Text

November 2008 Exam 05000325, 324/2008302 - Final Ro/Sro Written Exam References
	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)
	v • d • e

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, Attachment 1, 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

~ 280 -------..~~~~~........,......--...,..-~_r__r_~~.,...._."_~..........,....~....,.__....r__r___"r___r__....,...__.r_~___,,

o

~

260 '""~"'FI="'+"""'+~"f'''if~''4''''+'''f''+''*""t""""+~4"",,,+,,,,,~*...,"'-t'w"""!-="'+'''F'*~*'''f"+='f'""44"=,+.'""-t''''fw,=+,,,,"+,,,,,,+.=-t==f'='f-='~

250 -R"''.t'''+U''i-o'+''''f'+'''!-''+WW+'''~4W''{-w~""~f""~""""{--"'4"w*+""'+-""+='}-'4'"";;*=--~---+=+'~+Ww4=+~=~p...=+w"'""~

230 -t--+--~~~~~--4--+--+---+--+---+---4F-+-+--+-~~~ ~_?~ P~)IG 220 ~~EEm33~fiE~~nttm§§ERE """""""' ................

170 -.f-w-+-~~r---+-~~~+--+-~~~~~~~~~~+--+--~~~

VNmm g "

160 _+-I+-..- ", -+--+--;..-.....jf--+.-+--+--+-+--+--+--+-~ -+-+--tj -+--+- l

-f--f--+--+-f--+--+--+--+-+--+-+--4-::l~~.:.....-+--l----i 150 _u"-t--"'"wu=.,=,="~"!""",~",u",,,,-:uuu~,,,uu;,,,--+--+--w4Iwu=-,,,,u:"':=*~~i--+I**~*** *I-"'" -f- -+- +- -~+- t-l*I" "- "0I. . . . .". PIS- r~G" " "'I" " "-I-+-: : ;-+-: : : ~: : : : :*

u 1,000 2,000 3,000 4,.000 5,000 6,000 7,000 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 81 of 150 I

ATTACHMENT 5 Page 22 of 28 FIGURE 6 RHR NPSH Limit U. 290 0

~

W 280

~

)

40 PSIG I--------

270 -~ I--------

~ 260

~

W C.

~

250 W

I- 240 - ~ ........ ........ ~

20 PSIG -

~

W 230

~

3: 220 ------- - ----- -~

..J ~

10 PSIG I--------

0 210 0 ------ ~

C. 200 ~

Z I'

5 PSIG I--------

I--------

-enen 0 190

~

W 180 -~

~ """ """'" "-

C. 170 ""ll C.

~

0 PSIG I--------

f-----

)

en 160 I I I 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 L~vel at LL-4 (Minimum Steam Cooling Level) 0

-10

~ -20 CJ)

W J: -30 0

Z

~

-40

..J W

>W - 50 ':<A-.u.:.x,,~~.r ..-.:.,uu.r..~.:

..J 0 .. 60 ~ ............ .........

~ _~ REF LEG W ;

TEMP ABOVE

~ -70 .... ... ..._'"" .....

200(JF 0

-C

~~.,,_ ~

REF LEG

{ TEMP BELOW OR Z - 80 EQUAL TO 200'~F

-90

-100 1,1*50 00 300 500 900 1,100 60 200 400 1,000 REACTOR PRESSURE (PSIG)

WHEN REACTOR PRESSURE IS LESS THAN 60 PSIG, USE INDICATED LEVEL.

LL-4 IS -30.0 INCHES.

IOEOP-01 ~UG 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

OPENING BIAS 1000/0 WHEN TURBINE TRIP SLOW MEDIUM FAST ANY SPEED 60 90 180 SELECTED R~ T RiM TURBINE SPEED I ... 1.i.~~":;J

\: .

-<J::.;.:~~~:1**~ ...,

(-) (+)

K152 l S00

.f1RPM I.j500 RPM

'800 RPM "D RPM 00

+1400/0 ALL I

I lOvERSPEED

. "f-J VALVES CLOSED 7 TRIP TEST

. .1000/0 POWER LOAD UNBALANCE 'I SYNC SPEED (o.0!cJ)

~I .:~800 RPM RUNBACKS r-GENERATOR PCB CLOSED I CONTROL VALVE TSV<90o/~l. __ DEMAND (To DFGs)

PAM PRESSURE +/- Open Shell warm~,

0..1050 PSI ..

~.i.I.*.*N

~I . * . *.:*.*.:.*'.*.'**'.'*::

PRESSURE SET I ... IM~qr.(!):fRI 30 150-1050 PSI

~ ':':\.:,.'\l':. . LL.-....

BIAS 1.5 PSI

~ UV/o ~::1i~'::it:i~:=g~ UfO ... BYP VLV DEMAND PAM PRESSUR~t::: ~':>J I 10k T CLOSE 7" HG 0..1050 PSI +/- SM'ALL BYPASS LOW VACUUM CLOSE BIAS VALVE JA'CK TRIP

ATTACHMENT 1 Page 1 of 24 Emergency Action Levels Section Event Category 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 in increased drywell 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 pressure ~ 1250 psig.

OR 02.01.02 Inability to close an SRV with Reactor Coolant System pressure

~ 900 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 13 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

03.01.02 Gaseous A. Steam jet air ejector off-gas radiation monitor (D12-RM-K601 A 4

and B) reading of greater than 1.2 x 10 mR/hr.

B. Steam jet air ejector off-gas radiation monitor (D12-RM-K601 A 3

and B) increase of greater than 2.4 x 10 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) 5 reading of greater than 1.2 x 10 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 Page 14 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 Page 15 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 10CFR20, 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 (CDE) 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 approach~ng 212~.

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 of reactor heat 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 Primary Containment Pressure Limit (PCPL), and containment venting will be required within the next six (6) hours.

IOPEP-02.1 Rev. 50 Page 19 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 or required with 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 ~02 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 core damage. 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

- 20

30

..J -40 W

W

-J - 50 C

W

- 60

~

(.)

-c

70 z REF LEG TEMP ABOVE 200<lF

- 80 REF LEG TEtv'P BELOW OR

- 90 EQU.AL TO 200().F M100 1,150 100 300 500 700 900 1,100 60 200 400 600 800 1,000 REACTOR PRESSURE (PSIG)

WHEN REACTOR PRESSURE IS LESS THAN 60 PSIG, USE INDICATED LEVEL.

LL-5 IS -47.5 INCHES.

IOEOP-01-UG Rev. 51 Page 104 of 150 I

ATTACHMENT 5 Page 18 of 28 FIGURE 3 Heat Capacity Temperature Limit LL 0

'-""" 220 .... I I W ...

0:: 210 E --

1iIIl'aIl UNS F tt~

J SELECT E

~ 200 "-

""" ~'"

0::: !'-'- ."" "" "

W 190 " "" "- '-.

~ liIIIo..

a. 1'\

""llIiI """

~

2: ""- "-

~

I"'..

~

180 '- I"'..

W '"

.... '"Iil t- ""'- '- "' '-

liIIIo..

ill..

~

o::: 170

", ~

~

.........loo..

W " "-..

" I'.. "

r"..

~ "'" ""' ~

(-) 0.25 FT

~ (-) 1.25 FT

"- !lII.. ~ I"""'lI-...

160 ~

~

'"'" I' ~ ""

~ ~

'" '" """ '" i's.... ~

"'" ~ (-) 2.50 FT

..J 150 '" , ~ " ~ (-) 3.25 FT 0 '-.

!Il..

....... "' ......... I'..

0 140 "- (-) 4.25 FT

a. SAFE BELOW '-

Z 130 SELECTED LINE "'-

-enen 0

120

~ (-) 5.50 FT W

0::: 110 a.

a. 100

>> 1,150 en 100 300 500 700 900 1100 ,

o 200 400 600 800 1,000 REACTOR PRESSURE (PSIG)

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 potentiometer 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-LTM-N024A-1-1 or B21-LTM-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-LTS-N024A-1-2 or B21-LTS-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-LTM-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-LTS-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: B21-LT-N024A-1, B-1; B21-LT-N025A-1, B-1 INSTRUMENT NAME: Reactor Vessel Water Level Low TS

REFERENCE:

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: A1-N024A-1 A2-N025A-1 B1-N024B-1 B2-N025B-1 TRIP SYSTEM: A-A1 and A2 8-B1 and 82 TRIP LOGIC: A1 or A2 and 81 or 82 = Closes all MSIVs

=

A1 and B1 Closes B32-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 Closes G31-F001, R8 Vent, starts SBGT and B21-LTM-N024A-1-1 A71B-F1A H12-P609 A1 B21-LT-N024A-1 CREV B21-LTS-N024A-1-2 A71B-F61A H12-P609 Closes MSIVs, B32-F019, 821-F016 B21-LTM-N024B-1-1 A71 B-F1 B H12-P611 Same as A 1 master B1 B21-LT-N024B-1 B21-LTS-N024B-1-2 A718-F61 B H12-P611 Same as A 1 slave Closes G31-F004, RB Vent, starts SBGT and B21-LTM-N025A-1-1 A71B-F1C H12-P609 A2 B21-LT-N025A-1 CREV B21-LTS-N025A-1-2 A71 B-F61C H12-P609 Closes MSIVs, B32-F020, B21-F019 B21-LTM-N025B-1-1 A71 B-F1 D H12-P611 Same as A2 master B2 B21-LT-N025B-1 B21-LTS-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 OOI~1a I Rev. 57 I Page 13 of 10al

Primary Containment 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

NOTE-----------------------------------------------------------

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for channels inoperable. Functions 2.a, 2.b, and 6.b AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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

Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D.1 Isolate associated main 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND D.2.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. As required by Required E.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time for Condition F not met. AND OR G.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months As required by Required Action C.1 and referenced in Table 3.3.6.1-1.

(continued)

Brunswick Unit 2 3.3-50 Amendment No. 233

Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME H. As required by Required H.1 Declare associated standby 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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.

Brunswick Unit 2 3.3-51 Amendment No. 233

Primary Containment 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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

Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.6.1.8 ----------------------------------NOTES----------------------------

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 24 months on a RESPONSE TIME is within limits. STAGGERED TEST BASIS SR 3.3.6.1.9 Perform CHANNEL FUNCTIONAL TEST. 24 months Brunswick Unit 2 3.3-53 Amendment No. 233

Primary Containment 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

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 2 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

2. Primary Containment Isolation (continued)

c. Main Stack Radiation-High 1,2,3 F SR 3.3.6.1.2 SR 3.3.6.1.6 SR 3.3.6.1.7 SR 3.3.6.1.8

d. Reactor Building Exhaust Radiation- 1,2,3 G SR 3.3.6.1.1 High SR 3.3.6.1.2 SR 3.3.6.1.6 SR 3.3.6.1.7

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

Primary Containment 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 Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 5 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

6. RHR Shutdown Cooling System Isolation

a. Reactor Steam Dome Pressure-High 1,2,3 F SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.7

b. Reactor Vessel Water Level- Low 3,4,5 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 (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 fol~owing 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.

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 IOOP-06.4 Rev. 44 Page 29 of 107\

5.7.2 Procedural Steps

9. IF tank activity is within aDeM limits AND it is desired to D release Saltwater Release Tank #1, THEN PERFORM Section 5.9.

10. IF desired, THEN SHUT DOWN recirculation in accordance with:

a. Section 7.6, OR D

b. Section 7.5. D IOOP-06.4 Rev. 44 Page 30 of 1071

ML083380337

Text

REFERENCE:

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