and exhaust fans.
D.
During cold weather operation, the air wash associated valves should be lined
up for freeze protection. REFER TO 0-GOI-200-1 .
E.
The Refueling Zone supply and exhaust fans automatically shut down due to
anyone of the following:
1.
Refueling Zone high radiation
2.
Drywell high pressure
3.
Reactor low water level
4.
Refueling Zone high or low pressure
5.
Reactor Zone high radiation
F.
The Refueling Ventilation System is required to be in the refuel mode prior to
refueling operations. REFER TO the following:
1.
Mechanical Maintenance should be requested to manually adjust the
normally closed dampers, REACTOR CAVITY EXH DAMPER,
3-DMP-064-0501, and DRYER AND SEP STORAGE POOL EXH
DAMPER , 3-DMP-064-0504, to obtain to obtain desired flow rate
REFER TO Section 8.3 and 3-TI-218, or
2.
With permission from the Shift Manager (or his designated alternate)
continuous operation of the Standby Gas Treatment System (SGT) taking
suction from the Refueling Floor will satisfy the requirement for the
ventilation system being in the Refuel Mode.
G.
Whenever a ventilation fan is placed in operation, the fan and motor should be
checked. REFER TO 0-GOI-300-1.
H.
Dual speed fans should be run in SLOW speed during the heating season and
FAST speed during the cooling season. SLOW speed reduces heating load
during cold weather and FAST speed maximizes cooling effect during hot
weather.
19. SRO GENERIC 2.1.12 00l/C/A/T3/RPS/2/212000G2.1.12/2.9/4.0/SRO ONLY/
Given the following plant conditions:
Unit 1 is operating at 1000h power.
Unit 2 is operating at 100
% power.
Unit 3 is refueling (MODE 5), fuel movement is in progress.
Recirc pump 3A suction line work is in progress and has the potential to drain the RPV.
The 3ED DIG is out of service for an inspection.
During performance of the monthly SGT Surveillance, B SGT fails to start.
Which ONE of the following describes the required actions per Technical Specifications?
REFERENCE PROVIDED
A.
Enter LCO 3.0.3 on Unit 1 and 2 immediately. Suspend Unit 3 fuel movement immediately.
B. ttl
Enter LCO 3.0.3 on Unit 1 and 2 in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Initiate actions to suspend OPDRVs immediately.
C.
Be in Mode 3 on Unit 1 and 2 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Start A and C SGT trains in 4
hours.
D.
Be in Mode 3 on Unit 1 and 2 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Initiate actions to suspend
OPDRVs immediately.
KIA Statement:
Conduct of Operations
2.1.12 Ability to apply technical specifications for a system
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions and times to correctly determine the required actions per Technical Specifications.
References: Unit-3 Tech Specs
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.
SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility
operating limitations in the technical specifications and their bases.
0610 NRC SRO Exam
REFERENCE PROVIDED: U3 Tech Spec Sections 3.6 and 3.8
Plausibility Analysis:
In order to answer this question correctly, the candidate must determine the following:
1. Recognize that "C" SGT is only OPERABLE with 3ED DIG INOP as long as "A" and "B" SGT remain
OPERABLE. (supported system)
2. Recognize that "B" SGT becoming INOP also impact Units 1 & 2.
3. Recognize that Tech Specs allows 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from discovery to declare supported systems INOP.
A is incorrect.
The supported system LCO is not applied for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Fuel movement on Unit-3 is only
suspended when 2 or more U3 DIGs are INOP.
B is correct.
C is incorrect. These actions are required for failure to meet the LCO for 1 SGT train inop. Starting "A"
and "C" SGT trains is only required if "B" SGT has been INOP for longer than 7 days with no other
supported system issues.
D is incorrect. These actions are required for failure to meet the LCO for 1 SGT train inop.
SGT System
3.6.4.3
3.6
CONTAINMENT SYSTEMS
3.6.4.3 Standby Gas Treatment (SGT) System
Three SGT subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3,
During operations with a potential for draining the reactor vessel
(OPDRVs).
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION
TIME
A. One SGT subsystem
A.1
Restore SGT subsystem
7 days
to OPERABLE status.
B. Required Action and
B.1
Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
associated Completion
Time of Condition A not
AND
met in MODE 1, 2, or 3.
B.2
Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
(continued)
BFN-UNIT 3
3.6-51
Amendment No. 24-2, 249
September 27, 2004
ACTIONS (continued)
SGT System
3.6.4.3
CONDITION
REQUIRED ACTION
COMPLETION
TIME
C. Required Action and
C.1
Place two OPERABLE
Immediately
associated Completion
SGT subsystems in
Time of Condition A not
operation.
met during OPDRVs.
C.2
Initiate action to suspend
Immediately
D. Two or three SGT
D.1
Enter LCO 3.0.3.
Immediately
subsystems inoperable in
MODE 1, 2, or 3.
(continued)
BFN-UNIT 3
3.6-52
Amendment No. 2-+2-, 249
September 27, 2004
ACTIONS (continued)
CONDITION
REQUIRED ACTION
SGT System
3.6.4.3
COMPLETION
TIME
E. Two or three SGT
subsystems inoperable
during OPDRVs.
E.1
Initiate action to suspend
Immediately
BFN-UNIT 3
3.6-53
Amendment No. ~, 249
September 27, 2004
AC Sources - Shutdown
3.8.2
3.8
ELECTRICAL POWER SYSTEMS
3.8.2
AC Sources - Shutdown
The following AC electrical power sources shall be OPERABLE:
a. One qualified circuit connected between the offsite transmission
network and the onsite Class 1E AC electrical power
distribution subsystem(s) required by LCO 3.8.8, "Distribution
Systems - Shutdown";
b. Two of the four Unit 3 diesel generators (DGs) each capable of
supplying one 4.16 kV shutdown board of the onsite Class 1E
AC electrical power distribution subsystem(s) required by
LCO 3.8.8, "Distribution Systems - Shutdown"; and
c.
Unit 1 and 2 DGs capable of supplying the Unit 1 and 2 4.16 kV
shutdown boards required by LCO 3.8.8.
APPLICABILITY:
MODES 4 and 5,
During movement of irradiated fuel assemblies in the secondary
containment.
BFN-UNIT 3
3.8-14
Amendment No. 212
ACTIONS
AC Sources - Shutdown
3.8.2
CONDITION
REQUIRED ACTION
COMPLETION
TIME
A. (continued)
A.2.3
Initiate action to suspend
Immediately
operations with a
potential for draining the
reactor vessel (OPDRVs).
AND
A.2.4
Initiate action to restore
Immediately
required offsite power
circuit to OPERABLE
status.
B. One or more required
B.1.1
Suspend CORE
Immediately
Unit 3 DGs inoperable.
ALTERATIONS.
AND
B.1.2
Suspend movement of
Immediately
irradiated fuel assemblies
in secondary
containment.
AND
B.1.3
Initiate action to suspend
Immediately
AND
B.1.4
Initiate action to restore
Immediately
required Unit 3 DGs to
OPERABLE status.
(continued)
BFN-UNIT 3
3.8-16
Amendment No. 212
ACTIONS (continued)
CONDITION
REQUIRED ACTION
AC Sources - Shutdown
3.8.2
COMPLETION
TIME
C. One or more required
Unit 1 and 2 DGs
C.1
Declare affected SGT and
30 days
CREV subsystem(s)
AND
Immediately from
discovery of
Condition C
concurrent with
inoperability of
redundant
required
feature(s)
BFN-UNIT 3
3.8-17
Amendment No. 212
(
SGT System
3.6.4.3
3.6
CONTAINMENT SYSTEMS
3.6.4.3
Standby Gas Treatment (SGT) System
Three SGT subsystems shall be OPERABLE.
APPLICABI LITY:
MODES 1, 2, and 3,
During operations with a potential for draining the reactor vessel
(OPDRVs).
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION
TIME
A. One SGT subsystem
A.1
Restore SGT subsystem
7 days
to OPERABLE status.
B. Required Action and
B.1
Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
associated Completion
Time of Condition A not
AND
met in MODE 1, 2, or 3.
B.2
Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
(continued)
BFN-UNIT 3
3.6-51
Amendment No. ~, 249
September 27, 2004
SGT System
3.6.4.3
ACTIONS (continued)
CONDITION
REQUIRED ACTION
COMPLETION
TIME
C. Required Action and
C.1
Place two OPERABLE
Immediately
associated Completion
SGT subsystems in
Time of Condition A not
operation.
met during OPDRVs.
C.2
Initiate action to suspend
Immediately
D. Two or three SGT
D.1
Enter LCO 3.0.3.
Immediately
subsystems inoperable in
MODE 1, 2, or 3.
(continued)
BFN-UNIT 3
3.6-52
Amendment No. ~, 249
September 27, 2004
ACTIONS (continued)
CONDITION
REQUIRED ACTION
SGT System
3.6.4.3
COMPLETION
TIME
E. Two or three SGT
subsystems inoperable
during OPDRVs.
E.1
Initiate action to suspend
Immediately
BFN-UNIT 3
3.6-53
Amendment No. ~, 249
September 27, 2004
AC Sources - Shutdown
3.8.2
3.8
ELECTRICAL POWER SYSTEMS
3.8.2
AC Sources - Shutdown
The following AC electrical power sources shall be OPERABLE:
a. One qualified circuit connected between the offsite transmission
network and the onsite Class 1E AC electrical power
distribution subsystem(s) required by LCO 3.8.8, "Distribution
Systems - Shutdown";
b. Two of the four Unit 3 diesel generators (DGs) each capable of
supplying one 4.16 kV shutdown board of the onsite Class 1E
AC electrical power distribution subsystem(s) required by
LCO 3.8.8, "Distribution Systems - Shutdown"; and
c.
Unit 1 and 2 DGs capable of supplying the Unit 1 and 2 4.16 kV
shutdown boards required by LCO 3.8.8.
APPLICABILITY:
MODES 4 and 5,
During movement of irradiated fuel assemblies in the secondary
containment.
BFN-UNIT 3
3.8-14
Amendment No. 212
ACTIONS
CONDITION
REQUIRED ACTION
AC Sources - Shutdown
3.8.2
COMPLETION
TIME
A. One required offsite
circuit inoperable.
NOTE----------------
Enter applicable Condition and
Required Actions of LCO 3.8.8,
with any required 4.16 kV
shutdown board not energized
from a qualified source as a
result of Condition A.
A.1
Declare affected required
Immediately
feature(s) with no
qualified offsite power
available inoperable.
A.2.1
Suspend CORE
ALTERATIONS.
Immediately
A.2.2
Suspend movement of
Immediately
irradiated fuel assemblies
in secondary
containment.
(continued)
BFN-UNIT 3
3.8-15
Amendment No. 212
ACTIONS
AC Sources - Shutdown
3.8.2
CONDITION
REQUIRED ACTION
COMPLETION
TIME
A. (continued)
A.2.3
Initiate action to suspend
Immediately
operations with a
potential for draining the
reactor vessel (OPDRVs).
AND
A.2.4
Initiate action to restore
Immediately
required offsite power
circuit to OPERABLE
status.
B. One or more required
B.1.1
Suspend CORE
Immediately
Unit 3 DGs inoperable.
ALTERATIONS.
AND
B.1.2
Suspend movement of
Immediately
irradiated fuel assemblies
in secondary
containment.
AND
B.1.3
Initiate action to suspend
Immediately
AND
B.1.4
Initiate action to restore
Immediately
required Unit 3 DGs to
OPERABLE status.
(continued)
BFN-UNIT 3
3.8-16
Amendment No. 212
ACTIONS (continued)
CONDITION
REQUIRED ACTION
AC Sources - Shutdown
3.8.2
COMPLETION
TIME
C. One or more required
Unit 1 and 2 DGs
C.1-
Declare affected SGT and
30 days
CREV subsystem(s)
AND
Immediately from
discovery of
Condition C
concurrent with
inoperability of
redundant
required
feature(s)
BFN-UNIT 3
3.8-17
Amendment No. 212
(
20. SRO GENERIC 2.2.22 OOl/CIA/TECH SPECS/HPCVlG2.2.22/4.l/SRO/lO/27/2007
Given the following plant conditions:
Unit 2 is operating at 100% power with HPCI tagged out at 0730 hours0.00845 days <br />0.203 hours <br />0.00121 weeks <br />2.77765e-4 months <br />.
It is determ ined that a control power fuse had blown on MSRV 1-23, after noticing a loss of
position indication on Panel 9-3 vertical section, also at 0730 hours0.00845 days <br />0.203 hours <br />0.00121 weeks <br />2.77765e-4 months <br />.
Which ONE of the following describes the minimum required actions imposed by Tech Specs?
REEFERENCE PROVIDED
A.
Restore both the MSRV and HPCI to operable within 14 days or be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and
MODE 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
B. Restore either the MSRV Q! HPCI to operable within 14 days or be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and
MODE 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
(
C.
Enter LCO 3.0.3 immediately.
D~ Restore HPCI in 14 days or be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor pressure to less than or
equal to 150 psig in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
KIA Statement:
Equipment Control
2.2.22 Knowledge of limiting conditions for operations and safety limits.
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
equipment conditions and times to correctly determine operability issues and Limiting Conditions for
Operation.
References: Unit-2 Tech Specs
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.
SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility
operating limitations in the technical specifications and their bases.
0610 NRC SRO Exam
(
(
REFERENCE PROVIDED: Unit-2 Tech Spec section 3.4 & 3.5
Plausibility Analysis:
In order to answer this question correctly, the candidate must:
1- recall from memory that MSRV 1-23 is NOT an ADS valve on Unit-2. (ref. pg 1)
2- correctly apply the applicability of TS 3.5.1 excepting HPCI from operability ~ 150 psig. (ref. pg 2)
3- recognize that only 12 of 13 MSRVs are required to be operable for other than ADS function . (ref. pg 6)
Answer A is incorrect because MSRV 1-23 is not a LCO per TS 3.4.3 and HPCI is excepted from MODE
4 requirements. It is plausible if the candidate fails to apply #2 and #3 above.
Answer B is incorrect because MSRV 1-23 is not a LCO per TS 3.4.3 and HPCI is excepted from MODE
4 requirements. It is plausible if the candidate fails to apply #1, #2 and #3 above.
Answer C is incorrect because MSRV 1-23 is NOT an ADS valve, therefore TS 3.5.1.H does not apply.
This is plausible if the candidate fails to apply #1 above.
Answer D is the correct answer.
(
S/RVs
3.4.3
3.4
3.4.3 Safety/Relief Valves (S/RVs)
The safety function of 12 S/RVs shall be OPERABLE.
APPLICABILITY:
MODES 1,2, and 3.
ACTIONS
(
CONDITION
A. One or more required
S/RVs inoperable.
BFN-UNIT 2
REQUIRED ACTION
A.1
Be in MODE 3.
A.2
Be in MODE 4.
3.4-7
COMPLETION
TIME
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
36 hours
Amendment 253
(
ECCS - Operating
3.5.1
3.5
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE
ISOLATION COOLING (RCIC) SYSTEM
3.5.1
ECCS - Operating
Each ECCS injection/spray subsystem and the Automatic
Depressurization System (ADS) function of six safety/relief valves
shall be OPERABLE.
APPLICABILITY:
MODE 1,
MODES 2 and 3, except high pressure coolant injection (HPCI) and
ADS valves are not required to be OPERABLE with reactor
steam dome pressu re s 150 psig.
ACTIONS
NOTE--------------------------------------------------
LCO 3.0.4.b is not applicable to HPCI.
CONDITION
A. One low pressure ECCS
A.1
injection/spray subsystem
One low pressure coolant
injection (LPCI) pump in
both LPCI subsystems
REQUIRED ACTION
Restore low pressure
ECCS injection/spray
subsystemts) to
OPERABLE status.
COMPLETION
TIME
7 days'?
(continued)
(1) _ This Completion Time may be extended to 14 days on a one-time basis. This temporary approval
expires June 1, 2005.
(
BFN-UNIT 2
3.5-1
Amendment No. 253, 269, 286, 294
May 9,2005
ACTIONS (continued)
CONDITION
REQUIRED ACTION
ECCS - Operating
3.5.1
COMPLETION
TIME
B. Required Action and
B.1
Be in MODE 3.
associated Completion
Time of Condition A not
AND
met.
B.2
Be in MODE 4.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
36 hours
(continued)
BFN-UNIT 2
3.5-1a
Amendment No. 286
December 1, 2003
(
ECCS - Operating
3.5.1
ACTIONS (continued)
CONDITION
REQUIRED ACTION
COMPLETION
TIME
C. HPCI System inoperable.
C.1
Verify by administrative
Immediately
means RCIC System is
AND
C.2
Restore HPCI System to
14 days
OPERABLE status.
D. HPCI System inoperable.
D.1
Restore HPCI System to
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
OPERABLE status.
AND
Condition A entered.
D.2
Restore low pressure
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
ECCS injection/spray
subsystem to OPERABLE
status.
E. One ADS valve
E.1
Restore ADS valve to
14 days
OPERABLE status.
F. One ADS valve
F.1
Restore ADS valve to
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
OPERABLE status.
AND
Condition A entered.
F.2
Restore low pressure
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
ECCS injection/spray
subsystem to OPERABLE
status.
(continued)
BFN-UNIT 2
3.5-2
Amendment No. ass, 269
March 12, 2001
(
ECCS - Operating
3.5.1
ACTIONS (continued)
CONDITION
REQUIRED ACTION
COMPLETION
TIME
G. Two or more ADS valves
G.1
Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
AND
G.2
Reduce reactor steam
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
Required Action and
dome pressure to
associated Completion
~ 150 psig.
Time of Condition C, D,
E, or F not met.
H. Two or more low pressure
H.1
Enter LCO 3.0.3.
Immediately
ECCS injection/spray
subsystems inoperable
for reasons other than
Condition A.
HPCI System and one or
more ADS valves
BFN-UNIT 2
3.5-3
Amendment No. ~ 269
March 12, 2001
E
MINATION
REFERENCE
(,PROVIDED TO
CANDIDATE
(
S/RVs
3.4.3
3.4
3.4.3 Safety/Relief Valves (S/RVs)
The safety function of 12 S/RVs shall be OPERABLE.
APPLICABILITY:
MODES 1,2, and 3.
ACTIONS
(
CONDITION
A. One or more required
S/RVs inoperable.
BFN-UNIT 2
REQUIRED ACTION
A.1
Be in MODE 3.
A.2
Be in MODE 4.
3.4-7
COMPLETION
TIME
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
36 hours
Amendment 253
(
ECCS - Operating
3.5.1
3.5
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE
ISOLATION COOLING (RCIC) SYSTEM
3.5.1
ECCS - Operating
Each ECCS injection/spray subsystem and the Automatic
Depressurization System (ADS) function of six safety/relief valves
shall be OPERABLE.
APPLICABILITY:
MODE 1,
MODES 2 and 3, except high pressure coolant injection (HPCI) and
ADS valves are not required to be OPERABLE with reactor
steam dome pressure s 150 psig.
ACTIONS
NOTE--------------------------------------------------
LCO 3.0.4.b is not applicable to HPCI.
CONDITION
A. One low pressure ECCS
A.1
injection/spray subsystem
One low pressure coolant
injection (LPCI) pump in
both LPCI subsystems
REQUIRED ACTION
Restore low pressure
ECCS injection/spray
subsystem(s) to
OPERABLE status.
COMPLETION
TIME
7 days'"
(continued)
(1) _ This Completion Time may be extended to 14 days on a one-time basis. This temporary approval
expires June 1, 2005.
BFN-UNIT 2
3.5-1
Amendment No. 253, 269, 286, 294
May 9,2005
c
ACTIONS (continued)
CONDITION
REQUIRED ACTION
ECCS - Operating
3.5.1
COMPLETION
TIME
B. Required Action and
8.1
Be in MODE 3.
associated Completion
Time of Condition A not
AND
met.
8.2
Be in MODE 4.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
36 hours
(continued)
(
BFN-UNIT 2
3.5-1 a
Amendment No. 286
December 1, 2003
(
ECCS - Operating
3.5.1
ACTIONS (continued)
CONDITION
REQUIRED ACTION
COMPLETION
TIME
C. HPCI System inoperable.
C.1
Verify by administrative
Immediately
means RCIC System is
AND
C.2
Restore HPCI System to
14 days
OPERABLE status .
D. HPCI System inoperable.
D.1
Restore HPCI System to
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
OPERABLE status.
AND
Condition A entered.
D.2
Restore low pressure
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
ECCS injection/spray
subsystem to OPERABLE
status .
E. One ADS valve
E.1
Restore ADS valve to
14 days
OPERABLE status.
F. One ADS valve
F.1
Restore ADS valve to
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
OPERABLE status .
AND
Condition A entered.
F.2
Restore low pressure
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
ECCS injection/spray
subsystem to OPERABLE
status.
(continued)
(
BFN-UNIT 2
3.5-2
Amendment No. 2-&d, 269
March 12, 2001
(
ECCS - Operating
3.5.1
ACTIONS (continued)
CONDITION
REQUIRED ACTION
COMPLETION
TIME
G. Two or more ADS valves
G.1
Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
AND
G.2
Reduce reactor steam
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
Required Action and
dome pressure to
associated Completion
- s; 150 psig.
Time of Condition C, D,
E, or F not met.
H. Two or more low pressure
H.1
Enter LCO 3.0.3.
Immediately
ECCS injection/spray
subsystems inoperable
for reasons other than
Condition A.
HPCI System and one or
more ADS valves
BFN-UNIT 2
3.5-3
Amendment No. ~ 269
March 12, 2001
r
(
21. SRO GENERIC 2.2.24 OOl/CIA/T3/23112/GEN2.2.24/2.5/3.7/SRO/lO/27/07
Given the following plant conditions:
Unit 1 in Mode 5, initial fuel load in progress .
Unit 2 is 100% RTP.
Unit 3 is in Mode 5, CRD drive replacement in progress after 1st portion of fuel moves.
81 RHRSW pump is tagged for impeller replacement.
2C RHR Heat Exchanger tagged for eddy current testing.
The Outside AUO reports that both D RHRSW pump room sump pumps have failed to start
and water level is above the grate of the room sump.
Which ONE of the following describes the required actions for Unit 2?
REFERENCE PROVIDED
A.,;
7 day LCO for Suppression Pool cooling, Suppression Chamber sprays and Drywell sprays
30 day LCO for RHRSW system and Ultimate Heat Sink
B.
30 day LCO for Suppression Pool cooling, Suppression Chamber sprays and Drywell sprays
8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> LCO for RHRSW system and Ultimate Heat Sink
C.
7 day LCO for Suppression Pool cooling, Suppression Chamber sprays and Drywell sprays
8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> LCO for RHRSW system and Ultimate Heat Sink
D.
30 day LCO for Suppression Pool cooling, Suppression Chamber sprays and Drywell sprays
30 day LCO for RHRSW system and Ultimate Heat Sink
KIA Statement:
Equipment Control
2.2.24
Ability to analyze the affect of maintenance activities on LCO status.
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to correctly determ ine the LCO status resulting from maintenance activities .
References: U2 TSR Section 3.6 and 3.7
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.
SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility
operating limitations in the technical specifications and their bases.
0610 NRC SRO Exam
REFERENCE PROVIDED: U2 Tech Spec Section 3.6 and 3.7
Plausibility Analysis:
In order to answer this question correctly the candidate must know the following :
-
The 0 RHRSW Pump Room sumps result in 01 , 02 and 03 RHRSW Pumps being inoperable per
Tech Spec Bases.
-
RHRSW flow through the RHR HX is required for SP Cooling, Spray and OW Spray to be operable per
Tech Spec Bases. This leads to 2 RHRSW subsystems INOP for Unit 2.
-
Which RHRSW pumps provide flow to each RHRHX. (system knowledge)
-
Determine the LCO for RHRSW pumps, RHRSW subsystem, SP Cooling, SP Spray, and OW Spray.
A is correct.
B is incorrect. A 30 day LCO is required if only one RHRSW subsystem is INOP. In addition, the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
LCO for RHRSW system applies if 2 units are fueled. (six pumps required) Initial conditions state that only
one unit is fueled.
C is incorrect. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> LCO for RHRSW system applies if 2 units are fueled . (six pumps required)
Initial conditions state that only one unit is fueled.
D is incorrect. A 30 day LCO is required if only one RHRSW subsystem is INOP.
(
RHR Suppression Pool Cooling
3.6.2.3
3.6
CONTAINMENT SYSTEMS
3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling
Four RHR suppression pool cooling subsystems shall be
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION
TIME
A. One RHR suppression
A.1
Restore the RHR
30 days
pool cooling subsystem
suppression pool cooling
subsystem to OPERABLE
status.
B. Two RHR suppression
B.1
Restore one RHR
7 days
pool cooling subsystems
suppression pool cooling
subsystem to OPERABLE
status.
C. Three or more RHR
C.1
Restore required RHR
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
suppression pool cooling
suppression pool cooling
subsystems inoperable.
subsystems to
OPERABLE status.
(continued)
(
BFN-UNIT 2
3.6-31
Amendment No. ~272
June 8,2001
(
RHR Suppression Pool Cooling
3.6.2.3
ACTIONS (continued)
CONDITION
D. Required Action and
associated Completion
Time not met.
REQUIRED ACTION
D.1
Be in MODE 3.
AND
D.2
Be in MODE 4.
COMPLETION
TIME
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
36 hours
BFN-UNIT 2
3.6-32
Amendment No. ~272
June 8,2001
(
RHR Suppression Pool Cooling
3.6.2.3
SURVEILLANCE REQUIREMENTS
SR 3.6.2.3.2
SURVEILLANCE
Verify each RHR suppression pool cooling
subsystem manual, power operated , and
automatic valve in the flow path that is not
locked, sealed, or otherwise secured in
position is in the correct position or can be
aligned to the correct position.
Verify each RHR pump develops a flow rate
2 9000 gpm through the associated heat
exchanger while operating in the suppression
pool cooling mode.
FREQUENCY
31 days
In accordance
with the Inservice
Testing Program
(
BFN-UNIT 2
3.6-33
Amendment No. 253
(
RHR Suppression Pool Spray
3.6.2.4
3.6
CONTAINMENT SYSTEMS
3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray
Four RHR suppression pool spray subsystems shall be
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION
TIME
A, One RHR suppression
A,1
Restore the RHR
30 days
pool spray subsystem
suppression pool spray
subsystem to OPERABLE
status.
B. Two RHR suppression
B.1
Restore one RHR
7 days
pool spray subsystems
suppression pool spray
subsystem to OPERABLE
status.
C. Three or more RHR
C.1
Restore required RHR
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
suppression pool spray
suppression pool spray
subsystems inoperable.
subsystems to
OPERABLE status.
D. Required Action and
D.1
Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
associated Completion
Time not met.
AND
D.2
Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
BFN-UNIT 2
3.6-34
Amendment No. 253
(
RHR Suppression Pool Spray
3.6.2.4
SURVEILLANCE REQUIREMENTS
SURVEILLANCE
FREQUENCY
SR 3.6.2.4.2
Verify each RHR suppression pool spray
31 days
subsystem manual, power operated, and
automatic valve in the flow path that is not
locked, sealed, or otherwise secured in
position is in the correct position or can be
aligned to the correct position.
Verify each suppression pool spray nozzle is
5 years
unobstructed.
(
BFN-UNIT 2
3.6-35
Amendment No. 253
(
RHR Drywell Spray
3.6.2.5
3.6
CONTAINMENT SYSTEMS
3.6.2.5 Residual Heat Removal (RHR) Drywell Spray
Four RHR drywell spray subsystems shall be OPERABLE.
APPLICABI L1TY:
MODES 1, 2, and 3.
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION
TIME
A. One RHR drywell spray
A.1
Restore the RHR drywell
30 days
subsystem inoperable.
spray subsystem to
OPERABLE status.
B. Two RHR drywell spray
B.1
Restore one RHR drywell
7 days
subsystems inoperable.
spray subsystem to
OPERABLE status.
C. Three or more RHR
C.1
Restore required RHR
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
drywell spray subsystems
drywell spray subsystems
to OPERABLE status.
D. Required Action and
D.1
Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
associated Completion
Time not met.
AND
D.2
Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
BFN-UNIT 2
3.6-36
Amendment No. 253
(
RHR Drywell Spray
3.6.2.5
SURVEILLANCE REQUIREMENTS
SURVEILLANCE
FREQUENCY
SR 3.6.2.5.2
Verify each RHR drywell spray subsystem
31 days
manual, power operated, and automatic valve
in the flow path that is not locked, sealed , or
otherwise secured in position is in the correct
position or can be aligned to the correct
position.
Verify each drywell spray nozzle is
5 years
unobstructed.
BFN-UNIT 2
3.6-37
Amendment No. 253
(
3.7.1
3.7
PLANT SYSTEMS
3.7.1
Residual Heat Removal Service Water (RHRSW) System and Ultimate Heat
Sink (UHS)
NOTE----------------------------------------
The number of required RHRSW pumps may be reduced by one
for each fueled unit that has been in MODE 4 or 5 for ~ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Four RHRSW subsystems and UHS shall be OPERABLE with the
number of OPERABLE pumps as listed below:
1. 1 unit fueled - four OPERABLE RHRSW pumps.
2. 2 units fueled - six OPERABLE RHRSW pumps.
3. 3 units fueled - eight OPERABLE RHRSW pumps.
(
APPLICABILITY:
MODES 1, 2, and 3.
BFN-UNIT 2
3.7-1
Amendment No. 254
September 08, 1998
(
ACTIONS
CONDITION
REQUIRED ACTION
3.7.1
COMPLETION
TIME
A. One required RHRSW
pump inoperable.
A.1
NOTES-----------
1.
Only applicable for the
2 units fueled
condition.
2.
Only four RHRSW
pumps powered from
a separate 4 kV
shutdown board are
required to be
OPERABLE if the
other fueled unit has
been in MODE 4 or 5
for ~ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Verify five RHRSW
pumps powered from
separate 4 kV shutdown
boards are OPERABLE.
Immediately
A.2
Restore required RHRSW
30 days
pump to OPERABLE
status.
(continued)
(
BFN-UNIT 2
3.7-2
Amendment No. 254
September 08, 1998
(
3.7.1
ACTIONS (continued)
CONDITION
REQUIRED ACTION
COMPLETION
TIME
B. One RHRSW subsystem
8.1
NOTE-------------
Enter applicable
Conditions and Required
Actions of LCO 3.4.7,
- Hot Shutdown," for RHR
shutdown cooling made
inoperable by the
RHRSW system.
Restore RHRSW
30 days
subsystem to OPERABLE
status.
C. Two required RHRSW
C.1
Restore one inoperable
7 days
pumps inoperable.
RHRSW pump to
OPERABLE status.
D. Two RHRSW subsystems
D.1
NOTE-------------
Enter applicable
Conditions and Required
Actions of LCO 3.4.7, for
made inoperable by the
RHRSW System.
Restore one RHRSW
7 days
subsystem to OPERABLE
status.
(continued)
BFN-UNIT 2
3.7-3
Amendment No. 254
September 08, 1998
c.
3.7.1
ACTIONS (continued)
CONDITION
REQUIRED ACTION
COMPLETION
TIME
E. Three or more required
E.1
Restore one RHRSW
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
RHRSW pumps
pump to OPERABLE
status.
F. Three or more RHRSW
F.1
NOTE-------------
subsystems inoperable.
Enter applicable
Conditions and Required
Actions of LCO 3.4.7 for
made inoperable by the
RHRSW System.
Restore one RHRSW
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
subsystem to OPERABLE
status.
G. Required Action and
G.1
Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
associated Completion
Time not met.
AND
G.2
Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
BFN-UNIT 2
3.7-4
Amendment No. 254
September 08, 1998
)
E
MINATION
REFERENCE
.(;PROVIDED TO
CANDIDATE
(
RHR Suppression Pool Cooling
3.6.2.3
3.6
CONTAINMENT SYSTEMS
3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling
Four RHR suppression pool cooling subsystems shall be
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION
TIME
A. One RHR suppression
A.1
Restore the RHR
30 days
pool cooling subsystem
suppression pool cooling
subsystem to OPERABLE
status.
B. Two RHR suppression
B.1
Restore one RHR
7 days
pool cooling subsystems
suppression pool cooling
subsystem to OPERABLE
status.
C. Three or more RHR
C.1
Restore required RHR
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
suppression pool cooling
suppression pool cooling
subsystems inoperable.
subsystems to
OPERABLE status.
(continued)
(
BFN-UNIT 2
3.6-31
Amendment No. ~272
June 8,2001
(
RHR Suppression Pool Cooling
3.6.2.3
ACTIONS (continued)
CONDITION
D. Required Action and
associated Completion
Time not met.
REQUIRED ACTION
0 .1
Be in MODE 3.
AND
0 .2
Be in MODE 4.
COMPLETION
TIME
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
36 hours
BFN-UNIT 2
3.6-32
Amendment No. ~272
June 8,2001
(
RHR Suppression Pool Cooling
3.6.2.3
SURVEILLANCE REQUIREMENTS
SR 3.6.2.3 .1
BFN-UNIT 2
SURVEILLANCE
Verify each RHR suppression pool cooling
subsystem manual, power operated, and
automatic valve in the flow path that is not
locked, sealed, or otherwise secured in
position is in the correct position or can be
aligned to the correct position.
Verify each RHR pump develops a flow rate
~ 9000 gpm through the associated heat
exchanger while operating in the suppression
pool cooling mode.
3.6-33
FREQUENCY
31 days
In accordance
with the Inservice
Testing Program
Amendment No. 253
(
RHR Suppression Pool Spray
3.6.2.4
3.6
CONTAINMENT SYSTEMS
3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray
Four RHR suppression pool spray subsystems shall be
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION
TIME
A. One RHR suppression
A.1
Restore the RHR
30 days
pool spray subsystem
suppression pool spray
subsystem to OPERABLE
status.
B. Two RHR suppression
B.1
Restore one RHR
7 days
pool spray subsystems
suppression pool spray
subsystem to OPERABLE
status.
C. Three or more RHR
C.1
Restore required RHR
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
suppression pool spray
suppression pool spray
subsystems inoperable.
subsystems to
OPERABLE status.
D. Required Action and
D.1
Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
associated Completion
Time not met.
AND
D.2
Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
(
BFN-UNIT 2
3.6-34
Amendment No. 253
RHR Suppression Pool Spray
3.6.2.4
SURVEILLANCE REQUIREMENTS
SURVEILLANCE
FREQUENCY
SR 3.6.2.4.2
BFN-UNIT 2
Verify each RHR suppression pool spray
31 days
subsystem manual, power operated, and
automatic valve in the flow path that is not
locked, sealed, or otherwise secured in
position is in the correct position or can be
aligned to the correct position.
Verify each suppression pool spray nozzle is
5 years
unobstructed.
3.6-35
Amendment No. 253
(
RHR Drywell Spray
3.6.2.5
3.6
CONTAINMENT SYSTEMS
3.6.2.5 Residual Heat Removal (RHR) Drywell Spray
Four RHR drywell spray subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1,2, and 3.
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION
TIME
A. One RHR drywell spray
A.1
Restore the RHR drywell
30 days
subsystem inoperable.
spray subsystem to
OPERABLE status.
B. Two RHR drywell spray
B.1
Restore one RHR drywell
7 days
subsystems inoperable.
spray subsystem to
OPERABLE status.
C. Three or more RHR
C.1
Restore required RHR
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
drywell spray subsystems
drywell spray subsystems
to OPERABLE status.
D. Required Action and
D.1
Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
associated Completion
Time not met.
AND
D.2
Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
(
BFN-UNIT 2
3.6-36
Amendment No. 253
(
RHR Drywell Spray
3.6.2.5
SURVEILLANCE REQUIREMENTS
SURVEILLANCE
FREQUENCY
SR 3.6.2.5.2
Verify each RHR drywell spray subsystem
31 days
manual, power operated, and automatic valve
in the flow path that is not locked, sealed, or
otherwise secured in position is in the correct
position or can be aligned to the correct
position.
Verify each drywell spray nozzle is
5 years
unobstructed.
(
BFN-UNIT 2
3.6-37
Amendment No. 253
(
3.7.1
3.7
PLANT SYSTEMS
3.7.1
Residual Heat Removal Service Water (RHRSW) System and Ultimate Heat
Sink (UHS)
NOTE----------------------------------------
The number of required RHRSW pumps may be reduced by one
for each fueled unit that has been in MODE 4 or 5 for ~ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Four RHRSW subsystems and UHS shall be OPERABLE with the
number of OPERABLE pumps as listed below:
1. 1 unit fueled - four OPERABLE RHRSW pumps.
2. 2 units fueled - six OPERABLE RHRSW pumps.
3. 3 units fueled - eight OPERABLE RHRSW pumps.
APPLICABILITY:
MODES 1, 2, and 3.
BFN-UNIT 2
3.7-1
Amendment No. 254
September 08, 1998
(
ACTIONS
CONDITION
REQUIRED ACTION
3.7.1
COMPLETION
TIME
A. One required RHRSW
pump inoperable.
A.1
NOTES-----------
1.
Only applicable for the
2 units fueled
condition.
2.
Only four RHRSW
pumps powered from
a separate 4 kV
shutdown board are
required to be
OPERABLE if the
other fueled unit has
been in MODE 4 or 5
for ~ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Verify five RHRSW
pumps powered from
separate 4 kV shutdown
boards are OPERABLE.
Immediately
A.2
Restore required RHRSW
30 days
pump to OPERABLE
status .
(continued)
(
BFN-UNIT 2
3.7-2
Amendment No. 254
September 08, 1998
(
3.7.1
ACTIONS (continued)
CONDITION
REQUIRED ACTION
COMPLETION
TIME
B. One RHRSW subsystem
B.1
NOTE-------------
Enter applicable
Conditions and Required
Actions of LCO 3.4.7,
- Hot Shutdown," for RHR
shutdown cooling made
inoperable by the
RHRSW system.
Restore RHRSW
30 days
subsystem to OPERABLE
status.
C. Two required RHRSW
C.1
Restore one inoperable
7 days
pumps inoperable.
RHRSW pump to
OPERABLE status.
D. Two RHRSW subsystems
D.1
NOTE-------------
Enter applicable
Conditions and Required
Actions of LCO 3.4.7, for
made inoperable by the
RHRSW System.
Restore one RHRSW
7 days
subsystem to OPERABLE
status.
(continued)
(
BFN-UNIT 2
3.7-3
Amendment No. 254
September 08, 1998
3.7.1
ACTIONS (continued)
CONDITION
REQUIRED ACTION
COMPLETION
TIME
E. Three or more required
E.1
Restore one RHRSW
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
RHRSWpumps
pump to OPERABLE
status.
F. Three or more RHRSW
F.1
NOTE-------------
subsystems inoperable.
Enter applicable
Conditions and Required
Actions of LCO 3.4.7 for
made inoperable by the
RHRSW System.
Restore one RHRSW
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
subsystem to OPERABLE
status.
G. Required Action and
G.1
Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
associated Completion
Time not met.
AND
G.2
Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
BFN-UNIT 2
3.7-4
Amendment No. 254
September 08, 1998
22. SRO GENERIC 2.3.3 OOl IMEMlT3111GENERIC 2.3.3IISRO ONLY/ll/27/07 RMS
During performance of 0-SR-DCS3.1 .2.1, Spent Fuel Storage Inspection, you receive a report that a pile l
of leaves and other debris has been found at the base of Overpack BFN-0-CASK-079-01 00/8 .
(
Wh ich ONE of the following describes the required action(s)?
I
I;
A.
Notify Maintenance and Modifications Management within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. No further action is required.
B.
Notify the Maintenance Shift Manager to remove the debris and clear the blockage. Note that the
blockage was found in the Post Test Remarks.
C. Contact Facilities Management to remove the debris and clear the blockage.
No notations are
required since the blockage was cleared .
D~ Coordinate with Radiation Protection to remove the debris and clear the blockage. Note that the
blockage was found and cleared in the Post Test Remarks.
KIA Statement:
Radiation Control
2.3.3 Knowledge of SRO responsibilities for auxiliary systems that are outside the control room (e.g.,
waste disposal and handling systems).
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the required actions following discovery of a loss of radioactive material
control.
References: O-SR-DCS 3.1.2.1
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.
SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (4) Radiation
hazards that may arise during normal and abnormal situations, including maintenance activities and
various contam ination conditions.
0610 NRC SRO Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
(
In order to answer this question correctly, the candidate must:
1. Determine the appropriate action to take regarding the blockage.
2. Determine the documentation required based on Item 1.
A is incorrect. Maintenance and Modifications Management is only notified if the debris cannot be
cleared. This is plausible becuase the notification time required is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
B is incorrect. The Maintenance Shift Manager is only contacted if the blockage CANNOT be removed.
This is plausible because the documentation requirements are correct.
C is incorrect. The Facilities Department is not responsible for HI-STORM debris removal. This is
plausible since Facilities is responsible for general cleanliness inside the protected area EXCEPT the
ISFSI Pad area.
D is correct.
(
Spent Fuel Storage Inspection
0-SR-DCS3.1.2.1
Unit 0
Rev. 0004
Page 7 of 10
Date
7.2
Single HI-STORM Inspection Prior to Placement on the ISFSI
PAD: (continued)
[4]
IF the HI-STORM(s) located on the ISFI Pad will not be
inspected during the performance of this surveillance, THEN
MARK the HI-STORM(s) listed on Attachment 2 as N/A.
(Otherwise N/A this step.)
7.3
Inspection of HI-STORM(s) Located on the ISFSI Pad
[1]
PERFORM Attachment 2, HI-STORM Inspection Log.
[2]
IF any Inlet or Outlet Vents are found to have blockage, THEN
PERFORM the following: (Otherwise N/A this section.)
[2.1]
NOTIFY the Unit Supervisor which HI-STORM
ventilation ducts have blockage.
[2.2]
IF the blockage can be readily removed, THEN
PERFORM the following: (Otherwise N/A)
A..ffiIIF
nd COORDINATE wit~R-a"':"d-ia-ti-o--n'
Protectio
for debris removal.
.~
B.
REMOVE the blockage and debris from associated
HI-STORM(s).
C.
RECORD the HI-STORM UNID and specific vent(s)
that were blocked and cleared in the Narrative Log
for all HI-STORM(s) with blockage.
(
Spent Fuel Storage Inspection
0-SR-DCS3.1.2.1
UnitO
Rev. 0004
Page 8 of 10
Date
7.3
Inspection of HI-STORM(s) Located on the ISFSI Pad (continued)
[2.3]
IF the blockage cannot be readily removed, THEN
PERFORM the following: (Otherwise N/A)
A.
NOTIFY the Shift Maintenance Manager to
PERFORM applicable section of
MSI-O-079-DCS036 ISFSI Abnormal Conditions
Procedure.
B.
IF acceptance criteria are not met within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,
THEN
c.
NOTIFY Maintenance and Modifications
Management to start preparation for MPC up load
to HI-TRAC. (Otherwise N/A)
7.4
Completion and Notification
[1]
VERIFY the Inlet and Outlet Vents on all HI-STORM(s)
inspected on Attachment 2 are free of blockage.
[2]
COMPLETE Attachment 1, Surveillance Procedure Review
Form, up to Unit Supervisor review.
[3]
NOTIFY the Unit One, Two, and Three Unit Operators (UOs)
this Surveillance Procedure is complete.
[4]
NOTIFY the Unit Supervisor this Surveillance Procedure is
complete and PROVIDE status of any required Corrective
Action per SPP-8.1 or unsatisfactory performances.
8.0
ATTACHMENTS
Attachment 1: Surveillance Procedure Review Form
Attachment 2: HI-STORM Inspection Log
US
__(AC) I
(
23. SRO GENERIC 2.3.9 OOl/C/A/T3/CONTAINMENT//G2.3.9/2.5/3.4/R1TCK
Given the following plant conditions:
Unit 2 is commencing a scheduled reactor shutdown due to a leak in the Drywell.
The Operations Manager has directed that the Drywell and Torus be de-inerted so that an
entry team can inspect the Drywell prior to cold shutdown.
Containment entry is scheduled in 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />.
The unit is currently at 25% power.
Which ONE of the following describes the earliest time and method for purging the containment to allow
for Drywell entry?
A~ Immediately begin purging the Drywell. The Torus cannot be purged concurrently.
B.
Drywell purging cannot be commenced until the unit is <15% power. The Torus cannot be purged
concurrently.
C.
Immediately begin purging the Drywell and the Torus concurrently.
D. Drywell purging cannot be commenced until the unit is <15% power. The Torus can be purged
concurrently.
KIA Statement:
Radiation Control
2.3.9 Knowledge of the process for performing a containment purge.
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions and times to correctly determine the process for performing a containment purge.
References: 2-01-64, Rev.1 06, section 8.1
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 .
SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment
of facility conditions and selection of appropriate procedures during normal , abnormal, and emergency
situations.
0610 NRC SRO Exam
(
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly, the candidate must:
Plausibility Analysis:
The primary decisions that must be made to correctly answer this question are:
1. Can the purge be started immediately or must it commence after power is reduced below 15%?
2. Can the Drywell and Torus be purged simultaneously, or only one at a time?
Although the TS bases for 3.6.3.2 is clear in describing the ability to commence purging 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to
going less than 15% RTP, the wording in Applicability (b) ofTS 3.6.3.2 has been known to cause
confusion. This makes Answer Band C plausible.
The ability to simultaneously purge the Drywell and Torus is well within the capability of the system,
however this is prohibited due to an analysis of the potential for containment over-pressurization if a
LOCA occurred during the evolution . This makes Answer C and D plausible because the purge lineup is
available.
Answer A is the correct answer.
Answer B is incorrect since you don't have to wait until 15% power to start de-inerting. The shutdown is
scheduled, therefore you can begin de-inerting 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing power < 15%.
Answer C is incorrect since you cannot purge the Drywell and Torus concurrently when in Mode 1,2 or 3.
Answer D is incorrect since you don't have to wait until 15% power to start de-inerting and you cannot
purge the Drywell and Torus concurrently when in mode 1,2 or 3.
(
(
Containment Inerting System
2-01-76
Unit 2
Rev. 0062
Page 8 of 82
3.0
PRECAUTIONS AND LIMITATIONS
A.
During normal operation, Primary Containment oxygen concentration is
required to be maintained less than 4% (by volume) by dilution with pure
nitrogen. This will minimize the potential for combustion with the hydrogen that
could result from fuel damage following a Loss Of Cooling Accident (LOCA).
B.
During startup, the Drywell and Suppression Chamber is required to be inerted
within 24-hours of reaching 15% Thermal Power. De-inerting may commence
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing Thermal Power to <15% prior to a Reactor
Shutdown .
C.
H2/02 Analyzers are normally maintained in a standby configuration. Analyzers
may be placed in service to monitor the Drywell or the Suppression Chamber at
Operations discretion. If one of the analyzers is inoperable, then the other
analyzer should be utilized to satisfy TRM requirements. Operable analyzer
may be transferred to Suppression Chamber for a reading, then placed back to
Drywell as often as necessary. H2/02Analyzers flow rate outside 12.75
and 17.25 scth renders monitor inoperable.
D.
Containment Inerting System supplies makeup gas to accommodate
temperature changes and leakage during normal power generation.
E.
After Drywelll Suppression Chamber differential pressure has been established,
any subsequent makeup or venting of Primary Containment is required to be
logged in 2-SI-4.7.A.2.a, Primary Containment Nitrogen Consumption and
Leakage.
F.
Drywell to Suppression Chamber differential pressure should NOT exceed
1.3 psid.
G.
Minimum storage tank level for tank A is 140 inches H20 on level indicator,
0-L1-076-0003, which corresponds to 3,500 gallons of liquid nitrogen. Additional
supply is necessary when level drops to 3,500 gallons or prior to inerting.
H.
The 6320-gallon Liquid Nitrogen Storage Tank (tank A) is designed to store
6,000 gallons of liquid. This is equivalent to 550,000 standard cubic feet of gas
(one gallon of liquid nitrogen =93.11 SCF). Due to purging during inerting
operation, over one million SCF of nitrogen gas (two full tanks or one tank and
one supply transport) will be required to fully inert containment.
I.
Use of keylock bypass switch, 2-HS-76-69(79), on Panel 2-9-54(55) is only
permitted after a valid Group 6 Isolation with permission from Shift Manager.
J.
Drywell entry when a corrected value of oxygen content is <19.5% is permitted
only with use of a self-contained breathing apparatus (SCBA).
K.
Prior to admitting nitrogen for inerting, a "NO ENTRY" sign is required to be
posted at Drywell entrance.
(
Containment Inerting System
2-01-76
Unit 2
Rev. 0062
Page 9 of 82
3.0
PRECAUTIONS AND LIMITATIONS (continued)
L.
Liquid nitrogen freezes skin on contact. Injury can also result from handling
bare piping filled with liquid nitrogen. Caution should be used to prevent
exposure from leaks and spills of liquid and insulated gloves should be worn to
avoid direct contact with cold piping.
M.
BFN FSAR stipulates that 2-FCV-84-19 will be maintained closed except during
surveillance testing or when directed by EOls.
N.
The following valves are interlocked closed with Mode Switch in RUN unless
Division I and II RUN MODE BYPASS Switches, 2-HS-64-24 and 2-HS-64-25,
are in BYPASS position:
Div I (BYPASS SW 2-HS-64-24)
Div II (BYPASS SW 2-HS-64-25)
2-64-18
2-64-17
2-64-19
2-64-30
2-64-29
2-64-33
2-64-32
2-76-24
O.
TOE 0-97-064-0823 evaluated the impact of inerting or purging Suppression
Chamber and Drywell concurrently (Both FCV 64-19 and FCV 64-18 open at
the same time). This evaluation concluded there is a slight potential to over
pressurize primary containment in the event of a large break LOCA with both
FCV 64-19 and FCV 64-18 open at the same time with Reactor NOT in Cold
Shutdown. This situation could create a large bypass flow path between the
Drywell and the Suppression Chamber. Therefore, Suppression Chamber and
the Drywell are NOT allowed be inerted or purged at the same time when
Reactor is NOT in Cold Shutdown.
P.
Unless authorized by Shift Manager, applicable CAD TANK, level indicator's in
the main control rooms (Unit 1 or 3) is required to be indicating 100% prior to
filling/topping off Nitrogen tanks A or B (Refer to 2(3)-01-84 for filling CAD
TANKS)
Q.
Drywell O2 CONCENTRATION indicators on Panels 2-9-54 and 2-9-55 are no
longer calibrated in the high range. PIP-97-189 should be used to correct O2
CONCENTRATION value as a reference until a grab sample is taken. Grab
sample values of O2 concentration are expected to vary from high-range O2
analyzer values.
(
Primary Containment Oxygen Concentration
3.6.3.2
3.6
CONTAINMENT SYSTEMS
3.6.3.2 Primary Containment Oxygen Concentration
The primary containment oxygen concentration shall be < 4.0
volume percent.
APPLICABILITY:
MODE 1 during the time period:
a. From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is > 15% RTP
following startup, to
b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to < 15% RTP
prior to the next scheduled reactor shutdown .
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION
TIME
A.1
Restore oxygen
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
oxygen concentration not
concentration to within
within limit.
limit.
B. Required Action and
8.1
Reduce THERMAL
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
associated Completion
POWER to s 15% RTP.
Time not met.
BFN-UNIT 2
3.6-42
Amendment No. 253
(
24. SRO GENERIC 2.4.21 00lIC/A/T3/RHRIB15/GENERIC 2.4.21//SRO ONLY/l1l27/07 RMS
Given the following plant conditions:
During normal full power operation of Unit-3, a loss of 250VDC power causes the RHR
System I (Div I) Logic Power Failure alarm to occur.
I&C investigates and reports a blow fuse in the logic circuit.
Prior to any corrective action being taken, a Loss of Coolant Accident results in the followinq
plant conditions:
RPV Level
RPV Pressure
Drywell Pressure
8elowTAF
100 psig
21 psig
Which ONE of the following describes the response of Loop I RHR pumps and what subsequent actions
must be taken to address these plant conditions?
A.
RHR Pumps 3A and 3C should be manually started in DW Spray mode. 38 and 3D auto start in
LPCI mode and should remain there until RPV level is above TAF ..
B.
RHR Pump 3A only will auto start in LPCI mode. 3C can be manually started in LPCI mode and
should remain there until RPV level is above TAF. 38 and 3D should be manually started in DW
Spray mode.
C.
RHR Pump 3C only will auto start in LPCI mode. 3A can be manually started in LPCI mode and
should remain there until RPV level is above TAF. 38 and 3D should be manually started in LPCI
mode.
0 .01 All four (4) RHR pumps auto start in the LPCI mode and should remain there until RPV level is
above TAF.
KIA Statement:
Emergency Procedures IPlan
2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions including:1
Reactivity control 2. Core cooling and heat removal 3. Reactor coolant system integrity 4. Containment
conditions 5. Radioactivity release control
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the response of RHR system components during an emergency based on a
logic failure.
References:
OPL171.044, 3-ARP-9-3D (5), 1-ARP-9-3D (5), 2-ARP-9-3D (5)
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.
SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment
of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency
situations.
0610 NRC SRO Exam
REFERENCE PROVIDED: None
Plausibility Analysis:
In order to answer this question correctly, the candidate must determine the following :
1. Recognize how a 1055 of Div 1 logic power affects the RHR pump initiation logic.
2. Recognize the difference between Unit-3 logic and Unit 1/2 logic.
3. Determine the correct application of EOI requirements with RPV level below TAF.
A is incorrect. 3A and 3C RHR pumps auto start in LPCI mode. It is not appropriate to divert LPCI
injection for containment control until RPV level has been restored above TAF. This is plausible because
the Div 1 pumps on Unit 1/2 will not start automatically.
B is incorrect. 3C RHR pump will start in LPCI mode. It is not appropriate to divert LPCI injection for
containment control until RPV level has been restored above TAF. This is plausible because the CAS
logic for Unit 1/2 are divided into "preferred" and "non-preferred" pumps as well as Divisionalized initiation
logic.
C is incorrect. 3A RHR pump will auto start in LPCI mode. 3C and 3D RHR pumps were not affected by
the logic power failure and would start in LPCI mode. This is plausible because the CAS logic for Unit 1/2
are divided into "preferred" and "non-preferred" pumps as well as Divisionalized initiation logic.
D is correct.
OPL171.044
Revision 15
Page 49 of 159
INSTRUCTOR NOTES
4.
Pumps Trip/interlocks
a.
Electrical Faults
Obj. V.D.8.
b.
Loss of suction path
Obj. V.C.7.
c.
Each RHR pump has a NORMAUEMERGENCY
Obj. V.C.7.
switch at the breaker.
(1)
With the switch in the EMERGENCY position,
the pump can only be started from the breaker
(Le. automatic, local, and control room starts
are removed) and then only if the associated
pump drain valve is closed.
(2)
With switch in the EMERGENCY position, the
pump will trip only from operation of breaker
control switch, Shutdown board load shed, and
electrical faults.
d.
With an LPCI initiation signal present, the local
Obj. V.C.7.
station cannot be used to stop the pump.
(1)
The "white" light above the control switch in the
control room indicate one of the following:
(a)
The breaker is tripped with the control
switch in the Normal-Atter-Start or Start
position
(b)
Atter an automatic start, the white light
Red target on
would not come on if the pump tripped
control switch
unless the control switch was place to the
Start position.
(c)
The pump is running at "overload amp
condition".
e.
Since Unit 1 and Unit 2 share the U1/2 diesel as an
Obj. V.C.8.
onsite power supply, the possibility existed that the
TP-45, 46, 47,
diesels could be lost due to an overload conditions if
and 48
Unit 1 and Unit 2 equipment were to start on the
TP-61, 62, 63, 64
board at the same time.
To prevent this, each Unit was assigned
"Preferred and Non-preferred" pumps and logic
controls the pumps that are automatically started
based upon which unit(s) is/are experiencing an
accident.
(
OPL171.044
Revision 15
Page 50 of 159
INSTRUCTOR NOTES
Note:
Presently Unit 1 Accident signal will not affect Unit 2 due to DCN H2?35A that lifted wires
from relays. Unit 2 will still affect Unit 1.
However, the following represents modifications
to the inter-tie logic as it will be upon Unit 1 recovery.
f.
(1)
Unit 1 Preferred RHR pumps are 1A and 1C
(2)
Unit 2 Preferred RHR pumps are 28 and 2D
(3)
Unit 2 initiation logic is as follows: Div 1 RHR
logic initiates Div 1 pumps ( A and C), and
Div 2 logic initiates Div 2 pumps (B and D)
Accident Signal
(1)
LOCA signals are divided into two separate
signals, one referred to as a Pre Accident
Signal (PAS) and the other referred to as a
Common Accident Signal (CAS).
- PAS
-122" Rxwater level (Level 1)
2.45 psig OW pressure
-122" Rx water level (Level 1)
2.45 psig OW pressure AND <450
psig Rxpressure
(2)
If a unit receives an accident signal, then all
its respective RHR and Core Spray pumps
will sequence on based upon power source
to the SO Boards.
(3)
All RHR and Core Spray pumps on the non-
affected unit will trip (if running) and will be
blocked from manual starting for 60 seconds.
Obj. V.B.13.
Obj. V.C.3
Obj. V.C.?
Obj. V.D.6
Obj. V.E.II
Obj. V.B.13.
Obj. V.C.3
Obj. V.C.?
Obj. V.D.6
Obj. V.E.II
Note:
It should be clear
that the only
difference
between the two
signals is the
inclusion of Rx
pressure in the
CAS signal. The
PAS signal is an
anticipatory signal
that allows the
DG's to start on
rising OW
pressure and be
ready should a
CAS be received.
OPL171.044
Revision 15
Page 51 of 159
(
INSTRUCTOR NOTES
(4)
After 60 seconds all RHR pumps on the non-
Operator diligence
affected unit may be manually started.
required to
(5)
The non-preferred pumps on the non-
prevent
overloading SD
affected unit are also prevented from
boards/DG's
automatically starting until the affected unit's
accident signal is clear.
(6)
The preferred pumps on the non-affected
unit are locked out from automatically starting
until the affected unit accident signal is clear
OR the non-affected unit receives an
accident signal.
g.
4KV Shutdown Board Load Shed
Obj. V.C.8.
(1)
A stripping of motor loads on the 4KV boards
occurs when the board experiences an
undervoltage condition. This is referred to as a
4KV Load Shed. This shed prepares the board
for the DG ensuring the DG will tie on to the
bus unloaded and without faults.
(2)
The Load Shed occurs when an undervoltage
is experienced on the board i.e. or if the Diesel
were tied to the board (only source) and one
of the units experienced an accident signal
which trips the Diesel output breaker.
(3)
Then, when the Diesel output breaker
interlocks are satisfied, the DG output
breaker would close and, if an initiation signal
is present (CAS) the RHR, CS, and RHRSW
pumps would sequence on
(4)
Following an initiation of a Common Accident
Signal (which trips the diesel breaker), if a
subsequent accident signal is received from
another unit, a second diesel breaker trip on
a "unit priority" basis is provided to ensure
that the Shutdown boards are stripped prior to
starting the RHR pumps and other ECCS
loads
(5)
When an accident signal trip of the diesel
Occurs due to
breakers is initiated from one unit (CASA or
actuation of the
(
CASB), subsequent CAS trips of all eight
diesel breaker
diesel breakers are blocked.
TSCRN relay
(
SYS I
LOGIC
POWER FAILURE
I
5
REV 0025
3-XA- 55-3 D
SENSOR/TRI P POINT:
10A- K14 Relay
UNIT 3
3-ARP-9-3D
Page 6
Loss of 250V DC power
SENSOR LOCATION:
Panel 3-9-32,
Aux Instr Rm,
El 593'
PROBABLE CAUSE :
1.
Failed fuse .
AUTOMATI C ACTION:
2.
Loss of 25 0V DC power
s upply at 250V DC RMOV Bd 3B.
Non e
OPERATOR ACTION:
1.
DISPATCH personnel
to 250V DC RMOV Bd 3B,
Breaker 1E2 ,
t o verify position .
NOTE :
2.
DISPATCH personnel to Panel 3-9-32 t o check 10-AMP
fuses
3 .
REFER TO Te ch Spec 3.3.5.1 ,
3.5 . 1 ,
3.5 .2 ,
3 .6 .2 .3 ,
3.6 .2 .4, 3.6 .2.5,
and
TRM 3.3.3 .4.
IF alarm is valid ,
THEN t he fo llowing will
occur:
3A and 3C RHR Pumps wi l l
not receive an auto start
signal from Div I Logic.
3A and 3C RHR Pump wi l l
receive an auto start signal from Div II Logic .
SYS I Inboard Injection Valve will not receive an auto op en signal from DIV
I Logic.
SYS I Inboard Injection Valve will not manually open from the control room
due to loss of 450 psig logic from DIV I.
The SYS I
I nboa r d Injection Valve will
receive an
auto open signal from DIV II Logic .
REFERENCES :
3-45N620-2;
3- 45E712- 2 ;
FSAR 8 . 6 . 4 .2 ;
Tech Spec 3 .3 . 5 . 1 ,
Emergency Core Coo l i ng System
(ECCS)
Instrumentation;
3.5 .1,
ECCS - Operating;
3 .5.2,
ECCS -
Shutdown;
3 .6 . 2 .3 ,
Residual
Heat Removal
(RHR) Suppression Pool Cooling ,
3.6 .2.4 ,
Re sidual
Heat Removal
(RHR) Suppression Pool
Spray;
3 .6 .2.5 , Residual Heat Removal
(RHR) Drywell Spray,
TRM 3 .3.3 .4 ECCS and RCIC Trip System Bus Power Monitors.
(
Unit 1
SYS I LOGIC
POWER FAILURE
(Page 1 of 1)
XA-55-3D
Sensorrrrip Point:
10A-K1A Relay
1-ARP-9-3D
Rev. 0021
Page 8 of 43
Loss of 250V DC power
Sensor
Location:
Probable
Cause:
Automatic
Action:
Operator
Action:
1-PNLA-009-0032
Aux Instr. Rm, EI 593'
A. Cleared fuse.
B. Loss of 250V DC power supply.
None
A.
DISPATCH personnel to 250V DC Rx MOV Bd 1B, breaker 1E2, to
verify position. (Rx Bldg, EI593', R-1 Q-L1NE)
B. DISPATCH personnel to 1-PNLA-009-0032 to check fuses 10A-F1A
and 10A-F2A (10 amp).
C. REFER TO Tech Spec Sections 3.3.5.1, 3.5.1, 3.5.2, 3.6.2.3,
3.6.2.4, 3.6.2.5, TRM Section 3.3.3.4.
NOTE
n
o
o
IF alarm is valid, THEN the following will occur:
1A RHR Pump will NOT auto start.
1C RHR Pump will NOT auto start.
SYS I Inboard Injection Valve will NOT receive an auto open signal.
SYS I Inboard Injection Valve will NOT manually open from the control room due to loss of
450 psig logic from DIV I.
References:
1-45E602-2
1-45E712-2
1-730E920-4 and -12
Unit 2
RHR SYS I
LOGIC POWER
FAILURE
(Page 1 of1)
Panel 2-9-3
2-XA-55-3D
SensorlTrip Point:
2-RLY-074-10A-K1A
2-ARP-9-3D
Rev. 0025
Page 8 of 42
Loss of 250V DC power
Sensor
Location:
Probable
Cause:
Automatic
Action:
Operator
Action:
Panel 2-9-32, Aux Instr Rm, EI 593'
A. Cleared fuse.
B. Loss of 250V DC power supply, at 250V DC RMOV Board 2B.
None
A. DISPATCH personnel to Panel 2-9-32 to check 10-amp fuses
B. DISPATCH personnel to 250V DC RMOV Bd 2B Breaker 1E2, to
verify position.
C. REFER TO Tech Specs 3.3.5.1, 3.5.1, 3.5.2, 3.6.2.3, 3.6.2.4,
3.6.2.5, TRM 3.3.3.4.
o
o
o
NOTE
1)
IF alarm is valid, THEN the following will occur:
2A RHR Pump will NOT auto start.
2C RHR Pump will NOT auto start.
SYS I Inboard Injection Valve will NOT receive an auto open signal.
SYS I Inboard Injection Valve will NOT manually open from the control room due to loss of
450 psig logic from DIV I.
References:
2-45E620-2
2-45E712-2
2-45E765-4
Tech Specs 3.3.5.1, Emergency Core Cooling System (ECCS) Instrumentation
3.5.1, ECCS - Operating
3.5.2, ECCS - Shutdown
3.6.2.3, Residual Heat Removal (RHR) Suppression Pool Cooling
3.6.2.4, Residual Heat Removal (RHR) Suppression Pool Spray
3.6.2.5, Residual Heat Removal (RHR) Drywell Spray
TRM 3.3.3.4, ECCS and RCIC Trip System Bus Power Monitors
(
(
25. SRO GENERIC 2.4.30 001/CIA/REPIIIGENERIC 2.4.301ISRO ONLY/11/27/07 RMS
Given the following plant conditions:
You are the Shift Manager on dayshift attending the morning meeting.
During the meeting you are informed that both seals on 2A Recirc Pump have failed.
When you arrived in the Unit-2 control room, the following conditions were noted:
- Unit-2 reactor was manually scrammed.
- Actions were carried out in accordance with 2-AOI-68-1 to isolate the recirc pump
- Actions were carried out in accordance with 2-AOI-64-1 to vent the drywell.
- EOI-1 was entered, executed and exited.
- Reactor pressure is being controlled by bypass valves.
- Reactor water level is +32 inches being controlled by RFPT 2C in automatic.
- All four EDGs are running unloaded with an AUO monitoring their status locally.
- The drywell is being vented with pressure at 0.75 psig and lowering.
- 2A Recirc loop is isolated with CRD purge isolated.
Which ONE of the following describes the Emergency Classification Level and the appropriate action to
take?
REFERENCE PROVIDED
A.
An Alert (2.1-A) shall be declared to the ODS and the NRC, and then subsequently cancelled .
B.~
An Alert (2.1-A) shall be reported to the ODS and the NRC, but should not be declared.
C.
A NOUE (2.4-U) shall be reported to the ODS only and need not be reported to the NRC since the
emergency is resolved .
D.
An Alert (2.4-A) shall be reported to the NRC only and need not be reported to the ODS since the
emergency is resolved.
KJA Statement:
Emergency Procedures IPlan
2.4.30 Knowledge of which events related to system operations/status should be reported to outside
agencies
KJA Justification: This question satisfies the KIA statement by requiring the candidate to use specific
plant conditions to determine the emergency classification level and the associated reporting
requirements.
References:
EPIP-1
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.
SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment
of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency
situations .
0610 NRC SRO Exam
(
REFERENCE PROVIDED: EPIP-1
Plausibility Analysis:
In order to answer this question correctly , the candidate must determine the following :
1. Determine that current conditions do NOT indicate an EAL is being exceeded.
2. Determine that previous conditions, based on running DIGs, indicate that EAL 2.1-A or 2.4-A were
applicable at an earlier time but has been resolved before being declared .
3. Determine the correct notification requirement based on Item 2 above.
A is incorrect. The EAL should NOT be declared . This is plausible based on conservative decision
making, but is NOT in accordance with the directions in EPIP-1.
B is correct.
C is incorrect. The EAL is incorrect. A dual seal failure of a Recirc Pump will yield a significantly higher
leak rate than specified in EAL 2.4-U. In addition, the NRC should also be notified in accordance with
EPIP-1. This is plausible for a novice operator candidate.
D is incorrect. The ODS should also be notified in accordance with EPIP-1. This is plausible for a novice
operator candidate.
3.0
INSTRUCTION
(
BROWNS FERRY
EMERGENCY CLASSIFICATION PROCEDURE
EPIP-1
3.1
Following plant events or transients review EPIP-1 Section II, 1.0 through 8.0 and
determine if an event should be classified as an emergency.
NOTE
1. If an emergency action level for a higher classification was exceeded, but the present
situation indicates a lower classification, the fact that the higher classification occurred shall
be reported to the NRC and the CECC, if staffed, or ODS if the CECC is not staffed. The
higher classification should not be declared.
2. If an emergency action level was met but the emergency has been totally resolved, the
emergency class that was appropriate shall be reported to the ODS and the NRC but
should not be declared.
3.1.1
EPIP-1 Section 11,1 .0 through 8.0 captures events in eight major
categories as listed on the event classification index.
3.1.2
Each emergency action level (EAL) in a category is given an alpha-
numeric designator. The first numeric component of the EAL indicates the
section followed by a numeric designator for the specific EAL within the
section and an alpha numeric designator for the event class.
Example: 5.2-U
These designators provide for cross-reference between the specific EAL
and the basis document which provides technical supporting information
for the EAL and may aid the Shift Manager/SED in classifying events.
Curves, notes, or tables that support the EAL are located on the face
adjacent page within the matrix section of the procedure and are identified
within the event classification window on the information bar that precedes
the designator. The information bar contains the appropriate indication to
alert the user that a corresponding curve, note, or table applies to the
EAL.
Curves, notes, or tables that contain unit specific information will also be
identified within the event classification window by the letter "US" located
at the end of the EAL information bar. This information should alert the
user that the corresponding curve, note, or table contains unit specific
information.
Example
I
5.2-U I CURVE I NOTE I TABLE ~
PAGE 3 OF 201
REVISION 42
BROWNS FERRY
EMERGENCY CLASSIFICATION PROCEDURE
EVENT CLASSIFICATION MATRIX
EPlp*1
PRIMARY CONTAINMENT
PRESSURE
Description
Description
I
I
I
I
I
I
I
I
c::zc::enc::>>r-m
c:::mz
-4
2.1-A I
I
I TABLE I
I
I
I
I
Drywell pressure at or above 2.45 psig
AND
>>r-
Indication of Primary System leakage into
m
Primary Containment. Refer to Table 2.1-A.
- U
-4
OPERATING CONDITION:
Mode 1 or 2 or 3
2.1-S I CURVE I
I
I
2.2-S I
I
I
I
Suppression Chamber pressure can NOT be
Drywell or Suppression Chamber
en
maintained in the safe area of Curve 2.1-S.
hydrogen concentration at or above 4%
=im
AND
m
S
Drywell or Suppression Chamber
m
- U
oxygen concentration at or above 5%.
G)m
OPERATING CONDITION:
OPERATING CONDITION:
Z
0
Mode 1 or 2 or 3
Mode 1 or 2 or 3
-<
2.1-G I
I
I
I
2.2-G I
I
I
I
Suppression Chamber pressure can NOT be
Drywell or Suppression Chamber
maintained below 55 psig.
hydrogen concentration at or above 6%
G)mz
AND
m
Drywell or Suppression Chamber
~r-
oxygen concentration at or above 5%.
ms:m
- 0
(j)m
OPERATING CONDITION:
OPERATING CONDITION:
Z
0
Mode 1 or 2 or 3
Mode 1 or 2 or 3
-<
PAGE 26 OF 201
REVISION 42
(
BROWNS FERRY
II NOTES
CURVESITABLES:
EMERGENCY CLASSIFICATION PROCEDURE
EVENT CLASSIFICATION MATRIX
EPIP-1
II
TABLE 2.1-A
INDICATIONS OF PRIMARY SYSTEM LEAKAGE
INTO PRIMARY CONTAINMENT
Primary Containment Pressure Hiqh Alarm
Drvwsll Floor Drain Sump Pump Excessive Operation
Drvwell CAM Activity lncreasino
Orvwell Temperature Hloh Alarm
Chemistry Sample Radionuclide Comparison To Reactor Water
CURVE 2.1-5
PRESS SUPPR PRESS
35 __--r----r----r---,-----,--.-,------,------,-----,
-
ACTION
__ -
REOUIRED
_I.---
SAFE
(
14
15
16
SUPPRPL LVL(FT)
PAGE 25 OF 201
17
18
19
20
REVISION 42
DRYWELL INTERNAL
LOSS OF PRIMARY
LEAKAGE
CONTAINMENT
Description
Description
2.4-U I
I
I
I
2.5-U I
I
I TABLE I
Drywell unidentified leakage exceeds 10 gpm
Inability to maintain Primary Containment
c:
pressure boundary. Refer to Table 2.3/2.5-U.
Z
c:
C/)c:
Drywell identified leakage exceeds 40 gpm.
>>
rm
OPERATING CONDITION:
OPERATING CONDITION:
<:m
Mode 1 or 2 or 3
Mode 1 or 2 or 3
Z-t
2.4-A I
I
I
I
I
I
I
I
Drywell unidentified leakage exceeds 50 gpm.
>>
rm
- iO
-t
OPERATING CONDITION:
Mode 1 or 2 or 3
I
I
I
I
I
I
I
I
C/)
=imms:m
- iO
(j)mz
0-<
I
I
I
I
I
I
I
I
(j)mzm
~r-ms:m
- 0
(j)mzo-<
(
BROWNS FERRY
EMERGENCY CLASSIFICATION PROCEDURE
EVENT CLASSIFICATION MATRIX
PAGE 30 OF 201
EPIP-1
REVISION 42