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See also: [[ | See also: [[see also::IR 05000325/2008302]] | ||
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Revision as of 05:30, 12 July 2019
| ML083380337 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 12/03/2008 |
| From: | - No Known Affiliation |
| To: | Division of Reactor Safety II |
| References | |
| 50-324/08-302, 50-325/08-302 | |
| Download: ML083380337 (49) | |
See also: IR 05000325/2008302
Text
Final Submittal (Blue Paper)FINAL RO/SRO WRITTEN EXAMINATION
REFERENCES
BRUNSWICK NOVEMBER 2008 ExAM 05000325/2008302
&05000324/2008302
List of Reference Material for RO NRC Exam Steam Tables OEOP-01-UG, Attachment
5, Figure 5, Core Spray NPSH Limit OEOP-01-UG, Attachment
5, Figure 6, RHR NPSH Limit OEOP-01-UG, Attachment
6, Figure 18, Unit 1 Reactor Water Level at LL-4 20P-27, Figure 1 ,Estimated
Capability
Curves 201-03.2, Attachment
1, SJAE Off-Gas Radiation Monitors Channel Check Calculation
EHC Logic Diagram
List of Reference Material for SRO NRC Exam Steam Tables OEOP-01-UG, Attachment
5, Figure 5, Core Spray NPSH Limit OEOP-01-UG, Attachment
5, Figure 6, RHR NPSH Limit OEOP-01-UG, Attachment
6, Figure 18, Unit 1 Reactor Water Level at LL-4 20P-27, Figure 1, Estimated Capability
Curves 201-03.2, Attachment1, SJAE Off-Gas Radiation Monitors Channel Check Calculation
EHC Logic Diagram OPEP-02.1, Attachment
1, Emergency Action Levels OEOP-01-UG, Attachment
6, Figure 19, Unit 1 Reactor Water Level at LL-5 OE"OP-01-UG, Attachment
5, Figure 3, Heat Capacity Temperature
Limit OMST-PCIS21 Q, Acceptance
Criteria 001-18, B21-LT-N024A-1, 8-1;821-LT-N025A-1, B-1 TS 3.3.6.1, Primary Containment
Isolation Instrumentation
OOP-06.4, Recirculation
and Sampling of Saltwater Release Tank#1
ATTACHMENT
5 Page 21 of 28 FIGURE 5 Core Spray NPSH Limit 170
VNmm g" 7,000..........
260
...
250
230
220
"""""""'................
160_+-I+-..-",
-+-+--t j-+--+-l
.....-+--l----i
150_u"-t--"'"
1,000 2,000 3,000 4,.000 5,000 6,000280 oSUBTRACT 0.5 PSIG FROM INDICATED SUPPRESSION
CHAMBER PRESSURE FOR EACH FOOT OF WATER LEVEL BELOW A SUPPRESSION
POOL WATER LEVEL OF-31 INCHES (-2.6 FEET).*SUPPRESSION
CHAMBER PRESSURE (CAC-PI-1257-2A
OR CAC-PI-1257
-2B)I OEOP-01-UG
Rev.51 Page 81 of 150 I
.---.....290 U.0W 280:::)270260 W C.250W I-240W 230220 3:..J 0 210 0 C.200 Z 0 190-en en 180 WC.170 C.:::)160 en ATTACHMENT
5 Page 22 of 28 FIGURE 6 RHR NPSH Limit*40 PSIG I--------I---------........*20 PSIG........---------------....................
- 10 PSIG-I--------------.......-...........
....................
- 5 PSIG I--------I'I--------------'".............
""""""'""-""ll*0 PSIG I--------f-----III I I o 5,000 10,000 15,000 20,000 RHR PUMP FLOW (GPM)SUBTRACT 0.5 PSIG FROM INDICATED SUPPRESSION
CHAMBER PRESSURE FOR EACH FOOT OF WATER LEVEL BELOW A SUPPRESSION
POOL WATER LEVEL OF-31 INCHES (-2.6 FEET).*SUPPRESSION
CHAMBER PRESSURE (CAC-PI-1257-2A
OR CAC-PI-1257
-2B)I OEOP-01-UG
Rev.51 Page 82 of 150 I
ATTACHMENT
6 Page 14 of 19 FIGURE 18 Unit 1 Reactor Water
at LL-4 (Minimum Steam Cooling Level);REF LEG TEMP ABOVE 200(JF{REF LEG TEMP BELOW OR EQUAL TO
1,1*50 00 300 500 900 1,100 60 200 400 1,000 REACTOR PRESSURE (PSIG)0-10-20 CJ)W J:-30 0 Z..........-40..J W>-50
..-.:.,uu.r.. W..J 0..60.....................W-70 0....
......_'"".....-C Z-80-90-100 WHEN REACTOR PRESSURE IS LESS THAN 60 PSIG, USE INDICATED LEVEL.LL-4 IS-30.0 INCHES.I OEOP-01 Rev.51 Page 102 of 150 I
ATTACHMENT
1 Page 36 of 60 Table 1 SJAE OFF-GAS RADIATION MONITORS CHANNEL CHECK CALCULATION
RECORD SJAE Off-Gas SAT SUN MON TUE WED THUR FRI Radiation Monitor readings from Item 55&56.o 12-RM-K601A
o 12-RM-K601B
DIVIDE the highest reading monitor by 2.CONFIRM the lower reading monitor is greater than this value.IF the'lower reading monitor indication
is>%the higher reading monitor then the channel check is satisfactory.
IF the lower reading monitor indication
is,%the higher reading monitor contact E&RC Health Physics to obtain a local reading with an appropriate
survey instrument
for alternate criteria.NOTE: The survey instrument
should be positioned
near the operator aid[alternate
channel check survey point]located on the sample chamber.IF UTILIZED, RECORD local survey SAT SUN MON TUE WED THUR FRI instrument
reading MULTIPLY the local instrument
reading by.75.IF the lower reading monitor indication
is>.75 of the local survey instrument, the deviation is conservative
and the channel check is satisfactory.
Initiate a W/R to evaluate the deviation between the A and B monitors.IF the lower reading monitor indication
is.75 the local survey instrument, the deviation is non-conservative
and the instrument
should be declared inoperable.
1201-03.2 Rev.101 Page 41 of 781
RUNBACKS+1400/0 CONTROL VALVE DEMAND (To DFGs)T CLOSE 7" HG LOW VACUUM TRIPUV/o
UfO...BYP VLV DEMAND TURBINE TRIP...1000/0'I SYNC SPEED (o.0!cJ).
RPM r-GENERATOR
PCB CLOSED BYPASS VALVE JA'CK 1 0 k SM'ALL CLOSE BIAS I."f-J POWER LOAD UNBALANCE OPENING BIAS 1000/0 WHEN ANY SPEED SELECTED 30 I
l.__Open Shell FAST 180 RiM+/-+/-':':\.:,.'\l':.
.LL.-....1.5 PSI BIAS MEDIUM 90 T (+)SLOW 60(-)l S00 I.j 500'800"D 00.f1RPM RPM RPM RPM lOvERSPEED
TRIP TEST PAM
I 0..1050 PSI PAM PRESSURE 0..1050 PSI..
...***..*.:.*..*.*:.*'.*.'**'.'*:: PRESSURE SET I...
150-1050 PSITURBINE I...
..., SPEED\:.K152 ALL VALVES I CLOSED 7
Section Event Category ATTACHMENT
1 Page 1 of 24 Emergency Action Levels Page No.1.0 Abnormal Primary Leak Rate 11 2.0 Steam Line Break or Safety/Relief
Valve Failure.......................................
13 3.0 Abnormal Core Conditions
and Core Damage 14 4.0 Abnormal Radiological
Effluent or Radiation Levels..................................
16 5.0 Loss of Shutdown Functions:
Decay Heat and Reactivity.........................
18 6.0 Electrical
or Power Failures.......................................................................
20 7.0 Fire 21 8.0 Control Room Evacuation.........................................................................
22 9.0 Loss of Monitors or Alarms or Communication
Capability
23 10.0 Fuel Handling Accident 25 11.0 Security Threats........................................................................................
26 12.0 Fission Product Barriers and Specific LCOs 27 13.0 Hazards to Plant Operations.....................................................................
28 14.0 Natural Events 29 15.0 Shift Superintendent/Site
Emergency Coordinator
Judgments 31 IOPEP-02.1
Rev.50 Page 10 of 331
ATTACHMENT
1 Page 2 of 22 Emergency Action Levels 1.0 Abnormal Primary Leak Rate 1.1 Notification
of Unusual Event 01.01.01 Reactor Coolant System total leakage greater than 25 gpm averaged over the previous 24-hour period using the sum of drywell equipment drain integrator (G16-FQ-K603)
and drywell floor drain integrator (G16-FQ-K601), and the leakage rate has not been reduced to less than 25 gpm within eight hours, or plant shutdown is not achieved within required time period.OR 01.01.02 Unidentified
Reactor Coolant System leakage greater than 5 gpm averaged over the previous 24-hour period using the drywell floor drain integrator (G16-FQ-K601), and the leakage rate has not been reduced to less than 5 gpm within eight hours, or plant shutdown is not achieved within required time period.1.2 Alert 01.02.01 Small break LOCA with primary system leakage greater than 50 gpm.A LOCA is indicated by a significant
loss of reactor inventory to the drywell resulting inincreaseddrywell
pressure, temperature, and/or sump pump usage indicated by:*Low or falling Reactor Coolant System pressure with rising drywell pressure and temperature (C32-R608, CAC-PI-2685-1, CAC-TR-4426-1A, CAC-TR-4426-1
B, CAC-TR-4426-2A
and CAC-TR-4426-2B).
1.3 Site Area Emergency 01.03.01 Loss of coolant accident requiring the initiation
of Low Pressure Coolant Injection, Core Spray, or the Automatic Depressurization
System, AND REQUIRED FOR ADEQUATE CORE COOLING.IOPEP-02.1
Rev.50 Page 11 of 331
ATTACHMENT
1 Page 3 of 22 Emergency Action Levels 1.0 Abnormal Primary Leak Rate (Continued)
1.4 General Emergency 01.04.01 Loss of coolant accident requiring the initiation
of Low Pressure Coolant Injection, Core Spray, or Automatic Depressurization
System, AND REQUIRED FOR ADEQUATE CORE COOLING;AND Inability to provide makeup water to the Reactor Coolant System (i.e, failure of HPCI, Core Spray A and B, RHR Loops A and B, RCIC, condensate, and feedwater)
as indicated by falling or low reactor vessel level with attempts to inject water not successful.
IOPEP-02.1
Rev.50 Page 12 of 331
ATTACHMENT
1 Page 4 of 22 Emergency Action Levels 2.0 Steam Line Break or Safety/Relief
Valve Failure 2.1 Notification
of Unusual Event 02.01.01 Reactor Coolant System pressure1250 psig.OR 02.01.02 Inability to close an SRV with Reactor Coolant System pressure900 psig.2.2 Alert 02.02.01 Main Steam, HPCI or RCIC steam line break inside the primary containment
without (full)line isolation valve closure.2.3 Site Area Emergency 02.03.01 Main Steam, HPCI or RCIC steam line break outside primary containment
and line isolation valve(s)fail to close indicated by valid area radiation and/or temperature
alarms.2.4 General Emergency 02.04.01 N/A IOPEP-02.1
Rev.50 Page 1 3 of 331
ATTACHMENT
1 Page 5 of 22 Emergency Action Levels 3.0 Abnormal Core Conditions
and Core Damage 3.1 Notification
of Unusual Event 03.01.01 Liquid A.Reactor Coolant System (RCS)activity greater than 4.0 J.lCi/gm 1-131 dose equivalent.
B.RCS activity greater than 0.2 J.lCi/gm 1-131 dose equivalent
but less than limit above for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.03.01.02 Gaseous A.Steam jet air ejector off-gas radiation monitor (D12-RM-K601
A and B)reading of greater than 1.2 x 10 4 mR/hr.B.Steam jet air ejector off-gas radiation monitor (D12-RM-K601
A and B)increase of greater than 2.4 x 10 3 mR/hr in 30 minutes.3.2 Alert 03.02.01 Liquid Reactor coolant activity greater than 300 J.lCi/gm 1-131 dose equivalent.
03.02.02 Gaseous Steam jet air ejector off-gas radiation monitor (D12-RM-K601
A and B)reading of greater than 1.2 x 10 5 mR/hr.3.3 Site Area Emergency 03.03.01 Reactor Coolant System activity is greater than 4000 J.lCi/gm 1-131 dose equivalent.
IOPEP-02.1
Rev.50 Page14 of 331
ATTACHMENT
1 Page 6 of 22 Emergency Action Levels 3.0 Abnormal Core Conditions
and Core Damage (Continued)
3.4 General Emergency 03.04.01 Any two functional
high range drywell radiation monitors (D22-RI-4195, 4196, 4197, and 4198)reading greater than 5000 R/hr.IOPEP-02.1
Rev.50 Page15 of 331
ATTACHMENT
1 Page 7 of 22 Emergency Action Levels 4.0 Abnormal Radiological
Effluent or Radiation Levels 4.1 Notification
of Unusual Event 04.01.01 Liquid Release Any unplanned release from the liquid waste system resulting in activity levels in the discharge canal greater than those in 10CFR20, Appendix B, Table II, Column 2.04.01.02 Gaseous Release Any gaseous release which exceeds the dose limit specified in aDCM 7.3.7 (Le., exceeding the noble gas instantaneous
dose rate limit as evaluated by OE&RC-2020).
04.01.03 Any building evacuation
based on confirmed radiological
conditions (i.e., greater than 10 dac airborne[except precautionary
evacuations]
).4.2 Alert 04.02.01 Liquid Release Any liquid release resulting in activity concentration
levels in the discharge canal that are greater than 10 times those given in 1 OCFR20, Appendix B, Table II, Column 2 (10 times the concentration
listed in Unusual Event).04.02.02 Gaseous Release Any gaseous release which exceeds 10 times the dose rate limit specified in aDCM 7.3.7 (Le., exceeding 10 times the noble gas instantaneous
dose rate limit as evaluated by OE&RC-2020).
04.02.03 In-Plant Leak or Spill Unplanned, valid direct area radiation (gamma and/or neutron)reading(s)
increase by a factor of 1000 over normal levels.IOPEP-02.1
Rev.50 Page 16 of 331
ATTACHMENT
1 Page 8 of 22 Emergency Action Levels 4.0 Abnormal Radiological
Effluent or Radiation Levels (Continued)
4.3 Site Area Emergency.04.03.01 Projected dose exceeding 50 mRem Whole body (TEDE)OR exceeding 250 mRem Thyroid (CDE)at site boundary.04.03.02 Measured dose rate exceeding 100 mR/hr at site boundary.04.03.03 Measured 1-131 dose equivalent
concentration
exceeds 3.9E-7).lCi/cc at the site boundary.4.4 General Emergency 04.04.01 Offsite release resulting in a dose exceeding one (1)Rem Whole Body (TEDE)OR five (5)Rem Thyroid (CD E)at the Site Boundary as indicated by dose projection
or field data.04.04.02 Measured 1-131 Dose Equivalent
concentration
exceeding 3.9E-6).lCi/cc at the site boundary.IOPEP-02.1
Rev.50 Page 17 of 331
ATTACHMENT
1 Page 9 of 22 Emergency Action Levels 5.0 Loss of Shutdown Functions:
Decay Heat and Reactivity
5.1 Notification
of Unusual Event 05.01.01 N/A Alert 05.02.01 Complete loss of ability to maintain plant in cold shutdown: 1.Loss of essential service water loops, or Loss of RHR Loops A and B.AND 2.Loss of Condenser Condensate
System.AND 3.Either: a.Coolant temperature
exceeds 212°F, OR b.Uncontrolled
temperature
rise
05.02.02 Failure of the Reactor Protection
System to initiate and complete a scram, indicated on Panel A-5, which brings the reactor to a subcritical
condition as indicated by full core display panel P603 and neutron monitoring
instruments (APRM and IRM).IOPEP-02.1
Rev.50 Page 18 of 331
ATTACHMENT
1 Page 10 of 22 Emergency Action Levels 5.0 Loss of Shutdown Functions:
Decay Heat and Reactivity (Continued)
5.3 Site Area Emergency 05.03.01 Failure of the Reactor Protection
System to initiate and complete a scram as indicated by Section 05.02.02 above.AND Failure of standby liquid control to bring the reactor to a subcritical
condition.
05.03.02 Complete loss ofreactorheat
removal capability
indicated by inability to maintain Suppression
Pool below Heat Capacity Temperature
Limit curve.5.4 General Emergency 05.04.01 Site Area Emergency as indicated in Section 05.03.01 above lasting greater than 30 minutes.AND Loss of main condenser heat removal capability
indicated by MSIVs shut or loss of vacuum on condenser vacuum indicator.
AND EITHER 1.Failure of all low pressure coolant injection trains indicated on panel P601.OR 2.Failure of all service water trains necessary for decay heat removal indicated on panel P601 (RHR Service Water)and Panel XU2 (Nuclear and Conventional
Service Water).05.04.02 Containment
pressure approaching
Pressure Limit (PCPL), and containment
venting will be required within the next six (6)hours.IOPEP-02.1
Rev.50 Page19 of 331
ATTACHMENT
1 Page 11 of 22 Emergency Action Levels 6.0 Electrical
or Power Failures 6.1 Notification
of Unusual Event 06.01.01 Inability to power either 4 kV E Bus from off-site power.OR 06.01.02 Loss of all on-site AC power capability
indicated by failure of diesel generators
to start or synchronize.
6.2 Alert 06.02.01 Loss of all vital DC power.OR 06.02.02 Inability to power either 4 kV E Bus from off-site power.AND Loss of all on-site AC power capability
indicated by failure of diesel generators
to start or synchronize.
6.3 Site Area Emergency 06.03.01 Either Alert condition in Section 06.02.01 or 06.02.02 listed above AND lasting longer than 15 minutes.6.4 General Emergency 06.04.01 N/A IOPEP-02.1
Rev.50 Page 20 of 331
ATTACHMENT
1 Page 12 of 22 Emergency Action Levels 7.0 Fire 7.1 Notification
of Unusual Event 07.01.01 Fire located in or adjacent to the areas listed below NOT extinguished
within 15 minutes of alarm verification
or Control Room notification.
Areas: Emergency Diesel Generator Building Control Building Central Alarm Station/Secondary
Alarm Station Reactor Building Turbine Building Unit Intake Structures
Service Water Building 7.2 Alert 07.02.01 Fire which could potentially
affect vital safety-related
equipment.
7.3 Site Area Emergency 07.03.01 Any fire that impairs the operability
of any vital equipment which, in the opinion of the Site Emergency Coordinator, is essential to maintain the plant in a safe condition.
7.4 General Emergency 07.04.01 Any fire which in the opinion of the Site Emergency Coordinator
could cause massive common damage to plant systems.IOPEP-02.1
Rev.50 Page 21 of 331
ATTACHMENT
1 Page 13 of 22 Emergency Action Levels 8.0 Control Room Evacuation
8.1 Notification
of Unusual Event 08.01.01 N/A 8.2 Alert 08.02.01 Evacuation
of Control Room anticipated
orrequiredwith
control of shutdown established
from local stations.8.3 Site Area Emergency 08.03.01 Evacuation
of Control Room AND local control of shutdown is not established
in 15 minutes.8.4 General Emergency 08.04.01 N/A IOPEP-02.1
Rev.50 Page 22 of 331
ATTACHMENT
1 Page 14 of 22 Emergency Action Levels 9.0 Loss of Monitors or Alarms or Communication
Capability
9.1 Notification
of Unusual Event 09.01.01 Site communications
capability
impaired as determined
by loss of all of the following:
1.Both site Private Branch Exchanges (PBX's)2.All private phone lines (not routed through Plant Branch Exchange;Control Room, Security, Site Vice President Office)3.Selective Signaling 4.Decision Line 5.State and Local emergency management
radio system 6.Cellular phone system access 7.Satellite telephone 09.01 Unplanned loss of most or all annunciators
on Panels P601 , P603, XU-1, XU-2, XU-3, XU-51, and XU-80 for>15 minutes with the affected unit in Mode 1, 2, or 3;AND Compensatory (non-alarming)
indications
are available.
9.2 Alert 09.02.01 Unplanned loss of most or all annunciators
on Panels P601, P603, XU-1, XU-2, XU-3, XU-51, and XU-80 for>15 minutes with the affected unit in Mode 1, 2, or 3;AND Either;*Compensatory (non-alarming)
indications
are NOT available.
OR*A plant transient is in progress.IOPEP-02.1
Rev.50 Page 23 of 331
ATTACHMENT
1 Page 15 of 22 Emergency Action Levels 9.0 Loss of Monitors or Alarms or Communication
Capability (Continued)
9.3 Site Area Emergency 09.03.01 Unplanned loss of most or all annunciators
on Panels P601, P603, XU-1, XU-2, XU-3, XU-51, and XU-80 with the affected unit in Operational
Condition 1,2, or 3;AND*Compensatory (non-alarming)
indications
are NOT available.
AND*A plant transient is in progress.AND*Plant safety function indications (reactor power, reactor level, reactor pressure, containment
parameters)
are NOT available.
9.4 General Emergency 09.04.01 N/A IOPEP-02.1
Rev.50 Page 24 of 331
ATTACHMENT
1 Page 16 of 22 Emergency Action Levels 10.0 Fuel Handling Accident 10.1 Notification
of Unusual Event 10.01.01 N/A 10.2 Alert 10.02.01 Fuel handling accident involving damage to new or spent fuel indicated by: A.Observation/report
AND alarm on:1.Process Reactor Building ventilation
RAD monitor D12-K609A, B or D12-RR-R605.
OR 2.Reactor Building roof ventilation
monitor CAC-AIQ-1264-3.
OR 3.Refuel floor area monitor ARM channel 1-28 or 2-28.10.3 Site Area Emergency 10.03.01 Major damage to spent fuel indicated by:1.Observation
of substantial
damage to multiple fuel assemblies, or observation
that water level has dropped below the top of the fuel.AND 2.Indications
or alarms listed in Attachment
1, Section 10.02.01.A
above.10.4 General Emergency 10.04.01 N/A IOPEP-02.1
Rev.50 Page 25 of 331
ATTACHMENT
1 Page 17 of 22 Emergency Action Levels 11.0 Security Threats 11.1 Notification
of Unusual Event 11.01.01 Security threat or attempted entry (PA)or attempted sabotage.11.01.02 A credible site specific security threat notification.
11.01.03 A validated notification
from NRC providing information
of an aircraft threat.11.2 Alert 11.02.01 Ongoing security compromise (as determined
by security).
11.02.02 A validated notification
from NRC of an airliner attack threat less than 30 minutes away.11.02.03 A notification
from the site security force of an armed attack, explosive attack, airliner impact, or other HOSTILE ACTION within the OCA.11.3 Site Area Emergency 11.03.01 Imminent loss of physical control of the plant.11.03.02 A notification
from the site security force that an armed attack, explosive attack, airliner impact, or other HOSTILE ACTION is occurring or has occurred within the protected area.11.4 General Emergency 11.04.01 A HOSTILE FORCE has taken control of plant equipment such that plant personnel are unable to operate equipment required to maintain safety functions.
IOPEP-02.1
Rev.50 Page 26 of 331
ATTACHMENT
1 Page 18 of 22 Emergency Action Levels 12.0 Fission Product Barriers and Specific LeOs 12.1 Notification
of Unusual Event 12.01.01 Loss of containment
operability
requiring shutdown by Technical Specifications
and shutdown is not achieved within required time period.12.01.02 Loss of engineered
safety feature requiring shutdown by Technical Specifications
and shutdown is not achieved within req'uired time period.12.2 Alert 12.02.01 Loss of either Fuel Clad or the Reactor Coolant Boundary.12.3 Site Area Emergency 12.03.01 Loss of two-out-of-three
fission product barriers.12.4 General Emergency 12.04.01 Loss of any two-out-of-three
fission product barriers with a potential loss of the third barrier.IOPEP-02.1
Rev.50 Page 27 of 331
ATTACHMENT
1 Page 19 of 22 Emergency Action Levels 13.0 Hazards to Plant Operations
13.1 Notification
of Unusual Event 13.01.01 Non-hostile
Aircraft crash within site boundaries
with the potential to endanger safety-related
equipment.
13.01.02 Unplanned non-hostile
explosion within the site boundaries
with the potential to endanger safety-related
equipment.
13.01.03 Release of toxic or flammable gas that could endanger personnel.
13.01.04 Turbine rotating component failure causing rapid plant shutdown.13.2 Alert 13.02.01 Non-hostile
explosion, aircraft crash, or missile resulting in major damage to structures
housing safety-related
systems.13.02.02 Unplanned and uncontrolled
entry of toxic or flammable gases into vital areas in sufficient
quantities
to endanger personnel or the operability
of safety-related
equipment.
13.02.03 Turbine, failure causing penetration
of its outer casing.13.3 Site Area Emergency 13.03.01 Non-hostile
explosion, aircraft crash, or missile resulting in major damage to safe shutdown equipment with plant not in cold shutdown.13.03.02 Uncontrolled
entry of flammable or toxic gases into vital areas where lack of access constitutes
a safety problem with plant not in cold shutdown.13.4 General Emergency 13.04.01 Any major internal or external event substantially
beyond design basis which could cause massive common damage to plant systems.IOPEP-02.1
Rev.50 Page 28 of 331
ATTACHMENT
1 Page 20 of 22 Emergency Action Levels 14.0 Natural Events 14.1 Notification
of Unusual Event 14.01.01 Alarm on seismic monitor AND confirmation
of earthquake.
14.01.02 Hurricane warning issued.14.01.03 Tornado on site.14.2 Alert 14.02.01 Earthquake
registering
greater than 0.08g on seismic instrumentation.
14.02.02 Any adverse weather conditions
that causes a loss of function of two or more safety trains.14.02.03 Tornado striking inside protected area resulting in major damage to structures
housing safety-related
systems.14.02.04 Hurricane winds on site estimated:1.130 mph at 30 ft above ground level 2.180 mph at 300 ft above ground level 14.3 Site Area Emergency 14.03.01 Earthquake
registering
greater than 0.16g on seismic instrumentation
with plant not in cold shutdown.14.03.02 Flood, low water, or hurricane surge greater than design levels or failure to protect vital equipment at lower levels and plant not in cold shutdown.14.03.03 Plant not in cold shutdown with hurricane winds on site estimated:1.130 mph at 30 ft above ground level 2.180 mph at 300 ft above ground level IOPEP-02.1
Rev.50 Page 29 of 331
ATTACHMENT
1 Page 21 of 22 Emergency Action Levels 14.0 Natural Events (Continued)
14.4 General Emergency 14.04.01 Any major natural event substantially
beyond design basis which could cause massive common damage to plant systems.IOPEP-02.1
Rev.50 Page 30 of 331
ATTACHMENT
1 Page 22 of 22 Emergency Action Levels 15.0 Shift Superintendent/Site
Emergency Coordinator
Judgments When any condition exists which indicates a necessity for an increased level of awareness or readiness above previous plant conditions, the Shift Superintendent/Site
Emergency Coordinator
should use his judgment to declare the appropriate
emergency status for the plant.15.1 Notification
of Unusual Event 15.01.01 Plant conditions
exist that warrant increased awareness by plant staff such as exceeding any Technical Specification
safety limit.15.2 Alert 15.02.01 Plant conditions
exist that reflect a significant
degradation
in the safety of the reactor, but releases from this event would be small.15.3 Site Area Emergency 15.03.01 Plant conditions
exist that involve major failures of equipment and that will lead to coredamage.Unless
corrective
action is taken, significant
radiation releases may occur.15.4 General Emergency 15.04.01 Plant conditions
exist that make a release of a large amount of radioactivity
in a short time possible;any core melt situation.
IOPEP-02.1
Rev.50 Page 31 of 331
ATTACHMENT
6 Page 16 of 19 FIGURE 19 Unit 1 Reactor Water Level at LL-5 (Minimum Zero Injection Level)-10 REF LEG TEtv'P BELOW OR EQU.AL TO 200().F 1,150 900 1,100 1,000 700 600 800'REF LEG TEMP ABOVE 200<lF 100 300 500 60 200 400-20-60-40-50-80-90*30*70 M100..J W>W-J C W(.)-c z REACTOR PRESSURE (PSIG)WHEN REACTOR PRESSURE IS LESS THAN 60 PSIG, USE INDICATED LEVEL.LL-5 IS-47.5 INCHES.I OEOP-01-UG
Rev.51 Page 104 of 150 I
ATTACHMENT
5 Page 18 of 28 FIGURE 3 Heat Capacity Temperature
Limit-)5.50 FT-)0.25 FT-)1.25 FT-)2.50 FT-)3.25 FT-)4.25 FT 1,150..II......1iIIl'aIl UNS F E-'---SELECT E--"-""" ,.--""".,'-."""""!'-"'-......."liIIIo.."'-.""llIiI""" 1'\""-""'""""....""'"'""'-............""'-"-I"'...................."-"-'"'-I"'.."'"....'"Iil"'liIIIo..""""""-""'-'-'-ill..............
.............
...................
"'-"................,............................
...........
"""'"'-.r"....................loo..
("""'""-..I'..""'........."-!lII..I"""'lI-...
(,..........
'"'-."............
'"""'"(I'"""'"'"'"'" ," i's....('-.!Il.."'................SAFE BELOW"-I'..('-"-SELECTED LINE"'-"-('--100 300 500 700 900 1100 , o 200 400 600 800 1 ,000 REACTOR PRESSURE (PSIG)............
LL 0 220'-""" W 0:: 210::J200 0::: W 190 a.:2: 180 W t-o::: 170 W160..J 150 0 0 140 a.Z 130 0-en 120 en W 110 0::: a.a.100::::>>en SUPPRESSION
POOL WATER TEMPERATURE
IS DETERMINED
BY: CAC-TR-4426-1A, POINT WTR AVG OR CAC-TR-4426-2A, POINT WTR AVG OR COMPUTER POINT G050 OR COMPUTER POINT G051 OR CAC-TY-4426-1
OR CAC-TY-4426-2 SELECT GRAPH LINE IMMEDIATELY
BELOW SUPPRESSION
POOL WATER LEVEL AS THE LIMIT.I OEOP-01-UG
Rev.51 Page 78 of 150 I
4.0 PRECAUTIONS, LIMITATIONS
AND NOTES (Continued)
4.5.3 Trip point adjustment
on front of Master Trip Unit should not be rotated beyond its end points;otherwise, damage will occur to trip unit.4.5.4 False gross failure alarms may be caused by such actions as removing trip units from card file or pulling out the center knob of the calibration
unit and should be reset as they occur.4.5.5 If readout assembly has warmed up for ten minutes and is immediately
moved from a card file to another card file, no further warm up is required.5.0 SPECIAL TOOLS AND EQUIPMENT 5.1 510 DU readout assembly or equivalent
5.2 510 DU calibration
unit 5.3 510 DU extender board 5.4 Jumper 6.0 ACCEPTANCE
CRITERIA 6.1 Criteria stated in Step 6.1 must be met to satisfy Technical Specifications
requirements:
NOTE: Technical Specifications
allowable value for this LL2 trip is greater than or equal to 101 inches of water which relates to 11.70 mAdc.6.1.1 When a calibration
current of 12.00 (11.96 to 12.04)mAdc decreasing
is applied to Master Trip Units B21-L TM-N024A-1-1
or B21-L TM-N025A-1-1, Contacts 5-6 and 7-8 of the associated
relay (Relay A71 B-K1A or A71 B-K1 C)in Panel H12-P609 open.IOMST-PCIS21Q
Rev.3 Page 5 of 381
6.0 ACCEPTANCE
CRITERIA (Continued)
NOTE: Technical Specifications
allowable value for this LL3 trip is greater than or equal to 13 inches of water which relates to 4.99 mAdc 6.1.2 When a calibration
current of 7.43 (7.39 to 7.47)mAdc decreasing
is applied to Slave Trip Units B21-L TS-N024A-1-2
or B21-L TS-N025A-1-2,.Contacts
1-2 and 3-4 of the associated
Relay (Relay A71B-K1E or A71B-K1G)in Panel H12-P609 open.6.2 Technical Specifications
requirements
may be satisfied without meeting criteria stated in Step 6.2.6.2.1 When a calibration
current of 12.00 (11.96 to 12.04)mAdc decreasing
is applied to Master Trip Units B21-L TM-N024A-1-1
or B21-LTM-N025A-1-1, Annunciator
REACTOR VESS LO-LO WATER LEVEL SYS A (A-06 1-6)is in alarm.6.2.2 The reset differential
value of Master Trip Units B21-LTM-N024A-1-1
and-N025A-1-1
and Slave Trip Units B21-L TS-N024A-1-2
and-N025A-1-2
is within calibration
specifications
for pressures applied during test.I OMST-PCIS21Q
Rev.3 Page 6 of 381
INSTRUMENT
NUMBER: INSTRUMENT
NAME: TS REFERENCE:
B21-LT-N024A-1, B-1;B21-LT-N025A-1, B-1 Reactor Vessel Water Level Low 3.3.6.1,3.3.6.2;
and 3.3.7.1, TRM Tables 3.3.6.1-1.1a,-1.5g;3.3.6.2-1.1
TRIP CHANNEL: TRIP SYSTEM: A1-N024A-1
B1-N024B-1
A-A1 and A2 8-B1 and 82 A2-N025A-1
B2-N025B-1
TRIP LOGIC: A1 or A2 and 81 or 82=Closes all MSIVs A1 and B1=ClosesB32-F019,B21-F016, or G31-F001, RB Vent, starts SBGT and CREV A2 and B2=Closes B32-F020, B21-F019, or G31-F004, RB Vent, starts SBGT and CREV Place channel in tripped condition by: Pull fuse CHANNEL INSTRUMENT
NUMBER TRIP UNIT ACTION PANEL FUNCTION B21-LTM-N024A-1-1
A71B-F1A H12-P609 Closes G31-F001, R8 Vent, starts SBGT and A1 B21-L T-N024A-1 B21-LTS-N024A-1-2
A71B-F61A H12-P609 CREV Closes MSIVs, B32-F019, 821-F016 B1 B21-L T-N024B-1 B21-L TM-N024B-1-1
A71 B-F1 B H12-P611 Same asA1 master B21-L TS-N024B-1-2
A718-F61 B H12-P611 Same asA1 slave B21-L TM-N025A-1-1
A71B-F1C H12-P609 Closes G31-F004, RB Vent, starts SBGT and A2 B21-L T-N025A-1 B21-L TS-N025A-1-2
A71 B-F61C H12-P609 CREV Closes MSIVs, B32-F020, B21-F019 B2 B21-LT-N025B-1
B21-L TM-N025B-1-1
A71 B-F1 D H12-P611 Same as A2 master B21-L TS-N0258-1-2
A71 B-F61 D H12-P611 Same as A2 slave COMMENTS: If both channels in a trip system are inop, both channels must be tripped to assure all required functions occur.REFERENCE DRAWINGS: 1-FP-55109, 2-FP-50056
I
I Rev.57 I Page 13 of 10al
Isolation Instrumentation
3.3.6.1 3.3 INSTRUMENTATION
3.3.6.1 Primary Containment
Isolation Instrumentation
LCO 3.3.6.1 The primary containment
isolation instrumentation
for each Function in Table 3.3.6.1-1 shall be OPERABLE.APPLICABILITY:
According to Table 3.3.6.1-1.
ACTIONS-----------------------------------------------------------
NOT E-----------------------------------------------------------
Separate Condition entry is allowed for each channel.CONDITION REQUIRED ACTION COMPLETION
TIME A.One or more required A.1 Place channel in trip.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for channels inoperable.
Functions2.a,2.b, and 6.b AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Functions other than Functions 2.a, 2.b, and 6.b B.One or more Functions with B.1 Restore isolation capability.
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> isolation capability
not maintained.
C.Required Action and C.1 Enter the Condition Immediately
associated
Completion
Time referenced
in of Condition A or B not met.Table 3.3.6.1-1 for the channel.(continued)
Brunswick Unit 2 3.3-49 Amendment No.233
Isolation Instrumentation
3.3.6.1 CTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION
TIME D.As required by Required D.1 Isolate associated
main 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action C.1 and referenced
in steam line (MSL).Table 3.3.6.1-1.
OR D.2.1 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 D.2.2 Be in MODE 4.36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> E.As required by Required E.1 Be in MODE 2.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action C.1 and referenced
in Table 3.3.6.1-1.
F.As required by Required F.1 Isolate the affected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Action C.1 and referenced
in penetration
flow path(s).Table 3.3.6.1-1.
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 for Condition F not met.AND OR 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 /> As required by Required Action C.1 and referenced
in Table 3.3.6.1-1.
A (continued)
Brunswick Unit 2 3.3-50 Amendment No.233
Primary ContainmentIsolationInstrumentation
3.3.6.1 CTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION
TIME H.As required by Required H.1 Declare associated
standby 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Action C.1 and referenced
in liquid control subsystem Table 3.3.6.1-1.(SLC)inoperable.
OR H.2 Isolate the Reactor Water 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Cleanup (RWCU)System.I.As required by Required 1.1 Initiate action to restore Immediately
Action C.1 and referenced
in channel to OPERABLE Table 3.3.6.1-1.
status.OR 1.2 Initiate action to isolate the Immediately
Residual Heat Removal (RHR)Shutdown Cooling (SOC)System.A Brunswick Unit 2 3.3-51 Amendment No.233
Isolation Instrumentation
3.3.6.1 SURVEILLANCE
REQUIREMENTS
NOTES----------------------------------------------------------
1.Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment
Isolation Function.2.When a channel is placed in an inoperable
status solely for performance
of required Surveillances, entry into associated
Conditions
and Required Actions may be delayed as follows: (a)for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for Functions 2.c, 2.d, 3.a,3.b,3.e, 3.f, 3.g, 3.h, 4.a,4.b,4.e, 4.f, 4.g, 4.h, 4.i, 4.k, 5.a,5.b,5.e, 5.f, and 6.a;and (b)for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for all other Functions provided the associated
Function maintains isolation capability.
SURVEILLANCE
FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL
TEST.92 days SR 3.3.6.1.3 Calibrate the trip unit.92 days SR 3.3.6.1.4 Perform CHANNEL CALIBRATION.
92 days SR 3.3.6.1.5 Perform CHANNEL FUNCTIONAL
TEST.184 days SR 3.3.6.1.6 Perform CHANNEL CALIBRATION.
24 months SR 3.3.6.1.7 Perform LOGIC SYSTEM FUNCTIONAL
TEST.24 months (continued)
Brunswick Unit 2 3.3-52 Amendment No.233
Isolation Instrumentation
3.3.6.1 SURVEILLANCE
REQUIREMENTS (continued)
SR 3.3.6.1.8 SURVEILLANCE
NOT E S----------------------------
1.Radiation detectors are excluded.2.The sensor response time for Functions 1.a, 1.c, and 1.f may be assumed to be the design sensor response time.Verify the ISOLATION INSTRUMENTATION
RESPONSE TIME is within limits.FREQUENCY 24 months on a STAGGERED TEST BASIS SR 3.3.6.1.9 Perform CHANNEL FUNCTIONAL
TEST.Brunswick Unit 2 3.3-53 24 months Amendment No.233
Isolation Instrumentation
3.3.6.1 Table 3.3.6.1-1 (page 1 of 5)Primary Containment
Isolation Instrumentation
APPLICABLE
CONDITIONS
MODES OR REQUIRED REFERENCED
OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE
ALLOWABLE FUNCTION CONDITIONS
SYSTEM ACTION C.1 REQUIREMENTS
VALUE 1.Main Steam Line Isolation a.Reactor Vessel Water Level-Low 1,2,3 D SR 3.3.6.1.1 Level 3 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 SR 3.3.6.1.8 b.Main Steam Line Pressure-Low
E SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 c.Main Steam Line Flow-High 1,2,3 2 per D SR 3.3.6.1.1 MSL SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 SR 3.3.6.1.8 d.Condenser Vacuum-Low
1, D SR 3.3.6.1.1 2(3),3(3)SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 e.Main Steam Isolation Valve Pit 1,2,3 D SR 3.3.6.1.2 Temperature-High
SR 3.3.6.1.6 SR 3.3.6.1.7 Main Steam Line 2,3 D SR 3.3.6.1.1 Flow-High (Not in Run)SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 SR 3.3.6.1.8 2.Primary Containment
Isolation a.Reactor Vessel Water Level-Low 1,2,3 G SR 3.3.6.1.1 Level 1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 b.Drywell Pressure-High
1,2,3 G SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 (continued)(a)With any turbine stop valve not closed.Brunswick Unit 2 3.3-54 Amendment No.233
Isolation Instrumentation
3.3.6.1 Table 3.3.6.1-1 (page 2 of 5)Primary Containment
Isolation Instrumentation
FUNCTION 2.Primary Containment
Isolation (continued)
c.Main Stack Radiation-High
d.Reactor Building Exhaust
High APPLICABLE
MODES OR OTHER SPECIFIED CONDITIONS
1,2,3 1,2,3 REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS
REFERENCED
FROM REQUIRED ACTION C.1 F G SURVEILLANCE
REQUIREMENTS
SR 3.3.6.1.2 SR 3.3.6.1.6 SR 3.3.6.1.7 SR 3.3.6.1.8 SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.6 SR 3.3.6.1.7 ALLOWABLE VALUE 3.High Pressure Coolant Injection (HPCI)System Isolation a.HPCI Steam Line Flow-High 1,2,3 F SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 b.HPCI Steam Line Flow-High Time Delay 1,2,3 F SR 3.3.6.1.6 Relay SR 3.3.6.1.7 SR 3.3.6.1.9 c.HPCI Steam Supply Line Pressure-Low
1,2,3 F SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.7 d.HPCI Turbine Exhaust Diaphragm 1,2,3 F SR 3.3.6.1.2 Pressure-High
SR 3.3.6.1.6 SR 3.3.6.1.7 e.Drywell Pressure-High
1,2,3 F SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 HPCI Steam Line Area Temperature-
1,2,3 F SR 3.3.6.1.5 High SR 3.3.6.1.6 SR 3.3.6.1.7 g.HPCI Steam Line Tunnel Ambient 1,2,3 F SR 3.3.6.1.5 Temperature-High
SR 3.3.6.1.6 SR 3.3.6.1.7 h.HPCI Steam Line Tunnel Differential
1,2,3 F SR 3.3.6.1.5 Temperature-High
SR 3.3.6.1.6 SR 3.3.6.1.7 (continued)(b)Allowable Value established
in accordance
with the methodology
in the Offsite Dose Calculation
Manual.Brunswick Unit 2 3.3-55 Amendment No.233
Isolation Instrumentation
3.3.6.1 Table 3.3.6.1-1 (page 3 of 5)Primary Containment
Isolation Instrumentation
APPLICABLE
CONDITIONS
MODES OR REQUIRED REFERENCED
OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE
ALLOWABLE FUNCTION CONDITIONS
SYSTEM ACTION C.1 REQUIREMENTS
VALUE 3.HPCI System Isolation (continued)
HPCI Equipment Area Temperature-
1,2,3 F SR 3.3.6.1.5 High SR 3.3.6.1.6 SR 3.3.6.1.7 4.Reactor Core Isolation Cooling (RCIC)System Isolation a.RCIC Steam Line Flow-High 1,2,3 F SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 b.RCIC Steam Line Flow-High Time Delay 1,2,3 F SR 3.3.6.1.6 Relay SR 3.3.6.1.7 SR 3.3.6.1.9 c.RCIC Steam Supply Line Pressure-Low
1,2,3 F SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.7 d.RCIC Turbine Exhaust Diaphragm 1,2,3 F SR 3.3.6.1.2 Pressure-High
SR 3.3.6.1.6 SR 3.3.6.1.7 e.Drywell Pressure-High
1,2,3 F SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 RCIC Steam Line Area Temperature-
1,2,3 F SR 3.3.6.1.5 High SR 3.3.6.1.6 SR 3.3.6.1.7 g.RCIC Steam Line Tunnel Ambient 1,2,3 F SR 3.3.6.1.5 Temperature-High
SR 3.3.6.1.6 SR 3.3.6.1.7 h.RCIC Steam Line Tunnel and Area 1,2,3 F SR 3.3.6.1.5 Temperature-High
Time Delay SR 3.3.6.1.6 SR 3.3.6.1.7 RCIC Steam Line Tunnel Differential
1,2,3 F SR 3.3.6.1.5 Temperature-High
SR 3.3.6.1.6 SR 3.3.6.1.7 (continued)
Brunswick Unit 2 3.3-56 Amendment No.233
Brunswick Unit 2 3.3-57 Amendment No.233
Isolation Instrumentation
3.3.6.1 Table 3.3.6.1-1 (page 5 of 5)Primary Containment
Isolation Instrumentation
FUNCTION 6.RHR Shutdown Cooling System Isolation a.Reactor Steam Dome Pressure-High
b.Reactor Vessel Water Level-Low Level 1 APPLICABLE
MODES OR OTHER SPECIFIED CONDITIONS
1,2,3 3,4,5 REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS
REFERENCED
FROM REQUIRED ACTION C.1 F SURVEILLANCE
REQUIREMENTS
SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.7 SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 ALLOWABLE VALUE (d)In MODES 4 and 5, provided RHR Shutdown Cooling System integrity maintained, only one channel per trip system with an isolation signal available to one RHR shutdown cooling pump suction isolation valve is required.Brunswick Unit 2 3.3-58 Amendment No.233
5.7 Recirculation
and Sampling of Saltwater Release Tank#1 5.7.1 Initial Conditions
1.All applicable
prerequisites
as listed in Section 4.0 are D met.5.7.2 Procedural
Steps Saltwater Release Tank 1.PERFORM the following to ensure the Saltwater Release Tank can NOT receive inputs from the
sources during recirculation
and sampling: a.ENSURE the Unit 1 Breezeway North end West D side mop water drain tube is locked.b.DISCONNECT
electrical
supply to pipe tunnel dike D portable bilge pump.c.REMOVE Turbine Building portable sump pump D discharge hose from Saltwater Release Tank.d.DIVERT routing of air wash water to one of the D following in accordance
with 1 (2)OP-37.3:
-Unit 1 or Unit 2 TB Floor Drain Sumps OR-Unit 1 or Unit 2 TB Equipment Drain Sumps 2.IF the tank is being recirculated
through the Saltwater D Release System Filters in accordance
with Section 5.8, THEN GO TO Step 5.7.2.6.3.OPEN SAL TWA TER RELEASE SYSTEM D RECIRCULATION
VALVE, 1-SWR-V11.
IOOP-06.4 Rev.44 Page 28 of 1071
5.7.2 Procedural
Steps NOTE: Saltwater Release Tank level may drop 0-4°1b when placed in recirculation.
4.WHEN placing Saltwater Release Tank in recirculation, D THEN RECORD tank level prior to starting pump.IOOP-06.4 Tank Level.5.START Saltwater Release System Pump#1.D 6.SAMPLE Saltwater Release Tank#1 by completing
the following:
a.ALLOW Saltwater Release Tank to recirculate
for D 4 minutes for each percent of indicated tank volume.b.OPEN SAL TWATER RELEASE SYSTEM D SAMPLE STATION VALVE, 1-SWR-V17.
c.ALLOW sample to run for at least 5 minutes to D ensure a representative
sample is obtained.d.OBTAIN sample in accordance
with E&RC-2009.
D e.CLOSE SAL TWA TER RELEASE SYSTEM D SAMPLE VALVE, 1-SWR-V17.
7.IF tank activity is greater than aDCM limits, OR D additional
filtration
is desired, AND the tank is being recirculated
through the Saltwater Release System Filters in accordance
with Section 5.8, THEN CONTINUE recirculation.
8.IF tank activity is greater than allowed aDCM limits, OR additional
filtration
is desired, AND the tank is NOT being recirculated
in accordance
with Section 5.8, THEN: a.PERFORM Section 7.6 AND, D b.PERFORM Section 5.8 to conduct cleanup.D Rev.44 Page 29 of 1 07\
5.7.2 Procedural
Steps 9.IF tank activity is within aDeM limits AND it is desired to release Saltwater Release Tank#1, THEN PERFORM Section 5.9.D 10.IF desired, THEN SHUT DOWN recirculation
in accordance
with: IOOP-06.4 a.b.Section 7.6, OR Section 7.5.Rev.44 D D Page 30 of 1071