Text

Rev D to LGS Emergency Preparedness NUMARC Eals
	ML20195J165
Person / Time
Site:	Peach Bottom, Limerick
Issue date:	11/16/1998
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20195J148	List: ML20195J142; ML20195J154; ML20195J165;
References
	PROC-981116-01, PROC-981116-1, NUDOCS 9811240209
	Download: ML20195J165 (180)
	v • d • e

,

i i

I 1

i ATTACHMENT 2 Limerick Generating Station Units 1 and 2 I

Emergency Action Levels (Revised) 9811240209 991116 PDR ADOCK 05000277 F

PDR-i

.W T

PECO NUCLEAR LIMERICK GENERATING STATION Emergency Preparedness NUMARC EMERGENCY ACTION LEVELS NRC Submittal Copy i

Revision D i

i l

4 i

November 16,1998 i

l i

i

w m

Table of Contents e,n,my,a e

+w pn 3

"a O

r N f 9./f9 Contents

@geyn%pp,e n

C And g

2%

K.

Introduction

.u QF; :: A en U

M b a;) s h, y,- (

L:

m Technical Basis Manual

^ ' ' *

O NUMARC Comparison E

AdditionalInformation Response to NRC Request for O

b AVERY'*

RE ADY INDEX" (NDEXING SYSTEM

. _ = ~ - _. - -. - -,

-~.c~~

Contents / Introduction REV D. November 16,1998 v/

Page 1 of 1

\\

NUMARC EAL Submittal to NRC This package contains Revision D of the PECO Nuclear proposed EAL revision scheme using the NUMARC/NESP 007 methodology. Revision A was submitted in May 1995. Revision C was submitted in April 1998 and included use of the NUMARC suggested Fission Product Barrier Matrix, which was not included in the original submittal. This revision addresses issues raised in an NRC Request for Additional Information received in September 1998. This package includes:

Contents and Introduction - Yellow Tab 1 Technical Basis Manual-Orange Tab 2 Each EAL is listed as a separate number with the Initiating Condition, Applicable Operational Condition, Basis, Deviation, and References.

EAL Table - Red Tab 3 Ih EAL with associated initiating condition in table form.

)-

NUMARC Comparision - Magenta Tab 4 NUMARC EAL Number versus PECO EAL Number in matrix form.

Response to NRC Reauest for Additional Information - Violet Tab 5 RAI with PECO Nuclear response to each issue included in document.

I O

e

^

LGS EAL Technicd Basis Manual i>

REV o. November 16,1998 Page 1 of 126 l

f

LGS EAL Technical Basis Manual

(

Table of Contents l

Se ctio n 1 - I nt ro d u ctl o n........................................................................... 2 i

Section ll - Acronyms......

....4 Section ll1 - EAL Technical Basis;..............

..............6 1.0 Reactor Fuel l

1.1 Coolant Activity...

... 7 1.2 Irradiated Fuel or New Fuel..

................................................9 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level.........................................

. 19 2.2

~ Reactor Power..............

.. 23 3.0 Fission Product Barrier l

3.1 initiating Condition Matrix...

.... 29 3.2 Fuel Clad Barrier Thresholds.....................

...........................32 3.3 Reactor Coolant System Barrier Thresholds..................

..... 41 3.4 Primary Containment Barrier Thresholds......

.. 46 3.5 Fission Product Barrier Table.................

......................57 o

t i

4.0 Secondary Containment a

4.1 Main Steam Line.....

........59 5.0.

Radioactivity Release 5.1 Effluent Release and Dose..

... 61 5.2 in-Plant Radlation.......

... 71 6.0 Loss of Power 6.1 Loss of AC or DC Power..........................

...... 77 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation..

... 87 7.2 Loss of Decay Heat Removal Capability...

.90 7.3 Loss of Assessment / Communications Capability................

......... 94 8.0 External Events

[

8.1 Security Events.....

. 101 8.2 Fire / Explosion and Toxic / Flammable Gases..

........................ 1 06 8.3 Man-Made Events........................

. 113 8.4 Natural Events.................

..................................................116 l

9.0 Other 9.1 General.....

.... 123 f%

5 i

V I

lgs EAL Techrucal Basis Manual RElf o. November 16,1998 Page 2 of 126 1

Section 1 -Introduction 0'

This manual contains the technical basis for the Emergency Action Levels as utilized in ERP-101, Classification of Emergencies. The format and use of this manualls as follows.

1.

Heading and Sub Heading There are nine major headings each containing one or more sub-headings. These are j

i as follows:

1.0 Reactor Fuel 1.1 Coolant Activity 1.2 Irradiated Fuel or New Fuel 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level 2.2 Reactor Power 3.0 Fission Product Barrier 3.1 Initiating Condition Matrix 3.2 Fuel Clad Barrier Thresholds 3.3 Reactor Coolant System Barrier Thresholds 3.4 Primary Containment Barrier Thresholds 3.5 Fission Product Barrier Table 4.0 Secondary Containment 4.1 Main Steam Line 5.0 Radioactivity Release 5.1 Effluent Release and Dose 5.2 In-Plant Radiation 6.0 Loss of Power L

6.1 Loss of AC or DC Power 7.0 Intemal Events 7.1 Technical Specifications & Control Room Evacuation 7.2 Loss of Decay Heat Removal Capability 7.3 Loss of Assessment / Communications Capability 8.0 Extemal Events 8.1 Security Events I

8.2 Fire / Explosion and Toxic / Flammable Gases 8.3 Man-Made Events 8.4 Natural Events 9.0 Other 9.1 General l

lgs EAL Technic"A Basis Manual REV D, November 16,1998 Page 3 of 126

(

2.

Emergency Classification Level and Number Identification

(

The classifications range from Unusual Event through Alert, Site Area Emergency to General Emergency.

For each sub-heading, there may not be an EAL in every classification level. Each EAL is individually and uniquely numbered. No two numbers are the same.

3.

INITIATING CONDITION The Initiating Condition or IC (as described in NUMARC NESP-007) is contained in this section. ICs are a predetermined subset of conditions where either the potential exists for a radiological emergency or such an emergency has occurred. Additionally, ICs are the means by which EALs for different nuclear power plants are standardized.

4.

EAL Each Emergency Action Level exactly as it is contained in ERP-101.

5.

OPCON The operational condition (OPCON) that the EAL is applicable in is contained here.

There are six OPCONs (1,2,3,4 and 5 and defueled) that are used. LGS also uses mode switch position.

These positions are stated below and are Run, Startup, Shutdown and Refueling.

It should be noted that these OPCONs are entry level conditions. The EAL is applicable if the plant was in the OPCON at the start of the event. Subsequent positions of the mode selector switch should be ignored for purposes of classification.

f)

OPCON (MODE)

MODE SWITCH POSITION V

rumm Run um Startup

==

Shutdown (hot)

Shutdown (cold) emE Refueling

==

N/A (defueled) 6.

BASIS The technical basis of each EAL is contained in this section.

This includes any necessary calculations and also includes escalation references.

7.

DEVIATION Any deviations from the NUMARC NESP-007 methodology are contained in this section. If there are no deviations, NONE is used.

8.

REFERENCES All applicable references used in developing the technical basis for each EAL are contained in this section.

l I (S l

)

lgs EAL Technical Basis Manual REV D, November 16.1996 Page 4 of 126 Ol Section ll-Acronyms Attemating Current AC Automatic Depressurization System ADS Average Power Range Monitor APRM Attemate Rod Insertion ARI Area Radiation Monitor ARM Anticipated Transient Without Scram ATWS Bureau of Radiation Protection BRP Containment Atmosphere Control CAC Committed Dose Equivalent CDE Cubic Feet Per Minute CFM Code of Federal Regulations CFR Control Rod Drive CRD Core Spray CS Design Basis Accident DBA Direct Current DC Dose Equivalent lodine DEI Emergency Action Level EAL Emergency Core Cooling Systems ECCS Emergency Diesel Generator EDG Environmental Protection Agency EPA Emergency Response Procedure - Common ERP-C Emergency Service Water ESW Fuel Clad (Barrier)

FC Federal Telephone System FTS Gallons Per Minute GPM Heat Capacity Temperature Limit HCTL High Pressure Coolant injection HPCI Initiating Condition IC Intermediate Range Monitor IRM Kilovolt KV Limiting Condition for Operation LCO Limerick Generating Station LGS LOCA Loss of Coolant Accident Low Pressure Coolant Injection LPCI Miles Per Hour MPH Milli Roentgen Per Hour mR/hr MSIV Main Steam isolation Valve Normal Full Power Background NFPB NPSH Not Positive Suction Head Nuclear Regulatory Commission NRC Nuclear Management and Resources Council NUMARC ODCM Offsite Dose Calculation Manual Operating Condition OPCON Pennsylvania Emergency Management Agency PEMA Primary Containment (Barrier)

PC PCIS Primary Containment Isolation System Pounds Square Inch Gauge PSIG Reactor Enclosure Recire System RERS Reactor Coolant (Barrier)

RC Reactor Core Isolation Cooling RCIC Reactor Coolant System RCS

LGS EALTochrscal Basis Manual REV o. November 16,1998 -

l Page 5 of 126

- Residual Heat Removal -

[

P

V fiHR,

Reactor Protection System j

p RPS Reactor Pressure Vessel RPV-Redundant Reactivity Control System l

RRCS' SBO.

Station Blackout t

i Standby Gas Treatment System

.SGTS Steam Jet Air Ejector SJAE ?

SRM Source Range Monitor Safety Relief Valve j

SRV..

Top of Active Fuel i

TAF Total Protective Action Recommendation Dose TPARD

, TRIPS.

Transient Response implementation Plan Procedures l

Ci/cc Micro Curie Per Cubic Centimeter

. Micro Curie Per Gram -

L Ci/gm Updated Final Safety Analysis Report L:

UFSAR.

Volts Direct Current VDC r-i

.[

t- :

l l

1 l

l p

(-

l.

I l-f.~

z-

=

l LGS EAL Technical Br-is Mxnual REV 0, Novsmt:ct 16,1998 Page 6 of 126 1

Section 111 - EAL Technical Basis 9

O

lgs EAL Technical Basis Manual REV D. November 16,1998 Page 7 cf 126 q

1.0 Reactor Fuel j

V.

j 1,1 Coolant Activity

- UNUSUAL EVENT - 1.1.1.a IC Fuel Clad Degradation EAL l

Reactor Coolant activity > 4 pCl/gm Dose Equivalent lodine 131 OPCON

-m BASIS Coolant activity in excess of Technical Specifications (> 4 Ci/gm) is considered to be a precursor of more serious problems. The Technical Specification limit reflects a degrading or l

degraded core condition. This levelis chosen to be above any possible short duration spikes under normal conditions. An Unusual Event is only warranted when actual fuel clad damage is l('-

the cause of the elevated coolant sample (as determined by laboratory ' confirmation).

l -

However, fuel clad damage should be assumed to be the cause of elevated Reactor Coolant activity unless another cause is known, e.g.,

Reactor -Coolant. System chemical decontamination evolution (during shutdown) is ongoing with resulting high activity levels.

This event will be escalated to an Alert when Reactor Coolant activity exceeds 300 Ci/gm j -

Dose Equivalent lodine.131 per Fission Product Barrier Table.

I DEVIATION None l.

REFERENCES Technical Specification Section 3.4.5 l

NUMARC NESP-007, SU4.2 l

. t

?

4 4

lgs EAL Technical Basis Manual REV o. November 16,1998 Page 8 of 126 1.0 Reactor Fuel O

1.1 Coolant Activity UNUSUAL EVENT - 1.1.1.b IC Fuel Clad Degradation EAL SJAE Radiation (Offgas Monitor) > 2.1x10* mR/hr OPCON EEMBE BASIS The steam jet air ejector radiation monitor in the Control Room would be one of the first indicators of a degrading core. The high-high alarm is set at the Technical Specification limit of 2.1x10' mR/hr. This instrument takes a sample before the recombiner. This indicator of elevated activity is considered to be a precursor of more serious problems. The Technical Specification limit reflects a degrading or degraded core condition.

Escalation of this IC to the Alert levelis via the Fission Product Barrier Degradation Monitoring ICs.

DEVIATION The OPCON applicability [1,2,3) is a deviation from NUMARC [all) in that the SJAE Radiation Monitor and Main Steam Line Radiation Monitors will only be a valid indication of Fuel Clad Degradation in those OPCON's. At Limerick, there are no other monitors which can be an indicator of Fuel Clad Degradation. Degradation in cold shutdown or refueling will be first indicated by ventilation release monitofs which are covered by EAL on Effluent Release and Dose.

REFERENCES Technical Specifications Section 3.3.7.12,3.11.2.6 NUMARC NESP-007, SU4.1

lgs EAL Technical Basis Manual REV o, Nor.rnber 16,1998

{

Page 9 of 126 j

1.0 Reactor Fuel v

1.2 Irradiated Fuel or New Fuel UNUSUAL EVENT - 1.2.1.a IC Unexpected increase in Plant Radiation or Airborne Concentration.

EAL l

Uncontrolled water level decrease in the spent fuel pool with all irradiated fuel assemblies i

remaining covered by water OPCON BASIS This event tends to have a long lead time relative to potential for radiological release outside the site boundary, thus impact to public health and safety is very low.

i In light of Reactor Cavity Seal failure incidents at two different PWRs and loss of water in the O

Spent Fuel Pit / Fuel Transfer Canal at a BWR all occurring since 1984, explicit coverage of

'(

these types of events via this EAL is appropriate given their potential for increased doses to i

J plant staff. Classification as an Unusual Event is warranted as a precursor to a more serious event.

l This event will be escalated to an Alert as a result of uncovery of a fuel assembly and/or indication of high radiation levels on the refueling floor.

DEVIATION None REFERENCES NUMARC NESP-007, AU2.2 Technical Specifications i

i v.

4 i

lgs EAL Technic 1 Bast Manual REV D, November 16,1998 Page 10 of 126 1.0 Reactor Fuel 1.2 Irradiated Fuel or New Fuel UNUSUAL EVENT - 1.2.1.b IC Unexpected increase in Plant Radiation or Airborne Concentration.

EAL Unexpected Fuel Pool Storage low level alarm AND Visual observation of an uncontrolled water level decrease below the fuel pool skimmer surge tank inlet OPCON rwm l

BASIS A drop in the Spent Fuel Pool level or the RPV (when in refueling and flooded up with the gates removed) will result in a control room annunciator Fuel Pool Storage Lo Level Alarm.

This Control Room alarm directs an operator to be dispatched to a local alarm panel which will identify the reason for the alarm. This alarm is validated with visual observation of a decreasing Spent Fuel Pool level. If the spent fuel pool level decreases below the inlet to the skimmer surge tank, without a planned event such as removing a large piece of equipment, there must be a leak in the spent fuel pool or the RPV. This event has a long lead time relative to potential for radiological release outside the site boundary, thus the impact to public health and safety is very low. Classification as an Unusual Event is warranted as a precursor to a more serious event.

In light of Reactor Cavity Seal failure incidents at two different PWRs and loss of water in the Spent Fuel Pit / Fuel Transfer Canal at a BWR all occurring since 1984, explicit coverage of these types of events via this EAL is appropriate given their potential for increased doses to i

l plant staff. Classification as an Unusual Event is warranted as a precursor to a more serious 4

l event.

This event will be escalated to an Alert as a result of uncovery of a fuel assembly and/or indication of high radiation levels on the refueling floor.

DEVIATION l

1 None REFERENCES l

NUMARC NESP-007, AU2.1 l

l

lgs EAL Technical Basis Manual REV D. November 16,1998 Page 11 of 126 O

1.0 Reactor Fuel b

1.2 Irradiated Fuel or New Fuel ALERT - 1.2.2.a IC.

Major Damage to irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of irradiated Fuel Outside the Reactor Vessel EAL Unplanned general area radiation > 500 mR/hr on the refuel floor (Table 1-1)

OPCON i~'*

BASIS This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage, which is discussed in NUMARC/NESP-007 IC AU2,

" Unexpected increase in Plant Radiation or Airbome Concentration."

NUREG-0818, " Emergency Action Levels for Light Water Reactors," forms the basis for this gi)

EAL. The areas where irradiated fuel is located forms the basis for the radiation monitors listed in Table 1-1.

Unexpected radiation levels which are at least 100 times higher than the normal background will generally indicate a fuel handling accident or loss of water covering the irradiated fuel.

Readings may be from reftel floor Area Radiation Monitors or taken during a qualified radiological survey. Table 1-1 monitors are as follows:

Table 1-1 Refuel Floor ARMS RIS29-M1-1(2)K600, Drywell Head Laydown RIS30-M1-1(2)K600, Dryer /Seperator Area RIS31-M1-1(2)K600, Spent Fuel Pool RIS32-M1-1(2)K600, New Fuel Storage Vault RIS33-M1-1(2)K600, Pool Plug Laydown There is time available to take corrective actions, and there is little potential for substantial fuel damage. In addition, NUREG/CR-4982, " Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82," July 1987, indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and that risk of injury is low. In addition, NRC Information Notice No. 90-08, "Kr-85 Hazards from Decayed Fuel" presents the following in its discussion:

In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area

/G radius of one mile from the plant site) would be well below the Environmental Protection

~(/'

Agency's Protective Action Guides. Accordingly, it is important to be able to properly survey and monitor for Kr-85 in the event of an accident with decayed spent fuel.

lgs EAL Technical Bash Manual REv O, November 16.1998 Page 12 of 126 Licensees may wish to reevaluate whether Emergency Action Levels specified in the emergency plan and procedures governing decayed fuel-handling activities appropriately focus on concem for onsite workers and Kr-85 releases in areas where decayed spent fuel accidents could occur, for example, the spent fuel pool working Furthermore, licensees may wish to determine if emergency plans and floor.

corresponding exposures of onsite personnel who are in other areas of the plant.

Among other things, moving onsite personnel away from the plume and shutting off building air intakes downwind from the source may be appropriate.

Offsite doses during these accidents would be well below the EPA Protective Action Guidelines and the classification as an Alert is therefore appropriate. This radiation level could also be caused by an inadvertent criticality and is included even though the probability of this event occurring is low. Radiation increases above 500 mR/hr which were expected should not cause an Alert to be declared during a planned evolution. Additionally, surveys which identify

" hot spots" greater than 500 mR/hr should not cause an Alert to be declared.

Escalation, if appropriate, would occur via Effluent Release, in-plant radiation, or Emergency Director Judgement.

DEVIATION I

None l

REFERENCES NUMARC NESP-007, AA2.1 NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents l

l l

9 l

l 1

lgs EAL Technied B _ca Manual REv D. November 16.1998 Page 13 of 126 1.0 Reactor Fuel V

1.2 Irradiated Fuel or New Fuel ALERT - 1.2.2.b IC Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of irradiated Fuel Outside the Reactor Vessel EAL Report of visual observation of irradiated fuel uncovered OPCON tm e BASIS This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage, which is discussed in NUMARC/NESP-007 IC AU2,

" Unexpected increase in Plant Radiation or Airbome Concentration."

NUREG-0818, " Emergency Action Levels for Light Water Reactors," forms the basis for this EAL.

(V)

Studies of the loss of fuel pool water level indicate that a significant release may occur if rapid oxidation of the fuel clad occurs due to prolonged fuel uncovery. Offsite doses are not; however, expected to exceed EPA PAGs. In addition, NRC Information Notice No. 90-08, "Kr-85 Hazards from Decayed Fuel" presents the following in its discussion:

In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area radius of one mile from the plant site) would be well below the Environmental Protection Agency's Protective Action Guides. Accordingly, it is important to be able to properly survey and monitor for Kr-85 in the event of an accident with decayed spent fuel.

Licensees may wish to reevaluate whether Emergency Action Levels specified in the emergency plan and procedures goveming decayed fuel-handling activities appropriately focus on concern for onsite workers and Kr-85 releases in areas where decayed spent fuel accidents could occur, for example, the spent fuel pool working floor.

Furthermore, licensees may wish to determine if emergency plans and corresponding exposures of onsite personnel who are in other areas of the plant.

Among other things, moving onsite personnel away from the plume and shutting off building air intakes downwind from the source may be appropriate.

Thus, an Alert Classification for this event is appropriate. Escalation, if appropriate, would occur via Effluent Release, in-plant radiation, or Emergency Director Judgement.

(S DEVIATION None

' LGS EAL Tschnicsl Bisis M2nual REV D, November 16,1998 Page 14 of 126 O

REFERENCES NUMARC NESP-007, AA2.2 i

e l

l

)

l l

i l

1 0'

LGS EAL Technical Basis Manual REV o. November 16.1998 Page 15 of 126 h-1.0 Reactor Fuel u

1.2 Irradiated Fuel or New Fuel ALERT - 1.2.2.c

)

IC Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of irradiated Fuel Outside the Reactor Vessel EAL l

Water Level < 22 feet above RPV flange for the Reactor Refueling Cavity that will result in Irradiated Fuel uncovering OPCON mm BASIS This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage, which is discussed in NUMARC/NESP-007 IC AU2,

" Unexpected increase in Plant Radiation or Airbome Concentration."

l NUREG-0818, " Emergency Action Levels for Light Water Reactors," forms the basis for this f

(

EAL There is time available to take corrective actions, and there is little potential for substantial fuel damage. In addition, NUREG/CR-4982, " Severe Accident in Spent Fuel Pools in Support of i

Generic Safety issue 82," July 1987, indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and that risk of injury is low. In addition, NRC Information Notice No. 90-08, "Kr-85 Hazards from Decayed Fuel" presents the following in its discussion-In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area radius of one mile from the plant site) would be well below the Environmental Protection Agency's Protective Action Guides. Accordingly, it is important to be able to properly survey and monitor for Kr-85 in the event of an accident with decayed spent fuel.

Licensees may wish to reevaluate whether Emergency Action Level.s specified in the emergency plan and procedures goveming decayed fuel-handling activities appropriately focus on concern for onsite workers and Kr-85 releases in areas where decayed spent fuel accidents could occur, for example, the spent fuel pool working floor.

Furthermore, licensees may wish to determine if emergency plans and corresponding exposures of onsite personnel who are in other areas of the plant.

Among other things, moving onsite personnel away from the plume and shutting off building air intakes downwind from the source may be appropriate.

The value 22 feet above RPV flange is the Tech. Spec. Limit and an uncontrolled level p/

decrease that would uncover irradiated fuel is an indicator of a decrease in the level of safety y

of the plant.

Thus, an Alert Classification for this event is appropriate. Escalation, if appropriate, would occur via Effluent Release, in-plant radiation, or Emergency Director Judgement.

. LOS EAL Technical B1fs Manual REV D, Novemttr 16,1998 Page 10 of 126 DEVIATION None REFERENCES NUMARC NESP-007, AA2.3 l Tech Spec 3.9.8 O

r l

l l

9

lgs EAL Technied Basis Manua!

REV D. November 16,1998 Page 17 of 126 1.0 Reactor Fuel n

1.2 Irradiated Fuel or New Fuel ALERT - 1.2.2.d IC Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel EAL Water Level < 22 feet above seated irrad/ated Fuel for the Spent Fuel Pool that will result in Irradiated Fuel uncovering OPCON 4

BASIS This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage, which is discussed in NUMARC/NESP-007 IC AU2,

" Unexpected increase in Plant Radiation or Airbome Concentration."

NUREG-0818, " Emergency Action Levels for Light Water Reactors," forms the basis for this

' (O EAL.

/

v There is time available to take corrective actions, and there is little potential for substantial fuel damage. In addition, NUREG/CR-4982, " Severe Accident in Spent Fuel Pools in Support of Generic Safety issue 82," July 1987, indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and that risk of injury is low. In addition, NRC Information Notice No. 90-08, "Kr-85 Hazards from Decayed Fuel" presents the following in its discussion:

In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area radius of one mile from the plant site) would be well below the Environmental Protection Agency's Protective Action Guides. Accordingly, it is important to be able to properly survey and monitor for Kr-85 in the event of an accident with decayed spent fuel.

Licensees may wish to reevaluate whether Emergency Action Levels specified in the emergency plan and procedures governing decayed fuel-handling activities appropriately focus on concem for onsite workers and Kr-85 releases in areas where decayed spent fuel accidents could occur, for example, the spent fuel pool working floor.

Furthermore, licensees may wish to determine if emergency plans and corresponding exposures of onsite personnel who are in other areas of the plant.

Among other things, moving onsite personnel away from the plume and shutting off building air intakes downwind from the source may be appropriate.

The value 22 feet above seated Irradiated Fuel is the Tech. Spec. Limit and an uncontrolled (c) level decrease that would uncover irradiated fuel is an indicator of a decrease in the level of v

safety of the plant.

Thus, an Alert Classification for this event is appropriate. Escalation, if appropriate, would occur via Effluent Release, in-plant radiation, or Emergency Director Judgement.

=

I LGS EAL Technied Bas Manud REV D, Nevernber 10,1998 Page 18 of 126 f

DEVIATION None REFERENCES NUMARC NESP-007, AA2.4 l Tech Spec 3.9.9 O

O

LGS EAL Technical Bases Manual REV D, November 16.1996 Pa9e 10 of 126 2.0 Reactor Pressure Vessel 2.1 Reactor Pressure Boundary UNUSUAL EVENT - 2.1.1 IC Reactor Coolant System Leakage EAL The following conditions exist:

Unidentified Primary System Leakage > 10 ppm into the Drywell

.QB Identified Primary System Leakage > 25 ppm into the Drywell 1

OPCON cmm BASIS Utilizing the leak before break methodology, it is anticipated that there will be indication (s) of

]

minor reactor coolant system boundary integrity loss prior to this fault escalating to a major i

6' leak or rupture. Detection of low levels of leakage while pressurized is utilized to monitor for

._.(

the potential of catastrophic failures.

This EAL is included as an Unusual Event because it may be a precursor of more serious conditions and, as a result, it is considered to be a potential degradation of the level of safety

~ f the plant. The value of 10 gpm unidentified leakage is significantly higher than the expected o

pressurized leak rate from the reactor coolant system. The 10 gpm value for the unidentified pressure boundary leakage was selected as it is twice the Technical Specification value, indicating an increase beyond that assumed in Safety Analysis. It also is observable with

' normal control room indications. The EAL for identified leakage is set at a higher value (25 gpm) due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.

Technical. Specification LCO required actions would necessitate a plant shutdown and subsequent depressurization, unless the source of the leak can be isolated, identified, and/or stopped. Actions initiated by plant staff would include close monitoring of the calculated break 3

size such that any sudden or gradual increase in leak rate would be identified. A slow power l

reduction and gradual depressurization would be necessitated due to the possibility that a sudden power and/or pressure surge could potentially worsen the break or cause a catastrophic failure.

The leak rate of 10 gpm may cause a high drywell pressure indication. Other indications of a leak of this magnitude would include an increase in drywell temperature or radiation.

This event will escalate to an Alert based upon high Drywell pressure per Fission Product j

Barrier Table.

lgs EAL Technical Basis Manual REV D. November 16.1998 Page 20 of 126 O

DEVIATION NUMARC Example EAL SUS.1.a identifies pressure boundary leakage. There is no Limerick EAL listed for pressure boundary leakage specifically since it is a subset of unidentified leakage. Limerick Tech. Specs. requires a shutdown if any pressure boundary leakage is found.

REFERENCES NUMARC NESP-007, SUS Technical Specifications T-101, RPV Control T-102, Primary Containment Control 4

l O.

,. ~ _ -

, ~

.n LGS EAL Technied Basis Manual REV D. November 16,1996 Page 21 of 126

~

2.0 Reactor Pressure Vessel 2.1 Reactor Water Level i

SITE AREA EMERGENCY 2.1.3 -

IC.

Loss of Water Level in the Reactor Vessel That Has or Will Uncover fuel in the Reactor

- Vessel

~EAL RPV level < 161 "

1 I

OPCON BASIS -

l The indicator for " core is or will be uncovered" is Reactor Pressure Vessel. Water level below i

the Top of Active Fuel (TAF) -161 inches as indicated on RPV Fuel Zone Level Instruments.

Core submergence ensures adequate core cooling. When RPV level decreases below the top of active fuel the ability to remove the decay heat generated from the nuclear fuel becomes suspect and the Fuel Clad Fission. Product barrier can no longer be considered intact.

- Sustained partial. or total core uncovery can result in the release of a'significant amount of

' s fission products to the reactor coolant.

I Under the conditions specified by'this IC, severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured. It is intended to address concerns

~ raised by NRC Office for Analysis and Evaluation of Operational Data (AEOD) report AEOD/EG09,' "BWR Operating Experience involving inadvertent Draining of the Reactor Vessel," dated August 8,1986. This report states:

i In broadest terms, the dominant causes of inadvertent reactor vessel draining are related to the operational and design problems associated with the residual heat removal system when it is entering into or exiting form the shutdown cooling mode.

During this transitional period, water is drawn from the reactor vessel, cooled by the

]

residual heat removal system heat exchangers (from the cooling provided by the service water system), and retumed to the reactor vessel. First, there are piping and valves in the residual heat removal system which are common to both the shutdown cooling mode and other modes of operation such as low pressure coolant injection and suppression pool cooling. These valves, when improperly positioned, provide a drain j

L path for reactor coolant to flow from the reactor vessel to the suppression pool or the radwaste system.

Second, establishing or making such evolutions vulnerable to j

personnel'and procedural errors. Third, there is no comprehensive valve interlock arrangement for all shutdown cooling. Collectively, these factors have contributed to the inadvertent draining of the reactor vessel.

(

N Thus, declaration of a Site Area Emergency is warranted under the conditions specified by the IC. Escalation to a General Emergency is via effluent release EAL.

DEVIATION g

p.

t

-e a

lgs EAL Technical Basis Manual REtf D. November 16,1998 Page 22 of 126 During EAL review and approval process, it was determined that the condition stated in NUMARC NESP-007, SS5,1.a " Loss of all decay heat removal cooling as determined by (site-specific) procedure" is not necessary to conclude that the plant condition warrants a Site Area Emergency. Therefore, that sample NUMARC EAL was not included in this EAL.

REFERENCES NUMARC NESP-007, SS5 l

O O

LGS EAL Technical BIsis Manual REV D, November 10.1998 Page 23 of 126 2.0 Reactor Pressure Vessel

~\\

2.2 Reactor Power ALERT - 2.2.2 '

IC Failure of Reactor Protection System Instrumentation to Complete or initiate an -

Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been

' Exceeded and Manual Scram Was Successful EAL-Automatic RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS SCRAM to make Reactor shutdown OPCON BASIS Entry into this EAL is based on a reactor parameter actually exceeding a RPS setpoint and the l l reactor is not brought to a shutdown condition and maintained at that state with automatic RPS functions. The parameter must exceed the RPS setpoint by a significant margin eliminating

minor setpoint drifts which are accounted for in the Technical Specification Margin of Safety.

Subsequent manual scram actions were successful in bringing the reactor to a shutdown condition. Confirmation indications include control room annunciators, APRM/lRM/SRM power level, SRM period, and Control rod position indication.

A failure of the Reactor Protection System (RPS) to initiate and complete a reactor scram may indicate that the design limits of the nuclear fuel has been compromised. RPS is designed to automatically detect and generate a reactor scram signal when a limiting safety system

- setpoint is reached or exceeded. Control rod insertion following a scram signal is designed to be passive (i.e., system de-energizes, control rod motive energy source is previously charged).

l Assuming that shutdown (subcritical) conditions cannot be established / maintained, an automatic scram signal failure followed by a successful manual scram would still constitute a scram failure and should be classified under this event.

l Although the reactor may be brought initially suberitical based on partial control rod insertion, there is a possibility that positive reactivity may be introduced by a number of factors. Xenon decay and factors associated with cooldown, lower fuel temperature (doppler), lower moderator temperature, and a lower presence of steam bubbles (voids) may all contribute to cause a power increase.

_n

. Subcritical conditions can be assured even with the most reactive control rod fully withdrawn V

from the core if the remaining 184 control rods fully insert. Any other control rod pattem resulting from partial control rod int.ertion must be carefully analyzed and/or monitored to detect the possibility of re-criticality or local criticality.

1

lgs EAL Technical Basis Manual REV D, November 16,1998 Page 24 of126 Due to the buildup of Xenon in areas of the core that have previously been operating at high power levels, attention should be applied to the possibility that control rods which previously had low worth (e.g., peripheral control rods) may now have significant control rod worth.

When the reactor is not shutdown as identified in the Transient Response implementing Plan Procedures (TRIPS), then entry into this EAL is warranted. When partial control rod insertion occurs following a scram signal (either manual or automatic) judgement should be applied as to whether classification should occur. Multiple control rods failing to insert beyond notch position 02 may require actions to fully insert the control rods. However, the reactor has been made subcritical, and for all intent the reactor will remain suberitical. TRIP guidance will govern the insertion of these control rods.

This EAL would be escalated with a failure of both manual and automatic scram signals with l the Reactor remaining critical.

DEVIATION None REFERENCES NUMARC NESP-007, SA2 l T-101, RPV Control, RC-1 0

\\

l l

O

l lgs EAL Technical Basis Manual REV D, November 16,1998 I

i.

Page 25 of 126

! '. /]

2.0 Reactor Pressure Vessel V

2.2 Reactor Power SITE AREA EMERGENCY - 2.2.3 IC Failure of Reactor Protection System Instrumentation to Complete or initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful EAL Automatic RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%

OPCON ru m BASIS (O

A valid automatic and/or manual scram signal is present as indicted by control room indications and/or alarms and APRM indication is greater than 4% power. The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automatically). The system is " fail safe," that is, it de-energizes to function. An Anticipated Transient Without Scram (ATWS) event can be caused either by a failure of RPS (electrical failure) or a failure of the Control Rod Drive system to permit the control rods to insert (hydrauiic failure).

A failure of the Reactor Protection System to shut down the reactor (as indicated by reactor power remaining above 4%) is a degraded plant condition that together with suppression pool temperature approaching 110 F requires the injection of boron to shut down the reactor.

The RPV Control Trip Procedure establishes 4% power coincident with loss of scram capability as the initiating condition for various plant responses to ATWS events. With Reactor Power less than 4% the heat being generated in the core can be removed from the RPV and containment while actions are taken to bring the reactor subcritical.

A manual scram is defined as any set of actions by the reactor operator (s) at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical (i.e., mode switch to shutdown, manual scram push buttons, or manual ARI initiation). Taking the mode switch to shutdown as part of the actions required by trip procedure is considered a manual scram action, although the mode switch in shutdown will generate a scram signal.

While the plant is being shutdown, significant heat is being generated in the core and the heat y

up rate of the Suppression Pool (due to heat rejection through SRVs) can increase which could approach the Suppression Pool temperature limit prior to shutting down. As the Suppression Pool heat increases towards the limiting temperature, the probability of causing a major over-pressure event increases substantially.

B.GS EAL TG hnical Basis Manual REV D, November 16,1998 Page 26 of 126 After an ATWS event, there is a potential that the Main Steam isolation Valves (MSIV) will remain open. There is additional guidance in the Trip procedures to ensure that the MSIVs remain open even if RPV levelis intentionally lowered to below the normal MSIV isolation level.

This situation would allow the plant to remove heat and provide makeup through the normal steam / feed cycle, if this path is not available, or becomes unavailable during the transient, heat rejection will be to the Suppression Pool.

With Standby Liquid Control initiated and with partial or no control rod insertion, there is a possibility that the neutron flux profile in the reactor core may become uneven or distorted.

Localized clad damage is possible, if localized power levels increase significantly.

With reactor power remaining above 4% containment integrity is threatened, as the ability of systems to remove all of the heat transferred to the containment may be exceeded. As the energy contained in the containment increases there may be a degradation in the ability to remove heat generated by the "at power" reactor core. There is therefore a potential loss of the containment or the fuel cladding (caused by overheating).

This event will be escalated based on Suppression Pool Temperature exceeding 180 degrees F.

DEVIATION None REFERENCES NUMARC NESP-007, SS2 l T-101, RPV Control, RC/L-2 T-117, Level / Power Control O

LGS EAL Technical Bas's Manual REV D, November 16,1998 Page 27 of 126 2.0 Reactor Pressure Vessel 2.2 Reactor Power GENERAL EMERGENCY - 2.2.4 IC Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core EAL Automatic RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%

AND Suppression Pool Temperature is > 180 degrees F j

OPCON BASIS A valid automatic or manual scram signal is present as indicted by control room indications and/or alarms and APRM indication is greater than 4% power, in addition, control room instrumentation indicates that Suppression Pool temperature is > 180 F.

Failure of all automatic and manual trip functions coincident with a high Suppression Pool temperature will place the plant in a condition where reactivity control capability is jeopardized and heat removal capability is severely limited.

ECCS systems which may be used to cool the core, transfer heat from the reactor, and/or supply cooling water to the reactor all take a suction of the Suppression Pool. Operation with sustained high Suppression Pool temperatures may render these systems inoperable due to Net Positive Suction Head (NPSH) considerations.

The RPV Control Trip Procedure establishes 4% power coincident with loss of scram capability as the initiating condition for various plant responses to ATWS events. The timely initiation of Standby Liquid Control (prior to Suppression Pool temperature reaching 110 F) would bring the reactor to < 4 % power before Suppression Pool temperature approaches the heat capacity temperature limit curve limitations.

Under ATWS conditions, it is important to assure continuous, stable steam condensation i

capability. An elevated Suppression Pool temperature of 180 F would result in unstable steam condensation should rapid reactor depressurization occur (ADS activation).

180 F is the (q

,g SUPPRESSION POOL Heat Capacity Temperature Limit (HCTL). Maintaining the ability to condense steam will preclude the pressurization of the containment and prevent possible 4

i

. containment failure.

- lgs EAL Technkal Basis Manual HEV o. November 16,1998 Page 28 of 120 Containment over pressurization, which would be an eventual result of sustained operation with heat being added to the containment and Suppression Pool temperature above 180*F would result in the loss of containment integrity and the inability to remove the heat generated from the fuel. Fuel clad failure would result from the overheating of the 'uel DEVIATION None REFERENCES NUMARC NESP-007, SG2.1, SG2.2 l T-101, RPV Control T-117, Level / Power Control, RC/L-2 O

O P

lgs EAL Technic"J Basis Manual 1

REV o, November 16,1998 Page 29 of 126 3.0 Fission Product Barrier q-G 3,1 initiating Condition Matrix Determine which combination of the three barriers (Fuel Clad, Reactor Coolant, Primary Containment) are lost or have a potential loss and use the following key to classify the event.

'Also, an event for multiple events could occur which result in the conclusion that exceeding the loss or potential loss thresholds is IMMINENT (i.e., within 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ). In this IMMINENT LOSS situation, use judgement and classify as if the thresholds are exceeded.

UNUSUAL EVENT IC ANY loss or ANY Potential Loss of Containment EAL ANY loss or ANY Potential Loss of Containment ALERT IC ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS

(

EAL ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS SITE AREA EMERGENCY IC Loss of BOTH Fuel Clad AND RCS OR Potential Loss of BOTH Fuel Clad AND RCS OR Potential Loss of EITHER Fuel Clad OR RCS, and Loss of ANY Additional Barrier EAL Loss of BOTH Fuel Clad AND RCS OR Potential Loss of BOTH Fuel Clad AND RCS l

OR Potential Loss of EITHER Fuel Clad OR RCS, and Loss of ANY Additional Barrier fx

- G

lgs EAL Technical Basis Manual REV D. November 16,1998 Page 30 of 126 GENERAL EMERGENCY O

IC Loss of ANY Two Barriers AND Potential Loss of Third Barrier EAL Loss of ANY Two Barriers AND Potential Loss of Third Barrier OPCON cmm NOTES:

1.

Although the logic used for these initiating conditions appears overly complex, it is necessary to reflect the following considerations:

The Fuel Clad barrier and the RCS barrier are weighted more heavily than the Containment barrier. Unusual Event ICs associated with RCS and Fuel Clad barriers are addressed under the other plant condition EALs.

At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are from General Emergency. For example, if the Fuel Clad barrier and RCS barrier " Loss" EALs existed, this would indicate to the Emergency Director that, in addition to offsite dose assessments, must focus on continual assessments of radioactive inventory and containment integrity. If, on the other hand, both Fuel Clad barrier and RCS barrier " Potential Loss" EALs existed, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

The ability to escalate to higher emergency classes as an event gets worse must be maintained. For example, RCS leakage steadily increasing would represent an increasing risk to public health and safety.

2.

Fission Product Barrier ICs must be capable of addressing event dynamics. Thus, the EAL Reference Table states that IMMINENT (i.e., within 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) Loss or Potential Loss should result in a classification as if the affected threshold (s) are already exceeded, particularly for the higher emergency classes.

3.

The Fuel Clad barrier is the cladding tubes that contain the fuel pellets.

4.

The RCS Barrier is the reactor coolant system pressure boundary and includes the reactor vessel and a!! reactor coolant system piping up to the isolation valves.

5.

The Primary Containment Barrier includes the drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves.

i LGS EAL Tochtscd Basis Manual REV D, November 16,1996 Page 31 of 126

- 6.

If a " Loss" condition is satisfied, the " Potential Loss" category can be considered 3

satisfied. This is also applicable to conditions where this is a " Loss" indication with no corresponding " Potential Loss" condition.

7.-

For all conditions listed in Fission Product Barrier Table,' the barrier failure column is only satisfied if it fails when called upon to mitigate an accident. For example, failure of both containment isolation valves to isolate with a downstream pathway to the

- environment is only a concem during an accident. If this condition exists during normal power operations, it.will be an active Technical Specification -Action Statement.

However, during accident conditions, this will represent a breach of containment.

DEVIATION None REFERENCES NUMARC NESP-007, Recognition Category F, Table 3 I

O O

LGS EAL Technicd Basis Manual REV D, November 16.1998 Page 32 of 126 3.0 Fission Product Barrier O

3.2 Fuel Clad Barrier FC.1 Primary Coolant Activity Level EAL LOSS

(

l Reactor Coolant activity > 300 pCFgm Dose Equivalent lodine 131 POTENTIAL LOSS l

Not Applicable OPCON EEmm BASIS 1

A reactor coolant sample activity of greater than > 300 pCi/gm was determined to indicate significant clad heating and is indicative of the loss of the fuel clad barrier. This concentration is well above that expected for lodine spikes and corresponds to 2.8% clad damage. 2.8%

fuel clad damage is based upon NUREG-1228 core damage analysis.

Calculation of 300 Ci/cc equivalence to percent fuel clad damage is as follows (for purposes of this calculation, cc and gm are considered equivalent):

lodine isotope Dose Factors Ci/MWe Values (Time After Shutdown = 0) l (Rea Guide 1.109)

(NUREG-1228) l-131 4.39E-3 85000 l

1-132 5.23E-5 120000 1-133 1.04E-3 170000 1-134 1.37E-5 190000 1-135 2.14 E-4 150000 Time After Shutdown (T = 0) Ratios R 32 = 120000/85000(1-131) = 1.41(1-131) i l

Ri33 = 170000/85000(1-131) = 2.00(I-131)

Ri34 = 190000/85000(I-131) = 2.24(1-131) l Ri35 = 150000/85000(1-131) = 1.76(I-131) 1 Equation for Dose Equivalent lodine (Deli 3i)

, A in DF in + (R m) A in DF m + (R m) A in DF m + (R iu) A in DF m + (R m) A in DFin O

l

LGS EAL Technical Basis Maned REV D, November 16,1996 Page 33 of 126

. Solve for A 33 assuming Deli 33 = 300 Ci/cc q

i L..J 4.39E-3+1.41 A isi.23E-5+2.00 A isi.04E-3+ 2.24 A ssi.37E-5+1.76 A isi.14E 5

l l

2 300 = Ass 1 4.39E - 3

- Aisi 300=

4.39E - 3 Therefore: A 33 = 189 Ci/cc l-131 '

i

. Clad damage fraction (NUREG-1228, Table 4.1) =.02 Full Power = 1150 MWe Clad Activity 1-131

= (Ci/MWe) (MWe) (Clad Damage Fraction)

= (85000Ci/MWe) (1150MWe) ( 02)

= 1.96E6 Cl 1

Reactor Water Volume = 2.93E8 cc (ERP-C-1410)

(

Total Coolant Activity 1-131 = (Ai31)(Rx Water Volume) (Ci/ Ci) i

= (189 Ci/cc)(2.93E8cc)(1.0E-6Ci/ Ci)

= 5.54E4Ci i

Percent Clad Damage

= Total Coolant Activity / Clad Activity 1-131

= (5.54E4) / (1.53E6)

= 2.8%

This event will be escalated to an Site Area Emergency when additional fission product barriers are lost.

DEVIATION None REFERENCES l

NUMARC NESP-007, FC EAL #1 l-NUREG 1228 - Source Term Estimation During Incident Response to Severe Nuclear Power j-

. Plant Accidents, Table 2.2 Reg. Guide 1.109, Table E-9 l ERP-C-1410 n]

l t

1 lgs EAL Technical Basis Manual REv O, November 16,1998 Page 34 of 126 O) 3.0 Fission Product Barrier 3.2 Fuel Clad Barrier FC.2 Reactor Vessel Water Level EAL LOSS RPV level < -204 "

POTENTIAL LOSS RPV level < -161 "

OPCON men BASIS The " Loss" EAL -204 " value corresponds to the level which is used in the TRIPS to indicate challenge of core cooling. This is the minimum value to assure core cooling without further l

degradation of the clad. The " Potential Loss" EAL is the same as the RCS barrier " Loss" EAL 4 and corresponds to the fuel zone water level at the top of the active fuel. Thus, this EAL indicates a " Loss" of RCS barrier and a " Potential Loss" of the Fuel Clad Barrier. This EAL appropriately escalates the emergency class to a Site Area Emergency.

Core submergence is the preferred method of core cooling and as such, the failure to re-establish RPV water level above the top of active fuel for an extended period of time could lead to significant fuel damage. This condition, -204 ", could be indicative of a large break Loss Of Coolant Accident (LOCA)(where ECCS Systems are designed to maintain level at 2/3 core height) or a small LOCA with the inability of emergency core cooling systems to reflood the l

RPV. The value of -204" was chosen as it represents 2/3 core height.

l l

DEVIATION None 1

REFERENCES i

NUMARC NESP-007 FC EAL #2, RC EAL #4 T-101, RPV Control 1

. T-111, Level Restoration / Steam Cooling, LR-11 T-112, Rapid Depressurization T-117, Level / Power Control l

. T-116, RPV Flooding j

9l

LGS EAL Technical Basis Manual REV D. November 16,1996 Page 35 of 126 3.0 Fission Product Barrier 3.2 Fuel Clad Barrier FC.3 Drywell Radiation Monitoring EAL LOSS Drywell Rad Monitor reading > 4x10' R/hr POTENTIAL LOSS Not Applicable OPCON BASIS The 4x10' R/hr reading on a containment high range radiation monitor RR-26-191.291A, B, C,

' D, indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. The reading was calculated assuming an instantaneous release and dispersal of the Reactor Ccolant noble gas and iodine inventory into the Primary Containment (direct O

calculation is as follows:

reading not shine) at a coolant concentration of 300 Ci/gm Dose Equivalent lodine 131. This l Using Curve 3 [1%) of ERP-C-1410 Time after Shutdown = 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (more conservative due to lower value for EAL) 1% fuel clad damage the dose rate = 15,000 R/hr Extrapolating to 2.8%

(15,000 R/hr/1%)(2.8) = 42,000 R/hr This is rounded conservatively to 40,000 R/hr for human factors considerations 2.8% clad damage is based upon NUREG-1228 core damage analysis, and by virtue of its release into containment, the loss of the Reactor Coolant barrier (detailed calculations are contained in the Basis for Fission Product Barrier EAL FC #1).

Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage. This value is higher than that specified for RCS barrier Loss EAL #3. Thus, this EAL indicates a loss of both Fuel Clad barrier and RCS barrier.

~

There is no " Potential Loss" EAL associated with this item.

lgs EAL Technical Basis Manual REV D. November 16,1998 Page 36 of 126 DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #3 and RC EAL #3 NUREG 1228 - Source Term Estimation During Incident Response to Nuclear Power Plant Accidents l ERP-C-1410

(

i 1

O O

l LGS EAL Technical Basis Manual REV D, November 16,1998 Page 37 cf 126

- y-3.0 Fission Product Barrier 3.2 Fuel Clad Barrier

' FC.4 Other Indications EAL LOSS Not Applicable POTENTIAL LOSS Not Applicable -

OPCON rme BASIS There are no other indications at LGS for loss of the Fuel Clad Barrier.

DEVIATION

-(

None-REFERENCES NUMARC NESP-007, FC EAL #4 and RC EAL #5 i

LGS EAL Techical Basis Manual i

REV D, November 16,1998 Page 38 of 126 3.0 Fission Product Barrier ei 3.2 Fuel Clad Barrier FC.5 Emergency Director Judgement EAL Any condition in the judgement of the Emergency Director that indicates Loss or Potential Loss of the FUEL CLAD barrier OPCON 1-BASIS This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL, as a factor in Emergency Director judgement, that the barrier may be considered lost or potentially lost. (See also IC, " Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional information.)

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #5 O

lgs EAL Technical Basis Manual REV D, November 16,1998 Page 39 of 126

(]

3.0 Fission Product Barrier v

3.3 Reactor Coolant System Barrier RC.1 RCS Leak Rate EAL LOSS l

Not Applicable POTENTIAL LOSS RCS leakage >50 gpm M

Unisolable primary system leakage outside drywell as indicated by a T-103 Temperature Action Level is exceeded in ONE area requiring a SCRAM E

Unisolable primary system leakage outside drywell as indicated by a T-103 Radiation Action Level is exceeded in ONE area requiring a SCRAM OPCON rm -

BASIS r

(

l Potentialloss of RCS based on primary system leakage outside the drywellis determined from T-103 area temperatures or radiation levels.

TRIP guidance stipulates that when the Temperature or Radiation Action Level limits have been exceeded for one area, that the reactor be manually SCRAMMED.

There are two ways that the temperatures in the Secondary Containment can reach these levels; i.e., primary leak into secondary and a fire within the secondary containment. As the temperatures rise above normal conditions, the plant staff will isolate the containment and all systems, except those required for shutdown and cooling, at the Temperature Action Levels isolation levels. If the temperatures continue to rise to the Temperature Action Levels it is indicative that an unisolable leak has occurred. If the radiation levels rise above the Radiation Action Levels, it also indicates that an unisolable leak has occurred.

This event signifies that there is a direct path established for the transfer of main steam to inside the Turbine Building. Assumptions made in dose calculations regarding radioactive material transport (e.g., hold up, plate out, scrubbing, and retention) may be invalid.

Additionally the transport time associated with a radiological release may be significantly shortened and there may be a higher percentage of short lived radioisotopes in any release.

As both the reactor coolant pressure boundary and the primary containment are degraded; the i

extent of radioactive release is dependent on fuel clad integrity. Should a rapid reactor depressurization occur as a result of this event then there is a potential that a large amount of radiciodine may be released.

pV DEVIATION None I

r

LGS EAL Technical Basis Manual REV o, Noonber 16,1998 Page 40 of 126 O:

REFERENCES NUMARC NESP-007, RC EAL #1 PC EAL #2 T-103 Secondary Containment Control O

O

..~

.... ~...

LGS EAL Technical Basis Manual REV D. November 16,1998 Pf2e 41 of 126 TT 3.0 Fission Product Barrier

.V :

3.3 Reactor Coolant System Barrier RC.2 Drywell Pressure J

EAL LOSS Drywell Pressure > 1.68 psig.

AND Indication of a leak inside drywell POTENTIAL LOSS Not Applicable OPCON r-BASIS -

.The 1.68 psig drywell pressure is based on the drywell high pressure alarm set point and O

indicates a LOCA.

V' If drywell pressure-exceeds 1.68 psig, there is a clear indication that a leak of sufficient magnitude exists that prevents drywell pressure stabilization.

DEVIATION The NUMARC EAL contains only the drywell pressure. A qualifying:

"AND Indication of a leak inside drywell" was added as a human factor reminder to the Emergency Director that use of this EAL is for accident scenarios only. Thus, a Drywell pressure increase due to the loss of Drywell cooling will not require an emergency classification.

REFERENCES NUMARC NESP-007, RC EAL #2

~ T-101, RPV Control T-102, Primary Containment Control i

l Y

> v

lgs EAL Technical Basis Manual RDJ D. November 16,1998 Page 42 of 126 3.0 Fission Product Barrier O

3.3 Reactor Coolant System Barrier RC.3 Drywell Radiation Monitoring EAL LOSS Drywell Rad Monitor reading > 15 R/hr POTENTIAL LOSS Not Applicable OPCON BASIS l The 15 R/hr reading is a value which indicates the release of reactor coolant to the drywell.

The value assumes an it stantaneous release and dispersal of the reactor coolant noble gas l

and iodine inventory asswiated with concentrations corresponding to 0.001% Total Isotopic Distribution (TID) into the drywell atmosphere.

Using attachment 5 of ERP-C-1410, Curve 6 Time after Shutdown = 0.1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 0.001% TlD = 13 R/hr This is rounded to 15 R/hr for human factors considerations This reading is less than that specified for Fuel Clad Barrier EAL #3. Thus, this EAL would be indicative of a RCS leak only. If the radiation monitor reading increases to that value specified by Fuel Clad Barrier EAL #3, fuel damage would also be indicated.

There is no " Potential Loss" EAL associated with this item.

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #3 and RC EAL #3 NUREG 1228 - Source Term Estimation During Incident Response to Nuclear Power Plant Accidents l ERP-C-1410, Attachment 5 l

lgs EAL Tc:hnical Basis Manual REV D. November 16,1998 Page 43 of 126

,q 3.0 Fission Product Barrier j

ig 3.3 Reactor Coolant System Barrier i

RC.4 Reactor Vessel Water Level EAL-LOSS RPV level < -161 "

POTENTIAL LOSS Not Applicable OPCON EMBE BASIS This " Loss" EAL is the same as " Potential Loss" Fuel Clad Barrier EAL #2. The -161 " water level corresponds to the level which is used in TRIPS to indicate challenge of core cooling.

.This EAL appropriately escalates the emergency class to a Site Area Emergency. Thus, this EAL indicates a loss of the RCS barrier and a Potential Loss of the Fuel Clad Barrier.

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #2, RC EAL #4 T-101, RPV Control T-111, Level Restoration / Steam Cooling, LR-11 T-112, Rapid Depressurization i

T-117, Level / Power Control T-116, RPV Flooding O.

U

lgs EAL Technical Basis Manual REV o November 16,1998 Page 44 of 126 3.0 Fission Product Barrier O

3,3 Reactor Coolant System Barrier RC.5 Other Indications EAL LOSS Not Applicable POTENTIAL LOSS RPV level cannot be determined OPCON rram BASIS Inability to determine Reactor Pressure Vessel level may be due to reference leg boil-off, instrument power failure, or conflicting information on uncontrolled parameter oscillations.

TRIP procedure guidance will require the flooding of the Reactor Pressure Vessel, thus ensuring core submergence. Based on differences in calibration and design, all ranges of level instruments may not indicate exactly the same; this operational difference is expected and is not to be used for deciding that conflicting RPV level indication exists.

Multiple indications of level instruments pegged high is indication that the level is above the range and that it is known, also visual observation during refueling is indication of RPV water level.

If indeterminate Reactor Pressure Vessel levelis due to one of the reasons mentioned above, adequate core cooling would be rapidly assured using the guidance provided in the TRIP Procedures; however, if water level cannot be determined, it is conservative to assume that water level is actually below the top of active fuel and that both the Reactor Coolant System and Fuel Clad Fission Product Barriers are potentially lost.

Operator attention should be given to the possibility that under depressurized conditions, there is the possibility that gases may come out of solution and result in distorted RPV level indications. Operators should be attentive to observe multiple level indications (particularly those which utilize separate reference legs) to ensure that actual RPV level is known and displayed. Unexplained and/or sudden changes in specific levelindicaticas may be a result of degassification of the coolant contained in the levelinstrumentation.

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #4 and RC EAL #5 g

T-101, RPV Control, RC/L-1 W

T-112, Rapid Depressurization T-117, Level / Power Control T-110, RPV Flooding

.,.~_

- -. - -.. ~. _.

lgs EAL Technic'J Basis Manual REV D, November 16,1998 j

Page 45 of 126 3.0 Fission Product Barrier

^

3.3 Reactor Coolant System Barrier RC.6 Emergency Director Judgement 2

EAL 4

Any condition in the judgement of the Emergency Director that indicates Loss or Potential Loss of the RCS barrier i

OPCON BASIS This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL as a factor in Emergency Director judgement that the barrier may be considered lost or potentially lost. (See also IC, " Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional information.)

DEVIATION

, (

None e

l REFERENCES NUMARC NESP-007, RCS EAL #6 1

j bv l.

lgs EAL Technical Basis Manual REV D, November 16,1998 Page 46 of 126 3.0 Fission Product Barrier O

3.4 Primary Containment Barrier PC.1 Drywell Pressure EAL LOSS Rapid, unexplained decrease in Drywell Pressure following initial increase M

Drywell pressure response not consistent with LOCA conditions POTENTIAL LOSS Drywell Pressure > 44 ps/g and increasing M

Drywell Hydrogen > 6% AND Drywell Oxygen > 5%

OPCON cme BASIS Rapid unexplained loss of pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of containment integrity. Drywell pressure should increase as a result of mass and energy release into containment from a LOCA. Thus, drywell pressure not increasing under these conditions indicates a loss of containment integrity.

The 44 ps/g for potential loss of containment is based on the containment drywell design pressure. Existence of an explosive mixture means a hydrogen and oxygen concentration of at least the lower deflagration limit curve exists.

DEVIATION None REFERENCES NUMARC NESP-007, PC EAL #1 T-101, RPV Control T-102, Primary Containment Control T-103, Secondary Containment Control O'

lgs EAL Technicd Basis Manual REV D, November 16,1998 Page 47 of 126 3.0 Fissic,n Product Barrier pb 3.4 Primary Containment Barrier PC.2 Containment isolation Valve After Containment isolation EAL LOSS Failure of both valves in any one line to close AND downstream pathway to the environment exists Intentional venting per T-200 is required M

Unisolable primary system leakage outsico drywell as indicated by a T-103 Temperature Action Level is exceeded in ONE area requiring a SCRAM M

Unisolable primary system leakage outside drywell as indicated by a T-103 Radiation Action Level is exceeded in ONE area requiring a SCRAM POTENTIAL LOSS Not Applicable

~)

OPCON rme BASIS This EAL is intended to cover containment isolation failures allowing a direct flow path to the environment such as failure of both MSIVs to close with open valves downstream to the turbine or to the condenser. In addition, the presence of area radiation or temperature alarms indicating unisolable primary system leakage outside the drywell are covered. Also, an intentional venting of primary containment per TRIPS to the secondary containment and/or the environment is considered a loss of containment.

Loss of containment based on primary system leakage outside the drywell is determined from T-103 area temperatures or radiation levels.

TRIP guidance stipulates that when the Temperature or Radiation Action Level limits have been exceeded for one area, that the reactor be manually SCRAMMED.

There are two ways that the temperatures in the Secondary Containment can reach these levels; i.e., primary leak into secondary and a fire within the secondary containment. As the temperatures rise above normal conditions, the plant staff will isolate the containment and all systems, except those required for shutdown and cooling, at the Temperature Action Level isolation levels. If the temperatures continue to rise to the Temperature Action Levels it is g

indicative that an unisolable leak has occurred. If the radiation levels rise above the Radiation Action Levels, it also indicates that an unisolable leak has occurred.

LGS EAL VC:hnical BCsis Manual REV D, November 16,1996 Page 48 of 126 O

DEVIATION None REFERENCES NUMARC NESP-007, RCS EAL #1, PC EAL #2 T-103 Secondary Containment Control T-104, Radioactivity Release Control 9

1 1

0

~. - -..

lgs EAL Technical Basis Manual REV o, November 16.1998

)

Page 49 of 126 i

p 3.0 Fission Product Barrier G

3.4 Primary Containment Barrier PC.3 Significant Radioactive Inventory in Containment EAL LOSS Not Applicable-POTENTIAL LOSS Drywell Rad Monitor reading > 3xd R/hr OPCON BASIS 5

A containment high range radiation monitor RR-26-191/291A, B, C, D reading 3x10 R/hr indicates significant fuel damage, well in excess of that required for the loss of the RCS and n

Fuel Clad. As stated in Section 3.8 of NUMARC/NESP-007, a major release of radioactivity

(/

requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant.

Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted. NUREG-1228,

" Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents,"

indicates that such conditions do not exist when the amount of clad damage is less than 20%.

The reading was calculated assuming an instantaneous release of the Reactor Coolant volume into the Primary Containment (direct reading not shine) where the value corresponds to a l'

release of approximately 20% of the gap region. This calculation is as follows:

l Using Curve 3 [1%) of ERP-C-1410

. Time after Shutdown = 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (more conservative due to lower value for EAL) 1% fuel clad damage the dose rate = 15,000 R/hr Extrapolating to 20%

(15,000 R/hr/1%)(20) = 300,000 R/hr There is no " Loss" EAL associated with this item.

, f")

L g I

LGS EAL Technicd Basb Manual REif D. Rovember to.1998 Page 50 of 126 DEVIATION O

None REFERENCES NUMARC NESP-007, FC EAL #3, RC EAL #3 and PC EAL #3 NUREG 1228 - Source Term Estimation During incident Response to Severe Nuclear Power Plant Accidents l ERP-C-1410 1

O l

O

.n.

.. - ~

.-..-_..-.~.._...-._-_.--.,~--..~..,~..-.-.~..~.-.a.w_

LGS EAL Technical Basis Manual -

- REV D, November 16,1998 It Page 51 of 126 L

3.0 Fission Product Barrier -

3.4 Primary Containment Barrier -

' PC.4 Reactor Vessel Water Level

EAL LOSS Not Applicable POTENTIAL LOSS L/-

RPV level cannot be restored above -204 "

u 8hl.E l.

Maximum core uncovery time limit is in the UNSAFE region 3

- CPCON BASIS l

P In this EAL, the -204 " water level corresponds to the level which is used in the TRIPS to indicate challenge of core cooling. This is the minimum value to assure core cooling without l-further degradation of the clad.

The conditions in this potential loss EAL represent imminent melt sequences which, if not corrected, could lead to vessel failure and increased potential for containment failure. In conjunction with the level EALs in the Fuel and RCS barrier columns, this EAL will result in the declaration of a General Emergency on loss of two barriers and the potential loss of a third. If the TRIPS have been ineffective in restoring reactor vessel level within the maximum core uncovery time limit, there is not a " success" path.

Severe accident analysis (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation with the reactor vessel in a significant fraction of the core damage scenarios, and the likelihood of containment failure is very small in these events.

. Given this, it is appropriate to provide a reasonable period to allow TRIPS to arrest the core L

melt sequence. Whether or not the procedures will be effective should be apparent within the L

- time provided by the maximum core uncovery time limit. The Emergency Director should make il the declaration as soon as it is determined that the procedures have been, or will be, ineffective.

There is no " Loss" EAL associated with this item.

DEVIATION

' None

LGS EAL Technical Basis Manual REV o. November 16,1998 Page 52 of 126 O

REFERENCES NUMARC NESP-007, FC EAL #2, RC EAL #4 T-101, RPV Control T-111, Level Restoration / Steam Cooling, LR-11 T-112, Rapid Depressurization T-117, Level / Power Control T-116, RPV Flooding O'

O l

l

m LGS EAL Technical Basis Manual REV o. November 16,1996 Page 53 of 126 3.0 Fission Product Barrier

'p V

3.4 Primary Containment Barrier PC.5 Other Indications EAL LOSS Not Applicable POTENTIAL LOSS RPV level cannot be determined AND RPV Flooding cannot be established per T-116 OPCON mm BASIS The decision to enter RPV Flooding is made when RPV water level cannot be determined. This

, (q judgement consists of evaluating all plant indications which can influence the ability to maintain j

adequate core cooling. Entry to RPV flooding requires rapid RPV depressurization. The minimum RPV Flooding Pressure is defined as the lowest differential pressure between the RPV and the Suppression Pool at which steam flow through the SRVs will be sufficient to remove all of the generated decay heat. Operation at the minimum reactor flooding pressure requires that a sufficient amount of water reach the core to carry away decay heat by boiling, which in tum requires that RPV water levelincrease. So RPV flooding not established requires containment flooding.

Inability to determine Reactor Pressure Vessel level may be due to reference leg boil-off, instrument power failure, or conflicting information on uncontrolled indication oscillations. TRIP procedure guidance will require the flooding of the Reactor Pressure Vessel, thus ensuring core submergence. Based on differences in calibration and design, all ranges of level instruments may not indicate exactly the same; this operational difference is expected and is not to be used for deciding that conflicting RPV level indication exists. Level indication pegged high is indication that the level is above the range and that it is known, also visual observation during refueling is indication of RPV water level.

If indeterminate Reactor Pressure Vessel level is due to one of the reasons mentioned above, adequate core cooling would be rapidly assured using-the guidance provided in the Emergency Operating Procedures; however, if water level cannot be determined, it is conservative to assume that water level is actually below the top of active fuel and that both the Reactor Coolant System and Fuel Clad Fission Product Barriers are potentially lost.

,k l-The minimum RPV flooding pressure will ensure ' that adequate core cooling exists

- independent of RPV level indication. Failure to establish the differential pressure between the

Los EAL Technical Basis Manud REV D. November 16,1998 Page 54 of 126 RPV and the Suppression Pool in a timely manor can jeopardize the ability of the reactor coolant system to dissipate the decay heat generated.

Eventual clad failure may occur due to overheating of the nuclear fuel if RPV flooding pressure cannot be established in a timely manner. The heat produced from the fuel can cause additional core damage. If the cause of the RPV level problem was caused by a LOCA, then both the Clad and the Reactor Coolant have been lost. This will occur with heat being added to the containment. Thus there is a loss of the Fuel Clad and Reactor Coolant barriers with a potentialloss of the Containment barrier.

Ample time must be allotted for determining the failure of ECCS systems to pressurize the RPV. Control Room indications such as RPV level (used for trending), RPV Pressure, ECCS injection flow rates, Containment parameters, and injection system operability should all be used to gauge the effectiveness of the RPV Flood.

If the loss of level indication was caused by reference leg flashing, then level indicators can still be utilized to monitor the trend in RPV level. Actual RPV level will never be higher than indicated level.

In the event that the loss of level indication is only a result of degassification of the coolant contained in the level instrumentation piping, then it is anticipated that flooding pressure can be obtained.

RPV water level below the top of active fuel for a sustained period of time represents an early indicator that significant core damage is in progress while providing sufficient time to initiate public protective actions. For events starting from power operation, some core melting can be expected. Even under these conditions vessel failure and containment failure with resultant release to the public would not be expected for some time.

DEVIATION None REFERENCES NUMARC NESP-007, FC EAL #4, RCS EAL #5 and PC EAL #5 T-101, RPV Control T-111, Level Restoration / Steam Cooling, LR-11

)

T-112, Rapid Depressurization T-117, Level / Power Control T-116, RPV Flooding 0

___.m lgs EAL Technied Basis Manual REV o, Noember 16,1998 Page 55 of 126 f3 3.0 Fission Product Barrier

. Nj 3.4 Primary Containment Barrier PC.6 Emergency Director Judgement EAL' Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the CONTAINMENT barrier OPCON rme BASIS This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the Containment Barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL as a factor in Emergency Director judgement that the barriar may be considered lost or potentially lost. (See also IC, " Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional information.)

DEVIATION g

r None REFERENCES NUMARC NESP 007, PC EAL #6 l

l l

LGS EAL Technical Basis Mzaual REV D. Moecmtor 16,1998 Page $6 of 126 This page intentionally left blank l

9 I

O l

1 l

-~ n.

3.5 Fission Product Barrier Table

[ ';

Barrier Fuel Clad Reactor Coolant Sys3 Indicator Loss Potential Loss Loss PoteI N/A N/A N/A Reactor Coolant Reactor Coolant Acuvh activity > 300 pCFgm Dose Equivalent lodine 131 RPV level < -161 "

RPV level < -161 "

N/A RPV Level RPV level < -204 "

RPV Level Unknown N/A N/A N/A RPV levGf determin; h)

RCS Leak Rate N/A N/A N/A RCS lea %

>$0 gpm 9.!

Unisolabl system b outside d:

indicated Temperad Levelis G ONE arec SCRAM 9]

Unisolabl system b outside d:

indicated.

Radiatioa is exceed area reqv SCRAM l

1 j-1 l

i

.~ %

L-

3 LGS EAL Tcchnical Bnsis Manual REV D, Nov;mber 16,1998 Page 57 of 126 p

Primary Containment klLosa Loss Potential Loss N/A N/A UNUSUAL EVENT ANY Loss or ANY Potential Loss of N/A RPV level cannot be Containment restored above -204 "

AND Maximum core uncovery time limit is in the UNSAFE region ALERT

annot be N/A RPV level cannot be determined ANY Loss or ANY Potential Loss of AND EITHER Fuel Clad OR RCS RPV Flooding cannot

/4r'eRTUR@

m__

be established per T-H 6 CARD g2 N/A N/A Also Available ore SITE AREA EMERGENCY Ap irture Card Loss of BOTH Fuel Clad AND RCS Kag]

OR Potential Loss of BOTH Fuel Clad y a T 03 AND RCS tre Action OR

,ceed:d in Potential Loss of EITHER Fuel Clad

' requiring a OR RCS, and Loss of ANY Additional Barrier f prim:ry ik:g]

I/well as by a T-103 Action Lcvel GENERAL EMERGENCY l d in ONE b

ring a Loss of ANY Two Bcrriers AND Potential Loss of Third Barrier 9

0 o

6

3.5 Fission Product Barrier Table Barrier Fuel Clad Reactor Coolant Systq Indicator Loss Potential Loss Loss Potent Drywell Pressure N/A N/A N/A Diywell Pressure

> 1.68 psig AND Indication of a leak inside drywell i

N/A N/A Drywell Radiation Drywell Rad Monitor Drywell Rad Monitor l

reading > 4x1# R/hr reading > 15 R/hr Containment isolation N/A N/A N/A N/A p.

Imd Emergency Director Any condition in the judgement of the Any condition in the judgement of(

Judgement Emergency Director that indicates Loss or Emergency Director that indicates

(, _j Potential Loss of the FUEL CLAD barrier Potential Loss of the RCS barrier n -~

LGS EAL Tcchnic l Btsis Minu 1 l

REV D, Nov2mber 16,1998 l

Page 58 of 126

)

Primary Containment D Loss Loss Potential Loss l

Rapid, unexplained Drywell Pressure decrease in Drywell

> 44 psig and Pressure following increasing UNUSUAL EVENT initialincrease Drywell drogen ANY Loss or ANY Potential Loss of Drywell essure Containment response not

> 6% AND Drywell consistent with LOCA Oxygen > 5%

conditions l

N/A Drywell Rad Monitor reading > 3x10' R/hr ALERT N/A Failure of both valves in any one line to close AND ANY Loss or ANY Potential Loss of l

downstream pathway EITHER Fuel Clad OR RCS

' l,h' " ' "

AFERTUR@

ex ts CARD l

gg Intentional venting per Al'm M g y T-200 is required SITE AREA EMERGENCY Aperturo Card 9E Unisolable primary Loss of BOTH Fuel Clad AND RCS system leakage OR outside drywell as Potential Loss of BOTH Fuel Clad l

indicated by a T-103 AND RCS Temperature Action OR Levelis exceeded in Potential Loss of EITHER Fuel Clad ONE a ea requ,inng a OR S, and Loss of ANY Additional 98 l

Unisolable primary l

system leakage l

outside drywell as indicated by a T-103 GENERAL EMERGENCY j

Radiation Action Level l

Is exceeded in ONE Loss of ANY Two Barriers area requiring a AND SCRAM Potential Loss of Third Barrier te Any condition in the opinion of the Emergency Iss or Director that indicates Loss or Potential Loss of the CONTAINMENT barrier 6

a

i lgs EAL Technical Bas Mariuol _'

- REV o, November 16,19981-Page 59 of 126 H

4.0 Secondary Containment iQi 4.1-Main Steam Line UNUSUAL EVENT - 4.1.1 11C..

Fuel Ciad Degradation l

EAL'.

q Main Steam Line' HiHi Radiation (3xNFPB)

OPCOtt

' ann t.

L BASIS --

l' Main Steam Line High-High Radiation alarm (RE-41/42NBA,B,C,D) > 3 times normal full power background may be indicative of minor fuel cladding degradation and warrants the declaration l

I of an Unusual Event. This. level is set to preclude most spurious events including resin j

intrusion.

L l: n lThe main steam line high-high radiation condition requires a manual Main Steam isolation j (3 iValve closure and a reactor scram. This transient may result in the introduction of fission i

product gases (previously contained in the gap area) to be suddenly released into the coolant L due to the rapid down power transient and subsequent collapse of voids in the coolant.

L

{This level of steam line activity is indicative of the release of gap activity.to the coolant.

' however, this level is not indication of a major failure of the fuel clad The mechanics that caused main steam line radiation to increase to this level indicate there is a degradation of fuel clad integrity.

This event will escalate to an Alert based on the breach in the main steem line together with a failure of the MSIVs to isolate the main steam lines per Fission Product Barrier Table.

DEVIATION The OPCON applicability [1,2,3] is a deviation from NUMARC [all] in that, the SJAE Radiation Monitor and Main Steam Line Radiation Monitors will only be a valid indication of Fin' Clad Degradation in thosa OPCON's. At Limerick, there are no other monitors which can be an indicator of Fuel Clau Degradation. Degradation in cold shutdown or refueling will be first

indicated by ventilation release monitors and covered in Effluent Release section.

. REFERENCES NUMARC NESP-007, SU4.1 O

T-099, Post Scram Recovery IU

.T-101,' RPV Control i.

[

f

.,9

.,p

~,,,*----w--

w m

r w-+

g e---

~ - - - - - - - - - - - - - - _ - -

lgs EAL Techncti Basis Manual REV o. Novsmber 16.1998 Page 60 cf 126 4.0 Secondcry Containment 4.1 Main Steam Line ALERT - 4.1.2 IC RCS Leak Rate EAL Indication of a Main Steam Line Break:

Hi Steam Flow Annunciator AND Hi Steam Tunnel Temperature Annunciator 9.R_

Direct report of steam release OPCON mi85 BASIS Design basis accident analyses of a Main Steam Line Break outside of secondary containm shows that even if MStV closure occurs within design limits, dose consequences offsite from a

" puff" release would be in excess of 10 millirem.

Hi Steam Flow Annunciator and Hi Steam Tunnel Temperature Annunciator are both indicators of a Main Steam Line Break. Both parameters will cause an isolation of the MSIV's. Should both valves in any one line fail to isolate, this event would be considered a loss of Primary Containment and a potential loss of the RCS per the Fission Product Barrier Table and appropriately classified as a Site Area Emergency.

DEVIATION None REFERENCES NUMARC NESP-007, RC.1' T-101, RPV Control NUMARC Questions and Answers, June 1993, " Fission Product Barriers #7" O

lgs EAL Technicil B2is Manual REV D, Noember 16.1998 Page 61 of 126 5.0 Radioactivity Release 5.1 Effluent Release and Dose UNUSUAL EVENT - 5.1.1.a IC Any unplanned Release of Gaseous or Liquid Radioactivity to the Environment that l

Exceeds Two Times the Radiological Technical Specifications for 60 Minutes or Longer EAL A valid reading on one or more of the following radiatic1 monitors that exceeds TWO TIMES the HiHi alarm setpoint value for > 60 minutes:

North Stack, South Stack, Radwaste Discharge, Cooling Tower Blowdown Discharge AND Calculated maximum offsite dose rate using computer dose model exceeds 0.114 mrem /hr l

TPARD OR 0.342 mrem /hr child thyroid CDE based on a 60 minute average Note: If the required dose projections cannot be completed within the 60 minute period, then the declaration must be made based on the valid sustained monitor reading.

/G OPCON

=-

V BASIS The tenn " Unplanned", as used in this context, includes any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the appiicable permit.

Unplanned releases in excess of 0.114 mrem /hr TPARD or 0.342 mrem /hr CDE that continue for > 60 minutes represent an uncontrc:!ed situation and hence a potential degradation in the level of safety. The final integrated dose is very low and is not the primary concem. Rather it is the degradation in plant control implied by the ft:t that the release was not isolated within 60 minutes.

It is not intended that the release be averaged over 60 minutes, but exceed 0.114 mrem /hr TPARD or 0.342 mrem /hr CDE limits for 60 minutes. This EAL includes a 60 minute average for the dose projection with the release point radiation monitor above two times the HiHi alarm set point value for the entire 60 minutes. Also, it is intended that the event be declared as soon as it is determined that the release will exceed 0.114 mrem /hr TPARD or 0.342 mrem /hr CDE for greater than 60 minutes.

An indication or report is considered to be valid when it is verified by:

1.

An instrument channel check 2.

Indications on related or redundant instruments 3.

By direct observation by plant personnel Monitor indications are calculated based on the methodology of the site Offsite Dose Calculation Manual (ODCM). The HiHi alarm setpoints are set conservatively to indicate when a potential release may approach Technical Specification (ODCM) limits assuming multiple

lgs EAL Technical Bass Manual REv D. Novemt:316,1998 Page 62 of 126 release points. Use of this conservative setpoint in establishing a monitor read cause an inappropriate event classification since this EAL requires the magnitud monitor reading to be two times the setpoint, sustained for >60 minutes, and as dose projection indicating an offsite dose rate in excess of two times Technic (ODCM) limits. In the unlikely event that a dose projection cannot be comp minute period, the event will be declared based on the sustained monitor reading.

Total Protective Action Recommendation Dose (TPARD) is equal to Total Effective Dose Equivalent (TEDE) + 4 Day Deposition Dose. Committed Dose Equivalent (CDE) i the thyroid exposure due to iodine. The computerized dose model provides projec and CDE.

The Total Protective Action Recommendation Dose (TPARD) is calculated by dividing the yearly allowable Technical Specification limit (500 mrem /yr.) by the number of hours pe (8760 hr/yr.), and then multiplying by a factor of 2 times Technical Specifications [OD TPARD = 2x(Tech Spec Limit)/(hours per year)

= 2(500 mrem /yr.)/(8760 hr/yr.)

= 0.114 mrem /hr The Committed Dose Equivalent (CDE) is calculated by dividing the yearly allowable Technical Specification limit (1500 mrem /yr.) by the number of hcurs per year (8760 hr/yr.), and then multiplying by a factor of 2 times Technical Specifications [ODCM].

CDE

= 2x(Tech Spec Limit)/(hours per year)

&W

= 2(1500 mrem /yr.)/(8760 hr/yr.)

= 0.342 mrem /hr This event will be escalated to an Alert when effluents increase.

DEVIATION None REFERENCES NUMARC NESP-007, AU1.1 Offsite Dese Calculation Manual NUMARC Questions and Answers, June 1993, " Abnormal Rad Levels / Radiological Effluents

9" O

LGS EAL Technical Basis Manu:J.

REV o, Novembe' 16.1998 Page 63 cf 126 5.0 Radioactivity Release 5.1 Effluent Release and Dose UNUSUAL EVENT 5.1.1.b IC Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds Two Times Radiological Technical Specifications for 60 Minutes or Longer EAL-Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates exceeding TWO TIMES Tech Specs (ODCM 3.2.2 and 3.2.3) for > 60 minutes

.OPCON BASIS Releases in excess of two times technical specifications that continue for > 60 minutes

-represent an uncontrolled situation and hence a potential degradation in the level of safety.

The final integrated dose is very low and is not the primary concern.

Rather it is the degradation in plant control implied by the fact that the release was not isolated within 60 O

minutes.

It is not intended that the release be averaged over 60 minutes, but exceed two times technical specifications limits for 60 minutes. Further, it is intended that the event be declared as soon--

as it is determined that the release will exceed two times technical specifications for greater than 60 minutes.

t An indication or report is considered to be valid when it is verified by:

1.

An instrument channel check 2.

Indications on related or redundant instruments 3

. By direct observation by plant personnel The calculation called for in this EAL should also be conducted whenever a liquid release occurs for which a radioactive discharge permit wasn't prepared or that exceeds the conditions on the permit (e.g. minimum dilution, alarm setpoints, etc).

This event will be escalated to an Alert when effluents increase.

DEVIATION None REFERENCES

.p NUMARC NESP-007 AU1.2 Offsite Dose Calculation Manual

.T-104,. Radioactivity Release Control l

Los EAL Technscal Basis Manual REV D, November 10.1998 Page 64 of 126 5.0 Radioactivity Release O

5.1 Effluent Release and Dose ALERT - 5.1.2.a Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that IC Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer EAL A valid reading on one or more of the following radiation monitors that exceeds TWO HUNDRED TIMES the HiHi alarm setpoint value for > 15 minutes:

North Stack, South Stack, Radwaste Discharge, Cooling Tower Blowdown Discharge AND Calculated maximum offsite dose rate exceeds 11.4 mrem /hr TPARD OR 34.2 mre child thyroid CDE based on a 15 minute average If the required dose projections cannot be completed within the 15 minute period, Note:

i then the declaration must be made based on the valid sustained monitor reading.

l OPCON i

BASIS l

Releases in excess of 11.4 mrem /hr TPARD or 34.2 mrem /hr CDE that continue for > 15 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The primary concem is the final integrated dose [100 times greater than the Unusual i

Event) and the degradation in plant controlimplied by the fact that the release was not isolated within 15 minutes.

This EAL includes a 15 minute average for the dose projection with the release point radiation monitor above two hundred times the HiHi alarm set point value for the entire 15 minutes.

l Also, it is intended that the event be declared as soon as it is determined that the release will exceed 11.4 mrem /hr TPARD or 34.2 mrem /hr CDE for greator than 15 minutes.

l An indication or report is considered to be valid when it is verified by:

1.

An instrument channel check 2.

Indications on related or redundant instruments 3.

By direct observation by plant personnel Monitor indicaticos are calculated based on the methodology of the site Offsite Dose Calculation Manual (ODCM). The HiHi alarm setpoints are set conservatively to indicate when a potential release may approach Technical Specification (ODCM) limits assuming multiple release points. Use of this conservative setpoint in establishing a monitor reading will not cause an inappropriate event classification since this EAL requires the magnitude of the monitor reading to be two hundred times the setpoint, sustained for >15 minutes, and assessment by a dose projection indicating an offsite dose rate in excess of two hundred times Technical Sp scification (ODCM) limits. In the unlikely event that a dose projection cannot be

. ~... -....

~ ~

. -,. - -.. -. ~.

.....-_.-.a._~,1 1,

1 e.

LGS EAL Technical Basis Manual REV D, November 16,1998 '

Page 65 of 126 j

f completed within the 15 minute period, the event will be declared based'on the sustained monitor reading.

(

c Total Protective Action Recommendation Dose (TPARD) is equal to Total Effective Dose

~

i i

Equivalent (TEDE) + 4 Day Deposition Dose. Committed Dose Equivalent (CDE) is equal to

- the thyroid exposure due to iodine. The computerized dose model provides projected TPARD and CDE.'

The Total Protective Action ~Recomraendation Dose (TPARD) is calculated by dividing the j

yearly allowable Technical Specification limit (500 mrem /yr.) by the number of hours per year (8760 hr/yr.), and then multiplying by a factor of 200 times Technical Specifications (ODCM].

I l.

TPARD ~ = 200x(Tech Spec Limit)/(hours per year) "

i

= 200(500 mrem /yr.)/(8760 hr/yr.)

= 11.4 mrem /hr j

l' i

The Committed Dose Equivalent (CDE) is calculated by dividing the yearly allowable Technical Specification limit (1500 mrem /yr.) by the number of. hours per. year (8760 hr/yr.), and then i

multiplying by a factor of 200 times Technical Specifications (ODCM).

~

j i

.CDE.

.= 200x(Tech Spec Limit)/(hours per year) l

. = 200(1500 mrem /yr.)/(8760 hr/yr.)

= 34.2 mrem /hr This event will be escalated.to a Site Area Emergency when actual or projected doses are il determined to exceed 10CFR20 annual average population exposure limits.

DEVIATION None REFERENCES NUMARC NESP-007 AA1.1 Offsite Dose Calculation Manual NUMARC Questions and Answers, June 1993, " Abnormal Rad Levels / Radiological Effluents

9" L

l l

l O

L

lgs EAL Technical Bass Manusi REV D. Novrmber 16,1993 Page 66 of 126 5.0 Radioactivity Release O

5.1 Effluent Release and Dose ALERT - 5.1.2.b Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that IC Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer EAL Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates exceeding TWO HUNDRED TIMES Tech Specs (ODCM 3.2.2 and 3.2.3) for l

> 15 minutes OPCON BASIS Releases in excess of two hundred times technical specifications that continue for > 15 minutes represent an uncontrolled situation and hence a potential degradation in the level of The primary concem is the final integrated dose [100 times greater than the Unusual safety.

Event) and the degradation in plant control implied by the fact that the release was not isolated within 15 minutes.

It is not intended that the release be averaged over 15 minutes, but exceed two hundred times technical specifications limits for 15 rhinutes. Further, it is intended that the event be declared as soon as it is determined that the release will exceed two hundred times technical specifications for greater than 15 minutes.

An indication or report is considered to be valid when it is verified by:

1.

An instrument channel check 2.

Indications on related or redundant instruments 3.

By direct observation by plant personnel The calculation called for in this EAL should also be conducted whenever a liquid release occurs for which a radioactive discharge permit wasn't prepared or that exceeds the conditions on the permit (e.g. minimum dilution, alarm setpoints, etc).

This event will be escalated to higher classifications based on plant ccnditions.

DEVIATION None REFERENCES Ol NUMARC NESP-007 AA1.2 l

Offsite Dose Calculation Manual T-104, Radioactivity Release Control

lgs EAL Technical BIsis Manual REV o, November 16.1998 Page 67 of 126 5.0 Radioactivity Release 5.1 Effluent Release and Dose SITE AREA EMERGENCY - 5.1.3 IC Boundary Dose Resuit!ng from an Actual or imminent Release of Gaseous Radioactivity Exceeds 100 mR Whole Body or 500 mR Child Thyroid for the Actual or i

Projected Duration of the Release EAL A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes North Stack 4.16E+6 Ci/second South Stack 2.25E-3 Ci/cc AND Projected offsite dose using computer dose model exceeds 100 mrem TPARD QR_

500 mrem child thyroid CDE Note: If the required dose projections cannot be completed within the 15 minute period, j

/N then the declaration must be made based on the valid sustained monitor reading.

(,)

QR Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 100 mrem /hr expected to continue for more than one hour, QR_ Analysis of Field Survey results indicate child thyroid dose commitment of 500 mrem for one hour of inhalation OPCON BASIS Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

A monitor reading is considered to be valid when it is verified by:

1. An instrument check indicating the monitor has not failed; 2 Indications on related or redundant instrumentation; or, 3.

Direct observation by plant personnel.

l Total Protective Action Recommendation Dose (TPARD) is equal to Total Effective Dose l

Equivalent (TEDE) + 4 Day Deposition Dose.' Committed Dose Equivalent (CDE) is equal to the thyroid exposure due to iodine. Tne computerized dose model provides projected TPARD and CDE.

V An actual or projected dose of 100 mrem Total Protective Action Recommendation Dose r

l (TPARD)is based on the 10 CFR 20 annual average population exposure limit. This value also provides a desirable gradient (one order of magnitude) between the Site Area Emergency and General Emergency classifications. The 500 mrem integrated child thyroid dose was

r lgs EAL Techtucal Basis Manual REV D, Novemoer 16,1998 Page 68 of 126 estabfished in consideration of the 1:5 ratio of the EPA Protective Action Gu TPARD and Child Thyroid Committed Dose Equivalent (CDE). Actual meteorology is used, since it gives the most accurate dose projection.

Monitor indications are calculated using the computerized dose model with UFSAR source terms applicable to each monitored pathway in conjunction with annual average meteorology and a one hour release duraticn. The inputs are as follows:

North Stack South Stack Stability Class E

E Wind Speed 6.2 mph 6.2 mph Wind Direction 292*

292 Accident LOCA LOCA 4

Release Rate 4.16E+6 p/cc 2.25E-3 p/cc Child thyroid dose factors, rather than adult thyroid dose factors, are used for consistency with j

Pennsylvania Emergency Management. Agency (PEMA) / Bureau of Radiation Protection (BRP).

This event will be escalated to a General Emergency when actual or projected doses exceed EPA Protective Action Guidelines per EAL Section 5.1.4.

O DEVIATION lNone REFERENCES NUMARC NESP-007, AS1.1, AS1.3 and AS1.4 EPA 400 0

lgs EAL Technic:t Bans Minual REV o. November 16,1998 Page 69 of 128 5.0 Radioactivity Release 5.1 Effluent Release and Dose i

GENERAL EMERGENCY - 5.1.4 IC-Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1000 mR Whole Body or 5000 mR Child Thyroid for the Actual or Projected Duration of the Release Using Actual Meteorology EAL A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes:

North Stack 4.16E+7 Ci/second South Stack 2.25E-2 Ci/cc AND Projected offsite dose using computer dose model exceeds 1000 mrem TPARD Ql3 5000 mrem child thyroid CDE j

Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

,m SE Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 1000 mrem /hr expected to continue for more than one hour, QR Analysis of Field Survey results indicate child thyroid dose commitment of 5000 mrem for one hour of inhalation OPCON

~~

BASIS Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

A monitor reading is considered to be valid when it is verified by:

\\

1.

An instrument check indicating the monitor has not failed; 2 indications on related or redundant instrumentation; or,

3. Direct observation by plant personnel.

Total Protective Action Recommendation Dose (TPARD) is equal to Total Effective Dose l

Equivalent (TEDE) + 4 Day Deposition Dose. Committed Dose Equivalent (CDE) is equal to the thyroid exposure due to iodine. The computerized dose model provides projected TPARD and CDE.

':V The 1000 mR TPARD and the 5000 mR child thyroid integrated dose are based on the EPA protective action guidance. This is consistent with the emergency class description for a l-General Emergency. This level constitutes the upper level of the desirable gradient for the Site Area Emergency. Actual meteorology is specifically identified in the initiating condition j

since it gives the most accurate dose assessment.

I

l

\\

lgs EAL Technical BIsis Manual REV o, Nwemb:t 16.1998 l

Page 70 of 126 l

Monitor indications are calculated using the computerized dose model with UFSAR source terms applicable to each monitored pathway in conjunction with annual average meteorolo and a one hour release duration. The inputs are as follows:

l North Stack South Stack l

Stability Class E

E l

Wind Speed 6.2 mph 6.2 mph Wind Direction 292' 292 Accident LOCA LOCA Release Rate 4.16E+7 /cc 2.25E-2 /cc I

Child thyroid dose factors, rather than adult thyroid dose factors, are used for consistency with l

Pennsylvania Emergency Management Agency (PEMA) / Bureau of Radiation Protection l

l (BRP).

)

\\

l DEVIATION lNone REFERENCES NUMARC NESP-007, AG1.1, AG1.3 and AG1.4 EPA-400 9

lgs EAL Techncti Basis Manual REV o, Noember 13,1998 Page 71 of 126

'A 5.0 Radioactivity Release (I

5.2 In-Plant Radiation UNUSUAL EVENT - S.2.1 IC Unexpected increase in Plant Radiation or Airbome Concentration EAL Valid Direct Area Radiation Monitor readings increat,o by a factor of 1000 over normal

levels Normallevels can be considered as the highest reading in the past twenty-four hours excluding the current peak value.

OPCON BASIS Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

O An area monitor reading is considered to be valid when it is venfied by:

1.

an instrument channel check indicating the monitor has not failed; 2.

indications on related or redundant instrumentation; or 3.

direct observation by plant personnel This EAL addresses unplanned increases in in-plant radiation levels that represent a degradation in the control of radioactive material, and represents a potential degradation in the level of safety of the plant.

This event will be escalated to an Alert when radiation lovels increase in areas required for the safe shutdown of the plant resulting in impeded access.

DEVIATION None REFERENCES NUMARC NESP-007, AU2.4 T-103, Secondary Containment Control l

l rh 1

lgs EAL Technical Basis Manual REV D, November 10.1998 Pay 72 of 926 5.0 Radioactivity Release O

5.2 in-Plant Radiation ALERT - 5.2.2.a Release of Radioactive Material or increases in Radiation Levels Within the Facility IC That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown EAL Valid radiation level readings > 5000 mR/hrin areas requiring infrequent access to maintain plant safety functions as identified in procedure SE-1 or SE-6 AND Access is required for safe plant operation, but is impeded, due to radiation dose rates OPCON

'~ ~

BASIS Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

An area monitor reading is considered to be valid when it is verified by:

1. An instrument check indicating the monitor has not failed; 2 Indications on related or redundant instrumentation; or, 3.

Direct observation by plant personnel.

The single value of 5000 mR/hr was selected because it is based on radiation levels which result in exposure control measures intended to maintain doses within normal occupational exposure guidelines and limits (i.e.,10 CFR 20), and in doing so, will impede necessary access. Stay times for levels up to that value are, generally several minutes, enough time to enter an area and manually operate the equipment.

This EAL addresses increased radiation levels that impede necessary access to operating stations, or other areas containing equipment that must be operated manually, in order to maintain safe operation or perform a safe shutdown. These areas are identified in procedures SE-1 and SE-6. Use of these procedures willindicate the need to access the areas. It is this impaired ability to operate the plant that results in the actual or potential substantial degradation of the level of safety of the plant. The cause and/or magnitude of the increase in radiation levels is not a concem of this IC. The Emergency Director must consider the source or cause of the increased radiation levels and determine if any other iC may be involved. For example, a dose rate of 15 mR/hr in the control room or hi radiation monitor readings may also be indicative of high dose rates in the containment due to a LOCA. In this latter case, a SAE or GE may be indicated by the fission product barrier table.

This EAL could result in declaration of an Alert at one unit due to a radioactivity release or radiation shine resulting from a major accident at the other unit.

g.

LGS EAL Technical Basis Manual REV D. Hovember 16,1998 Page 73 (f 126 T~'%.

V This EAL is not meant to apply to increases in dryv/e!I radiation monitors, as these are events which are addressed in the fission product barrier table. Nor is it' intended to apply to

~

anticipated temporary increases due to planned events (e.g., incore detector movement, radwaste container movement, depleted resin transfers, etc.) -

This event will be escalated to a Site Area Emergency when loss of control of radioactive materials cause significant offsite doses.

DEVIATION None REFERENCES NUMARC NESP-007, AA3.2 T-103, Secondary Containment Control SE-1, Remote Shutdown.

SE-6, Alternate Remote Shutdown 10

.V I

.f

lgs EAL Technical Basis Manual REV D. Maember 16,1998 Page 74 of 126 5.0 Radioactivity Release O

S.2 in-Plant Radiation ALERT - 5.2.2.b Release of Radioactive Material or increases in Radiation Levels Within the Facility IC That impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown EAL Valid Control Room Oil Central Alarm Station radiation reading > 15 mR/hr OPCON BASIS Valid means that a radiation monitor reading has been confirmed by the operators to be correct.

An area monitor reading is considered to be valid when it is verified by:

1. An instrument check indicating the monitor has not failed; 2 Indications on related or redundant instrumentation; or, l

3. Direct observation by plant personnel.

l The EAL address radiation levels which would impede operation of systems required to l

maintain safe operations or to establish or maintain cold shutdown. Radiation levels could be indicated by ARM or radiological survey.

l l

Plant normal and emergency procedures may be implemented without requiring any areas except the Control Room and Central Alarm Station to be continuously occupied.

The Radwaste Control Room is not required to be continuously occupied in order to maintain plant l

safety functions since inputs to radwaste will be isolated with a secondary containment I

isolation and releases can only be performed manually.

l The value of 15 mR/hr is derived from the GDC 19 value of 5 REM in 30 days with adjustment for expected occupancy times. Although Section Ill.D.3 of NUREG-0737, " Clarification of TMI l

Action Plan Requirements", provides that the 15 mR/hr value can be averaged over the 30 l

days, the value is used here without averaging, as a 30 day duration implies an event potentially more significant than an Alert.

This event will be escalated to a Site Area Emergency when loss of control of radioactive l

materials cause significant offsite doses.

O

... ~. ~... ~. _.. -. _

LGS EAL Technics! Basis Manual REV D, November 16,1998 Pige 75 of 126 DEVIATION Mone REFERENCES NUMARC NESP-007 AA3.1 4

O

. LGS EAL Tschnicsi BMs Manual REV D, November 16,1996 Page 76 of 126 This page intentionally left blank O

O

..--~_._....-..~...--~_,--.-.,~.-~.-.-,._...~...~..-~.-.,_u_.

L I

LGS EAL Technied Basis Manual REV D, November 10.1996 j:

Page 77 of 126 6

1 6.0 Loss of Power 6.1 Loss of AC or DC Power-UNUSUAL EVENT - 6.1.1.a IC Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes,

.EAL

.The following conditions exist:

1 Loss of Power to 101 and 201 Safeguard Transformers for >15 minutes 8.NQ

. l At least Two Diesel Generators are supplying power to their respective 4 KV -

emergency busses OPCON BASIS This EAL addresses the loss of offsite AC power supplying the station. Offsite power is fed through 101 and 201 Safeguard Transformers. Loss of offsite power wi'l cause a reactor O

scram and a containment isolationc All four (4) emergency Diesel Generators will be available to carry the essential loads for each unit (the four Diesel Generators are shared between each L

l-unit). - Balance of Plant systems that would assist in plant operations (i.e., condensate pumps, etc.) may be unavailable due the loss of power, Prolonged loss of AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete Loss of AC Power

~

-(Station Blackout). Fifteen minutes was selected as a threshold to exclude transient or

. momentary power losses.

Escalation of this event to an Alert would be based on having a loss of all offsite AC power coincident with onsite AC power being reduced to a single power source in Modes 1,2, and 3 or having a loss of all offsite and onsite AC power in Modes 4 or 5.

I i-DEVIATION p

None REFERENCEE NUMARC' NESP-007, SU1 E-10/20, Loss of Offsite Power

?~

s lgs EAL Tahnical Basis Manual REV D, November 16,1998 Page 78 of 126 O

6.0 Loss of Power 6,1 Loss of AC or DC Power UNUSUAL EVENT 6.1.1.b Unplanned Loss of Required DC Pov/er During Cold Shutdown or Refueling Mode for IC Greater than 15 Minutes EAL The following conditions exist:

Unplanned Loss of ALL safety related DC Power indicated by < 105 VDC bus voltage l

indications for DC Panels 1(2)FA, B, C, D AND Failure to restore power to at least one required DC bus within 15 minutes from the time of the loss OPCON amas BASIS O

The purpose of this EAL is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling operations. This l

EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. The safety related 125 volt DC Distribution Panels are as follows:

l 1(2)FA, Division i Safeguard 125/250 DC Bus 1(2)FA 1(2)FB, Division il Safeguard 125/250 DC Bus 1(2)FB 1(2)FC, Division ll1 Safeguard 125 DC Bus 1(2)FC 1(2)FD, Division IV Safeguard 125 DC Bus 1(2)FD 105 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is near the t

minimum voltage selected when battery sizing is performed.

Unplanned is included in this IC and EAL to preclude the declaration of an emergency as a result of planned maintenance activities. Routinely, plants will perform maintenance on a Train l

related basis during shutdown periods. It is intended that the loss of the operating (operable) train is to be considered. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will occur.

O l

l.

..-~_m_.....

.._.. _ _.__ _ _. _ _. _.... _ _. _ -, _ _. _ _,. ~. -.

i LGS EAL Technied Basia Manual REV D Normber 16,1998

. Page 70 of 126 O

, DEVIATION -

j None

.' REFERENCES

' NUMARC NESP-007, SU7 E-1FA,- Loss of Division i Safeguard 125/250 DC Bus 1FA

- E-1FB,- Loss of Division 11 Safeguard 125/250 DC Bus 1FB E-1FC, Loss of Division 111 Safeguard 125 DC Bus 1FC E-1FD, Loss of Division IV Safeguard 125 DC Bus 1FD j

lO i

1 l'

I:

e l'

LGS EAL Technical Basis Manual REV D, November 16,1998 Page 80 of 126 O

6.0 Loss of Power 6.1 Loss of AC or DC Power ALERT - 6.1.2.a AC power capability to essential busses reduced to a single power source for greater IC than 15 minutes such that any additional single failure would result in station blackout EAL The following conditions exist:

Loss of Power to 101 and 201 Safeguard Transformers for >15 minutes AND Only One 4 KV emergency bus powered from a Single Onsite Power Source due to the Loss of: Three of Four Division Diesel Generators, D/G Output Breakers, or 4 KV Emergency Busses as indicated by bus voltage OPCON rna l

BASIS This EAL is intended to provide an escalation from " Loss of offsite Power for greater than 15 minutes." This condition is a degradation of the offsite and onsite power systems such that any additional failure would result in a station blackout.

Fifteen (15) minutes has been l

selected to exclude transient or momentary power losses. However, an Alert should be declared in less than 15 minutes if it can be determined in less than 15 minutes that the power l

loss is not transient or momentary.

f Depending on the 4 KV AC bus that remains energized there is a disparity in the systems that may be available. The ability to remove heat from the containment via Suppression Pool cooling may be lost due to the need to operate the remaining available RHR pump in other than Suppression Pool cooling (e.g., LPCI). As such there is a decrease in the systems l

available to remove heat transferred to the containment and there is an ongoing release of energy from the reactor to the containment (via SRVs, HPCI and/or RCIC operation). The l

ability to cool the nuclear fuel, remove decay heat, and control containment parameters is j

severely limited. Should equipment be unavailable prior to the loss of power, functions necessary to maintain the plant in a cold shutdown condition may be threatened.

Escalation of this event would be based on the loss of the remaining Emergency Diesel Generator.

DEVIATION l

None REFERENCES NUMARC NESP-007, SAS l

E-1, Loss of All AC Power (Station Blackout)

I

,-~

~

~.

lgs EAL Technical Basis M:nual REV D, November 16,1998

)

Page 81 of 126

. fi 6.0 Loss of Power

. 'y/

6.1 Loss of AC or DC Power 4

. ALERT - 6.1.2.b IC-Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses Dunng Cold Shutdown Or Refueling Mode I

EAL The following conditions exist:

Loss of Power to 101 and 201 Safeguard Transformers AND l

Failure to restore power to at least One 4 KV emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power l

OPCON EIEEE BASIS

- (

Loss of all.AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink. When in cold shutdown, refueling, or defueled mode, the event can be classified as an Alert, because of the significantly reduced decay-heat, lower temperature and pressure, increasing the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL. Escalating to Site Area Emergency, if appropriate, is be Effluent Release /In-Plant Radiation, or Emergency Director Judgement.

Fifteen (15) minutes has been selected to exclude transient or momentary power losses.

However, an Alert should be declared in less than 15 minutes if it can be determined in less than 15 minutes that the power loss is not transient or momentary, DEVIATION None REFERENCES NUMARC NESP-007, SA1 E-1, Loss of All AC Power (Station Blackout) gd

lgs E AL Technical Basis Manual REV D. Rovember 16.1998 Page 82 of 126 O

6.0 Loss of Power 6.1 Loss of AC or DC Power SITE AREA EMERGENCY - 6.1.3.a Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses f

IC EAL The following conditions exist:

Loss of Power to 101 and 201 Safeguard Transformers AND l

Failure to restore power to at least One 4 KV emergency bus within 15 minutes from the time of loss of both offsite and onsite AC OPCON Nmm BASIS Control Room annunciators would indicate that all offsite and onsite AC power feeds have been lost. Loss of all AC power compromises all plant safety system.s requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal, RHR Service g

Water, and Emergency Service Water. Although instrumentation (supplied through instrument inverters) and DC power loads would be available, their operability would be limited to the amount of stored energy contained in their respective batteries. Instrumentation, communication equipment, and in-plant lighting and ventilation will be significantly hampered by the loss of all AC power.

Fifteen (15) minutes has been selected to exclude transient or momentary power losses.

However, an Alert should be declared in less than 15 minutes if it can be determined in less than 15 minutes that the power loss is not transient or momentary.

Escalation of this event would be based on the time that the Emergency Diesel Generator are unavailable.

DEVIATION None REFERENCES l

NUMARC NESP-007, SS1 E-1, Loss of All AC Power (Station Blackout)

O

_.,._._.__.m.

LGS EAL Technical Basis Manual REV o, November 16,1998 Page 83 of 126 '

P 6.0 Loss of Power -

6.1 Loss of AC or DC Power

. SITE AREA EMERGENCY - 6.1.35b IC -

. Loss of All Vital DC Power j-EAL i

Loss of ALL Safety Related DC Power indicated by < 105 VDC on DC Panels 1(2)FA, B, C, D

for > 15 minutes OPCON.

N" BASIS:

I A loss of all DC power compromises the ability to monitor and control plant functions.125 Volt 4'

DC system provides control power to engineered safety features valve actuation, diesel generator auxiliaries, plant alarm and indication circuits as well as the control power for the

' associated load group. If 125 Volt DC power is lost for an extended period of time (greater 4

. than 15 minutes) critical plant functions such as RPS Logic, Attemate Rod Insertion, Emergency Service Water Indication, 4KV Breaker Controls, HPCI, RCIC and RHR pump controls required to maintain safe plant conditions may not operate and core uncovery with subsequent reactor coolant system and primary containment failure might occur. The 125 volt DC Main Distribution Panel Busses are as follows:

1(2)FA, Division i Safeguard 125/250 DC Bus 1(2)FA 1(2)FB, Division 11 Safeguard 125/250 DC Bus 1(2)FB 1(2)FC, Division lli Safeguard 125 DC Bus 1(2)FC 1(2)FD, Division IV Safeguard 125 DC Bus 1(2)FD Loss of all DC Power causes the loss of the following equipment:

e Altemate Rod insertion ADS HPCI e

RCIC Normal Recirculation Pump Trip Normal EDG Control e

e Containment instrument Gas Compressors e

j.

Other 4KV Circuit Breakers (e.g., RHR, CS, CRD)

Loss of ADS creates a loss of low pressure ECCS due to the inability to depressurize the reactor. In addition, loss of these buses will eventually lead to MSIV closure and reactor trip due to the loss of the Containment Instrument Gas Compressor as a result of suction valve

. closure.' Subsequent to MSIV closure, much of the equipment noted above will be required for plant stabilization and shutdown.

O

'A sustained loss of DC power will threaten the ability to remove heat from the reactor core, i.

resulting in eventual fuel clad damage. The loss of DC power will also result in the loss of the

" ability to remove heat from the containment SRVs will remain operable in the relief mode and

i Los EAL Technical Basis Manual GEV D. November 16,1998 Page 84 of 126 the heat addition to the containment could result in a loss of the primary containment as a fission product release barrier.

105 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of This voltage is near the operation before the onset of inability to operate those loads.

minimum voltage selected when battery sizing is performed.

DEVlATION None REFERENCES NUMARC NESP-007, SS3 T-101, RPV Control T-102, Primary Containment Control E-1, Loss of All AC Power (Station Blackout)

O t

S 1

. ~

lgs EAL Ttchnical Bass Minual REV D, Normber 16.1998 Page G3 of 126 O

6.0 Loss of Power V-6.1 Loss of AC or DC Power GENERAL EMERGENCY - 6.1.4 IC Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power EAL s

i Prolonged loss of all offsite and onsite AC power as indicated by:

Loss of Power to 101 and 201 Safeguard Transformers AND l

Failure of ALL Emergency Diesel Generators to supply power to 4 KV emergency busses i

AND At least one of the following conditions exist:

Restoration of at least One 4 KV emergency bus within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is NOTlikely g

Reactor Water Level cannot be maintained > 161 "

sQ N

O Suppression Pool temperature is greater than the Heat Capacity Temperature e

Limit (HCTL)

OPCON ram BASIS Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power willlead to loss of fuel clad, RCS, and containment. The two hours to restore AC power is base (! on the site blackout coping analysis as described below. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response.

10 CFR 50.2 defines Station Blackout (SBO) as complete loss of AC power to essential and non-essential buses. SBO does not include loss of AC Power to busses fed by station batteries through inverters, nor does it assume a concurrent single failure or design basis accident.

Successful SBO coping maintains the following key parameters within given acceptable limits:

1.

Reactor water level > -161" (TAF) f]

2.

Suppression Pool level low enough to prevent HPCI and/or RCIC steam exhaust line U

flooding 3.

Reactor pressure >150 psig to maintain HPCI and RCIC operable 4.

Containment pressure < 62.5 psig, design limit

lgs EAL Technical B! sis Manual REv o. Nov%mber 16,1998 Pege 86 of 126 6.

Drywell temperature

<200 degrees F indefinitely

<250 degrees F 99 days

<320 degrees F 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months

<340 degrees F 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Successful extended SBO coping depends on ability to keep HPCl/RCIC available for injection, and ability to maintain RPV depressurized for low pressure injection should HPCI and RCIC become unavailable. Control power for HPCI, RCIC and SRVs is provided by 125V DC.

The parameters listed above can be maintained as long as the batteries are intact. Two hours is the earliest the batteries would fail, and thus is the basis for the time limit in this EAL.

The significance of a station blackout relative to the loss of fission product release barriers is that all three barriers will eventually be lost due to the inability to remove heat from the fuel and the containment. Although the RCS will be intact the longest, eventually SRVs will operate in the relief mode due to RPV over pressurization and if the containment has already failed then there is a direct bypass of the RCS boundary.

DEVIATION None REFERENCES NUMARC NESP-007, SG1 E-1, Loss of All AC Power (Station Blackout)

T-101, RPV Control T-102, Primary Containment Control T-104, Radioactivity Release Control O

l l

LGS EAL Technicd Basis Manual REV D Nov mber 16.1998 Page 87 of 126 I

7.0 Internal Events -

7.1 Technical Specification & Control Room Evacuation UNUSUAL EVENT - 7.1.1 IC.

Inability to Reach Required Shutdown Within Technical Specification Limits EAL Inability to reach required shutdown mode within Tech. Spec. LCO required action completion time.

OPCON Estama BASIS Limiting Conditions of Operation (LCOs) require the plant to be brought to a required shutdown mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may of may not be an emergency or precursor to a more severe condition.

h any case,.the initiation of plant shutdown required by the site Technical

. Specifications requires a one hour report under 10 CFR 50.72 (b) Non-emergency events.

The plant is within its safety envelope when being shut down within the allowable action

.I statement time in the Technical Specifications. An immediate Notification of an Unusual Event is required when it is determined that the plant cannot be brought to the required operating mode witnin the allowable action statement time in the Technical Specifications. Declaration of an Unusual Event is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other various EAL Sections.

DEVIATION None REFERENCES NUMARC NESP-007, SU2 Technical Specifications

lgs EAL Techncal BMs 6Aanual REv D, Nc< ember 16,1998 Page 88 of 126 7.0 Internal Events 7.1 Technical Specification & Control Room Evacuation ALERT - 7.1.2 IC Control Room Evacuation Has Been Initiated EAL Entry into SE-1 or SE-6 procedure for Control Room evacuation l

OPCON t""~

BASIS Control Room evacuation requires establishment of plant control from outside the control room (e.g., lccal control and remote shutdown panel) and support from the Technical Support Center Control Room evacuation represents a and/or othcr emergency facilities as necessary.

serious plant situation since the level of contrnt is not as complete as it would be without evacuation. The establishment of system control outside of the Control Rocm will bypass many protective trips and interlocks. In addition, much of the instrumentation and assessment tools available in the Control Room will not be available.

This event will be escalated to an Alert if control cannot be established within fifteen minutes.

DEVIATION None REFERENCES NUMARC NESP-007.

1 SE-6, Attemate Remc< Shutdown SE-1, Remote Shutdown l

l i

_ ~. -

._~

h lgs EAL Tcchnical B:sia Manual REV D. November 16.1998' Page G9 of 126

! n 7.0 Internal Events 7.1 Technical Specification & Control Room Evacuation l

SITE AREA EMERGENCY - 7.1.3 IC Control Room Evacuation Has Been initiated and Plant Control Cannot Be Established EAL The following conditions exist:

Control room evacuation has been initiated AND Control of the plant cannot be established per SE-1 or SE-6 within 15 minutes OPCON

' * " ~

BASIS Transfer of safety system control has not been performed in an expeditious manner but it is unknown if any damage has occurred to the fission product barriers. The 15 minute time limit ip for transfer of controlis based on a reasonable time period for personnel to leave the control x

room, arrive at the remote shutdown area, and reestablish plant control to preclude core uncovery and/or core damage. During this transitional period the function of monitoring and/or

)

controlling parameters necessary for plant safety may not be occurring and as a result there i

may be a threat to plant safety.

This event will be escalated based upon system malfunctions or damage consequences.

DEVIATION 4

None REFERENCES NUMARC NESP-007, HS2

. SE-6, Altemate Remote Shutdown SE 1, Remote Shutdown

LGS EAL Technical Basis Manual REV D, No" ember 10.1998 Page 90 of 126 7.0 Internal Events O

7.2 Loss of Decay Heat Removal Capability ALERT - 7.2.2 IC Inability to Maintain Plant in Cold Shutdown t

EAL

' The following conditions exist:

l Loss of all decay heat removal cooling as determined by procedure GP-6.2 AND Uncontrolled Temperature increase that either:

Exceeds 200 *F a

9B Results in temperature rise approaching 200 "F OPCON

==

BASIS This EAL addresses complete loss of functions required for core cooling during refueling and J

cold shutdown modes. Escalation to Site Area Emergency or General Emergency would be via Effluent Release /In-Plant Radiation or Emergency Director Judgement ICs.

Procedure GP-6.2, " Shutdown Operations - Shutdown Condition Tech. Spec. Actions," directs the methods for establishing decay heat removal and guidance on amount of time available to prevent excessive heat-up as a function of time after shutdown.

"Uncontrolleo" means that system temperature increase is not the result of planned actions by the plant staff.

The EAL guidance related to uncontrol led temperature rise is necessary to preserve the anticipatory philosophy of NUREG-0654 for events starting from temperatures much lower than the cold shutdown temperature limit.

This EAL is concemed with the ability to keep the reactor core temperature less than 200 *F.

The criteria of uncontrolled Reactor Coolant temperature increase > 200 *F is met as soon as it becomes known that sufficient cooling cannot be restored in time to maintain the temperature

< 200 F, regardless of the current temperature. The inability to establish attemate methods of decey heat removal indicates that either attemate methods are unavailable to cool the core in the RPV or when the steam is transferred to the Suppression Pool, Suppression Pool cooling is unavailable. Loss of Suppression Pool cooling will result in a continuing, uncontrolled increase in reactor coolant tem.oerature.

l

.,m...

_.. _ i.._..

i.

LGS EAL Tcchnical Basis Manual i

l REV D, Novemby 16,1998 i

Page 91 of 126 -

!~

i ' 'e -

Escalation to the Site Area Emergency is by EAL IC, " Loss of Water Level in the Reactor Vessel that has or will uncover Fuel in the Reactor Vessel," or by Effluent Release /in-Plant

Radiation ICs.

I i

l~

DEVIATION

.None l

REFERENCES NUMARC NESP-007, SA3

GP-6.2., Shutdown Operations - Shutdown Conditions Tech. Spec. Actions

~: Technical Specifications L

t i

\\

l i

l..

-D I

o

~

i.--

I E..

lgs EAL Technical Baso Manual REV o, November 16,1998 Page 92 of 126 7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability SITE AREA EMERGENCY - 7.2.3 IC Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown EAL

)

Loss of Main Condenser as a heat sink J

AND Loss of SUPPRESSION POOL heat sink capabilities as evidenced by T-102 legs requiring an Emergency Blowdown AND Either of the following conditions:

l RPV level < -161 "

9.R_

Reactor Power > 4%

OPCON E2 min BASIS:

This EAL addresses complete loss of functions, including ultimate heat sink and reactivity control, required for hot shutdown with the reactor at pressure and temperature. Under these conditions, there is an actual major failure of a system intended for protection of the public.

Thus, declaration of a Site Area Emergency is warranted. Escalation to General Emergency would be via Effluent Release /In-Plant Radiation, Emergency Director Judgement, or Fission Product Barrier Degradation ICs.

The normal method for rejecting heat during operation is via the Main Condenser. If the Main Condenser is not available, heat may be rejected directly to the SUPPRESSION POOL utilizing SRVs. The number of SRVs required to reduce pressure will be dependent upon reactor pressure and power. A low SUPPRESSION POOL level would result in Heat Capacity Temperature Limit (HCTL) being exceeded if a full power blowdown occurred at water level in the SUPPRESSION POOL. A h!gh SUPPRESSION POOL temperature would result in the SUPPRESSION POOL being at the HCTL whereby it can no longer function as a heat sink. If the SUPPRESSION POOL Levelis at a high level the SUPPRESSION POOL cannot handle a full power blowdown. T-102 requires an Emergency Blowdown before these SUPPRESSION POOL conditions are reached to ensure the transfer of the energy to the SUPPRESSION POOL. Without an Emergency Blowdown, reactor pressure cannot be reduced to the shutdown cooling pressure interlock of 75 psig and shutdown cooling cannot be established.

Once the interlock is cleared, shutdown cooling can be utilized to reduce reactor coolant temperature to below 200 *F.

O

LGS EAL Technca! BIsis Manual REV D. November 16,1996 Pgge 03 of 126

' DEVIATION None REFERENCES NUMARC NESP-007, SS4-e-

h T-102, Primary Containment Control, SP/L-8 O

O I --.

i!

-?

g

lgs EAL Tecnnaf Baso Manuil REV D, Normber 16.1998 Page 94 of 126 O

7.0 Internal Events 7.3 Loss of Assessment / Communication Capability UNUSUAL EVENT - 7.3.1.a Unplanned Loss of Most or All Safety System Annunciation or Indication in The Contro IC Room for Greater Than 15 Minutes EAL Unplanned loss of most or all safety system annunciators (Table 7-1) QR indicators (Table 7-2) for > 15 minutes requiring increased surveillance to safely operate the unit (s).

OPCON Cmm BASIS l

This EAL recognizes the difficulty associated in monitoring conditions without normal l

annunciators.

In the opinion of the Shift Supervisor this loss of annunciators requires l

l increased surveillance to safely operate the plant. It is not intended that a detailed count of l

instrumentation be performed, but that only a rough approximation be used to determine the severity of the loss. The Plant Monitoring System (PMS) is available to provide compensatory indication. Fifteen minutes is used as a threshold to exclude transient or momentary power l

losses.

Unplanned loss of annunciators excludes scheduled maintenance and testing activities. Control Room panels with annunciators and direction for response are included in ON-122, Loss of Main Control Room Annunciators.

Table 7-1 indicates those system annunciator panels considered to be safety related:

Table 7-1 Safety System Annunciators ECCS Containment Isolation Reactor Trip Process Radiation Monitoring Table 7-2 indicates those indications important for monitoring:

Table 7-2 Safety Function Indicators Reactor Power Decay Heat Removal Containment Safety Functions l

Reportability of Technical Specification imposed shutdowns, or the inability to comply with Technical Specification action statements is covered in EAL section, Technical Specifications.

This EAL is not applicable in cold shutdown or refueling modes due to the limited number of safety systems required for operation.

This event will be escalated to an Alert if a transient is in orogress or if compensatory indications become unavailable.

.......__......m__...._.

_..._..m_~.___.

~

LGS EAL Technical Basia Manual REV D. Novemb2r 16,1998 Page 95 of 126 9

e -

-f DEVlATION 1.

't None e

1.

REFERENCES NUMARC NESP-007, SU3 i

- ON-122, Loss of Main Control Room Annunciators AIT A0004447, EP Self Assessment on Salem Loss of Annunciators I. -

J M

4 5

k 7'

l e

1 J

Les EAL Technical Basis Manual REV D, November 16,1998 Page 96 of 126 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability UNUSUAL EVENT - 7.3.1.b Unplanned Loss of All Onsite or Offsite Communications Capabilities IC EAL Loss of ALL Onsite communications (Table 7-3) affecting the ability to perform routine

}

operations OR l

Loss of2L Offsite communications (Table 7-3)

OPCON BASIS This EAL recognizes a loss of communication ability that significantly degrades the plant operations staff's ability to perform tasks necessary for plant operations or the ability to communicate with offsite authorities.

This EAL is separated into two groups of communications, Onsite and Offsite. A complete loss of either group is so severe, that the l Unusual Event declaration is warranted. Table 7-3 is identified as follows:

l Table 7-3 Communications Onsite Offsite Site Phones (omension 2000)

X X

PRELUDE System X

X Plant Public Address X

Station Radio X

NRC (FTS-2000)

X PA State Police Radio X

County Police Radio X

Load Dispatcher Radio X

PECO Dial Network X

There is no escalation to an Alert for loss of communications, although there is escalation to higher classifications if other communications for plant assessment is lost.

DEVIATION None REFERENCES NUMARC NESP-007, SU6 Nuclear Emergency Plan

.-..m_

___,_m.__._

m-l lgs EAL Technical Ersis Manual REV D. November 16.1998 Page 97 of 126 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability ALERT - 7.3.2 -

IC Unplanned Loss of Most or All Safety System Annunciation or Indication In Control Room With Either (1) a Significant Transient in Progress, or (2) Compensatory Non--

Alarming indicators are Unavailable EAL l

l~

Unplanned loss of most or all safety system annunciators (Table 7-1) QR indicators (Table l

7 2) for > 15 minutes requiring increased surveillance to safely operate the unit (s)

AND EITHER l

A si nificant p! ant transient is in progress (Table 7-4) QR the plant monitoring system (PMS) 0 is unavailable.

OPCON cm m l

BASIS 3

This EAL recognizes the difficulty. associated ' in monitoring conditions without normal (V

annunciators.

In.the opinion of the Shift Supervisor this loss of annunciators requires l increased surveillance to safely operate the plant. This EAL represents an increase in severity above 7.3.1.a in that the Plant Monitoring System (PMS) can~ not provide compensatory l

indication, or that a significant transient is in progress.

l Table 7-1 indicates those system annunciator panels considered to be safety related:

l p

Table 7-1 Safety System Annunciators ECCS l-Containment Isolation Reactor Trip Process Radiation Monitoring l.

Table 7-2 indicates those indications important for monitoring:

i-Table 7-2 Safety Function Indicators Reactor Power Decay Heat Removal Containment Safety Functions L

. Table 7-4, significant plant transients include response to automatic or manually initiated j

actions including:

1 L

~ - - - - - - - - - - - - - - _ -

lgs EAL Technical Basis Manual REV o. Novomcor 16.1998 Page 98 of 126

~l Table 7-4 Plant Transients SCRAM Recire runbacks > 25% thermal power Thermal power oscillations of 10% or greater Stuck open relief valves ECCS injection Fifteen minutes is used as a threshold to exclude transient or momentary power loses.

Control Room panels with annunciators and direction for restoration is included in ON-122, Loss of Main Control Room Annunciators.

Reportability of Technical Specification imposed shutdowns, or the inability to comply with Technical Specification action statements is covered in EAL section, Technical Specifications.

This EAL is not applicable in cold shutdown or refueling modes due to the limited number of safety systems required for operation.

This event will be escalated to a Site Area Emergency if a transient is in progress, the Plant Monitoring System is unavailable and a loss of annunciators occurs.

DEVIATION None REFERENCES NUMARC NESP-007, SA4 ON-122, Loss of Main Control Room Annunciators O

l

LGS EAL Techrucd Basis Manual REV o, November 16,1998 Page 99 of 126 7.0 Internal Events

(~]

'w/

7.3 Loss of Assessment / Communication Capability SITE AREA EMERGENCY - 7.3.3 IC Inability to Monitor a Significant Transient in Progress EAL Loss of safety system annunciators (Table 7-1) l AND indicators (Table 7-2)

AND PMS l

AND a significant plant transient is in progress. (Table 7-4)

OPCON N=

BASIS This EAL recognizes the difficulty associated in monitoring conditions without normal annunciators.

In the opinion of the Shift Supervisor this loss of annunciators requires l increased surveillance to safely operate the plant. This EAL represents an increase in severity

(

above 7.3.2 in that the Plant Monitoring System can not provide compensatory indication, and k

that a significant transient is in progress.

Table 7-1 indicates those system annunciator panels considered to be safety related:

Table 7-1 Safety System Annunciators ECCS Containment Isolation Reactor Trip Process Radiation Monitoring Table 7-2 indicates those indications important for monitoring:

Table 7-2 Safety Functior. Indicators Reactor Power Decay Heat Removal Containment Safety Functions l Table 7-4 significant plant transients include response to automatic or manually initiated actions including:

l Table 7-4 Plant Transients SCRAM

,q.

Recirc runbacks >25% thermal power change Q

Thermal power oscillations of 10% or greater Stuck open relief valves ECCS injection

lgs EAL Technid Basis Manual REV D, November 16,1998 Page 100 of 126 Planned maintenance or testing activities are included in this EAL due to the significance of this event. Control Room panels with annunciators and the restoration is included in ON-122, Loss of Main Control Room Annunciators.

DEVIATION None REFERENCES NUMARC NESP-007, SS6 ON-122, Loss of Main Control Room Annunciators O

O

__m

(

t LGS EAL Tschnic*.J Bsr.is Manual REV D, November 16.1998 Page 101 of 126 8.0 External Events l

8.1 Security Events L

UNUSUAL EVENT - 8.1.1 l

l lC Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Piant L

EAL L

Credible sabotage or bomb threat within the Protected Area

~E Credible intrusion and attack threat to the Protected Area 2

Attempted intrusion and attack to the Protected Area E

l Attempted sabotage discovered within the Protected Area 2

Hostage / Extortion situation that threatens normal plant operations l

OPCON

"""~

O BASIS A security threat that is identified as being directed towards the station and represents a potential degradation in the level of. safety of the plant. A security threat is satisfied if physical i

evidence supporting the threat exists, if information independent from the actual threat exists, or if a specific group claims responsibility for the threat. The Shift Management will declare an l Unusual Event subsequent to consulting with the on shift Security representative to determine the credibility of the security event.

Security threats which meet the threshold for declaration of an Unusual Event are:

1. Credible sabotage or bomb threat within the Protected Area

2. Credible intrusion and attack threat to the Protected Area

3. Attempted intrusion and attack to the Protected Area l

l 4. Attempted sabotage discovered within the Protected Area 5.-

Hostage / Extortion situation that threatens normal plant operations Security events which do not represent a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or 10 CFR 50.72 and will not cause an Unusual Event to be declared.

This event will be escalated to an Alert based upon a hostile intrusion or act within the

~ Protected Area.

f i

I lgs EAL Technical Basis Manual QEV D, November 16,1996 Page 102 of 126 O

DEVIATION A bomb device discovered within Plant Protected Area and ou Alert declaration as determined per the site Safeguards Contingency Plan an included as an Unusual Event in the EAL scheme.

REFERENCES NUMARC NESP-007. HU4.1 and HU4.2 Safeguards Contingency Plan Physical Security Plan l

O I

t l

l l

l i

l 9

f

..,....~._- --

-......,. _ _~ _..__.~..._

-. -._. _.. _. ~..-.-.... _.., -.

LGS EAL Tcchnical B: sis Manual REV D, November 16.1998 Page 103 of 126

/m 8,0 External Events 8.1 Security Events ALERT i 8.1.2 IC Security Event in a Plant Protected Area EAL-Intrusion into plant protected area by a hostile force

.Q8 Confirmed bomb, sabotage or sabotage device discovered in the Protected Area i

OPCON-m>w

BASIS l

This class of security event represents an escalated threat to the level of safety of the plant.

This event is satisfied if physical evidence supporting the hostile intrusion or attack exists. The Shift Management will declare an Alert subsequent to consulting with the on shift Security representative to determine the validity of the entry conditions.

]

Security threats which meet the threshold for declaration of an Aled are:

1 1 Intrusion into plant protected area by a hostile force

2. Confirmed bomb, sabotage or sabotage device discovered within the Protected Area l

This event will be escalated to a Site Area Emergency based upon a hostile intrusion or act in

~

plant Vital Areas.

DEVIATION None REFERENCES NUMARC NESP-007, HA4.1 and HA4.2

. Safeguards Contingency Plan

' Physical Security Plan 4

s O

W v

.r Y

F'89' W

Cf' e-

--ym?

--v-t-F

+y---

?

y-r y-wr+-m ye m'

y-Cmq 4--*+u-4 m---

v

lgs EAL Technical Basio Manual REV D. November 16,1996 Page 104 of 126 O

8.0 External Events 8.1 Security Events SITE AREA EMERGENCY - 8.1.3 IC Security Event in a Plant Vital Area EAL Intrusion into plant Vital area by a hostile force 9.R Confirmed bomb, sabotage or sabotage device discovered in a Vital Area OPCON BASIS This class of security event represents an escalated threat to plant safety above that contained in an Alert in that a hostile intrusion or attack has progressed from the Protected Area to a Vital Area. The Vital Areas are within the Protected Area and are generally controlled by key card readers. These areas contain vital equipment which includes any equipment, system, device or material, the failure, destruction or release of could directly or indirectly endanger the public health and safety by exposure to radiation. Equipment or systems which would be required to function to protect health and safety following such failure, destruction or release are also considered vital.

Security threats which meet the threshold for declaration of a Site Area Emergency are:

1. Intrusion into plant Vital area by a hostile force

2. Confirmed bomb, sabotage or sabotage device discovered in a Vital Area This event will be escalated to a General Emergency based upon the loss of physical control of the Control Room or Remote Shutdown Capability DEVIATION None REFERENCES NUMARC NESP-007, HS1.1 and HS1.2 Safeguards Contingency Plan Physical Security Plan O

lgs EAL Technical Basis Manual REV D. Nwember 10,1998 Page 105 of 126 r

8.0 External Events t

8.1 Security Events GENERAL EMERGENCY - 8.1.4 IC Security Event Resulting in Loss of Ability to Reach and Maintain Cold Shutdown EAL Loss of physical control of the control room due to security event 9.R Loss of physical control of the remote shutdown capability due to security event OPCON o""'~

BASIS This class of security event represents conditions under which a hostile force has taken physical control of areas required to reach and maintain cold shutdown. Loss of Remote Shutdown Capability would occur if the control function of the Remote Shutdown Panels was lost.

f')

Secunty events which meet the threshold for declaration of a General Emergency are physical U

loss of the Control Room or the Remote and Altemate Shutdown Panels.

This situation leaves the plant in a very unstable condition with a high potential of multiple barrier failures.

DEVIATION None REFERENCES NUMARC NESP-007, HG1.1 and HG1.2 Safeguards Contingency Plan Physical Security Plan

lgs EAL Technical Basis Manual REV D, November 16.1998 Page 106 of 126 8.0 External Events O

8.2 Fire / Explosion and Toxic / Flammable Gases UNUSUAL EVENT - 8.2.1.a IC Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes of Detection EAL Fire within SE-8 Plant Vital Structures (Table 8-1) which is not extinguished within 15 minutes of control room notification or verification of a control room alarm OPCON BASIS The pu@ose of this IC is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems. This excludes such items as fires within administration buildings, waste-basket fires, and other small fires of no safety consequence.

This IC applies to buildings and areas contiguous to plant vita! areas or other significant buildings or areas. The intent of this IC is not to include buildings (e.g., warehouses) or areas that are not cont'guous or immediately adjacent to plant vital areas. Verification of the alarm in this context means ihose actions taken in the control room to determine that the control room alarm is not spuriot.s.

This EAL addresses fires in Plant Vital Structures that house safety systems. These fires may be precursors to damage to safety systems contained in these structures. There are no areas / buildings contiguous to Plant Vital Structures which could effect a safety system in one of the listed Plant Vital Structures except for those already on the list. Therefore, no additional areas / buildings are considered for this EAL. Verification that a fire exists is by operator actions to confirm that fire alarms received in the Control Room are not spurious or by any verbal notification by plant personnel. Fifteen minutes has been established to allow plant staff to respond and control small fires or to verify that no fire exists. Table 8-1 Plant Vital Structurec are as follows:

Table 81 Plant Vital Structures Reactor Enclosure Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network This event will be escalated to an Alert if the fire damages redundant trains of plant safety systems required for the current operating condition.

O

....... ~.... _... -..

.. - _............ ~.. - -... ~. -. -.. -..

- -.. -. ~

LGS EAL Technical Batis Wnual REV D, Novsmbw 16,1998 Page 107 of 126 DEVIATION J

None REFERENCES NUMARC NESP 007, HU2 i

i 4.

O

i lgs EAL Technied Basis Manual REV D, November 16.1998 Page 108 of 126

)

i 8.0 External Events l

8.2 Fire / Explosion and Toxic / Flammable Gases UNUSUAL EVENT - 8.2.1.b Release of Toxic or Flammable Gasses Deemed Detrimental to Safe Operation of the

)

IC i

Plant EAL Report or detection of toxic or flammable gases that could enter within the site area boundary in amounts that can affect normal operation of the plant OR Report ELocal, County or State Officials for potential evacuation of site personnel based on offsite event OPCON i"' ~

BASIS This EAL addresses toxic / flammable gas releases within the Protected Area in concentrations high enough to affect health of plant personnel or the safe operation of the plant. This includes releases that originate both onsite and offsite. A toxic / flammable gas is considered to be any substance that is dangerous to life or limb by reason of inhalation or skin contact. A gas release is considered to be impeding normal plant operations if concentrations are high enough to restrict normal operator movements. It also includes areas where access is only possible with respiratory equipment, as this equipment restricts normal visibility and mobility. It should not be construed to include confined spaces that must be ventilated prior to entry or situation involving the Fire Brigade who are using respiratory equipment during the performance of their duties unless it also affects personnel not involved with the Fire Brigade.

An offsite event (such as a tanker truck accident or train derailment releasing toxic gases) may place the Protected Area within the evacuation area. This evacuation is determined from the DOT Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials.

DEVIATION None REFERENCES NUMARC NESP-007, HU3.1 and HU3.2 i

O l

l l

lgs EAL Techne.at Basis Manual REV D, November 16,1998 Page 109 of 126 8.0 External Events I

8.2 Fire / Explosion and Toxic / Flammable Gases

, UNUSUAL EVENT - 8.2.1.c IC Natural and Destructive Phenomena Affecting the Protected Area EAL l

Report by plant personnel of an unanticipated explosion within protected area boundary resulting in visible damage to permanent structure or equipment OPCON i-"

1

-BASIS l

The protected area boundary is typically that part within the security isolation zone and is defined in the site security plan.

Only those explosions of sufficient force to damage permanent structures or equipment within

-the protected area should be considered. As used here, an explosion is a rapid, violent, c

(

unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to near-by structures and materials. No attempt is made in this EAL to assess the actual magnitude of the damage. The occurrence of the explosion with reports of evidence of damage (e.g., deformation, scorching) is sufficient for declaration.

The Emergency Director also needs to consider any security aspects of the explosion, if applicable.

Any security aspects of this event should be considered under EAL Section 8.1, Security

)

Events.

This event will be escalated to an Alert if the explosion damages one or more redundant trains l

of plant safety systems required for the current operating condition.

l DEVIATION l

l None l'

REFERENCES l

NUMARC NESP-007, HU1.5 L-m l

lgs EAL Techrucal Basis Manual REV D. November 16.1998 Page 110 of 126 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases ALERT - 8.2.2.a Fire or Explosion Affecting the Operability of Plant Safety Systems Required to IC Establish or Maintain Safe Shutdown EAL The following conditions exist:

Fire or explosion which makes inoperable:

Two or More subsystems of a Safe Shutdown System (Table 8-2) QR Two or More i

Safe Shutdown Systems,OR Plant Vital Structures containing Safe Shutdown Equipment AND Safe Shutdown System or Plant Vital Structure is required for the present Operational Condition OPCON BASIS The primary concern of this EAL is the magnitude of the fire and the effects on Safe Shutdown Systems required for the present Operational Condition. A Safe Shutdown System is defined as any system required to maintain safe operation or to establish or maintain Cold Shutdown.

A system being " inoperable" means that it is incapable of performing the design function. For example, the LPCI System is intended to maintain adequate core cooling by covering the core i

to at least 2/3 core height following a DBA LOCA. In order for the system to be unable to maintain its intended function, multiple loops would need to be disabled by the fire.

Table 8 2 Safe Shutdown Systems i

l Diesel Generators 4KV Safeguard Buses ADS l

HPCI RCIC RHR (All Modes) l Core Spray RHR Service Water ESW j

SGTS RERS CAC l

PCIS Control Room Ventilation Safe Shutdown Analysis is consulted to determine systems required for the applicable mode.

Two examples of applying this methodology are as follows:

Diesel Generators and 4 KV Safeguard Buses l

The fire disables multiple Diesel Generators or 4 KV Safeguard Buses so that the number of emergency power systems available would be decreased to below what would be required to mitigate an accident under the current operating conditions.

For 100% power, this could be conservatively interpreted as at least two Diesel Generators or 4 KV Buses disabled.

m _.. _ _

_..m lgs EAL Technied Basis Manual REV o, November 16,1998 Pag) 111 of 126 l. r

(

RHR - LPCI Mode The fire disables multiple loops of LPCI so that adequate core submergence could not be assured following a DBA LOCA. For 100% power, this could also be conservatively interpreted as at least two loops disabled.

l The EAL includes the condition that the fire must make "TWO OR MORE" subsystems or

. TWO OR MORE" systems inoperable. In those cases where it is believed that the fire may have caused damage to Safety Systems, then an Alert declaration is warranted, since the full extent of the damage may not be known. For Plant Vital Structure damage, classification is required under this EAL if the structure houses or otherwise supports Safety Systems required for the present Operational Condition.

Degraded system performance or observation of damage that could degrade system performance is used as the indicator that the safe shutdown system was actually affected or made inoperable. A report of damage should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of damage. The occurrence of the fire or explosion with reports of damage l

(e.g., deformation, scorching) is sufficient for declaration.

Fire is defined as combustion characterized by the generation of heat and smoke. Sources of smoke such as overheated electrical equipment and slipping drive belts, for example, do not constitute fires. Observation of a flame is preferred, but is NOT required if large quantities of smoke and heat are observed.

This event will be escalated to higher classifications based upon damage consequences covered under other various EAL Sections.

DEVIATION j

None REFERENCES NUMARC NESP-007, HA2 LGS Safe Shutdown Analysis NUMARC Questions and Answers, June 1993, " Hazards Question #7" l

lgs EAL Techrucal Basis Manual REV D, November 16,1998 Page 112 of 126 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases ALERT - 8.2.2.b IC Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown EAL Report or detection of toxic gases within Plant Vital Structures (Table 8-1) in concentrations that will be life threatening to plant personnel O.B Report or detection of flammable gases within Plant Vital Structures (Table 8-1) in concentrations affecting the safe operation of the plant OPCON

~ -

j BASIS i

This EAL recognizes that toxic / flammable gases have entered Plant Vital Structures and are affecting safe operation of the plant by impeding operator access to the safety systems that must be operated manually in these structures. The cause and/or magnitude of the gas concentrations is not a concem, but rather that access is required to an area and is impeded.

Plant Vital Structures that must be accessed are as follows:

Table 8-1 Plant Vital Structuras Reactor Enclosure Controi Enclosure Turbine Enclosure Diesel Generator Enclosure l

Spray Pond Pump House / Spray Network The intent of this IC is not to include bdidings (e.g., warehouses) or other areas that are not contiguous or immediately adjacern to plant Vital Areas. It is appropriate that increased l

monitoring be done to ascertain whether consequential damage has occurred. This event will be escalated to higher classificat;ons based upon damage consequences covered under other various EAL Sections.

DEVIATION None REFERENCES g

l NUMARC NESP 007, HA3.1 and HA3.2 l

l l

l

.=.

.. ~..

~

lgs EAL Technical Basis Manual REV D, November 16,1998 Page 113 of 126 8.0 External Events v

8.3 Man Made Events UNUSUAL EVENT - 8.3.1.a IC Natural and Destructive Phenomena Affecting the Protected Area EAL l

Vehicle crash within protected area boundary that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant.

1 OPCON Le m BASIS This EAL is intended to address such items as plane, helicopter, or train crash that mcy potentially damage plant structures containing functions and systems required for safe shutdown of the plant. If the crash is confirmed to affect a plant vital area, the event may be escalated to Alert.

DEVIATION None REFERENCES

- NUMARC NESP-007, HU1.4 T

4 4

NA

,..e-.

LGS EAL Technical Basa Manual REV D. November 16,1998 Page 1140f 126 8.0 External Events 8.3 Man Made Events UNUSUAL EVENT. 8.3.1.b IC Natural and Destructive Phenomena Affecting the Protected Area EAL Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

OPCON BASIS This EAL is intended to address main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustible fluids (e.g., lubricating oils) and gases (e.g., hydrogen) to the plant environs. Actual fires and flammable gas build up are appropriately classified via other EALs. This EAL is consistent with the definition of an Unusual Event while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment. Escalation of the emergency classification is based on potential damage done by missiles generated by the failure or by the radiological releases and would be classified by the radiological ICs or Fission Product Barrier ICs.

Turbine failure of sufficient magnitude to cause observable damage to the turbine casing or seals of the turbine generator increases the potential for laakage of combustible fluids and gases (Hydrogen cooling) to the Turbine Enclosure.

The damage should be readily observable and should not require equipment disassembly to locate.

DEVIATION None REFERENCES NUMARC NESP 007, HU1.6 0

LGS EAL Technicd Bais Manual REV D. November 10.1998 Page 115 cf 126 8.0 External Events t

8.3 Man-Made Events ALERT - 8.3.2 IC

~ Destructive Phenomena Affecting the Plant Vital Area EAL Vehicle crash affecting Plant Vital Structures (Table 8-1) 9.8 Turbine failure generated missiles result in any visible structural damage to or penetration of any Plant Vital Structures (Table 8-1) -

OPCON-om-BASIS -

This EAL address crashes of vehicles or missile impacts that have caused damage to Plant

- Vital Structures, and thus.. damage may be assumed to have occurred to safe shutdown systems. No attempt should be made to assess the magnitude of damage to Plant Vital Structures prior to classification. The evidence of damage is sufficient for declaration. A vehicle crash includes aircraft and large motor vehicles, such as a crane. Missile impacts including flying objects from offsite, onsite rotating equipment or turbine failure causing casing penetration. Table 8-1 Plant Vital Structures are as follows:

Table 8-1 Plant Vital Structures Reactor Enclosure Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network This event will be escalated to higher classifications based upon damage consequences covered under other various EAL Sections.

DEVIATION None REFERENCES NUMARC NESP-007, HA1.5 and HA1.6

lgs EAL Technical 80 sis Manual REV D, November 16.1998 Page 116 of 126 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.a Nat:-al and Destructive Phenomena Affecting the Protected Area IC EAL Earthquake >.005 g as determined by procedure SE-5 OPCON

' " + -

BASIS This EAL addresses a sensed earthquake. The magnitude of.005g is the lowest detectable earthquake measured on LGS seismic instrumentation per SE-5. An earthquake of this magnitude may be sufficient to cause minor damage to plant structures or equipment within the Protected Area. Damage is considered to be minor, as it.would not affect physical or structural integrity.

This event is not expected to affect the capabilities of plant safety functions.

This event will be escalated to an Alert if the earthquake reaches an Operating Basis Earthquake, DEVIATION None REFERENCES NUMARC NESP-007, HU1.1 SE 5, Earthquake UFSAR, section 3.7.4.2.1 0

.~..._ _

..m.

~

.m LGS EAL Technical Basis Manual REV D, November 16,1998 Page 117 of 126 i

j-8.0. Extemal Events 8.4 Natural Events l

UNUSUAL EVENT - 8.4.1.b -

l' i

j IC

- Natural and Destructive Phenomena Affecting the Protected Area l

- EAL i

)

l Report by plant personnel of tomado striking within protected area j

[

-QR-l Wind speeds > 75 mph as indicated on site Meteorological data for > 15 minutes

' OPCOW

=tw l

. BASIS A tomado touching down within the Protected Area or wind speeds > 75 mph within the owner controlled Area are of sufficient velocity to have the potential to cause damage to Plant Vital o

~ Structures. The value of 75 mph was selected to maintain consistency with plant value and to coincide with the Beaufort Scale for Hurricane wind speed winds of 73-136 mph. These conditions are indicative of unstable weather conditions and represent a potential degradation -

in the level of safety of the plant. Venfication of a tomado will be by direct observation and

~

reporting by station personnel. Verification of wind speeds > 75 mph will be via meteorological J

l data in the control room.. For purposes of this EAL, sustained !s > 15 minutes j

This event will be escalated to an Alert if the tomado or high wind speeds strike Plant Vital Structures. If it is determined that the torWC or high wind speeds have caused a loss of shutdown cooling, then escalation will be by EAL IC, Loss of Decay Heat Removal Capability.

l

. DEVIATION None q

1 REFERENCES NUMARC NESP-007, HU1.2 and HU1.7 4

LO 3

L k

++

4

+

-w 2,

,,, -- -.-- +.., i -

e-ew r

LGS EAL Technica: Basis Manual REV D, November 16.1998 Page 118 of I?6 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.c IC Natural and Destructive Phenomena Affecting the Protected Area EAL-Assessment by the control room that an event has occurred. (Natural and Destructive Phenomena Affecting the Protected Area)

OPCON BASIS This EAL allows for the control room to determine that an event has oc'.:urred and take appropriate action based on personal ascessment as opposed to verification (s.g., an earthquake is felt but does not register on any plant-r'ecific instrumentation, etc.)

DEVIATION l

None REFERENCES NUMARC NESP-007, HU1.3 O

. - ~,..

lgs EAL Technical Basis Manual REV D, November 16.1998 Page 119 of 126 i r~N -

8.0 External Events

' d.

8.4 Natural Events I ALERT - 8.4.2.a IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL 1

Earthquake >.075 g (Operating Basis Earthquake OBE) as determined by procedure.

SE-5 OPCON

- ma u BASIS This EAL addresses an earthquake that exceeds the Operaung Basis Earthquake level of

.075g and is beyond design basis limits. An earthquake of this magnitude may be sufficient to cause damage to safety related systems and functions.

l l r

.The Max Credible Earthquake for LGS is 0.15g per UFSAR section 3.7, therefore this EAL is

(],/

conservative and warrants an Alert classification.

This event will be escalated to a higher emergency classification based upon damage l

consequences covered under other various EAL Sections.

i DEVIATION f

i None-REFERENCES NUMARC NESP-007, HA1.1 SE-5, Earthquake l

UFSAR section 3.7 f

4 a

v'

.c

lgs E AL Technical Basis Manual REV D, Wovember 16.1998 Page 120 of 126 O

8.0 External Events 8.4 Natural Events ALERT 8.4.2.b IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL Tornado or wind speeds > 75 mph striking Plant Vital Structures (Table 8-1)

OPCON uumua BASIS This EAL is based on FSAR design basis. Wind loads of this magnitude can cause damage to safety functions.

This EAL addresses events where Plant Vital Structures have been struck with high winds, and thus damage may have occurred to safe shutdown systems. No attempt should be made to assess the magnitude of damage to Plant Vital Structures prior to classification. Table 8-1 Plant Vital Structures are as follows:

Table 8-1 Plant Vital Structures Reactor Enclosure Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network This event will be escalated to a higher emergency classification based upon damage consequences covered under other various EAL Sections.

DEVIATION None REFERENCES NUMARC NESP-007, HA1.2 0

lgs EAL Technicil Basis Manual REV o. November 16.1998 Page 121 of 126 7

8.0 External Events

~

\\

8.4 Natural Events ALERT - 8.4.2.c IC-Natural and Destructive Phenomena Affecting the Plant Vital Area EAL l

Report of any visible structural damage to any Plant Vital Structure (Table 8-1)

'OPCON

""m~"'

BASIS This EAL specifies the Plant Vital Structures which contain systems and functions required for safe shutdown of the plant. Table 8-1 Plant Vital Structures are as follows:

Table 8-1 Plant Vital Structures Reactor Enclosure Contro! Enclosure s

Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network Other site structuret, listed in the NUMARC document are not plant vital structures and are not required for safe shutdown. Those are: Schuykill River Pumphouse, RWST, CST.

This event will be escalated to a higher emergency classification based upon damage consequences covereo under other various EAL Sections.

DEVIATION None REFERENCES NUMARC NESP 007, HA1.3 OV I

f'

LGS EAL Technical Basis M:nual REV D, November 16,1998 Page 122 of 126 This page intentionally left blank O.

l l

1 I

O l

...,. ~..

.-....~....._...-.

..-.........~.~.. -

- ~.... -. -

LGS EAL Technmal Basis M2tual REV D. N:vember 16,1996 Page 123 of 126 9.0 Other

< 9.1 General UNUSUAL EVENT - 9.1.1 -

~ lC.

Other Conditions Existing Which in the Judgement of the Emergency Director Warrant i

- Declaration of an Unusual Event EAL Other conditions exist which in the judgement of the Emergency Director indicate a potential degradation of the level of safety of the plant J OPCON

'~ ~ -

BASIS l

l This EAL allows the Shift Management to declare an Unusual Event 'upon the determination that the level of safety'of the plant has degraded. Where the degradation is associated with-equipment or. system malfunctions,-the decision.that it is degraded should be~made upon-functionality, not operability. A system, subsystem, train,. component or device,- though lp

- degraded in equipment condition or configuration, should be considered functional if it is

'y capable of maintaining respective system parameters within acceptable design limits.-

l l

Releases of radioactive materials requiring offsite response or monitoring are not expected to occur at this level unless further degradation of safety systems occurs. However, if one does -

occur, it will be classified under " Radioactivity Releases."

L DEVIATION None l

l REFERENCES

.NUMARC NESP-007. HUS l,-

h t,

- 3 4

j 1

e

...-m.,..

lgs EAL Technical Basis Manual REtf o. November 16,1998 Page 124 of 126 9.0 Other 9.1 General ALERT - 9.1.2 Other Conditions Existing Which in the Judgement of the Emergency Director Warrant IC Declaration of an Alert EAL Other conditions exist which in the Judgement of the Emergency Director indicate that plant safety systems may be degraded and that increased monitoring of plant functions is warranted.

OPCON

~

BASIS This EAL allows the Shift Management to declare an Alert upon tha determination that the level of safety of the plant has substantially degraded but is not explicitly addressed by other EALs. This includes a determination by Shift Management that the TSC and OSC should be activated and command and control functions should be transferred for the event to be effectively mitigated. Transfer of command and control functions is used as an initiator since an event significant to warrant transfer is a substantial reduction in the level of safety of the plant. Other examples are:

Intemal flooding affects the operability of plant safety systems required to establish or maintain cold shutdown.

Releases that are expected will be limited to a small fraction of the EPA Protective Action Guidelines and will be classified under " Radioactivity Releases."

DEVIATION None REFERENCES NUMARC NESP-007, HA6 O

lgs EAL Technical Basis Manual REV D, November 16,1998 Page 125 of 126 9.0 Other 9.1 General SITE AREA EMERGENCY - 9.1.3 IC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant Declaration of Site Area Emergency l

EAL Other conditions exist which in the Judgement of the Emergency Director indicate actual cr likely major failures of plant functions needed for protection of the public OPCON

'~ ~

BASIS This EAL' allows the Shift Management _to declare a Site Area Emergency upon the determination of an actual or likely major failure of plant functions needed for protection of the public, but is not explicitly addressed by other EALs.

Releases are not expected to result in exposure levels which exceed the EPA Protective Action Guidelines except within the site boundary and will be classified under " Radioactivity Releases,"

DEVIATION None

. REFERENCES NUMARC NESP-007, HS3 O

v

~-

lgs EAL Technical Basis Manual REV D. November 16,1998 Page 126 of 126 9.0 Other O

9.1 General GENERAL EMERGENCY - 9.1.4 Other Conditions Existing Which in the Judgement of the Er.lergency Director Warrant IC Declaration of General Emergency EAL st which in the Judgement of the Emergency Director indicate: (1)

Other conditions actual or immmain substantial core degradation with potential for loss of containment, or (2) potential for uncor, trolled radionuclide releases. These releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the site boundary

?

I OPCON BASIS i

This EAL allows the Shift Management to declare a General Emergency upon the determination of an actual or imminent substantial core degradation or melting with the potential for loss of containment integrity, but is not explicitly addressed by other EALs.

Releases may exceed the EPA Protective Action Guidelines for more than the immediate site area and will be classified under" Radioactivity Releases."

DEVIATION None REFERENCES NUMARC NESP-007, HG2 O'

_ -, ~........ _ _ _ m. m.. ___ m - _....

-[

(

yg LGS EAL Table.

. REV D, November 16,1998

. Page 1 of 24 I T, LGS EAL Table j

- Table of Contents 1.0 Reactor Fuel:

1.1 ~ : Coolant A ctivity.-.......................................,................................................. 2 1

1.2 Irradiated Fuel or New Fuel.......................................................................... 3

. 2.0 ;

Rea'etor Pressure Vessel.

'2.1 -

R eacto r Wate r Level............-................................................. c............ 4

' 2.2.

R e a cto r P owe r................................................................................. o l'

' 3.0 Fission' Product Barrier k

o

, 3.1 L Initiating Condition Matrix............................................................................ 6 l

3.2

~ Fission Product Barrier Table................................................................. 7 4.0

. Secondary Containment.

N 4.1

- M a i n Stea m Li n e.............................................................................. 9 j

$ 5.0 -

Radioactivity Release.

5.1 Effluent Release and Dose................................

.................10 5.2

' In-Plant Radiation..'...........................

................12-h 6.0 -

Loss of Power:

- 6.1 Lo ss of A C o r DC P ower.................................................................. 13 p

< +

t

4 17.0'

-Internal Events -

7.1.

Technical Specifications &l Control Room Evacuation..................................15 l

7.2

. Loss of Decay Heat Removal Capabililty.........-........................................'16 -

7.3 -. Loss of Assessment / Communications Capabililty.....................................17 L

8.0

~ External Events 8.1 : ; Secu rity Events............................................................................... 19 8.2

. Fire / Explosion and Toxic / Flammable Gases.................

..................20 8.3 Ma n-M a de Events..................................................................... 2 2 8.4 Natural Events.........................

.......23 9.0

'Other..

p

' 9.1 :

General............

. 24

},'

OPCON (MODE)-

MODE SWITCH POSITION rumma Run y

Startup_

p

' n "

= -

Shutdown (hot) r.

n== tan Shutdown (cold) umurla Refueling N

N/A (defueled) kp\\

s t...

.'q' lL i ^

l i,

4

.---,-o

-~+- -

-- ~ -

'=

- '~

" ^ ~ ~ -

LGS EAL Table REV D. Nonmber 16,1998 Page 2 of 24 A

1.0 Reactor Fuel

%J 1.1 Coolant Activity CLASSIFICATION EMERGENCY ACTION LEVEL IC Fuel Clad Degradation UNUSUAL EVENT 1.1.1.a mm Reactor Coolant activity > 4 pC#gm Dose Equivalent lodine 131 1.1.1.b mm SJAE Radiation (Offgas Monitor) > 2.1x10 mR/hr ALERT None SITE AREA None EMERGENCY GENERAL None EMERGENCY

^

LGS EAL Table REV D, November 16,1998 Page 3 of 24 r

(N 1.0 Reactor Fuel l - ()

1.2 Irradiated Fuel or Now Fuel CLASSIFICATION EMERGENCY ACTION LEVEL IC Unexpected increase in Plant Radiation or Airbome Concentration.

UNUSUAL EVENT 1.2.1.a EEE Uncontrolled water level decrease in the spent fuel pool with all irradiated fuel assemblies remaining covered by water 1.2.1.b Unexpected Fuel Pool Storage low level alarm AND Visual observation of an uncontrolled water level decrease below the fuel pool skimmer surge tank inlet IC Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will ALERT Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel 1.2.2.a

'"">'"*1 Unplanned general area radiation > 500 mR/hr on the refuel floor (Table 1-1)

'l I

1.2.2.b l

Report of visual observation of irradiated fuel uncovered l

1.2.2.c H>m a Water Level < 22 feet above RPV flange for the Reactor Refueling Cavity that will result in Irradiated Fuel uncovering 1.2.2.d

-~

l Water Level < 22 fect above seated lerad/ated Fuel for the Spent Fuel Pool that will result in irradiated Fuel uncovering l

SITE AREA None

[

EMERGENCY I

GENERAL None EMERGENCY Table 1-1 Refuel Floor ARMS RIS29-M1-1(2)K600, Drywell Head Laydown RIS30-M1-1(2)K600, Dryer /Seperator Area RIS31-M1-1(2)K600, Spent Fuel Pool r

RIS32-M1-1(2)K600, New Fuel Storage Vault (3/

RIS33-M1-1(2)K600, Pool Plug Laydown

lgs EAL Table REV D, November 16,1998 Page 4 of 24 r^3 2.0 Reactor Pressure Vessel

.( /

2.1 Reactor Water Level CLASSIFICATION EMERGENCY ACTION LEVEL IC Reactor Coolant System Leakage UNUSUAL EVENT 2.1.1 man The following conditions exist:

Unidentified Primary System Leakage > 10 ppm into the Drywell

_O_R_

Identified Primary System Leakage > 25 gpm into the Drywell ALERT None IC Loss of Water Levelin the Reactor Vessel That Has or Will Uncover fuel SITE AREA in the Reactor Vessel EMERGENCY 2.1.3 man RPV level < -161 "

GENERAL None EMERGENCY l

r

C/

lgs EAL Table REV D, November 16,1998 l

Page 5 of 24 O('"s 2.0 Reactor Pressure Vessel 2.2 Reactor Power CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT None IC Failure of Reactor Protection System Instrumentation to Complete or ALERT Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful 2.2.2 m'am Automatic RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS SCRAM to make Reactor shutdown IC Failure of Reactor Protection System Instrumentation to Complete or SITE AREA EMERGENCY initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful 2.2.3 Omm Automatic RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%

IC Failure nf the Reactor Protection System to Complete an Automatic GENERAL Scram and Manual Scram was NOT Successful and There is indication of EMERGENCY an Extreme Challenge to the Ability to Cool the Core 2.2.4 E2m Automatic RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%

AND Suppression Pool Temperature is > 180 degrees F G,l

LGS EAL Table REV D, November 16.1998 Page 6 of 24 7'

3.0 Fission Product Barrier Table (Q

3.1 initiating Condition Matrix i

CLASSIFICATION EMERGENCY ACTION LEVEL 311 "

UNUSUAL EVENT ANY Loss or ANY Potential Loss of Containment 3.1.2 ALERT ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS 3.1.3 -

SITE AREA L ss of BOTH Fuel Clad AND RCS EMERGENCY OR Potential Loss of BOTH Fuel Clad AND RCS OR Potential Loss of EITHER Fuel Clad OR RCS, and Loss of ANY Additional Barrier 3.1.4 emie GENERAL L ss of ANY Two Barriers EMERGENCY AND

-l I

Potential Loss of Third Barrier NOTES:

1.

If a " Loss" condition is satisfied, the " Potent;al Loss" category can be considered satisfied.

. 2.

For all conditions listed in Fission Product Barrier Table, the barrier failure column is only satisfied if it fails when called upon to mitigate an accident. For example, failure of both containment isolation valves to isolate with a downstream pathway to the environment is only a concern during an accident.

If this condition exists during normal power operations, it will be an active Technical Specification Action Statement. However, during accident conditions, this will represent a breach of containment.

4 ig

-1

c~,

n l

l 3.2 Fission Prodo l f,,.'

Barrier Fuel Clad Reactor Coolant Syst Indicator Loss Potential Loss Loss Poten Reactor Coolant Reactor Coolant N/A N/A N/A Activity activity > 300 pCFgm Dose Equivalent lodine 131 RPV Level RPV level < -204 "

RPV level < -161 "

N/A RPV Level Unknown N/A N/A N/A RPV leve determin:

>k RCS Leak Rate N/A N/A RCS lea N/A

>50 gp.m OJ Unisolab]

system t outside C indicated Tempera-Level is C ONE are:

SCRAM d

Unisolab]

system t outside 6 indicated Radiatio 3 is exceed area reql SCRAM a.

m

_e lgs EALTabio Bt Barrier Tcble

" "T,1ej om Primary Containment klLoss Loss Potential Loss N/A N/A

(

UNUSUAL EVENT ANY Loss or ANY Potential Loss of N/A RPV level cannot be Containment restored above -204 "

AND Maximum core uncovery time limit is in the UNSAFE region ALERT I cannot be N/A RPV level cannot be d

determined ANY Loss or ANY Potential Loss of AND EITHER Fuel Clad OR RCS RPV Flooding cannot be established per APERTURE T-116 CARD aga N/A N/A

+4100 Available on SITE AREA EMERGENCY Aport 1re Card i

' P" L ss of BOTH FuelClad AND RCS sk' g OR peuas Potential Loss of BOTH Fuel Clad by a T-103 AND RCS ura Actiop OR xceeded in Potential Loss of EITHER Fuel Clad i requiring a OR RCS, and Loss of ANY Additional Barrier g

e primary sk:ge well as by a T-103 GENERAL EMERGENCY Action Level ed in ONE Loss of ANY Two Barriers iring a AND Potential Loss of Third Barrier r

2, l

l u

7-

3.2 Fission ProdI g(. -

Barrier Fuel Clad Reactor Coolant Sys2 indicator Loss Potential Loss Loss Poten Drywell Pressure N/A N/A N/A Drywell Pressure

> 1.68 psig AND Indication of a leak inside drywell Drywell Radiation Drywell Rad Monitor Drywell Rad Monitor N/A N/A l

reading > 4x10' R/hr reading > 15 R/hr Containment isolation N/A N/A N/A N/A b

l 1

I l

Emergency Director Any condition in the judgement of the Any condition in the judgement of t gN Judgement Emergency Director that indicates Loss or Emergency Director that indicates l Potential Loss of the FUEL CLAD barrier Potential Loss of the RCS barrier l

M M4 L..

..m-

{t Barrier Tcble LGS EALTable

]

REV D, Nowry1ber 16,1990 I

Page 8 of 24 i

l pm.

Primary Containment ial Loss Loss Potential Loss l

Rapid, unexplained Drywell Pressure decrease in Drywell

> 44 ps/g and Pressure following increasing UNUSUAL EVENT initialincrease OR Drywell drogen ANY loss or ANY Potential Loss of Drywell essure Containment i

response not

> 6% AND Drywell consistent with LOCA Oxygen > 5%

conditions N/A Drywell Rad Monitor l

reading > 3x10' R/hr ALERT N/A l

Failure of both valves in any one line to close AND ANY Loss or ANY Potential Loss of l

downstream pathway EITHER Fuel Clad OR RCS l

to the environment APERTURI exists E

Intentional venting per Also Available T-200 is required SITE AREA EMERGENCY _

E I

Unisolable primary Loss of BOTH Fuel Clad AND RCS l

system leakage OR L

outside drywell as Potential Loss of BOTH Fuel Clad l

Indicated by a T-103 AND RCS l

Temperature Action OR I

Levelis exceeded in Potential Loss of EITHER Fuel Clad ONE a ea requin,ng a OR RCS, and Loss of ANY Additional C

Barrier E

l Unisolable primary l

system leakage outside drywell as

)

l indicated by a T-103 GENERAL EMERGENCY L

Radiation Action Level is exceeded in ONE Loss of ANY Two Barriers area requinng a AND SCRAM Potential Loss of Third Barrier ne Any condition in the opinion of the Emergenc/

l.oss cr Director that indicates Loss or Potential Lou of the CONTAINMENT barrier l

N lA [V V

T l

i L

i

. ~.

LGS EAL Table REV o, November 16,1998 Page 9 of 24 l

fs,.

4.0 Secondary Containment q) 4.1 Main Steam Line CLASSIFICATION EMERGENCY ACTION LEVEL IC Fuel Clad Degradation UNUSUAL EVENT 4.1.1 c a nn Main Steam Line HiHi Radiation (3xNFPB)

ALERT IC RCS Leak Rate 4.1.2 aman Indication of a Main Steam Line Break:

Hi Steam Flow Annunciator AND Hi Steam Tunnel Temperature Annunciato-9_8 Direct report of steam release SITE AREA ~

None EMERGENCY GENERAL None EMERGENCY l

4 4

4 4 'a

REV D. No m

,i Page 10 of 24 1

l A

5.0 Radioactivity Release

" 'uf 5.1 Effluent Release arid Dose

. CLASSIFICATION EMERGENCY ACTION LEVEL IC '

Any Unplanned Release of Gaseous or Liquid Radioactivity to the UNUSUAL EVENT Environment that Exceeds Two Times the Radiological Technical Specifications for 60 Minutes or Longer 5.1.1.a EEHEEE A valid reading on one or more of the following radiation monitors that exceeds TWO TIMES the HiHi alarm setpoint value for > 60 minutes:

North Stack, South Stack, Radwaste Discharge, Cooling Tower Blowdown Dischargc AND Calculated maximum offsite dose rate using computer dose model exceeds 0.114 mrem /hr TPARD QR 0.342 mrem /hr child thyrold CDE based on a j

60 minute average Note: If the required dose projections cannot be completed within the 60 minute period, then the declaration must be made based on the valid sustained monitor reading.

- I I

5.1.1.b -i Ccnfirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates exceeding iWC TIMES Tech Specs (ODCM

. l 3.2.2 and 3.2.3) for > 60 minutes IC

. Any Unplanned Release of Gaseous or Liquid Radioactivity to the ALERT.

Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes'or Longer 5.1.2.a mmm c,

A valid ieading on one or more of the following radiation monitors that exceeds TWO HUNDRED TIMES the HiHi alarm setpoint value for > 15 minutes:

North Stack, South Stack, Radwaste Discharge, Cooling Tower Blowdown Discharge AND Calculated maximum offsite dose rate exceeds 11.4 mrem /hr TPARD Ql3 34.2 mrem /hr child thyroid CDE based on a 15 minute average Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

5.1.2.b q

Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates exceeding TWO HUNDRED TIMES Tech Specs Il (ODCM 3.2.2 and 3.2.3) for > 15 minutes L..-

r LGS EAL Table REV o, November 16,1998 Page 11 of 24 IC Boundary Dose Resulting from an Actual or Imminent Release of SITE AREA EMERGENCY Gaseous Radioactivity Exceeds 100 mR Whole Body or 500 mR Child Thyroid for the Actual or Projected Duration of the Release 5.1.3 rum A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes:

North Stack 4.16E+6 Ci/second South Stack 2.25E-3 Ci/cc AND 4

Projected offsite dose using computer dose model exceeds 100 mrem TPARD

@ 500 mrem child thyroid CDE Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 100 mrem /hr expected to continue for more than one hour, M Analysis of Field Survey results indicate child thyroid dose commitment of 500 mrem for one hour of inhalation IC Boundary Dose Resulting from an Actual or Imminent Release of GENERAL EMERGENCY Gaseous Radioactivity that Exceeds 1000 mR Whole Body or 5000 mR Child Thyroid for the Actual or Projected Duration of the Release Using Actual Meteorology 5.1.4 i~ u A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes:

North Stack 4.16E+7 Ci/second South Stack 2.25E-2 Ci/cc AND Projected ef' site dose using computer dose model exceeds 1000 mRom TPARD M 5000 mrem child thyroid CDE Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

M Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 1000 mrem /hr expected to continue for more than one hour, M Analysis of Field Survey results indicate child thyroid dose commitment of 5000 mrem for one hour of inhalation NOTE:

CDE Committed Dose Equivalent

=

TPARD

=

Total Protective Action Recommendation Dose

lgs EAL Table REV D, November 16,1998 Page 12 of 24

^'t 5.0 Radioactivity Release

5000 mR/hrin areas requiring infrequent access to maintain plant safety functions as identified in procedure SE-1 or SE-6 l

I AND Access is required for safe plant operation, but is impeded, due to radiation dose rates 5.2.2.b mm Valid Control Room OR Central Alarm Station radiation reading > 15 mR/hr SITE AREA None EMERGENCY GENERAL None EMERGENCY t

i I

l lgs EAL Table REV D, N;vember 16,1998 Page 13 of 24 6.0 Loss of Power V(N 6.1 Loss of AC or DC Power CLASSIFICATION EMERGENCY ACTION LEVEL IC Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes UNUSUAL EVENT 6.1.1.a The following conditions exist:

Loss of Power to 101 and 201 Safeguard Transformers for >15 minutes AND At least Two Diesel Generators are supplying power to their respective 4 KV emergency busses IC Unplanned Lo.es of Required DC Power During Cold Shutdown or Refueling Mode for Greater than 15 Minutes 6.1.1.b ennttu The following conditions exist:

Unplanned Loss of ALL safety related DC Power indicated by < 105 VDC bus voltage indications for DC Panels 1(2)FA, B, C, D AND Failure to restore power to at least one required DC bus within 15 minutes from the time of the loss IC AC power capability to essential busses reduced to a single power source ALERT for greater than 15 minutes such that any additional single failure would result in station blackout 6.1.2.a mma The following conditions exist:

Loss of Power to 101 and 201 Safeguard Transformers for >15 m/nutes AND l

Only One 4 KV emergency bus powered from a Single Onsite Power Source due to the Loss of: Three of Four Division Diesel Generators, D/G l

Output Breakers, or 4 KV Emergency Busses as indicated by bus voltage IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Shutdown Or Refueling Mode 6.1.2.b EHDEEE The following conditions exist:

Loss of Power to 101 and 201 Safeguard Transformers AND l

Failure to restore power to at least One 4 KV emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power 5

--- - =

1 1'

lgs EAL Table REV D, November 16,1998 Page 14 of 24 IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential SITE AREA Busses EMERGENCY 6.1.3.a rama The following conditions exist:

Loss of Power to 101 and 201 Safeguard Transformers AND l

Failure to restore power to at least One 4 KV emergency bus within 15 minutes from the time of loss of both offsite and onsite AC IC Loss of All Vital DC Power 6.1.3.b EEEEE Loss of ALL Safety Related DC Powerindicated by < 105 VDC on DC Panels 1(2)FA, B, C, D for > 15 minutes IC Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC GENERAL Power EMERGENCY 6.1.4 Le ahm Prolonged loss of all offsite and onsite AC power as indicated by:

(

)

Loss of Power to 101 and 201 Safeguard Transformers AND l

Failure of ALL Emergency Diesel Generators to supply power to 4 KV emergency busses AND At least one of the following conditions exist:

Restoration of at least One emergency bus within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is NOT likely M

Reactor Water Level cannot be maintained > -161 "

=

M Suppression Pool temperature is greater than the Heat Capacity Temperature Limit (HCTL)

J

=

LGS EAL Table REV D, Novemtr> 16,1998 Page 15 of 24 7N 7.0 Internal Events V

7.1 Technical Specification & Control Room Evacuation CLASSIFICATION EMERGENCY ACTION LEVEL IC Inability to Reach Required Shutdown Within Technical Specification UNUSUAL EVENT -

Limits 7.1.1 ENas 4

Inability to reach required :,nutdown mode within Tech. Spec. LCO required action completion time.

IC Control Room Evacuation Has Been initiated ALERT i

7.1.2 umw_w Entry into SE-1 or SE-6 procedure for Control Room evacuation IC Control Room Evecuation Has Been initiated and Plant Control Cannot Be SITE' AREA Established EMERGENCY 7.1.3 uwim l

I The following conditions exist:

Control room evacuation has been initiated AND Control of the plant cannot be established per SE-1or SE-6 within 15 minutes GENERAL.

None EMERGENCY

lgs EAL Table

' REV D, November 16,1998 i

Page 16 of 24 '

T T-7.0 Internal Events G

7.2 Loss of Decay Heat Removal Capability CLASSIFICATION EMERGENCY ACTION LEVEL UNdSUAL EVENT None IC Inability to Maintain Plant in Cold Shutdown ALERT 7.2.2 EnlE The following conditions exist:

'~

Loss of all decay heat removal cooling as determined by procedure GP-6.2 1

AND Uncontrolled Temperature increase that either, Exceeds 200 'F e

m Results in temperature rise approaching 200 F IC Complete Loss of Function Needed to Achieve or Maintain Hot Shutoown SITE AREA EMERGENCY-7.2.3 cme Loss of Main Condenser as a heat sink AND Loss of SUPPRESSION POOL heat sink capabilities as evidenced by T-102 legs requiring an Emergency Blowdown AND Either of the following conditions:

]

RPV level < -161 "

m Reactor Power > 4%

GENERAL None EMERGENCY.

v h

/

l-I

lgs EAL Table REV D, November 16,1998 Page 17 of 24

(]

7.0 Internal Events 7.3 Loss of Assessment / Communication Capability CLASSIFICATION EMERGENCY ACTION LEVEL IC Unplanned Loss of Most or All Safety System Annunciation or Indication UNUSUAL EVENT in The Control Room for Greater Than 15 Minutes 7.3.1.a rum Unplanned loss of most or all safety system annunciators (Table 7-1) M l

> wicators (Table 7-2) for > 15 minutes requiring increased surveillance to safely operate the unit (s).

IC Unplanned Loss of All Onsite or Offsite Communications Capabilities 7.3.1.b mem Loss of ALL Onsite communications (Table 7-3) affecting the ability to perform routine operations OR l'

Loss ofTLL Offsite communications (Table 7-3)

IC Unplanned Loss of Most or All Safety System Annunciation or Indication ALERT in Control Room With Either (1) a Significant Transient in Progress, or (2)

Compensatory Non-Alarming Indicators are Unavailable 7.3.2 rme Unplanned loss of most or all safety system annunciators (Table 7-1) M l

indicators (Table 7-2) for > 15 minutes requiring increased surveillance to safely operate the unit (s)

AND EITHER l

A significant plant transient is in progress (Table 7-4) @ the plant monitoring system (PMS) is unavailable.

IC Inability to Monitor a Significant Transient in Progress SITE AREA EMERGENCY 7.3.3 EDME Loss of safety system annunciators (Table 7-1) l AND indicators (Table 7-2)

AND PMS l

AND a significant plant transient is in progress. (Table 7-4)

{

pENERAL None EMERGENCY l

LGS EAL Table REV D, November 16,1998 Page 18 of 24 Table 7-1 Safety System Annunciators CCS.

' Containment isolation R: actor Trip Process Radiation Monitoring Table 7-2 Safety Function Indicators "i

R: actor Power Decay Heat Removal

~ Containment Safety Functions l Table 7-3 Communications Onsite Offsite Sits Phones (Dimension 2000)

X X

PRELUDE System X

X PI:nt Public Address X

Station Radio X

NRC (FTS-2000)

X PA State Police Radio X

i County Police Radio X

Lord Dispatcher Radio X

PECO Dial Network X

able 7-4 Plant Transients SCRAM R cire Runbacks > 25% thermal power Th:rmal power oscillations > 10%

Stuck open relief valve (s)

ECCS injection 7g

' \\.j

LGS EAL Table REV D, November 16,1998 Page 19 of 24 iT

.8.0 External Events L):

8.1 Security Threats i

CLASSIFICATION

' EMERGENCY ACTION LEVEL IC.

Confirmed Security Event Which Indicates a Potential Degradation in the UNUSUAL EVENT Level of Safety of the Plant 8.1.1 r-J Credible sabotage or bomb threat within the Protected Area M

Credible intrusion and attack threat to the Protected Area Attempted intrusion and attack to the Protected Area OR l

AttempUsabotage discovered within the Protected Area M

Hostage / Extortion situation that threatens normal plant operations IC Security Event in a Plant Protected Area ALERT.

{

F 8.1.2 Enam intrusion into plant protected area by a hostile force M

Confirmed bomb, sabotage or sabotage device discovered in the Protected Area IC Security Event in,a Plant Vital Area SITE AREA EMERGENCY 8.1.3 umm intrusion into plant Vital area by a hostile force M

Confirmed bomb, sabotage or sabotage device discovered in a Vital Area IC Security Event Resulting in Loss of Ability to Reach and Maintain Cold GENERAL Shutdown EMERGENCY 8.1.4 Loss of physical control of the control room due to security event E-

. Loss of physical control of the remote shutdown capability due to security event

'I I

/

LGS EAL Table REV D, Novemb;r 16.1998 Page 20 of 24 (Vl 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases CLASSIFICATION EMERGENCY ACTION LEVEL IC Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes UNUSUAL EVENT of Detection 8.2.1.a Fire within SE-8 Plant Vital Structures (Table 8-1) which is not extinguished within 15 minutes of control room notification or verification of a control room alarm IC Release of Toxic or Flammable Gasses Deemed Detrimental to Safe Operation of the Plant 8.2.1.b Report or detection of toxic or flammable gases that could enter within the site area boundary in amounts that can affect normal operation of the plant 9.8 Report by Local, County or State Officials for potential evacuation of site personnel based on offsite event l

l IC Natural and Destructive Phenomena Affecting the Protected Area 8.2.1.c Report by plant personnel of an unanticipated explosion within protected area boundary resulting in visible damage to permanent structure or equipment IC Fire or Explosion Affecting the Operability of Plant Safety Systems ALERT Required to Establish or Maintain Safe Shutdown 8.2.2.a Nm The following conditions exist:

Fire or explosion which makes inoperable:

Two or More subsystems of a Safe Shutdown System (Table 8-2) QB Two or More Safe Shutdown Systems QR Plant Vital Structures containing Safe Shutdown Equipment AND l

Safe Shutdown System or Plant Vital Structure is required for the present Operational Condition

,l I

L

LGS EAL Table REV D, Novemb;r 16,1998 Page 21 of 24 IC Release of Toxic or Flammable Gases Within a Facility Structure Which

~

Jeopardizes Operation of Systems Reauired to Maintain Safe Operations or to Establish or Maintain Cold Shutdown j

8.2.2.b F"!*

j Report or detection of toxic gases within Plant Vital Structures (Table 8-1) in concentrations that will be life threatening to plant personnel 9.R Report or detection of flammable gases within Plant Vital Structures (Table 8-1) in concentrations affecting the safe operation of the plant SITE AREA None i

EMERGENCY GENERAL None j

EMERGENCY Tr ble 8-1 Plant Vital Structures O

Reactor Enclosure

'( /

Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network Trble 8-2 Safe Shutdown Systems Diesel Generators 4KV Safeguard Buses ADS HPCI RCIC RHR (All Modes)

Core Spray RHR Service Water ESW SGTS RERS CAC PCIS Control Room Ventilation (V

lgs EAL Table REV D, November 16,1998 Page 22 of 24 E

V).O External Events 8.3 Man-Made Events CLASSIFICATION EMERGENCY ACTION LEVEL IC Natural and Destructive Phenomena Affecting the Protected Area UNUSUAL EVENT 8.3.1.a

""M Vehicle crash within protected area boundary that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant.

8.3.1.b ama Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

IC Destructive Phenomena Affecting the Plant Vital Area ALERT 8.3.2 tauM Vehicle crash affecting Plant Vital Structures (Table 8-1)

OR I

I TurbineElure generated missiles result in any visible structural damage to or penetration of any Plant Vital Structures (Table 8-1)

SITE AREA None EMERGENCY GENERAL None EMERGENCY Table 8-1 Plant Vital Structures Reactor Enclosure Control Enclosure Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network

/~'N U

)

lgs EAL Table REV D, Novemb:r 13.1998 Page 23 of 24 (N

8.0 External Events -

L/

]

8.4 Natural Events-CLASSIFICATION -

EMERGENCY ACTION LEVEL IC Natural and Destructive Phenomena Affecting the Protected Area UNUSUAL EVENT 8.4.1.a mm Earthquake >.005 g as determined by procedure SE-5 8.4.1.b

'-M Report by plant personnel of tornado striking within protected area 9.8 Wind speeds > 75 mph as indicated on site Meteorological data for > 15 minutes 8.4.1.c m hm Assessment by the control room that an event has occurred. (Natural and Destructive Phenomena Affecting the Protected Area)

{

$LERT Natural and Destructive Phenomena Affecting the Plant Vital Area IC 8.4.2.a mm Earthquake >.075 g (Operating Basis Earthquake OBE) as determined by procedure SE-5 5-8.4.2.b me Tomado or wind speeds > 75 mph striking Plant Vital Structures (Table 8-1) 8.4.2.c em Report of any visible structural damage to any Plant Vital Structure (Table 8-1)

SITE AREA None EMERGENCY GENERAL None EMERGENCY Table 8-1 Plant Vital Structures Reactor Enclosure

(

)

Control Enclosure v

~ Turbine Enclosure Diesel Generator Enclosure Spray Pond Pump House / Spray Network

.. ~. -.

lgs EAL Title REV D. Nocmber 16,1998 Page 24 of 24 3

9.0 Other J

9.1 General CLASSIFICATION EMERGENCY ACTION LEVEL IC Other Conditions Existing Which in the Judgement of the Emergency UNUSUAL EVENT Director Warrant Declaration of an Unusual Event 9.1.1 Other conditions exist which in the judgement of the Emergency Director indicate a potential degradation of the level of safety of the plant IC Other Conditions Existing Which in the Judgement of the Emergency ALERT Director Warrant Declaration of an Alert 9.1.2 Ersm Other conditions exist which in the Judgement of the Emergency Director indicate that plant safety systems may be degraded and that increased monitoring of plant functions is warranted.

IC Other Conditions Existing Which in the Judgement of the Emergency SITE AREA Director Warrant Declaration of Site Area Emergency EMERGENCY 9.1.3 Other conditions exist which in the Judgement of the Emergency Director indicate actual or likely major failures of plant functions needed for protection of the public IC Other Conditions Existing Which in the Judgement of the Emergency GENERAL Director Warrant Declaration of General Emergency EMERGENCY 9.1.4

'+ -

Other conditions exist which ir. the Judgement of the Emergency Director indicate: (1) actual or imminent substantial core degradation with potential for loss of containment, or (2) potential for uncontrolled radionuclide releases.

These releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the site boundary O

lgs EAL NUMARC Compirison REV D, Nov:mber 16,1998 Page 1 of 4 pg LGS EAL NUMARC Comparison ij NUMARC LGS COMMENTS / EXCEPTIONS AU1,1 5.1.1.a AU1.2 5.1.1.b AU1.3 N/A LGS does not have telemetered perimeter monitors.

AU1.4 N/A LGS does not use automatic initiation of real time dose assessment.

AU2.1 1.2.1.b i

Unexpected" and " Uncontrolled" added to EAL for clarity.

AU2.2 1.2.1.a Fuel transfer canal is not applicable to LGS.

AU2.3 N/A LGS does not have dry fuel storage capabilities.

AU2.4 5.2.1 AA1.1 5.1.2.a AA1.2 5.1.2.b AA1.3 N/A LGS does not have telemetered perimeter monitors.

AA1.4 N/A LGS does not use automatic initiation of real time dose

>)

assessment t

1.2.2.a AA2.1 4

AA2.2 1.2.2.b AA2.3 1.2.2.c AA2.4 1.2.2.d Fuel transfer canalis not applicable to LGS.

AA3.1 5.2.2.b AA3.2 5.2.2.a AS1.1 5.1.3 AS1.2 N/A LGS does not have teleme'ered perimeter monitors.

AS1.3 5.1.3 4

AS1.4 5.1.3 AG1.1 5.1.4 AG1.2 N/A LGS does not have telemetered perimeter monitors.

AG1.3 5.1.4 AG1.4 5.1.4

('

HU1.1 8.4.1.a

LGS EAL NUMARc Comparison REV o, November 16.1998 l'

Page 2 of 4 1

i NUMARC LGS COMMENTS / EXCEPTIONS p

HU1.2 8.4.1.b HU1.3 8.4.1.c

-l HU1.4 8.3.1.a Wording from Basis added to EAL for clarity.

HU1.5' '

8.2.1.c HU1.6 8.3.1.b HUi.7 8.4.1.b HU2 8.2.1.a j

HU3.1 8.2.1.b y

HU3.2 8.2.1.b HU4.1 8.1.1 A bomb device discovered within Plant Protected Area and outside the Plant Vital Areas is an Alert declaration as j

determined per the site Safeguards Contingency Plan and i

therefore is not included as an Unusual Event in the EAL

.I scheme.

HU4.2 8.1.1 p

HUS 9.1.1 HA1.1 -

8.4.2.a i

HA1.2 8.4.2.b HA1.3 ~

8.4.2.c HA1,4 N/A LGS does not have any additional indications in the Control Room of Natural and Destructive Phenomena Affecting the Plant Vital Area HA1.5 8.3.2 HA1.6 8.3.2 HA1.7 N/A LGS does not have any additional Natural and Destructive Phenomena Affecting the Plant Vital Area HA2 8.2.2.a Wording of EAL changed for clarity. No change in intent.

HA3.1 8.2.2.b HA3.2 8.2.2.b "That will affect" changed to "affecting" for added clarity per basis.

HA4.1 8.1.2 HA4.2 8.1.2 C,/

HA5 7.1.2

LGS EAL NUMARC Comparison REV D, November 16.1998 '

Page 3 of 4 NUMARC LGS COMMEN~i S I EXCEPTIONS

%)

HA6 9.1.2 HS1.1 8.1.3 HS1.2 8.1.3 HS2 7.1.3 HS3 9.1.3 HG1.1 8.1.4 l

HG1.2 8.1.4 HG2 9.1.4 SU1 6.1.1.a SU2 7.1.1 SU3 7.3.1.a Wording of EAL simplified. No deviation in intent.

1 i

SU4.1 1.1.1.b These indicators are only valid in OPCONS [1,2,3]. In 4.1.1 OPCONS (4,5], the first indication of fuel clad degradation would be via release monitors and covered in Effluent Realease and Dose EALs.

(,)

SU4.2 1.1.1.a SU5 2.1.1 There is no pressure boundary leakage EAL. This is included in unidentified leakage.

SU6 7.3.1.b SU7 6.1.1.b SA1 6.1.2.b Wording of EAL simplified. No deviation in intent.

SA2 2.2.2 SA3 7.2.2

" Uncontrolled" moved to provide added clarity.

SA4 7.3.2 Wording of EAL simplified. No deviation in intent.

SAS 6.1.2.a Wording of EAL simplified. No deviation in intent.

SS1 6.1.3.a Wording of EAL simplified. No deviation in intent.

SS2 2.2.3 SS3 6.1.3.b SS4 7.2.3 R._,

l I

LGS EAL NUMARC Comparison REV D, November 16,1998 Page 4of 4

.rh()

NUMARC LGS COMMENTS / EXCEPTIONS SS5 2.1.3 During EAL review and approval process, it was determined that the condition stated in NUMARC NESP-007, SS5,1.a

" Loss of all decay heat removal cooling as determined by (site

-specific) procedure"is not necessary to conclude that the plant condition warrants a Site Area Emergency, Theiefore, that sample NUMARC EAL was not included in this EAL.

SS6 7.3.3 Wording of EAL simplified. No deviation in intent.

SG1 6.1.4 SG2.1 2.2.4 SG2.2 2.2.4 FC.1 3.2-FC.1 FC.2 3.2-FC.2 FC.3 3.2-FC.3 FC.4 3.2-FC.4 FC.5 3.2-FC.5 RC.1.

3.3-RC.1

(]

4.1.2 i

RC.2 3.3-RC.2 Indication of a leak inside the drywell was added to avoid classification for non-accident conditions. (e.g., loss of drywell cooling)

RC.3 3.3-RC.3 RC.4 3.3-RC.4 RC 5 3.3-RC.5 RC.6 3.3-RC.6 PC.1 3.4-PC.1 PC.2 3.4 PC.2 PC.3 3.4-PC.3 PC.4 3.4-PC.4 PC.5 3.4-PC.5 PC.6 3.4-PC.6 O()1

,

ML20195J165

Text

Navigation menu

Search