Text

Attachment 4 - NEI 99-01, Rev. 6, Deviations and Differences, Seabrook Station Nuclear Power Plant - Unit 1
	ML16068A137
Person / Time
Site:	Seabrook
Issue date:	02/27/2016
From:	; NextEra Energy Seabrook
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML16068A128	List: ML16068A130; ML16068A131; ML16068A137; ML16068A139; SBK-L-15120, Independent Spent Fuel Storage Installation (ISFSI) Ics/Eals;
References
	SBK-L-15120 NEI 99-01, Rev. 6
	Download: ML16068A137 (73)
	v • d • e

ATTACHMENT 4 NEI 99-01, Rev. 6, Deviations and Differences, Seabrook Station Nuclear Power Plant -

Unit 1

NEI 99-01 Rev 6 Deviations and Differences Seabrook Station Nuclear Power Plant - Unit I

N ET99-01 Rev 6 ________________earo__atin ___l ___o____ l__ t_____________________iii ii~ii i...

___________li!

References BWRs Deleted_____BWR ___references_______as__appropriate_____

Uses A for radiological effluent/radiation level ICs Uses R for radiological effluent/radiation level ICs Uses E-HU for ISFSI ICs Uses______EU____for_____ISFSI_______I_____

Uses S for System Malfunction ICs Uses____M __for___System_____Malfunction_______I__s Emergency Classification ICs are presented in ascending order (NOUB - GE) Emergency Classification ICs are presented in descending order (GE -

All NOTEs made site specific by identifying the STED/ED as the user.

Site specific information is highlighted in yellow.

No deviations were indicated to complete the upgrade to Revision 6.

1

ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS RGI: INITIATING CONDITIONS Difference/fJustification None

~THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant 2

ABNORMAL RAD LEVELS /RADIOLOGICAL EFFLUENT ICS/EALS (1) Reading on ANY of the following radiation monitors greater (l) Reading on ANY of the following radiation monitors greater than the than the reading shown for 15 minutes or longer: reading shown for 15 minutes or longer:

(site-specific monitor list and threshold values) Monitor Reading (2) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at RM-6528-4 (WRGM rate) 2.85E+8 uCi/sec or beyond (site-specific dose receptor point). Time After Shutdown Reading (3) Field survey results indicate EITHER of the following at or

<lhr >lhrto <2hrs beyond (site-specific dose receptor point):

Closed window dose rates greater than 1,000 mR/hr RM-6481-1" (MSL A) 1310 mR/hr ! 060mR/hr expected to continue for 60 minutes or longer. RM-6482- * (MSL B) 1310 mR/hr 1060OmR/hr

Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation. RM-6482-2" (MSL C) 1310 mR/hr 1060 mR/hr RM-I6481-2" (MSL D) 1310 mR/hr 1060OmR/hr

With release path to the environment from affected steam line, e.g.,

open ASDV or SRV, line is faulted, open steam supply to 1 FW P 37A.

(2) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the site boundary.

(3) Field survey results indicate EITHER of the following at or beyond the site boundary:

Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.

Difference /Justification RGI.I: Site specific information, see V3 EPCALC-06 Effluent Monitor Values for R EALs RG1.2 & 3: Site specific information, see V4 ODCM and TS Basis for Site Boundary Receptor Point 3

ABNORMAL RAD LEVELS /RADIOLOGICAL EFFLUENT ICS/EALS RG2: INITIATING CONDITIONS NET 99-01 Rev 6 Seabrook Station Nuclear Power Plant Spent fuel pool level cannot be restored to at least (site-specific Level 3 Spent fuel pool level cannot be restored to at least 1.5 ft. (Level3) for 60 description) for 60 minutes or longer. minutes or longer.

Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Spent fuel pool level cannot be restored to at least (site-specific (1) Spent fuel pool level cannot be restored to at least 1.5 ft. above the fuel Level 3 value) for 60 minutes or longer, racks for 60 minutes or longer as indicated by SF-LI-26 16 (MPCS computer point A4172) or SF-LI-2617 (MPCS computer point A4220)

Difference /Justification RG2. 1: Site specific information, see V2 SFP Levels Drawing 4

ABNORMAL RAD LEVELS /RADIOLOGICAL EFFLUENT ICS/EALS RSI: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant 5

ABNORMAL RAD LEVELS /RADIOLOGICAL EFFLUENT ICS/EALS (1) Reading on ANY of the following radiation monitors (1) Reading on ANY of the following radiation monitors greater than the reading greater than the reading shown for 15 minutes or longer: shown for 15 minutes or longer:

(site-specific monitor list and threshold values)

(2) Dose assessment using actual meteorology indicates Monitor Reading doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond (site-specific dose receptor point). RM 6528 4 (WRGM rate) 2.85E+7 uCi/sec (3) Field survey results indicate EITHER of the following at Time After Shutdown Reading or beyond (site-specific dose receptor point):

Closed window dose rates greater than 100 mR/hr <lhr >lhrto <2hrs expected to continue for 60 minutes or longer. RM-648l1-"(MSL A) 130OmRihr 100 mR/hr

Analyses of field survey samples indicate thyroid RM-6482-1" (MSL B) 130 mR/hr 100 mRihr CDE greater than 500 mrem for one hour of inhalation. RM-6482-2" (MSL C) 130 mR/hr 100 mR/hr RM-6481-2" (MSL D) 130 mR/hr 100 mR/hr

With release path to the environment from affected steam line, e.g., open ASDV or SRV, line is faulted, open steam supply to 1-FW-P-37A, etc.

(2) Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the site boundary.

(3) Field survey results indicate EITHER of the following at or beyond the site boundary:

Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.

Difference /Justification RSI.I: Site specific information, see V3 EPCALC-06 Effluent Monitor Values for R EALs RS1.2 & 3: Site specific information, see V40ODCM and TS Basis for Site Boundary Receptor Point 6

C/EL ABNOMALRAD EVES2 INIRADIOLGICALNEFFLUNTS RS2: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NET 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Lowering of spent fuel pooi level to (site-specific Level 3 value). (1) Lowering of spent fuel pooi level to 1.5 ft above the fuel racks as indicated by SF-LI-26 16 (MPGCS computer point A4 172) or SF-LI-26 17 (MPC S computer point A4220).

Difference /Justification

[RS2.1: Site specific information, see V2 SFP Levels Drawing 7

ABNORMAL RAD LEVELS /RADIOLOGICAL EFFLUENT ICS/EALS RAl: INITIATING CONDITIONS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant Release of gaseous or liquid radioactivity resulting in offsite dose Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE greater than 10 mrem TEDE or 50 mrem thyroid CDE. or 50 mrem thyroid CDE.

Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 /Seabrook Station Nuclear Power Plant 8

ABNORMAL RAD LEVELS /RADIOLOGICAL EFFLUENT ICS/EALS (1) Reading on ANY of the following radiation monitors greater than the reading shown (1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: for 15 minutes or longer:

(site-specific monitor list and threshold values)

(2) Dose assessment using actual meteorology indicates doses Monitor Readings greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site-specific dose receptor point). RM-6528-4 (WRGM rate) 2.85E+6 uCi/sec (3) Analysis of a liquid effluent sample indicates a concentration RM-6481-1" (MSL A) 10 mR/hr or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site-specific RM-6482-1" (MSL B) 10 mR/hr dose receptor point) for one hour of exposure. RM-6482-2" (MSL C) 10 mR/hr (4) Field survey results indicate EITHER of the following at or RM-6481-2" (MSL D) 10 mR/hr beyond (site-specific dose receptor point):

With release path to the environment fr'om affected steam line, e.g., open ASDV or

Closed window dose rates greater than 10 mRihr expected SRV, line is faulted, open steam supply to 1-FW-P-37A, etc.

to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE (2) Dose assessment using actual meteorology indicates doses greater than 10 mrem greater than 50 mrem for one hour of inhalation.

TEDE or 50 mrem thyroid CDE at or beyond the site boundary.

(3) Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the site boundary for one hour of exposure.

(4) Field survey results indicate EITHER of the following at or beyond the site boundary:

Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.I A.

Difference /Justification RAI.I: Site specific information, see V3 EPCALC-06 Effluent Monitor Values for R EALs RA1.2, 3, 4: Site specific information, see V40ODCM and TS Basis for Site Boundary Receptor Point 9

ABNORMAL RAD LEVELS /RADIOLOGICAL EFFLUENT ICS/EALS RA2: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Uncovery of irradiated fuel in the REFUELING PATHWAY. (I) Uncovery of irradiated fuel in the REFUELING PATHWAY.

(2) Damage to irradiated fuel resulting in a release of radioactivity from (2) Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following radiation monitors: the fuel as indicated by high-alann, or reading in excess of the (site-specific listing of radiation monitors, and the associated current high-alarm setpoint on ANY of the following radiation readings, setpoints and/or alarms) monitors:

(3) Lowering of spent fuel pooi level to (site-specific Level 2 value). I RM-65 18-1, FSB High Range

[See Developer Notes] I RM-6562-1, FSB Vent I RM-6535A-1, Manip Crane I SRM-6535B-1, Manip Crane (3) Lowering of spent fuel pool level to 12 ft. 3 inches above the fuel racks on SF-LI-2616 (MPCS computer point A4172) or SF-LI-2617 (MIPCS computer point A4220).

Difference /Justification RA2.1: Site specific information, see V6 Refueling pathway RU2 RA2 RA2.2: Site specific information, see V5 UFSAR Table 12.3 CTMT Post-LOCA Range RA2.3: Site specific information, see V2 SFP Levels Drawing 10

ABNORMAL RAD LEVELS /RADIOLOGICAL EFFLUENT ICS/EALS RA3: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant 11

ABNORMAL RAD LEVELS /RADIOLOGICAL EFFLUENT ICS/EALS (1) Dose rate greater than 15 mR/hr in ANY of the following areas: (1) Dose rate greater than 15 mR/hr in ANY of the following areas:

Control Room SControl Room

Central Alarm Station SCentral Alarm Station (CAS) by survey

(other site-specific areas/rooms)

[Secondary Alarm Station (SAS) by survey (2) An UNPLANNED event results in radiation levels that prohibit or impede access to any of the following plant rooms or areas: (2) An UNPLANNED event results in radiation levels that prohibit or (site-specific list of plant rooms or areas with entry-related mode IMPEDE access to any of the following plant rooms or areas:

applicability identified) Table HI Area Mode Primary Aux Building 25 ft elevation1,34 7 ft elevation 1 ,3

- 26 ft elevation Turbine Building 1, 2, 3 Switchgear Rooms Essential 1, 2,3, 4 Non-essential Steam and Feedwater Pipe chases 1, 2, 3 Waste Process Building 25 ft elevation1,23

-3 ft elevation1,23

-31 ft elevation Containment 3, 4 Equipment Vaults 3, 4 Difference /Justification Table HI: Site specific information, see V7 - Table Hi Procedure References 12

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS RUI: INITIATING CONDITIONS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant Release of gaseous or liquid radioactivity greater than 2 times the (site-specific Release of gaseous or liquid radioactivity greater than 2 times the effluent release controlling document) limits for 60 minutes or longer. ODCM limits for 60 minutes or longer.

Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Reading on ANY effluent radiation monitor greater than 2 times the (site- (1) Reading on ANY non-isolated effluent path radiation monitor specific effluent release controlling document) limits for 60 minutes or greater than 2 times the ODCM limits for 60 minutes or longer: longer: ____________

(site-specific monitor list and threshold values corresponding to 2 times the IRM-6509- 1 (WTT Disch) 1 controlling document limits) IRM-6521-1 (TB Sump) 1 (2) Reading on ANY effluent radiation monitor greater than 2 times the alarm t M6I1- S setpoint established by a current radioactivity discharge permit for 60 minutes RM-619do (Sn)

(3) Sample analysis for a gaseous or liquid release indicates a concentration or RM-6473-1 (WT LIQ release rate greater than 2 times the (site-specific effluent release controlling EFF) document) limits for 60 minutes or longer. [RM-6528-4 (WRGM rate) ]

(2) Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.

(3) Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer.

Difference /Justification RUI. 1: Site specific information, see V8 012_Table 03 UFSAR WRGM Ranges and V32 UFSAR Table 11.5-1 13

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS 14

COLSHTDON /REUELIITAINGSYSDTEMMAFNCTOSCSEL RU2: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. UNPLANNED water level drop in the REFUELING (1) a. UNPLANNED water level drop in the REFUELING PATHWAY as PATHWAY as indicated by ANY of the following: indicated by ANY of the following:

(site-specific level indications). l-SF-LI-2607 (Spent Fuel Pool Level)I AND [__-F-LI2_29or_1SF-LT-2_9-1_ReacorRfuelCaviyLeel)_

b. UNPLANNED rise in area radiation levels as indicated by I S-I22 rI-F-L-69 Ratr eulCvt'Lvl ANY of the following radiation monitors. AND

b. UNPLANNED rise in area radiation levels as indicated by ANY of (site-specific list of area radiation monitors) the following radiation monitors:

I RM-653 5-A-i, Containment Manipulator Crane I

[ RM-6535-B- 1, Containment Manipulator Crane I

[ RM-6549-1, FSB Spent Fuel Range Low I

[RM-6518-1, FSB Spent Fuel Range Hi]

Difference/fJustification CGI.la: Site specific information, see V9 - SFP level CGI.lb: Site specific information, see VlI0 UFSAR Table 12.3 CTMT Post-LOCA Range 15

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS CGI: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NET 99-01 Rev 6 [Seabrook Station Nuclear Power Plant 16

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS (1) a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level less than (1) a. RVLIS Full Range <255% (-141.5 in) for 30 minutes or (site-specific level) for 30 minutes or longer. longer.

AND AND

b. ANY indication from the Containment Challenge Table (see b. ANY indication from the Containment Challenge Table C2.

below). (2) a. Reactor vessel/RCS level cannot be monitored for 30 minutes or (2) a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be longer.

monitored for 30 minutes or longer. AND AND) b. Core uncoverv is indicated by ANY of the following:

b. Core uncovery is indicated by ANY of the following: RM-6535A- 1 (Manipulator Crane) reading greater than 9500 mRihr

(Site-specific radiation monitor) reading greater than (site-specific value) RM-6535SB-I (Manipulator Crane) reading greater than 9500 mRihr

Erratic source range monitor indication [PWR] Erratic source range monitor indication

UNPLANNED increase in (site-specific sump and/or tank) UNPLANNED increase in Containment Sumps A or B levels of levels of sufficient magnitude to indicate core uncovery sufficient magnitude to indicate core uncovery.

(Other site-specific indications) Visual observation.

AND AND C. ANY indication from the Containment Challenge Table (see c. ANY indication from the Containment Challenge Table C2.

below). Containment Challenge Table C2 Containment Challenge Table

CONTAINMENT CLOSURE not established*

CONTAINMENT CLOSURE not established*

Containment H2 concentration > 6%

(Explosive mixture) exists inside containment

UNPLANNED increase in containment pressure

UNPLANNED increase in containment pressure

Secondary containment radiation monitor reading above (site-

If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is specific value) [BWR] not required.

If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

Difference /Justification CGI.la: Site specific information, see VI11 EPCALC-06 RVLIS Values CG1.2b: Site specific information, see VIO0 UFSAR Table 12.3 Manipulator Crane Monitor Range and V 13 Containment Sumps CG1.2c: Site specific information, see V14 H2 concentration in containment 17

COLDSHUTDOWN INREFUELINGSYSDTEMMAFNCTOSCSEL CS1: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. CONTAINMENT CLOSURE not established. (1) a. CONTAINMENT CLOSURE not established.

AND AND

b. (Reactor vessel/RCS [PWR] or RPV [BWRI) level less than b. RVLIS Full Range < 63% (-101.9 in).

(site-specific level). (2) a. CONTAINMENT CLOSURE established.

(2) a. CONTAINMENT CLOSURE established. AND AND b. RVLIS Full Range <55% (-141.5 in).

b. (Reactor vesseliRCS [PWR] or RPV [BWR]) level less than (3) a. Reactor vesselIRCS level cannot be monitored for 30 minutes or (site-specific level), longer.

(3) a. (Reactor vessel/RCS [PWRI or RPV [BWR]) level cannot be AND monitored for 30 minutes or longer. b. Core uncovery is indicated by ANY of the following:

AND RM-6535A-1 (Manipulator Crane) reading greater than 9500 mR/hr

b. Core uncovery is indicated by ANY of the following: RM-6535SB- 1 (Manipulator Crane) reading greater than 9500 mR/hr

(Site-specific radiation monitor) reading greater thanEraisoceanem itrndain (site-specific value)Eraisorernem itrndain

Erratic source range monitor indication [PWR] UNPLANNED increase in Containment Sumps A or B levels of

UNPLANNED increase in (site-specific sump and/or sufficient magnitude to indicate core uncovery.

tank) levels of sufficient magnitude to indicate core Visual observation.

uncovery

(Other site-specific indications)

Difference/fJustification 18

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS CSl.lb CSl.3b: &Site specificSite CS1.2b: specific information, information, see V11Table see V12 UFSAR EPCALC-06-04 - RVLIS 12.3 CTMT ValuesIRange and V13 Containment Sumps Post-LOCA 19

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS CAI: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NET 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Loss of (reactor vesselIRCS [PWR] or RPV [BWR]) inventory as (1) Loss of reactor vessel/RCS inventory as indicated by RVLIS full range indicated by level less than (site-specific level). < 64% (-96.9 in).

(2) a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be (2) a. Reactor vessel/RCS level cannot be monitored for 15 minutes or monitored for 15 minutes or longer longer.

AND AND

b. UNPLANNED increase in (site-specific sump and/or tank) b. UNPLANNED increase in Containmnent Sumps A or B levels due levels due to a loss of (reactor vessel/RCS [PWR] or RPV to a loss of reactor vessel/RCS inventory.

[BWR ]) inventory.

Difference /Justification CAI.lb: Site specific information, see V1 1 EPCALC-06 RVLIS Values CAI .2b: Site specific information, see VI13 Containment Sumps 20

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS CA2: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NET 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Loss of ALL offsite and ALL onsite AC Power to (site-specific NOTE: For a bus to be considered energized from SEPS, both SEPS diesel emergency buses) for 15 minutes or longer, generator sets must be functional.

(1) Loss of ALL offsite and ALL onsite AC Power to BOTH AC emergency buses E5 AND E6 for 15 minutes or longer.

Difference /Justification Added NOTE for consideration of an additional non-safety power supply Supplemental Emergency Power System (SEPS) 21

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS CA3: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) UNPLANNED increase in RCS temperature to greater than (site- (1) UNPLANNED increase in RCS temperature to greater than 2000 F for specific Technical Specification cold shutdown temperature limit) greater than the duration specified in the following table.

for greater than the duration specified in the following table.

Table C1 - RCS Heat-up Duration Thresholds Table: RCS Heat-up Duration ThresholdsCOTIMN RCS Status Containment Closure Heat-up RCS Status CLSR tts Heat-up Duration Status Duration COUESau invntat(unoty atWRedce Not applicable 60 minutes* INTACT and reactor Not applicable 60 minutes*

invetor [PR])vessel>_ -36 inches Not intact (or at reduced Established 20 minutes*

inventor [PWR) Not Established 0 minutes Not INTACT or reactor Established 20 minutes*

If an RCS heat removal system is in operation within this time frame and vessel < -36 inches Not Established 0 minutes RCS temperature is being reduced, the EAL is not applicable.

If RHR is in operation within this time frame and RCS temperature is (2) UNPLANNED RCS pressure increase greater than (site-specific being reduced, the EAL is not applicable.

pressure reading). (This EAL does not apply during water-solid plant conditions. [PWR]) (2) UNPLANNED RCS pressure increase greater than 25 psig. (This EAL does not apply during water-solid plant conditions.)

Difference /Justification CA3.1: Site specific information, see VI15 Cold SD Temp Limit TS and V1I6 - Reduced Inventory CA3.2: Site specific information, see VI 7 RCS Pressure range 22

COLDSHUTDOWN INREFUELINGSYSDTEMMAFNCTOSCSEL CA6: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. The occurrence of ANY of the following hazardous events: (1) a. The occurrence of ANY of the following hazardous events:

Seismic event (earthquake) [ Seismic event (earthquake)

Internal or external flooding event 1Internal or external flooding event

High winds or tornado strike [Hg id rtraosrk

FIRE

EXPLOSION [FIRE

(site-specific hazards) [ EXPLOSION

Other events with similar hazard characteristics as [ Other events with similar hazard characteristics as determined by]

determined by the Shift Manager [the Shift Manager AND AND

b. EITHER of the following: b. EITHER of the following:

1. Event damage has caused indications of degraded 1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM performance in at least one train of a SAFETY SYSTEM needed for the current operating mode. needed for the current operating mode.

OR OR

2. The event has caused VISIBLE DAMAGE to a SAFETY 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current SYSTEM component or structure needed for the current operating mode. operating mode.

Difference/fJustification 23

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS None 24

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS CUt: INITIATING CONDITIONS NEI 99-01 Rev 6 Sea brook Station Nuclear Power Plant UNPLANNED loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory UNPLANNED loss of reactor vessellRCS inventory for 15 minutes or longer.

for 15 minutes or longer.

Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) UNPLANNED loss of reactor coolant results in (reactor vessel/RCS (1) UNPLANNED loss of reactor coolant results in reactor vessel/RCS level

[PWR] or RPV [BWR]) level less than a required lower limit for 15 less than a required lower limit of an operating band, specified by an minutes or longer, operating procedure for 15 minutes or longer.

(2) a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored. (2) a. Reactor vesselIRCS level cannot be monitored.

AND AND

b. UNPLANNED increase in (site-specific sump and/or tank) b. UNPLANNED increase in Containment Sump A or B level.

levels.

Difference /Justification CU1.2b: Site specific information, see V13 Containment Sumps 25

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS CU2: INITIATING CONDITIONS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant Loss of all but one AC power source to emergency buses for 15 minutes or Loss of all but one AC power source to emergency buses for 15 minutes or longer. longer.

Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (I) a. AC power capability to (site-specific emergency buses) is Note: For power restoration fr'om the SEPS, both SEPS diesel generator reduced to a single power source for 15 minutes or longer, sets must be functional.

AND (1) a. AC power capability to BOTH AC emergency buses E5 AND E6 is

b. Any additional single power source failure will result in reduced to a single power source for 15 minutes or longer.

loss of all AC power to SAFETY SYSTEMS. AND

b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.

Difference /Justification Added NOTE to clarify that both SEPS constitute a single power source.

26

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS CU3: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) UNPLANNED increase in RCS temperature to greater than (site- (1) UNPLANNED increase in RCS temperature to greater than 2000 F.

specific Technical Specification cold shutdown temperature limit). (2) Loss of ALL RCS temperature and reactor vessel/RCS level indication (2) Loss of ALL RCS temperature and (reactor vessel/RCS [PWR] or for 15 minutes or longer.

RPV [BWRI) level indication for 15 minutes or longer.

Difference /Justification CU3.1: Site specific information, see V1 5 Cold SD Temp Limit TS 27

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS CU4: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NET 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Indicated voltage is less than (site-specific bus voltage value) on (1) Indicated voltage is less than 105V on required Vital DC buses required Vital DC buses for 15 minutes or longer, associated with the Protected Train for 15 minutes or longer.

I Train A- 1A and 11C ITrain B-llIB and llD Difference /Justification CU4.1: Site specific information, see V18 UFSAR 8.3.2 - DCV 105 limit 28

COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS CUS5: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NET 99-01 Rev 6 Sea brook Station Nuclear Power Plant (1) Loss of ALL of the following onsite communication methods: (1) Loss of ALL of the following onsite communication methods:

(site-specific list of communications methods) jIn-Plant (PBX) Telephones I (2) Loss of ALL of the following ORO communications methods: Gai-Tronics]

(site-specific list of communications methods) iPatRdoSse (3) iLosspeofi Als oftefloigNCcommunications methods : (2) Loss of ALL of the following ORO communications methods:

(sie-seciic ommnictios istof mthos)Nuclear Alert System (NAS) I IBackup NAS IAll plant telephones 1 Cellular telephones]

(3) Loss of ALL of the following NRC communications methods:

I Emergency Notification System (ENS) 1 SAll plant telephones 1 I FTS telephones in the TSC I ICellular telephones Difference/fJustification Provided site specific communications methods 29

INDEPENDENT SPENT FUEL STORAGE FACILITY (ISFSI) ICS/EALS Difference /Justification None NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Damage to a loaded cask CONFINEMENT BOUNDARY as (I) Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than (2 times the indicated by ANY of the following on-contact surface radiation site-specific cask specific technical specification allowable radiation readings greater than:

level) on the surface of the spent fuel cask. 1600 mrem/hr at the front bird screen 4 mrem/hr at the door centerline 4 mrem/hr at the end shield wall exterior Difference /Justification Added NOTE pulled from the basis allowing calculating surface dose from a distance dose EUI.I: Site specific information, see VI9 NUHOMS HSM Dose Rates Technical Specification 30

FISSION PRODUCT BARRIER ICS/EALS PWR FISSION PRODUCT BARRIERS MATRIX - INITIATING CONDITIONS/THRESHOLDS NET 99-01 Rev 6 FA1 - Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier.

[ FS1 - Loss or Potential Loss of any two barriers. FG1 - Loss of any two barriers and Loss or Potential Loss of the third barrier.

. .... .... ~Seabrook Station Nuclear Power Plant . . ..

FG1 - Loss of any two barriers and Loss or Potential Loss of the third barrier.

FS1 - Loss or Potential Loss of any two barriers. TFAI jClad - Any Loss or any Potential Loss of either the Fuel or RCS barrier.

Difference /Justification None Fuel Clad Barrier RCS Barrier Cnanet are Loss Potential LossLosPtnilLsLssoetalos NEI 99-01 Rev 6

1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage Not Applicable A. RCS/reactor A. An automatic or A. Operation of a standby A. A leaking or Not Applicable vessel level less manual ECCS (SI) charging (makeup) RUPTURED SG is than (site-specific actuation is required pump is required by FAULTED outside of level), by EITHER of the EITHER of the containment.

following: following:

UNISOLABLE 1. UNISOLABLE RCS leakage RCS leakage OR OR

SG tube 2. SG tube leakage.

RUPTURE.

OR 31

FISSION PRODUCT BARRIER ICS/EALS B. RCS cooldown rate greater than (site-specific pressurized thermal shock criteria/limits defined by site-specific indications).

~Seabrook Station Nuclear Power Plant Not Applicable A. Core Cooling A. An automatic or A. Operation of a second A. A leaking or Not Applicable (C) CSF - ORANGE manual SI actuation is charging pump in the RUPTURED SG is entry conditions met required by EITHER of normal charging mode FAULTED outside of the following: is required by EITHER containment.

UNISOLABLE of the following:

RCS leakage 1. UNISOLABLE OR RCS leakage

SG tube RUPTURE. OR

2. SG tube leakage.

OR B. RCS Integrity (P) CSF

- RED entry conditions met with RCS press >

300 psig.

Difference /Justification Fuel Clad Barrier Potential Loss 1.A: Site specific information, see V20 CSFST Core Cooling RCS Barrier Potential Loss I.B: Site specific information, see V21I CSFST Integrity 32

FISSION PRODUCT BARRIER ICS/EALS NEI 99-01 Rev 6

2. Inadequate Heat Removal 2. Inadequate Heat Removal 2. Inadequate Heat Removal A. Core exit A. Core exit Not Applicable A. Inadequate RCS heat Not Applicable A. 1. (Site-specific thermocouple thermocouple removal capability via criteria for readings greater readings greater steam generators as entry into core than (site-specific than (site-specific indicated by (site- cooling temperature temperature specific indications), restoration value), value). procedure)

OR AND B. Inadequate RCS 2 etrto heat removal p.Roeuesnortio capability via effectivre withi steam generators 15etvewti as indicated by 15minutes.

(site-specific indications).

~Seabrook Station Nuclear Power Plant A. Core Cooling (C) A. Core Cooling (C) Not Applicable A. Heat Sink (H) CSF - Not Applicable A. Core Cooling (C)

CSF - RED entry CSF - ORANGE RED entry conditions CSF - RED entry conditions met. entry conditions met, conditions met for met. 15 minutes or OR longer.

B. Heat Sink (H)

CSF - RED entry conditions met.

Difference /Justification Fuel Clad Barrier: Loss 2.A, Potential Loss 2.A and Containment Barrier Potential Loss 2.A: Site specific information, see V20 CSFST Core Cooling RCS Barrier: Potential Loss 2.A: Site specific information, see V22 CSFST Heat Sink NEI 99-01 Rev 6 33

FISSION PRODUCT BARRIER ICS/EALS

3. RCS Activity / Containment Radiation 3. RCS Activity / Containment Radiation 3. RCS Activity / Containment Radiation A. Containment Not Applicable A. Containment radiation Not Applicable Not Applicable A. Containment radiation monitor monitor reading radiation monitor reading greater greater than (site- reading greater than than (site-specific specific value). (site-specific value).

value).

OR B. (Site-specific indications that reactor coolant activity is greater than 300 ptCi/gm dose equivalent 1-131).

~Seabrook Station Nuclear Power Plant A. Post LOCA Not Applicable A. Post LOCA Radiation Not Applicable Not Applicable A. Post LOCA Radiation Monitors Radiation Monitors Monitors RM 6576A-1 or RM RM 6576A-1 or RM RM 6576A-1 or 6576B-1 6576B-1 RM 6576B-1 >_16 R/hr. >_1,305 R/hr.

> 95 R/hr.

OR B. RCS activity >

300 uCi/grn Dose Equivalent I1131 as determined per Procedure CS0925.01, Reactor Coolant Post Accident Sampling.

Difference/IJustification 34

FISSION PRODUCT BARRIER ICS/EALS All Barriers: Loss & Potential Loss 3.A: Site specific information, see V23 EPCALC-06-01I -Rad Values for Fission Product Barrier Matrix NEI 99-01 Rev 6

4. Containment Integrity or Bypass 4. Containment Integrity or Bypass 4. Containment Integrity or Bypass Not Applicable Not Applicable Not Applicable Not Applicable A. Containment isolation is A. Containment required pressure greater than AND (site-specific value)

EITHER of the OR following: B. Explosive mixture

1. Containment exists inside integrity has been containment lost based on OR Emergency C. 1. Containment Director judgment. pressure greater OR than (site-

2. UNISOLABLE specific pressure pathway from the setpoint) containment to the AND environment exists. 2. Less than one OR full train of B. Indications of RCS (site-specific leakage outside of system or containment, equipment) is operating per design for 15 minutes or longer.

~Seabrook Station Nuclear Power Plant 35

FISSION PRODUCT BARRIER ICS/EALS Not Applicable Not Applicable Not Applicable Not Applicable A. Containment isolation is A. Containment (Z) required CSF - RED entry AND conditions met.

EITHER of the OR following: B. Cnmt. hydrogen

3. Containment concentration >_6%

integrity has been OR lost based on C. 1. Containment STED/SED pressure > 18 judgment.

psig OR AND

4. UNISOLABLE

2. Less than one pathway from the full train of containment to the Cnmt. Building environment exists.

Spray (CBS) is OR operating per B. Indications of RCS design for 15 leakage outside of minutes or containment. longer.

Difference /Justification Containment Barrier: Potential Loss 4.A: Site specific information, see V24 CSFST Containment Containment Barrier: Potential Loss 4.B: Site specific information, see V14 H2 concentration in containment Containment Barrier: Potential Loss 4.C1: Site specific information, see V25 Containment Spray Setpoint NEI 99-01 Rev 6

5. Other Indications 5. Other Indications 5. Other Indications A. (site-specific as A. (site-specific as A. (site-specific as A. (site-specific as A. (site-specific as A. (site-specific as applicable) applicable) applicable) applicable) applicable) applicable)

Seabrook Station Nuclear Power Plant Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable 36

FISSION PRODUCT BARRIER ICS/EALS Difference /Justification None NEI 99-01 Rev 6

6. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment A. ANY condition in A. ANY condition in A. ANY condition in the A. ANY condition in the A. ANY condition in the A. ANY condition in the opinion of the the opinion of the opinion of the opinion of the opinion of the the opinion of the Emergency Emergency Emergency Director Emergency Director Emergency Director Emergency Director Director that Director that that indicates Loss of that indicates Potential that indicates Loss of that indicates indicates Loss of indicates Potential the RCS Barrier. Loss of the RCS the Containment Potential Loss of the the Fuel Clad Loss of the Fuel Barrier. Barrier. Containment Barrier.

Barrier. Clad Barrier.

Sebrook Station Nuclear Power Plant A. ANY condition in A. ANY condition in A. ANY condition in the A. ANY condition in the A. ANY condition in the A. ANY condition in the opinion of the the opinion of the opinion of the opinion of the opinion of the the opinion of the STED/SED that STED/SED that STED/SED that STED/SED that STED/SED that STED/SED that indicates Loss of indicates Potential indicates Loss of the indicates Potential Loss indicates Loss of the indicates Potential the Fuel Clad Loss of the Fuel RCS Barrier, of the RCS Barrier. Containment Barrier. Loss of the Barrier. Clad Barrier. Containment Barrier.

Difference /Justification None 37

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS Ii:TiTANGCONIIONS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (I) a. A HOSTILE ACTION is occurring or has occurred within (1) a. A HOSTILE ACTION is occurring or has occurred within the the PROTECTED AREA as reported by the (site-specific PROTECTED AREA as reported by the security shift supervision.

security shift supervision). AND AND b. EITHER of the following has occurred:

b. EITHER of the following has occurred: I. ANY of the following safety functions cannot be controlled or

1. ANY of the following safety functions cannot be maintained.

controlled or maintained. ]Reactivity control I

Reactivity control [Core cooling I

Core cooling [PWR] / RPV water level [BWR] RSha eoa

RCS heat removal ]RSha eoa OR OR 2. Damage to spent fuel due to damaged SFP cooling system or loss of SEP integrity has occurred or is IMMINENT.

2. Damage to spent fuel has occurred or is IMMINENT.

Difference /Justification Added SEP cooling and integrity from the basis to clarify threshold.

38

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HG7: INITIATING CONDITIONS Difference/fJustification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Other conditions exist which in the judgment of the Emergency (1) Other conditions exist which in the judgment of the STED/SED Director indicate that events are in progress or have occurred which indicate that events are in progress or have occurred which involve involve actual or IMMINENT substantial core degradation or actual or IMMINENT substantial core degradation or melting with melting with potential for loss of containment integrity or HOSTILE potential for loss of containment integrity or HOSTILE ACTION ACTION that results in an actual loss of physical control of the that results in an actual loss of physical control of the facility.

facility. Releases can be reasonably expected to exceed EPA Releases can be reasonably expected to exceed EPA Protective Protective Action Guideline exposure levels offsite for more than the Action Guideline exposure levels offsite for more than the immediate site area. immediate site area.

Difference /Justification None 39

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HSI: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) A HOSTILE ACTION is occurring or has occurred within the Note: This Initiating Condition and EAL do not apply to an attack solely on PROTECTED AREA as reported by the (site-specific security shift the Dry Fuel Storage Protected Area. An attack on the Dry Fuel supervision). Storage Facility Protected Area should be considered an attack within the Owner Controlled Area and classified as an Alert per Initiating Condition HA 1 (1) A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the security shift supervision.

Difference /Justification Added NOTE: from the basis to clarify~ the threshold.

40

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HS6: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NET 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. An event has resulted in plant control being transferred (1) a. An event has resulted in plant control being transferred from the from the Control Room to (site-specific remote shutdown Control Room to the Remote Safe Shutdown components.

panels and local control stations). AND AND b. Control of ANY of the following key safety functions is not

b. Control of ANY of the following key safety functions is not reestablished within 15 minutes.

reestablished within (site-specific number of minutes). I Reactivity control

Reactivity control j Core cooling

Core cooling [PWR] I RPV water level [BWR] RSha eoa

RCSCheattremoval Difference /Justification HS6.1lb: Default time per developer notes 41

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HS7: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Other conditions exist which in the judgment of the Emergency (1) Other conditions exist which in the judgment of the STED/SED Director indicate that events are in progress or have occurred which indicate that events are in progress or have occurred which involve involve actual or likely major failures of plant functions needed for actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public, effective access to equipment needed for the protection of the Any releases are not expected to result in exposure levels which public. Any releases are not expected to result in exposure levels exceed EPA Protective Action Guideline exposure levels beyond the which exceed EPA Protective Action Guideline exposure levels site boundary. beyond the site boundary.

Difference /Justification None 42

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HAl: INITIATING CONDITIONS Seabrook Station Nuclear Power Plant NEI 99-01 Rev 6 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. attack threat within 30 minutes.

Difference /Justification None THRE*SHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) A HOSTILE ACTION is occurring or has occurred within the (1) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the (site-specific OWNER CONTROLLED AREA or the Dry Fuel Storage Facility security shift supervision), as reported by the security shift supervision.

(2) A validated notification from NRC of an aircraft attack threat within (2) A validated notification from NRC of an aircraft attack threat within 30 minutes of the site. 30 minutes of the site.

Difference /Justification HAl.l: Added the Dry Fuel Storage Facility (I SF SI) for clarification.

43

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HA5: IJNITIATING CONDITIONS Difference/fJustification None THRESHOLDS NEI 99-01 Rev 6 [Seabrook Station Nuclear Power Plant 44

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS (1) a. Release of a toxic, corrosive, asphyxiant or flammable gas (1) a. Release of a toxic, corrosive, asphyxiant or flammable gas into any of the following plant rooms or areas: into any Table Hi rooms or areas.

(site-specific list of plant rooms or areas with entry-related AND mode applicability identified) b. Entry into the room or area is prohibited or impeded.

AND " ~Table Hi"

b. Entry into the room or area is prohibited or impeded. Area Mode Primary Aux Building 25 ft elevation1,234 7 ft elevation 1 ,3

- 26 ft elevation Turbine Building 1, 2, 3 Switchgear Rooms Essential 1, 2,3, 4 Non-essential Steam and Feedwater Pipe chases 1, 2, 3 Waste Process Building 25 ft elevation1,3

-3 ft elevation1,23

-31 ft elevation Containment 3, 4 Equipment Vaults 3, 4 Differencie spectific ifration , e 7 al 1PoeueRfrne HA5.lb: Site specific information, see V7 - Table Hl Procedure References 45

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HA6: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) An event has resulted in plant control being transferred from the (1) Entry into Procedure 0S1200.02 for control room evacuation Control Room to (site-specific remote shutdown panels and local resulted in plant control being transferred from the Control Room to control stations). Remote Safe Shutdown Components.

Difference /Justificatioa HA6.1 Site specific information, see OS 1200.02, SAFE SHUTDOWN AND COOLDOWN FROM THE REMOTE SAFE SHUTDOWN FACILITIES 46

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 1147: INITIATIN'~G CONDITIONS NET 99-01 Rev 6 Seabrook Station Nuclear Power Plant Other conditions exist which in the judgment of the Emergency Director Other conditions exist which in the judgment of the STED/SED warrant warrant declaration of an Alert, declaration of an Alert.

Difference /Jnstification None

~THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Other conditions exist which, in the judgment of the Emergency (1) Other conditions exist which, in the judgment of the STED/SED, Director, indicate that events are in progress or have occurred which indicate that events are in progress or have occurred which involve involve an actual or potential substantial degradation of the level of an actual or potential substantial degradation of the level of safety of safety of the plant or a security event that involves probable life the plant or a security event that involves probable life threatening threatening risk to site personnel or damage to site equipment risk to site personnel or damage to site equipment because of because of HOSTILE ACTION. Any releases are expected to be HOSTILE ACTION. Any releases are expected to be limited to limited to small fractions of the EPA Protective Action Guideline small fractions of the EPA Protective Action Guideline exposure exposure levels, levels.

Difference/fJustification None 47

HAZARDANDOHER: CONDTITIONS AFFECTINGPLNTSAEYC/AL HUl: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) A SECURITY CONDITION that does not involve a HOSTILE (I) A Code Yellow is reported by the Security Shift Supervision.

ACTION as reported by the (site-specific security shift supervision)

(2) Notification of a credible security threat directed at the site. (2) Notification of a credible security threat directed at Seabrook (3) vaidaed he RC poviinginfomaton ~Station.

fom ntifcaton f aicattra.(3) A validated notification from the NRC providing information of an aircraft threat Difference /Justification HUI.I: Code Yellow is a Security Condition 48

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HiU2: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Seismic event greater than Operating Basis Earthquake (OBE) as (1) Seismic event greater than Operating Basis Earthquake (OBE) as indicated by: indicated by:

(site-specific indication that a seismic event met or exceeded OBE a. The red "EVENT" light is lit on seismic monitoring control limits) panel 1-SM-CP-58.

AND

b. The yellow "OBE" light is lit on seismic monitoring control panel 1-SM-CP-58.

OR (2) a. Seismic monitoring system out of service AND

b. Control Room personnel feel an actual or potential seismic event AND

c. The occurrence of a seismic event is confirmed in a manner deemed appropriate by the Shift Manager Difference /Justification HU2.1a and b: Site specific information, see V27 EC282 184 - Seismic Monitoring System Upgrade 49

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HU3: INITIATING CONDITIONS Difference /Justification None THRE*SHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) A tornado strike within the PROTECTED AREA. (1) A tornado strike within the PROTECTED AREA.

(2) Internal room or area flooding of a magnitude sufficient to require (2) Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode. component needed for the current operating mode.

(3) Movement of personnel within the PROTECTED AREA is impeded (3) Movement of personnel within the PROTECTED AREA is due to an offsite event involving hazardous materials (e.g., an offsite IMPEDED due to an offsite event involving hazardous materials.

chemical spill or toxic gas release). (4) A hazardous event that results in on-site conditions sufficient to (4) A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.

prohibit the plant staff from accessing the site via personal vehicles.

(5) (Site-specific list of natural or technological hazard events)

Difference /Justification None 50

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HU4: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 [Seabrook Station Nuclear Power Plant 51

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS (1) a. A FIRE is NOT extinguished within 15-minutes of ANY of NOTE: A containment fire alarm is considered valid upon receipt of an the following FIRE detection indications: actuated alarm on CP 376. combined with any of the following:

Report from the field (i.e., visual observation) [CP 376 panel - MUTIPLE ZONES ACTUATED

Receipt of multiple (more than 1) fire alanms or [Plant equipment - SPURIOUSLY OPERATING indications SContainment temperature - INCREASING

Field verification of a single fire alarm SContainment particulate radiation - INCREASING AND (1) a. A FIRE is NOT extinguished within 15-minutes of ANY of the

b. The FIRE is located within ANY of the following plant following FIRE detection indications:

rooms or areas: [ Report from the field (i.e., visual observation) I (site-specific list of plant rooms or areas) I Receipt of multiple (more than 1) fire alarms or indications (2) a. Receipt of a single fire alarm (i.e., no other indications of a [Field verification of a single fire alarm '

FIRE).

AND AND

b. The FIRE is located within ANY Table H2 plant rooms or areas:

b. The FIRE is located within ANY of the following plant Table H2 rooms or areas: Condensate Storage Tank Enclosure Fuel Storage Building (site-specific list of plant rooms or areas) Containment Primary Auxiliary Building AND Control Building Service Water Pump House

c. The existence of a FIRE is not verified within 30-minutes Cooling Tower Steam and Feedwater Pipe of alarm receipt. Chases (3) A FIRE within the plant or ISFSI [for plants with an ISFSI outside Diesel Generator Building North Tank Farm the plant ProtectedArea] PROTECTED AREA not extinguished Emergency Feedwater Pump House Startup Feedwater Pump Area within 60-minutes of the initial report, alarm or indication. Equipment Vault______ _______

(4) A FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant ProtectedArea] PROTECTED AREA that requires (2) a. Receipt of a single fire alarm (i.e., no other indications of a firefighting support by an offsite fire response agency to extinguish. FIRE).

AND

b. The FIRE is located within ANY of the Table H2 plant rooms or areas except Containment. (see note above)

AND

c. The existence of a FIRE is not verified within 30-minutes of alarm receipt.

(3) A FIRE within the plant PROTECTED AREA or Dry Fuel Storage Facility not extinguished within 60-minutes of the initial report, alarm or indication.

(4) A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.

52

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS Difference /Justification Added NOTE to clarifyi the containment fire alarm HU4.1b: Site specific information, see V28 Verification of Fire Areas HU4.2b: Containment is excepted but is covered by the second note. EAL 1 would be applicable. Entry into the Containment to perform verifications within 30 minutes is a challenge.

HU4.4: Dry Fuel Storage Facility was deleted. Seabrook Station relies on offsite fire response for any fire within the Dry Fuel Storage facility regardless of severity or potential to affect plant safety. The 60 minute duration of EAL #3 would indicate potential of a fire within the Dry Fuel Storage facility to degrade level of safety of the plant.

53

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HU7: INITIATING CONDITIONS Difference/fJustification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Other conditions exist which in the judgment of the Emergency (1) Other conditions exist which in the judgment of the STED/SED Director indicate that events are in progress or have occurred which indicate that events are in progress or have occurred which indicate indicate a potential degradation of the level of safety of the plant or a potential degradation of the level of safety of the plant or indicate indicate a security threat to facility protection has been initiated. No a security threat to facility protection has been initiated. No releases releases of radioactive material requiring offsite response or of radioactive material requiring offsite response or monitoring are monitoring are expected unless further degradation of safety systems expected unless further degradation of safety systems occurs.

Occurs.

Difference /Justification None 54

SYSTEM MALFUNCTIONS MGI: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. Loss of ALL offsite and ALL onsite AC power to (site- (1) a. Loss of ALL offsite and ALL onsite AC power to BOTH AC specific emergency buses). emergency buses E5 AND E6.

AND AND

b. EITHER of the following: b. ANY of the following:

Restoration of at least one AC emergency bus in less Restoration of at least one AC emergency bus in less than 4 than (site-specific hours) is not likely,] hours is not likely.

(Site-specific indication of an inability to adequately Core Cooling (C) CSF - RED entry conditions met.I remove heat from the core)

Difference /Justification MGI.lb: Site specific information, see V26 - SBO Coping and V20 - CSFST Core Cooling 55

SYSTEM MALFUNCTIONS MG8: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. Loss of ALL offsite and ALL onsite AC power to (site- Note:

For a bus to be considered energized from SEPS, both SEPS diesel specific emergency buses) for 15 minutes or longer, generator sets must be functional.

AND (1) a. Loss of ALL offsite and ALL onsite AC power to BOTH AC

b. Indicated voltage is less than (site-specific bus voltage value) emergency buses E5 AND E6 for 15 minutes or longer.

on ALL (site-specific Vital DC buses) for 15 minutes or AND longer. b. Indicated voltage is less than 105 V on ALL Vital DC buses 11A, 11B, 11C and l1D for 15 minutes or longer.

Difference/fJustification Added NOTE for consideration of an additional non-safety power supply Supplemental Emergency Power System (SEP S)

MG8.1b: Site specific information, see V18 UFSAR 8.3.2 - DCV 105 limit 56

SYSTEM MALFUNCTIONS MSI: INITIATING CONDITIONS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant Loss of all offsite and all onsite AC power to emergency buses for 15 minutes Loss of all offsite and all onsite AC power to emergency buses for 15 or longer. minutes or longer.

Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (I) Loss of ALL offsite and ALL onsite AC power to (site-specific Note: For a bus to be considered energized from SEPS, both SEPS diesel emergency buses) for 15 minutes or longer, generator sets must be functional.

(1) Loss of ALL offsite and ALL onsite AC power to BOTH AC emergency buses E5 AND E6 for 15 minutes or longer.

Difference /Justification Added NOTE for consideration of an additional non-safety power supply Supplemental Emergency Power System (SEPS) 57

SYSINTEMIALFUNGCODTIONS MS5: INITIATING CONDITIONS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant Inability to shut down the reactor causing a challenge to (core cooling [PWRI Inability to shutdown the reactor to neutron flux < 5% causing a challenge to

/ RPV water level [BWR]) or RCS heat removal, core cooling or RCS heat removal.

Difference /Justification Added site specific value for determining the reactor is shutdown.

THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. An automatic or manual (trip [PWR] / scram [BWR]) did not (1) a. An automatic or manual trip did not shutdown the reactor to shutdown the reactor. neutron flux < 5%.

AND AND b. All manual actions to shutdown the reactor have been unsuccessful.

b. All manual actions to shut down the reactor have been AND

c. EITHER of the following conditions exist:

unsuccessful.

I Core Cooling (C) CSF - RED entry conditions met.

AND [Heat Sink (H) - RED entry conditions met.]

c. EITHER of the following conditions exist:

(Site-specific indication of an inability to adequately remove heat from the core)

(Site-specific indication of an inability to adequately remove heat from the RCS)

Difference /Justification MS51.a: Site specific information, see V29 CSFST Subcriticality MS5.1c: Site specific information, see V20 CSFST Core Cooling and V22 CSFST Heat Sink 58

SYSTEM MALFUNCTIONS MS8: INITIATING CONDITIONS Difference /Justification None THRESHOLDS Difference /Justification MS8.1: Site specific information, see V18 UFSAR 8.3.2 - DCV 105 limit 59

SYSTEM MALFUNCTIONS MAI: INITIATING CONDITIONS NEI 99-0 1 Rev 6 Seabrook Station Nuclear Power Plant Loss of all but one AC power source to emergency buses for 15 minutes or Loss of all but one AC power source to emergency buses for 15 minutes or longer. longer.

Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. AC power capability to (site-specific emergency buses) is (1) AC power capability to BOTH AC emergency buses E5 AND E6 is reduced to a single power source for 15 minutes or longer, reduced to a single power source for 15 minutes or longer.

AND

b. Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.

Difference /Justification MAI.lb - deleted as not necessary to make classification. The intent of the EAL remains the same and the timing of classification is not affected.

60

SYSTEM MALFUNCTIONS MA2: INITIATING CONDITIONS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant UNPLANNED loss of Control Room indications for 15 minutes or longer UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress. with a significant transient in progress.

Difference /Justification None THRESHOLDS NEI 9-01 ev 6Seabrook Station Nuclear Power Plant 61

SYSTEM MALFUNCTIONS (1) a. An UNPLANNED event results in the inability to monitor (1) a. An UNPLANNED event results in the inability to monitor one one or more of the following parameters from within the or more of the following parameters from within the Control Control Room for 15 minutes or longer. Room for 15 minutes or longer.

Reactor Power

[BWR parameterlist] [PWR parameterlist] Pressurizer Level Reactor Power Reactor Power RCS Pressure RPV Water Level RCS Level Core Exit Temperature RPV Pressure RCS Pressure Levels in at least two steam generators Primary Containment In-Core/Core Exit Steam Generator Emergency Feed Water Flow Pressure Temperature AND Suppression Pool Level Levels in at least (site- b. *bNY of the following transient events in progress.

specific number) steam SAutomatic or manual runback greater than 25%

generators thennal reactor power Suppression Pool Steam Generator Auxiliary ] lectrical load rejection greater than 25% fUllelcrclla Temperature or Emergency Feed Water Flow [ Reactor trp 1 AND [ ECCS (SI) actuation

b. ANY of the following transient events in progress.

Automatic or manual runback greater than 25%

thermal reactor power

Electrical load rejection greater than 25% full electrical load

Reactor scram [BWR] / trip [PWR]

ECCS (SI) actuation

Thermal power oscillations greater than 10% [BWR]

Diferne /utfcto None 62

SYSTEM MALFUNCTIONS MA5: INITIATING CONDITIONS NEI 99-01 Rev 6 Sea brook Station Nuclear Power Plant Automatic or manual (trip [PWR] / scram [BWR]) fails to shut down the Automatic or manual trip fails to shutdown the reactor to neutron reactor, and subsequent manual actions taken at the reactor control consoles flux < 5%, and subsequent manual actions taken at the Main Control Board are not successful in shutting down the reactor. are not successful in shutting down the reactor.

Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (I) a. An automatic or manual (trip [PWR] / scram [BWR]) did (I) a. An automatic or manual trip did not shutdown the reactor to not shutdown the reactor. neutron flux <5%.

AND AND b. Manual actions taken at the MCB are not successful in shutting down the reactor.

b. Manual actions taken at the reactor control consoles are not successful in shutting down the reactor.

Difference /Justification MA5.1: Site specific information, see V29 CSFST Subcriticality 63

SYSTEM MALFUNCTIONS MA9: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. The occurrence of ANY of the following hazardous events: (1) a. The occurrence of ANY of the following hazardous events:

Seismic event (earthquake) Seismic event (earthquake)

Internal or external flooding event jInternal or external flooding event

High winds or tornado strike FIRE High winds or tornado strike

EXPLOSION IFIRE 1

(site-specific hazards) EXPLOSION 1

Other events with similar hazard characteristics as Other events with similar hazard characteristics as determined determined by the Shift Manager by the Shift ManagerI AND AND

b. EITHER of the following: b. EITHER of the following:

1. Event damage has caused indications of degraded 1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM performance in at least one train of a SAFETY SYSTEM needed for the current operating mode. needed for the current operating mode.

OR OR

2. The event has caused VISIBLE DAMAGE to a SAFETY 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current SYSTEM component or structure needed for the current operating mode. operating mode.

Difference /Justification 64

SYSTEM MALFUNCTIONS LNone]

MUl: INITIATING CONDITIONS NEI 99-01 Rev 6 ]Seabrook Station Nuclear Power Plant Loss of all offsite AC power capability to emergency buses for 15 minutes or Loss of all offsite AC power capability to emergency buses for 15 minutes or longer., longer.

Difference /Justification None THRESHOLDS Difference /Justification None 65

MUYSINTEMIALFUNGCODTIONS MU2: INITIATING CONDITIONS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant

.4-(1) a. An UNPLANNED event results in the inability to monitor (1) a. An UNPLANNED event results in the inability to monitor one or one or more of the following parameters from within the more of the following parameters from within the Control Room for 15 Control Room for 15 minutes or longer. minutes or longer.

Reactor Power

[B WR parameter list] [P WR parameter list] Pressurizer Level Reactor Power Reactor Power RCS Pressure Core Exit Temperature RPV Water Level RCS Level Level in at least two steam generators RPV Pressure RCS Pressure Steam Generator Emergency Feed Water Flow Primary Containment In-Core/Core Exit Pressure Temperature Suppression Pool Level Levels in at least (site-specific number) steam generators Suppression Pool Steam Generator Temperature Auxiliary or Emergency Feed Water Flow Difference /Justification None 66

SYSTEM MALFUNCTIONS MU3: INITIATING CONDITIONS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant Reactor coolant activity greater than Teclnical Specification allowable limits. Reactor coolant activity greater than Technical Specification allowable limits.

Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) (Site-specific radiation monitor) reading greater than (site-specific (I) RM-6520-1reading greater than 2,670 mR/hr.

value).

(2) Sample analysis indicates that a reactor coolant activity value is (2) Sample analysis indicates that a reactor coolant activity value is greater than the Limiting Condition for Operation (LCO) specified greater than an allowable limit specified in Technical Specifications. in Technical Specification 3.4.8 Reactor Coolant System Specific Activity.

Difference /Justification MU3.I: Site specific information, see V30 EPCALC-06 Letdown Monitor Value MU3.2: Site specific information, see V31 TS 3.4.8 Specific Activity 67

SYSINTEMIALFUNGCODTIONS MiU4: INITIATING CONDITIONS Difference /Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) RCS unidentified or pressure boundary leakage greater than (site- (1) RCS unidentified or PRESSURE BOUNDARY LEAKAGE greater specific value) for 15 minutes or longer. than 10 gpm for 15 minutes or longer.

(2) RCS identified leakage greater than (site-specific value) for 15 minutes or longer. (2) RCS IDENTIFIED LEAKAGE greater than 25 gpm for 15 minutes (3) Leakage from the RCS to a location outside containment greater or longer.

than 25 gpm for 15 minutes or longer.

(3) Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.

Difference/(Justification Retained leakage values.

68

SYSTEM MALFUNCTIONS MU5: INITIATING CONDITIONS Difference /Justification THRESHOLDS NET 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. An automatic (trip [PWR] / scram [BWR]) did not shutdown (1) a. An automatic trip did not shutdown the reactor to neutron flux the reactor. < 5%.

AND AND

b. A subsequent manual action taken at the reactor control b. A subsequent manual action taken at the MCB is successful in consoles is successful in shutting down the reactor, shutting down the reactor.

(2) a. A manual trip ([PWR] / scram [BWR]) did not shutdown the OR reactor. (2) a. A manual trip did not shutdown the reactor to neutron flux <

AND 5%.

b. EITHER of the following: AND

1. A subsequent manual action taken at the reactor control b. EITHER of the following:

consoles is successful in shutting down the reactor. 1. A subsequent manual action taken at the MCB is successful OR in shutting down the reactor.

2. A subsequent automatic (trip [PWR] / scram [BWR]J) is OR successful in shutting down the reactor. 2. A subsequent automatic trip is successful in shutting down the reactor.

Difference /Justification MU5.1a and 2a: Site specific information, see V29 CSFST Subcriticality 69

SYSTEM MALFUNCTIONS MUJ6: INITIATING CONDITIONS Difference /Justification None THPESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Loss of ALL of the following onsite communication methods: (1) Loss of ALL of the following onsite communication methods:

(site-specific list of communications methods) I[In Plant (PBX) Telephones (2) Loss of ALL of the following ORO communications methods: [ Gai Tronics (site-specific list of communications methods)

(3) Loss of ALL of the following NRC communications methods: [__________________

Plant Radio System (site-specific list of communications methods) (2) Loss of ALL of the following ORO communications methods:

[ Nuclear Alert System (NAS) f Backup NAS

[All plant telephones

[Cellular telephones]

(3) Loss of ALL of the following NRC communications methods:

[ Emergency Notification System (ENS) I

[All plant telephones[

[FTS telephones in the TSC

]Cellular telephones Difference /Justification Provided site specific communications methods 70

MUJ7: INITIATING CONDITIONS Difference/fJustification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. Failure of containment to isolate when required by an (1) a. Failure of containment to isolate when required by an actuation actuation signal. signal.

AND AND

b. ALL required penetrations are not closed within 15 minutes of

b. ALL required penetrations are not closed within 15 the actuation signal.

minutes of the actuation signal. (2) a. Containment pressure greater than 18 psig.

AND (2) a. Containment pressure greater than (site-specific pressure). b. Less than one full train of Containment Building Spray (CBS) is AND operating per design for 15 minutes or longer.

b. Less than one full train of (site-specific system or equipment) is operating per design for 15 minutes or longer.

Difference /Justification MU7.2a: Site specific information, see V25 Containment Spray Setpoint 71

ATTACHMENT 4 NEI 99-01, Rev. 6, Deviations and Differences, Seabrook Station Nuclear Power Plant -

Unit 1

NEI 99-01 Rev 6 Deviations and Differences Seabrook Station Nuclear Power Plant - Unit I

N ET99-01 Rev 6 ________________earo__atin ___l ___o____ l__ t_____________________iii ii~ii i...

___________li!

References BWRs Deleted_____BWR ___references_______as__appropriate_____

Uses A for radiological effluent/radiation level ICs Uses R for radiological effluent/radiation level ICs Uses E-HU for ISFSI ICs Uses______EU____for_____ISFSI_______I_____

Uses S for System Malfunction ICs Uses____M __for___System_____Malfunction_______I__s Emergency Classification ICs are presented in ascending order (NOUB - GE) Emergency Classification ICs are presented in descending order (GE -

All NOTEs made site specific by identifying the STED/ED as the user.

Site specific information is highlighted in yellow.

No deviations were indicated to complete the upgrade to Revision 6.

1

ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS RGI: INITIATING CONDITIONS Difference/fJustification None

~THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant 2

ABNORMAL RAD LEVELS /RADIOLOGICAL EFFLUENT ICS/EALS (1) Reading on ANY of the following radiation monitors greater (l) Reading on ANY of the following radiation monitors greater than the than the reading shown for 15 minutes or longer: reading shown for 15 minutes or longer:

(site-specific monitor list and threshold values) Monitor Reading (2) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at RM-6528-4 (WRGM rate) 2.85E+8 uCi/sec or beyond (site-specific dose receptor point). Time After Shutdown Reading (3) Field survey results indicate EITHER of the following at or

<lhr >lhrto <2hrs beyond (site-specific dose receptor point):

Closed window dose rates greater than 1,000 mR/hr RM-6481-1" (MSL A) 1310 mR/hr ! 060mR/hr expected to continue for 60 minutes or longer. RM-6482- * (MSL B) 1310 mR/hr 1060OmR/hr

Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation. RM-6482-2" (MSL C) 1310 mR/hr 1060 mR/hr RM-I6481-2" (MSL D) 1310 mR/hr 1060OmR/hr

With release path to the environment from affected steam line, e.g.,