ML20195J154

Rev D to PBAPS Emergency Preparedness NUMARC Eals

ML20195J154
Person / Time
Site:	Peach Bottom
Issue date:	11/16/1998
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20195J148	List: ML20195J142 ML20195J154 ML20195J165
References
PROC-981116, NUDOCS 9811240208
Download: ML20195J154 (180)

Text

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

l-r l

l-L l

l 1

l I

i t

l l

l l l

! ATTACHMENT 1 l t .

Peach Bottom Atomic Power Station Units 2 and 3 1 Emergency Action Levels (Revised)

I i

1 l

l 1

j.

i i

t 9811240208 981116 F l PDR ADOCK 05000277 Y PDR l-l a

. [-. ,,

y m g - -

-y--

,.e y - --- ,s e

l l-1 A

N i N i

M i

M N

M

~- ,

l PECO t

NUCLEAR i

i PEACH BOTTOM ATOMIC POWER STATION i Emergency Preparedness i.

NUMARC EMERGENCY ACTION LEVELS i

v

) NRC Submittal Copy Revision D i

i 4

i

} November 16,1998 i

)

};

3

,, go 7 p e OF Ontents e ,ny-emma a,o "a a r wm , ;w 91 g O Contents 7 sy

.PJ:i%.:

2dg[p'Q J}And W

(gs:f J MK3 f y-  : eggs Introduction m Jy W:h e

4:16 n}Ll' % , yn9e a ssia ~

Technical Basis Manual EAL Table NUMARC Comparison E

Response to NRC O Request for AdditionalInformation b AVE.4T('" READY INDEX' If4DEXtNG SYSTEM

_ - , - , - - - - - - - - - , - - - -,.----- - - - - - , - , - - - , - , - - - - , -,-----m-- -------- - - - , - - - - - -,- - , , - -- - -

'te i.- ,

f i,&f f ..

F ti g, QAa .+

+v.m.

wr,.*4*E A ef e sa eea -

%. .w.... y 9'4M4

~

M% t wm4i,%.r Mac- se- A va W *We W- et*p+wd 4 pe.nei c , mum e- .#w mw.--s,wmww+-, +

g- -gano--g,,e,,, -h -

EJ'ptym,.a e'-' he v e*sa.,7- S 9% y$ 4fc na-wwe W *=

.4ppee mp n=sv .J-' e++-w *w

4em e w-Mr neeu M-t'a. 'fa.-EM,= + * -

M'me n.& p.4e'm- A

h. Mag 4Tm a aa*W+e-4* +*

• I i - . ~

I pssm m 9%-<.-ro u-==W-*

A .a ser d

• e < As em # *e_ -

96ep A -- mw.-

%q:upap.e.-www.=At eM -e r" m A w e.L  % 4-WaDGE'M' a% sdikJBiWW W- '- es'

Contents / introduction REV D, November 16,1998 Page1of1 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 AdditionalInformation 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 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 Roauest for Additional Information - Violet Tab 5 RAI with PECO Nuclear response to each issue included in document.

PBAPs EAL Technical Basis Manu<.J REV D. November 16,1998 Page 1 of 128 PBAPS EAL Technical Basis Manual t/'i Table of Contents V

Section 1 - Introduction. . .. . ....... ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........2 Section 11 - Acronyms . . . . . . . . . . . . . . . . . . . . . . . . .. .. .. . . . . . . . .4 Section lli - EAL Technical Basis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . .6 1.0 : Reactor Fuel 1.1 Coolant Activity.. . . . . . . . . . . .. .. ........ ..... .. .. . ./

1.2 Irradiated Fuel or New Fuel. . . . . . . . . . . . . . . . .. . . . . . . . ....9 2.0 Reactor Pressure Vessel 2.1 Reactor Water Level . . . . . . .. . .. . . . . . . .. . . . . . . .. 1 9 2.2 Reactor Power... . . . . . . . . . . . . . .. . . . . . . . . ... . .23 3.0 Fission Product Barrier 3.1 Initiating Condition Matrix. .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 29

3.2 FuelClad Barrier Thresholds.. . . . . . . . . . . . . . . . . . . . . . . . . . .. 32 3.3 Reactor Coolant System Barrier Thresholds . . ... . . . . . . . , . . . . . . . . 39 3.4 Primary Containment Barrier Thresholds. . . . . ... .... . .. . . . . . . . ... . 46 3.5 Fission Product Barrier Table. .. . . . . . . .. . . . .. .. 57 4.0 Secondary Containment 4.1 Main Steam Line . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . 59

, 5.0 Radioactivity Release l 5.1 Effluent Relea se and Dose . . . . . . .. ................. . ... . . . .. . .. . .. .. . . ... .. 61 5.2 In-Plant Radiation.. . . . .. .. .. ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 71 l

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 l

8.0 Extemal Events I

8.1 Security Events . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01 8.2 Fire / Explosion and Toxic / Flammable Gases . . . . .. .. .. .106 8.3 Man-Made Events . ... ... . . . . . . . . . . . . . . . . . .. . . . . .. . 113 8.4 Natural Events. .. . . . . . . . . . . . . .. ... .116 9.0 Other 9.1 General .. . . .... .... ..... . . . . . . . . .........................................125 n

. v)

, t I-

PBAPs EAL Tecnrucol BTo Manual REV D. November 16.1998 Page 2 of 126 Section I-Introduction This manual contains the technical basis for the Emergency Action Levels as utilized in ERP-O' 101, Classification of Emergencies. The format and use of this manualis as follows.

1. Heading and Sub Heading There are nine major headings each containing one or more sub-headings. These are 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 6.1 Loss of AC or DC Power 7.0 Internal 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 8.2 Fire / Explosion and Toxic / Flammable Gases 8.3 Man-Made Events

( 8.4 Natural Events l

9.0 Other 9.1 General i

l

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

l PBAPs EAL Technical Basis Manual REV D, November 16.1998 Pags 3 of 128

2. Emergency Classification Level and Number Identification O 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 l classification level. Each EAL is individually and uniquely numbered. No two numbers l are the same.

3. INITIATING CONDITION l The Initiating Condition or IC (as described in NUMARC NESP-007) is contained in this j section. ICs are a predetermined subset of conditions where either the potential exists I for a radiological emergency or such an emergency has occurred. Additionally, ICs are l 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 l The operational condition (OPCON) that the EAL is applicable in is contained here. l There are six OPCONs (1, 2,3,4 and 5 and defueled) that are used. PBAPS 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.

, O OPCON (MODE) n'um amenac MODE SWITCH POSITION Run Startup ms- Shutdown (hot) muuru Shutdown (cold) meurt Refueling mamura 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.

O

PBAPS EAL Technical Basis Manual QEV D, November 16,1998 Page 4 of 12G Section11- Acronyms Altemating Current O

AC -

Automatic Depressurization System ADS -

APRM - Average Power Range Monitor ARI - Altemate Rod Insertion ARM - Area Radiation Monitor ATWS - Anticipated Transient Without Scram BRP - Bureau of Radiation Protection CAC - Containment Atmosphere Control CAD - Containment Atmosphere Dilution CDE - Committed Dose Equivalent CFM - Cubic Feet Per Minute CFR - Code of Federal Regulations CRD - Control Rod Drive CS - Core Spray DBA - Design Basis Accident DC - Direct Current DEI - Dose Equivalent lodine EAL - Emergency Action Level ECCS - Emergency Core Cooling Systems ECW - Emergency Cooling Water EDG - Emergency Diesel Generator Environmental Protection Agency EPA -

Emergency Response Procedure - Common g

ERP-C -

Emergency Service Water W

ESW -

FC -

Fuel Clad (Barrier)

FTS - Federal Telephone System GPM - Gallons Per Minute HCTL - Heat Capacity Temperature Limit HPCI - High Pressure Coolant injection HPSW - High Pressure Service Water IC - Initiating Condition IRM - Intermediate Range Monitor KV - Kilovolt LCO - Limiting Condition for Operation LOCA - Loss of Coolant Accident LPCI - Low Pressure Coolant injection MPH - Miles Per Hour mR/hr - Milli Roentgen Per Hour MSIV - Main Steam isolation Valve NFPB - Normal Full Power Background NPSH - Net Positive Suction Head NRC - Nuclear Regulatory Commission NUMARC - Nuclear Management and Resources Council ODCM - Offsite Dose Calculation Manual OPCON - Operating Condition PBAPS -

Peach Bottom Atomic Power Station PEMA - Pennsylvania Emergency Management Ager4cy PC - Primary Containment (Barrier)

PCIS - Primary Containment Isolation System PSIG - Pounds Square Inch Gauge RC -

Reactor Coolant (Barrier)

_ _ , . . _ . . _ _ _ _ . _ . . _ .. , __..,__.____m_... _.__ . . ,___.__._.,,____._m__. . _ _ . . . _ _ . _ .

t' PBAPS EAL Technical Basia Manual REV D. Hocmber 16,1998 Page 5 of 128

!l ' s - RCIC - Reactor Core Isolation' Cooling RCS - . Reactor Coolant System RHR. - Residual Heat Removal RPS -- . Reactor Protection System -

RPV -

Reactor Pressure Vessel SBGTS - Standby Gas Treatment System ..

SBO --

Station Blackout SJAE -

Steam Jet Air Ejector SRM -

Source Range Monitor SRV - Safety Relief Valve TAF -- Top of Active Fuel

-TPARD --

Total Protective Action Recommendation Dose TRIPS --

Transient Response implementation Plan Procedures

Ci/cc -

Micro Curie Per Cubic Centimeter

Cilgm -

Micro Curie Per Gram L. UFSAR -

Updated Final Safety Analysis Report l VDC -

Volts Direct Current t

7)

!b i

l.

l l

i i

lO 1

4 t

PBAPS EAL Technical B: sis Manual REV D, Novsmber 16,1998 Page 6 of 128 Section lli- EAL Technical Basis 1

1 l

1 0

i i

O

PBAPS EAL Tschnical B: sis Manual REV D, November 16.1998 Page 7 of 128 1.0 Reactor Fuel 1.1 Coolant Activity UNUSUAL EVENT - 1.1.1.a IC Fuel Clad Degradation EAL Reactor Coolant activity > 4 pCl/gm Dose Equivalent lodine 131 OPCON 'A BASIS Coolant activity in excess of Technical Specifications (> 4 Cl/gm) is considered to be a precursor of more serious problems. The Technical Specification limit reflects a degrading or degraded core condition. This level is chosen to be above any possible short duration spikes under normal conditions. An Unusual Event is only warranted when actual fuel clad damage is the cause of the elevated coolant sample (as determined by laboratory confirmation).

However, fuel clad damage should be assumed to be the cause of elevated Reactor Coolant "O. 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 Dose Equivalent lodine 131 per Fission Product Barrier Table.

DEVIATION None REFERENCES Technical Specification Section 3.6.8 NUMARC NESP-007, SU4.2 O

r PBAPD EAL Technical Basis Manual REV o. November 16.1998 Page 8 of 128 1.0 Reactor Fuel 1.1 Coolant Activity 0

UNUSUAL EVENT - 1.1.1.b l

IC Fuel Clad Degradation EAL I l

SJAE Radiation (Offgas Monitor) > 2.5x10' mR/hr l

l I

OPCON ENE I l

BASIS l l

The steam jet air ejector (Offgas) radiation monitor RR-2(3)-17-152 in the Control Room would l be one of the first indicators of a degrading core. The high-high alarm is set at the Technical Specification limit of 2.5x10 mR/hr. This instrument takes a sample before the recombiner. l 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 l 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 Peach Bottom, 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 monitor's which are covered by EAL on Effluent Release and Dose.

REFERENCES Technical Specifications Section 3.8.C.7.a NUMARC NESP-007, SU4.1 ,

l l

l l

l 9

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

I-PBAPS EAL Technical Basis Mr.nual REV o, November 16,1998 Page 9 of 128 .,

!e 1.0 Reactor Fuel 1.2: Irradiated Fuel or New Fuel UNUSUAL EVENT - 1.2.1.a )

IC Unexpected increase in Plant Radiation or Airborne Concentration.

EAL r 1 I

Uncontrolled water level decrease in the spent fuel pool with all irradiated fuel assemblies remaining covered by water l

" " + -

OPCON l BASIS

-This event tends to have a long lead time relative to potential for radiological release outside i the site boundary, thus impact to public health and safety is very low. ]

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 EWR all occurring since 1984, explicit coverage of )

these types of events via this E/.L is appmpriate given their potential for increased doses to plant staff. Classification as an Unusuc! 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

(-

l REFERENCES NUMARC NESP-007, AU2.2 Technical Specifications t

[O

. . ~ .

PBAPS EAL Technical Basis Manual REV D. November 16,1998 Page 10 of 128 1.0 Reactor Fuel O

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

EAL Unexpected Skimmer Surge Tank low level alarm AND Visual observation of an uncontrolled water level decrease below the fuel pool skimmer surge tank inlet l l

"":>~

OPCON 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 Cooling and Cleanup System Trouble Alarm. This Control Room alarm directs an operator to be dispatched to a >

local alarm panel which will identify the Skimmer Surge Tank low level 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.

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

l DEVIATION 1

i None REFERENCES l NUMARC NESP-007, AU2.1

._..n..-- %_.- .m._.. . _ _ _ . _ _ . - _ . . _ _ _ . . . . _ _ . _ _ . . . _ . _ _ _ _ . . . _ _ _ . _ _ _ _ _ _ . _

PBAPS EAL Technical Basis MInual REV D, Novemtar 16,1998 Page 11 of 128 1.0 Reactor Fuel 1.2 Irradiated Fuel or New Fuel ALERT - 1.2.2.a 3

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 <

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. 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 refuel floor Area Radiation Monitors or taken during a qualified radiological survey. Table 1-1 monitors are as follows:

Table 1-1 Refuel Floor ARMS 3-7 (7-9) Steam Separator Pool 3-8 (7-10) Refuel Slot 3-9 (7-11) Fuel Pool 3-10 (7-12) Refueling Bridge

. 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:

4 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.

PBAPs EAL Technical Basis Manual REV D. November 16.1998 Page 12 of 128 Licensees may wish to reevaluate whether Emergency Action Levels specified in th emergency plan and procedures governing decayed fuel handling activities appropriately focus on concern for onsite workers and Kr-85 releases in areas decayed spent fuel accidents could occur, for example, the spent fuel pool workin 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.

' Offsite doses during these accidents 'vould 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 None REFERENCES l

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

)

Plant Accidents I

i O

PBAPs EAL Technical Bisa Manual REv D, November 16,1998 Pagi 13 of 128 n 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 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.

[ \

U Studies of the loss of fuel pool water levelindicate 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 appropnately 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.

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

(3 V

l

PBAPS EAL Tgchnical Basis Manual REV D. November 16,1998 Page 14 of 128 DEVIATION None REFERENCES NUMARC NESP-007, AA2.2 O

O

l l PBAPs EAL Technical Basis Manual REV o. Novtmber 16.1998 Page 15 of 128 m 1.0 Reactor Fuel lU 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 Water Level < 458 " above RPVinstrument zero for the Reactor Refueling Cavity that will result in Irradiated Fuel uncovering OPCON r-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.

(.

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 goveming 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.

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

q l The value 458" above RPV instrument zero is the Tech. Spec. Limit and an uncontrolled leve'

!(/ decrease that would uncover irradiated fuel is an indicator of a decrease in the level of safety

of the plant.

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

l

PBAPS EAL Technd Bacis ManaJf REV D.Noonber 16,1998 Page 16 of 128 DEVIATION g

None-REFERENCES NUMARC NESP-007, AA2.3 l Tech Spec 3.9.6 l

l 9

f i

i i

i 1

O

PBAPs EAL Technic 51 Baris Manual REV D. November 16,1998 Pag 317 of 120 l l

1.0 Reactor Fuel p

V 1.2 Irradiated Fuel or New Fuel ,

I l ALERT - 1.2.2.d .

l f IC Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the i i

Uncovering of Irradiated Fuel Outside the Reactor Vessel l

f EAL 1

Water Level < 232 ft 3 inches plant elevation for the Spent Fuel Pool that will result in irradiated Fuel uncovering l

=m OPCON BASIS l 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

/'N EAL.

Gl 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 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 232 ft 3 inches plant elevation is the Tech. Spec. Limit and an uncontrolled level (m~) of the plant.

decrease that would uncover irradiated fuel is an indicator of a decrease in the level of safety Thus, an Alert Classification for this event is appropriate. Escalation, if appropriate, would occur via Effluent Release, in-plant radiation, or Emergency Director Judgement.

PBAPS EAL Technical 81 sis Manual REV D, Novsmb;r 16.1996 Page 18 of 128 DEVIATION O None REFERENCES NUMARC NESP-007, AA2.4 l Tech Spec 3.7.7 O

O

PBAPs EAL Techncal Basis Manual REV o, Nov1mber 16,1998 Page 19 of 128 O 2.0 Reactor Pressure Vessel N) 2.1 Reactor Pressure Boundary UNUSUAL EVENT - 2.1.1 IC Reactor Coolant System Leakage EAL 4

The following conditions exist:

Unidentified Primary System Leakage > 10 ppm into the Drywell

.O_B Identified Primary System Leakage > 25 ppm into the Drywell OPCON N=

BASIS Utilizing the leak before break methodology, it is anticipated that there wiu be indication (s) of minor reactor coolant system boundary integrity loss prior to this fault escalating to a major leak or rupture. Detection of low levels of leakage while pressurized is utilized to monitor for O 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 of the plant. The value of 10 gpm unidentified leakage is significantly higher than the expected pressurized leak rate from the reactor coolant system. The 10 gpm value for the unidentified pressure boundary leakcoe was selected as it is twice the Technical Specification value, indicating an increaso 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 size such that any sudden or gradual increase in leak rate would be identified. A slow power 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.

b" This event will escalate to an Alert based upon high Drywell pressure per Fission Product Barrier Table.

PBAPs EAL Technical Basis Manual REV o November 16,1998 Page 20 of 128 DEVIATION NUMARC Example EAL SUS.1.a identifies pressure boundary leakage. There is no Peach O

Bottom EAL listed for pressure boundary leakage specifically since it is a subset of unidentified leakage. Peach Bottom Tech. Specs. requires a shutdown if any pressure boundary leakage is found.

REFERENCES NUMARC NESP-007, SUS Technical Specifications 3.6.C.1 T-101, RPV Control T-102, Primary Containment Control O

l l

l l

l l

\

1 O

I

.._m _ ._ .-. ~ . _ _ _ . _ . . . _ _ _ _ . . - _ ~.-.._ .- _ _.m _ . -_ . _ . _ _ _

0 PBAPS EAL Technical B: sis Manual REV D, November 16,1998 Page 21 of 128

-]

2.0 Reactor Pressure Vessel 2.1 Reactor Water Level SITE AREA EMERGENCY - 2.1.3 I

IC Loss of Water Levelin the Reactor Vessel That Has or Will Uncover fuelin the Reactor

. Vessel EAL RPV level < -172 "

OPCON e j BASIS

.l The indicator for " core is or will be uncovered" is Reactor Pressure Vessel Water level below 1 the Top of Active Fuel (TAF) -172 inches as indicated on RPV Fuel Zone LevelInstruments Ll-2(3)-02-3-091 or Li-2(3)-02-3-113. Core submergence ensures adequate core cooling. When

RPV level' decreases below the top of active fuel the ability to remove the decay heat l generated from the nuclear fuel becomes suspect and the Fuel Clad Fission Product barrier j can no longer be considered intact. Sustained partial or total core uncovery can result in the release of a significant amount of fission products to the reactor coolant. l

[.

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 concems 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:

l 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 l

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 ,

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

i 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.

I

PBAPs EAL Techrucal Basb Manual REV D, November 16,1998 Page 22 of 928 DEVIATION l

During EAL review and approval process, it was determined that the condition stated in i

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 9

O

PBAPS EAL Technical Basis Manual )

REV D, Nonmber 16.1998 Prgi 23 of 128 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 rmm BASIS Entry into this EAL is based on a reactor parameter actually exceeding a RPS setpoint and the l reactor is not brought to a suberitial condition and maintained at that state with automatic RPS O. 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 subcriticak 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 (suberitial) 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 subcritical 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, e

I Suberitical conditions can be assured even with the most reactive control rod fully withdrawn from the' core if the remaining 184 control rods fully insert. Any other control rod pattem ,

. resulting from partial contro! rod insertion must be carefully analyzed and/or monitored to detect the possibility of re-criticality or local criticality.

PBAPs EAL Technical Basis Manual REV o. November 16,1998 Page 24 of 128 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 prev 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 O

i O

l

__ _ . .. . . _ _ . _ _ _ __ _~ _ . . _ _ _ _ _ _ _ _ _ . . .

I PBAPs EAL Tcchnical Basis Manual REV D, November 16,1998 Page 25 of 128 i I

i p 2.0 Reactor Pressure Vessel C/ .

2.2 Reactor Power l

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 1'

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 < 3%

OPCON rm m 1 I

BASIS A valid automatic and/or manual scram signal is present as indicted by control room indications i l b and/or alarms and APRM indication is greater than 3% power. The Reactor Protection System l (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 l 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 (hydraulic failure).

A failure of the Reactor Protection System to shut down the reactor (as indicated by reactor power remaining above 3%) 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 3% power coincident with loss of scram capability

as the initiating condition for various plant responses to ATWS events. With Reactor Power less than 3% the heat being generated in the core can be removed from the RPV and j containment while actions are taken to bring the reactor suberitical.

A manual scram is defined as any set of actions by the reactor operator (s) at the reactor l 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 l

procedure is considered a manual scram action, although the mode switch in shutdown will generate a scram signal.

i p While the plant is being shutdown, significant heat is being generated in the core and the heat 1

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

i PBAPs EAL Technical Basis Mcnual REV o. Woonber 16,1998 Page 26 o(128 After an ATWS event, there is a potential that the Main Steam isolation Valves (MSlV) will remain open. There is additional guidance in the Trip procedures to ensure that the MSIVs remain open even if RPV level is 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 Torus.

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 3% 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 Torus Temperature exceeding 180 degrees F.

DEVIATION None REFERENCES NUMARC NESP-007, SS2 T-100, Scram T-101, RPV Control, RC/L-2 T-117, Level / Power Control O

PBAPS EAL T'chnical B sis Manual REV D, Novemtnf 16,1998 l Pay 27 of 128 o 2.0 Reactor Pressure Vessel i j 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 < 3%

AND Torus Temperature is > 180 degrees F OPCON -

() BASIS V

A valid automatic or manual scram signal is present as indicted by control room indications and/or alarms and APRM indication is greater than 3% power. In addition, control room instrumentation indicates that Torus temperature is > 180 F.

Failure of all automatic and manual trip functions coincident with a high Torus 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 Torus. Operation with sustained high Torus temperatures may render these systems inoperable due to Net Positive Suction Head (NPSH) considerations.

The RPV Control Trip Procedure establishes 3% 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 Torus temperature reaching 110 F) would bring the reactor to

< 3 % power before Torus temperature approaches the heat capacity temperature limit curve limitations.

Under ATWS conditions, it is important to assure continuous, stable steam condensation capability. An elevated Torus temperature of 180 F would result in unstable steam p

t condensation should rapid reactor depressurization occur (ADS activation). 180 F is the

/ TORUS Heat Capacity Temperature Limit (HCTL). Maintaining the ability to condense steam will preclude the pressurization of the containment and prevent possible containment failure.

I l Containment over-pressurization, which would be an eventual result of sustained operation l with heat being added to the containment and Torus temperature above 180 F would result in l

~~ - ------- - _ _ _ __ _

PBAPS EAL Technical Basis Manual REV D. November 16,1998 Paga 28 of 128 I 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 fuel.

l DEVIATION None REFERENCES NUMARC NESP-007, SG2.1, SG2.2 T-100. Scram T-101, RPV Control T-117, Level / Power Control, RC/L-2 O

O I

l

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

PBAPs EAL Technical Basis Manual l REV D. Novtmber 16,1998 l Page 29 of 128 1

fn ' 3.0 Fission Product Barrier '

L) 3.1 Initiating Condition Matrix l

Determine which combination of the three barriers (Fuel Clad, Reactor Coolant, Primary Containment) are lost or have a potentialloss 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this IMMINENT LOSS situation, use judgement and classify as if the thresholds are exceeded.

UNUSUAL EVENT I

IC ANY Loss or ANY Potential Loss of Containment EAL 1 l

ANY Loss or ANY Potential Loss of Containment ALERT-IC ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS

,n

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

I l

PBAPS EAL Technical Bas'a Manual REV D. November 16.1998 Page 30 of 128 GENERAL EMERGENCY (C Loss of ANY Two Barriers O

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

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

l

. 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 O

assess how far present conditions are from General Emergency. For example, l

if the Fuel Clad barrier and RCS barrier " Loss" EALs existed, this would indicate I to the Emergency Director that, in addition to offsite dose assessments, must I focus on continual assessments of radioactive inventory and containment I integrity. If, on the other hand, both Fuel Clad barrier and RCS barrier " Potential l

Loss" EALs existed, the Emergency Director would have more assurance that l 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 l

EAL Reference Table states that IMMINENT (i.e., within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) Loss or Potential l 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 all reactor coolant system piping up to the isolation valves. i

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

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

PBAPS EAL Technical BIsis Manual REV D. November 16.1998 Page 31 of 126 If a " Loss"' condition is satisfied, the '" Potential Loss" category can be considered 6.

l 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 concern during an accident. If this condition exists during normal j power operations, it will be an active Technical Specification Action Statement. ,

j However, during accident conditions, this will represent a breach of containment. 1 ll DEVIATION None l l REFERENCES 1 l

l NUMARC NESP-007, Recognition Category F, Table 3 I

l l

l i-I 1

I i

f a

i

( .

PBAPs EAL Technical Ba1ps Manual REV o. November 16.1998 Page 32 of 128 3.0 Fission Product Barrier 3.2 Fuel Clad Barrier FC.1 Primary Coolant Activity Level EAL LOSS Reactor Coolant activity > 300 pCugm Dose Equivalent fodine 131 POTENTIAL LOSS Not Applicable OPCON cma BASIS A reactor coolant sample activity of greater than > 300 Ci/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 2.6% clad damage. 2.6%

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)

(Rea Guide 1.109) LNUREG-1228) 1-131 4.39E-3 85000 1-132 5.23E-5 120000 1-133 1.04E-3 170000 1-134 1.37E-5 190000 l-135 2.14E-4 150000 Time After Shutdown (T = 0) Ratios Rm = 120000/85000(1-131) = 1.41(I-131)

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

Rm = 190000/85000(1-131) = 2.24(I-131)

R$3s = 150000/85000(1-131) = 1.76(I-131)

Equation for Dose Equivalent lodine (Deli 33)

A in DF in + (R m) A in DF m + (R m) A in DF m + (R m) A in DF m + (R m) A in -

DF2 in =

DF m O

Solve for At31 assuming Deli 3, = 300 Ci/cc

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

_ _ . . _ . . . _ . _ . . _ . _ . . . . _ . . _ _ . . _ _ _ . _ . . . . . . _ . . . .__ . . . _.__ .~ .._. .._...

PBAPS EAL Techrical Basis Manual REV D, November 16,1998 Page 33 of 128

.i 4 $

300 = A ni .39E-3+1.41 A ni .23E-5+2.00 A nJ.04E-3+2.24 A nd.37E-$+1.76 A 4.39E - 3 300 , - Ani 4.39E - 3 Therefore: A 3i = 189 Ci/cc l-131 3

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

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

= 1.96E6 Cl l 1

l Reactor Water Volume = 2.67E8 cc (ERP-C-1410) J

' Total' Coolant Activity 1-131 = (A 33) (Rx Water Volume) (Ci/ Ci) 3 l

= (189 Ci/cc)(2.67E8cc)(1.0E-6Ci/ Ci)  :!

= 5.05E4Ci '

Percent Clad Damage = Total Coolant Activity / Clad Activity 1-131

= (5.05E4) / (1.96E6) .

= 2.6%

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

! DEVIATION None REFERENCES-NUMARC NESP-007, FC EAL #1

NUREG 1228 - Source Term Estimation During incident Response to Severe Nuclear Power i Plant Accidents, Table 2.2 Reg. Guide 1.109, Table E-9 l ERP-C-1410 l

O r , , . , .,.m, , _ . _ _ , _ _ ,_ _w, .v., __ .

PBAPs EAL Technical Basis Marwal REV D. November 16.1998 Page 34 of 128 3.0 Fission Product Barrier 3.2 Fuel Clad Barrier FC.2 Reactor Vessel Water Level EAL LOSS RPV level < 226 " l POTENTIAL LOSS RPV level < -172 "

l 1

OPCON ENE  !

BASIS The " Loss" EAL -226 " 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 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, -226 " as read on instruments LI-2(3)-02-3-091 or Ll2(3)-02 3-113, 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 RPV. The value of -226" was chosen as it represents 2/3 core height.

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

.. _ .. _ . m .. . m._. - _ _ _ , _ . _ - _ . - . _ . . _ _ ._ . -m._ . _ . . . . . _ . . _ . . _

PBAPS EAL Technical Basis Manual I

REV D. November 18.1998 Page 35 of 128 3.0 Fission Product Barrier

'A 3.2 - Fuel Clad Barrier ,

FC.3 Drywell Radiation Monitoring EAL LOSS Drywell Rad Monitor reading > 8x1/ R/hr POTENTIAL LOSS l-Not Applicable OPCON mm BASIS The 8x10' R/hr reading on a containment h' igh range radiation monitor RI-8(9)103A,B C,D is a value which indicates the release of reactor coolant, with elevated activity indicative of_ fuel 1

damage, into the drywell. The reading was calculated assuming an instantaneous release and dispersal of the Reactor Coolant noble gas and iodine inventory into the Primary Containment

-(direct reading not shine) at a coolant concentration of 300 Ci/gm Dose Equivalent lodine O' 131. This calculation is as follows:

l Using Curve 3 [1%) of ERP-C-1410 Time after Shutdown = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (more conservative due to lower value for EAL) 1% fuel clad damage the dose rate = 30,000 R/hr Extrapolating to 2.6% -

(30,000 R/hr/1%)(2.6) = 78,000 R/hr This is rounded conservatively to 80,000 R/hr for human factors considerations 2.6% 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 lodine spiking) allowed within technical specifications and are i .therefore indicative of fuel damage. This value is h!gher than that specified for RCS barrier l' Loss EAL #3.' Thus, this EAL indicates a loss of both Fuel Clad barrier and RCS barrier.

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

1 i

, _ _ . . ~ , _

PBAPs EAL Techtred Basis Manual REV o, November 16,1998 Page 36 of 128 DEVIATION None O

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 l

l 1

l i

\ O I

l l

l O

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

PBAPS EAL Tcchnical Basis Manual REV D, Nov:mb:r 16.1998 l P:-ge 37 of 120 e 3.0 Fission Product Barrier i l

! 3.2 Fuel Clad Barrier )

j FC.4 Other indications l l l

EAL l 4

l l

LOSS

!. Not Applicable l

POTENTIAL LOSS l Not Applicable OPCON N l BASIS There are no other indications at PBAPS for loss of the Fuel Clad Barrier. i I,

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

l.

I O

l

.- - . . .. . ~ - . .

I PBAPs EAL Technical Basis Manua!

REV D. November 16,1998 l

Page 38 of 128 j

! 3.0 Fission Product Barrier 3.2 Fuel Clad Barrier FC.S Emergency Director Judgement EAL t

Any condition in the judgement of the Emergency Director that indicates Loss or Potential l

! Loss of the FUEL CLAD barrier OPCON w ooa 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 l

l information.)

DEVIATION I

None REFERENCES NUMARC NESP-007, FC EAL #5 9

- - ..- - - .+. . .-....-..__--.n.. - - - - - _ _ . - , .-

PBAPS EAL Technical Bms Manual J- REV D, November 16,1998 j Page 39 of 128 l l

, f- 3.0 Fission Product Barrier - I

1

' 3.3 Reactor Coolant System Barrier

) RC.1 RCS Leak Rate EAL-i LOSS 1 Not Applicable l

r POTENTIAL LOSS i RCS leakage >50 ppm I _

E j Unisolable primary system leakage outside drywell as indicated by a T-103 Temperature i Action Level is exceeded in ONE area requiring a SCRAM 3

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

]

BASIS I I .s l

Potential loss of RCS based on primary system leakage outside the drywell is determined from -  !

3 T-103 area ' temperatures or radiation levels. TRIP guidance stipulates that when the l 4

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. lf 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 anJ there may be a higher percentage of short lived radioisotopes in any release.

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

P8APS EAL TCchnical Basis GAanual REV D. November 10,1998 Page 40 of 12G DEVIATION None-REFERENCES 1 NUMARC NESP-007, RC EAL #1 PC EAl. #2 T-103 Secondary Containment Control O

O

PBAPs EAL Technical Basis Manual REV D, November 16,1998 Page 41 of 128

(~'T 3.0 Fission Product Barrier Q) 3.3 Reactor Coolant System Barrier RC.2 Drywell Pressure EAL LOSS Drywell Pressure > 2.0 ps/g AND Indication of a leak inside drywel!

POTENTIAL LOSS Not Applicable OPCON Nm BASIS The 2.0 psig drywell pressure is based on the drywell high pressure alarm set point and

( indicates a LOCA.

If drywell pressure exceeds 2 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 Contro!

,rh V

PBAPs EALTechnical Basis Manual i REV o, November 16,1998 Page 42 of 128 l

3.0 Fission Product Barrier 3.3 Reactor Coolant System Barrier O\

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 instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with concentrations corresponding to 0.001% Total isotopic Distribution (TlD) into the drywell atmosphere.

Using attachment 5 of ERP-C-1410, Curve 6 Time after Shutdown = 0.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 0.001% TID = 17 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

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

PBAPs EAL Technical Basis Mant,at REV o, November 16,1998 Page 43 of 128 3.0 Fission Product Barrier 3.3 Reactor Coolant System Barrier RC.4 Reactor Vessel Water Level EAL LOSS RPV level < -172 "

POTENTIAL LOSS Not Applicable OPCON. = = = -

BASIS This " Loss" EAL is the same as " Potential Loss" Fuel Clad Barrier EAL #2. The 172 " 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.

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

I O

V l

l l

PBAPs EAL Technical 8xis Manual

," EV o. November 16,1998 Page 44 of 128 3.0 Fission Product Barrier 3.3 Reactor Coolant System Barrier O l l

RC.5 Other Indications \

EAL LOSS Not Applicable POTENTIAL LOSS RPV level cannot be determined OPCON I"=

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 level is 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.

l Operator attention should be given to the possibility that under depressurized conditions, there j 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 levelindications may be a result of degassification of the coolant contained in the level instrumentation.

DEVIATION None REFERENCES l NUMARC NESP-007, FC EAL #4 and RC EAL #5 T-101, RPV Control, RC/L-1 T-112, Rapid Depressurization T-117, Level / Power Control T-116, RPV Flooding

Pb. CS EAL Technical Basis Manual REV D, November 16,1996 Page 45 of 128 3.0 Fission Product Barrier 3.3 Reactor Coolant System Barrier RC.6 Emergency Director Judgement EAL-Any condition in the judgement of the Emergency Director that indicates Loss or Potential Loss of the RCS barrier

.OPCON mm 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 REFERENCES NUMARC NESP-007, RCS EAL #6 -

O

PBAPs EAL Technicd Basis Manual REV D, November 16.1998 Page 46 of 128 3.0 Fission Product Barrier 3.4 Primary Containment Barrier O

PC.1 Drywell Pressure EAL LOSS Rapid, unexplained decrease in Drywell Pressure following initial increase 8 l l

Drywell pressure response not consistent with LOCA conditions

/

l POTENTIAL LOSS ,

Drywell Pressure > 49 ps/g and increasing 98 Drywell Hydrogen > 6% AND Drywell Oxygen > 5%

OPCON Nw 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 49 pc/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 ON-110, Loss of Primary Containment j T-101, RPV Control T-102, Primary Containment Control T-103, Secondary Containment Control O

PBAPs EAL Tcchnical Basis Manual RE ' D, Nov:mMr 16,1998 Page 47 of 128 q 3.0 Fission Product Barrier b 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 M

Intentional venting per T-200 is required M

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

Unisolable primary system leakage outside drywell as indicated by a T-103 Radiation Action Levelis exceeded in ONE area requiring a SCRAM POTENTIAL LOSS

( Not Applicable L]J OPCON EEDusE 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 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 Level isolation levels, if the temperatures continue to rise to the Temperature Action Levels it is g) 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.

~

PBAPS EAL Tcchnic 1 B: sis Manual REV D, November 16.1998 l Page 48 of 128 DEVIATlON None O!

REFERENCES NUMARC NESP-007, RCS EAL #1, PC EAL #2 T-103 Secondary Containment Control T-104, Radioactivity Re{ ease Control O

I O

1 PBAPs EAL T:chnical Basis Manual REV D. No'r.mber 16,1998 Page 49 of 128

,e m 3.0 Fission Product Barrier e i V

3.4 Primary Containment Barrier  !

1 PC.3 Significant Radioactive Inventory in Containment l EAL l

LOSS Not Applicable POTENTIAL LOSS 1 8

Drywell Rad Monitor reading > 6x10 R/hr I

q OPCON C2aiEE BASIS l l

A containment high range radiation monitor 9RI-8(9)103A,B,C,D reading 6x10' R/hr indicates  !

significant fuel damage, well in excess of that required for the loss of the RCS and Fuel Clad. j fm As stated in Section 3.8 of NUMARC/NESP-007, a major release of radioactivity requiring i

( offsite protective actions from core damage is not possible unless a major failure of fuel l 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 l 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 20E l 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 i 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (more conservative due to lower value for EAL) 1% fuel clad damage the dose rate = 30,000 R/hr Extrapolating to 20%

(30,000 R/hr/1%)(20) = 600,000 R/hr There is no " Loss" EAL associated with this item.

]

PBAPS EAL Technic"J Basia Manual REV D, f!ovember 16,1998 Page 50 of 128 DEVIATION None O

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 9

l l

l O

l

\

l

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

1 PBAPS EAL Technical Basis Manual l REV D. November 16.1998 i Page 51 of 128 j 3.0 Fission Product Barrier 3.4 Primary Containment Barrier j PC.4 Reactor Vessel Water Level EAL l

LOSS Not Applicable POTENTIAL LOSS RPV level cannot be restored above -226 "

AND Maximum core uncovery time limit is in the UNSAFE region OPCON cm -

BASIS i O.

V In this EAL, the -226 " 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 further degradation of the clad.

4 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 melt sequence. Whether or not the procedures will be effective should be apparent within the time provided by the maximum core uncovery time limit. The Emergency Director should make 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 I

None

-ve,, e ,, a n.. .--.-n.--s , - - - w - , -m. --, r . - - - - - - - - - , - , - - . -,- - - - -

l PBAPS EAL Technical Basis Manual l REU D, November 16,1998 Page 52 of 128 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 9

I l

O

_ _. . _ . . _ .._m m _. ___..___._m___ __ .. _ _ _. _. __ _ . . _ _ ~ . . _ , _.

l PBAPs EAL Tschnical Basis Manual REV D, November 16,1998 Page s3 of 120 l

l 3.0 Fissicn Product Barrier 3.4 Primary Containment Barrier

! PC.5 Other Indications EAL l

Loss Not Applicable POTENTIAL LOSS l RPV level cannot be determined AND RPV Flooding cannot be established per T-116 OPCON rw - .

BASIS The decision to enter RPV Flooding is made when RPV water level cannot be determined. This

judgement consists of evaluating all plant indications which can influence the ability to maintain 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 Torus 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 turn requires that RPV water level increase. 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 l conservative to assume that water level is actually below the top of active fuel and that both l- the Reactor Coolant System and Fuel Clad Fission Product Barriers are potentially lost.

[U f

' 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 RPV and the Torus in a timely manor can jeopardize the ability of the reactor coolant system to dissipate the decay heat generated.

a -'

k-

~~ - ----- _ _ _ _ _ _ __

P3APS EAL Technical Bisa Manual REv o. Novemtar 16,1998 Page 54 of 128 Eventual clad failure may occur due to overheating of the nuclear fuelif RPV flooding pressure The heat produced from the fuel can cause cannot be established in a timely manner.

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 barr 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 1

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 waterlevel 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 9,

l

PBAPs EAL Ti,chnical Basis M:Enual REV D. Novemtra 16.1998 Page 55 of 128

~

3.0 Fission Product Barrier

(}

's.J 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 uum 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 alsn 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, PC EAL #6 i

i i

p*

k V)

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

PBAPS EAL Technical Basis Manual REV D. November 16,1998 Page 56 of 128 l

This page intentionally left blank 1

0 I

t O

i

. . . . . ~ _ . _ _ . _ , _ ._ . . . _ . _ . _ _ . . _ _ _ ._..__.~_.____....._.___._.___...._.m.____

PBAPS EAL Technical Basia Manual REV D, Novsmber 16,1998 f Page 57 of 128 l

l I

i This page intentionally left blank l

l l

l l

l I

I i

I 1 1

i l

i I-

+

i 4

i l

I h

l.

i. _ ._. _ _ . . _ . - _ _ _ _ _ ._. . . __ __

PBAPS EAL Technical Bass Manual REV D, Novsmber 16,1998 Page 58 of 120 This page intentionally left blank O

l

[

(

i

.r J

O

~,~

3.5 Fission Product Barrier Table Barrier Fuel Clad Reactor Coolant Systi Indicator Loss Potential Loss Loss Poten Reactor Coolant Reactor Coolant N/A N/A N/A

^* Y activity > 300 pCUgm Dose Equivalent lodine 131 RPV Level RPV level < -172 " RPV level < -172 " N/A RPV level < -226 "

RPV Level Unknown N/A N/A N/A RPV lev 21 determine RCS Leak Rate N/A N/A N/A RCS leak

>50 gpm1 E

Unisolabz system 13 outside dl indicated !

Tempera]

Levelis o ONE aret SCRAM l 9

Unisolabl!

system L9 outside d indicated '

Radiatioa is exceed area regi SCRAM e<#

.n - -.

PBAPS EAL Tcchnicil B sis Manurl REV D, Nov mber 16,1998 Page 57 of 128 ri Primary Containment alLos3 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 -226 "

AND Maximum core uncovery time limit is in the UNSAFE region ALERT cannot be N/A RPV level cannot be I determined ANY Loss or ANY Potential Loss of AND EITHER Fuel Clad OR RCS AE I::DTIIDE N s-s1 a u m RPV Flooding cannot be established per CARD T-116 Also Available or Aperture Card iga N/A N/A_

SITE AREA EMERGENCY l

Loss of BOTH Fuel Clad AND RCS

'k 3 OR

' well as Potential Loss of BOTH Fuel Clad by a T-103 AND RCS bra Action OR kceed:d in Potential Loss of EITHER Fuel Clad requiring a OR RCS, and Loss of ANY Additional Barrier

{

e primary aktg3 ywell as by a T-103 Action L; vel GENERAL EMERGENCY ed in ONE iring a Loss of ANY Two Barriers AND Potential Loss of Third Barrier w

a/f v i _ .

c_. --

3.5 Fission Product Barrier Table

- Il'

~

indicator Barrier Loss Fuel Clad Reactor Coolant Systi Potential Loss Loss Poten Drywell Pressure N/A N/A N/A Drywell Pressure l > 2.0 psig AND Indication of a leak l inside drywell l

Drywell Radiation Drywell Rad Monitor N/A N/A Drywell Rad Monitor l reading > 8x10' R/hr reading > 15 R/hr l

Containment isolation N/A N/A N/A N/A t

i l

l l

l l

l i

I l Emergency Director Any condition in the judgement of the Any condition in the judgement of; i , . . Judgement Emergency Director that indicates Loss or Emergency Director that indicate 0 l(

l j Potential Loss of the FUEL CLAD barrier Potential Loss of the RCS barrier l

- s.

I

PBAPS EAL Tcchnic l B: sis Manual REV D, Nov:mber 16,1998 Page 58 of 128 em Primary Containment tial Loss Loss Potential Loss Rapid, unexplained Drywell Pressure decrease in Drywell > 49 psig and Pressure following increasing UNUSUAL EVENT ini

, tialincrease OR ANY Loss or ANY Potential Loss of Drywell7 essure Drywell Hydrogen Containment response not > 6% AND Drywel consistent with LOCA Oxygen > 5%

conditions N/A Drywell Rad Monitor reading > 6x10* R/hr N/A Failure of both valves in any one line to close AND ANY Loss or ANY Potential Loss of downstream pathway EITHER Fuel Clad OR RCS

.m.

to the environment exists areRTURE m CARD Intentional venting per 8 aHaNoon:

T-200 is required SITE AREA EMERGENCY Apedure Card E

Unisolable primary Loss of BOTH Fuel Clad AND RCS system leakage OR outside drywell as Potential Loss of BOTH Fuel Clad indicated by a T-103 AND RCS Temperature Action OR Levelis exceeded in Potential Loss of EITHER Fuel Clad ONE a ea requ,n,ng i a OR RCS, and Loss of ANY Additional Barrier E

Unisolable primary system leakage outside drywell as indicated by a T-103 GENERAL EMERGENCY Radiation Action Level is exceeded in ONE Loss of ANY Two Barriers area requiring a AND SCRAM Potential Loss of Third Barrier tha Any condition in the opinion of the Emergency Lo:s or Director that indicates Loss or Potential Loss of the CONTAINMENT barrier

\

v _

___________________J

PBAPs EAL Tcchnical Basis Manual REV D. Normber 16.1998 Page 59 of 128 4.0 Secondary Containment

{v 4.1 Main Steam Line UNUSUAL EVENT - 4.1.1 IC Fuel Clad Degradation I

EAL Main Steam Line HiHi Radiation (10xNFPB)

OPCON E5m BASIS Main Steam Line High-High Radiation alarm (2(3)-252,A,B,C,D and 2(3)-251,A,B,C,D) > 10 times normal full power background may be indicative of minor fuel cladding degradation and warrants the declaration of an Unusual Event. This level is set to preclude most spurious events including resin intrusion.

, .9 The main steam line high-high radiation condition requires a manual Main Steam Isolation

(" ) Valve closure and a reactor scram. This transient may result in the introduction of fission product gases (previously contained in the gap area) to be suddenly released into the coolant due to the rapid down power transient and subsequent collapse of voids in the coolant.

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 l clad integrity.

This event will escalate to an Alert based on the breach in the main steam 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 Fuel Clad Degradation in those OPCON's. At Peach Bottom, 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 monitors and covered in Effluent Release section.

REFERENCES NUMARC NESP-007, SU4.1

('V ) T-099, Post Scram Recovery T-101, RPV Control l

l l

1

PBAPs EAL Technical Basis Manual l ret / D. November 16,1998 Page 60 of 128 4.0 Secondary Containment O,

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 QR i

Direct report of steam release OPCON Emi!IE BASIS Design basis accident analyses of a Main Steam Line Break outside of secondary containment shows that even if MSIV 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.

DEVIAT!ON None REFERENCES NUMARC NESP-007, RC.1 T-101, RPV Control NUMARC Questions and Answers, June 1993, " Fission Product Barriers #7" l

l l O

l PBAPs EAL Technical Basis Manual l-REV D, Novsmber 16.1998 i Page 61 of 126 5.0 Radioactivity Release l 5.1 Effluent Release and Dose l-UNUSUAL EVENT - 5.1.1.a IC Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds Two Times the Radiological Technical Specifications for 60 Minutes or Longer f

EAL A valid reading on one or more of the following radiation monitors that exceeds TWO TIMES

! the HiHi alarm setpoint value for > 60 minutes:

l Main Stack, Vent Stack, Radwaste Discharge, Service Water Discharge l AND Calculated maximum offsite dose rate using corrputer dose model exceeds 0.114 mrem /hr l TPARD QR 0.342 mrem /hr child thyroid CDE bred on a 60 minute average Note: If the required dose projections cannot be completed within the 60 mlnute period, then the declaration must be made based on the valid sustained monitor reading.

l l

OPCON

! BASIS t

j- The term " Unplanned", as used in this context, includes any release for which a radioactive q l 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 applicable permit. I

' Unplanned releases in excess of 0.114 mrem /hr TPARD or 0.342 mrem /hr CDE 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 concem. Rather it l' is the degradation in plant control implied by the fact that the release was not isolated within 60 l minutes.

i 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 o set point value for the entire 60 minutes. Also, it is intended that the event be declared as l- 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.

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

I 1. An instrument channel check j 2. Indications on related or redundant instruments By direct observation by plant personnel 3.

l 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 L

a potential release may approach Technical Specification (ODCM) limits assuming multiple

PBAPs EAL Techrucat Basis Manual REV D. Nommber 16,1998 Page 62 of 128 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 times the setpoint, sustained for >60 minutes, and assessment by a dose projection indicating an offsite dose rate in excess of two times Technical Specification (ODCM) limits. In the unlikely event that a dose projection cannot be completed within the 60 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) is equal to the thyroid exposure due to iodine. The computerized dose model provides projected TPARD 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 per year (8760 hr/yr.), and then multiplying by a factor of 2 times Technical Specifications [ODCM].

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 Techn, al Specification limit (1500 mrem /yr.) by the number of hours per year (8760 hr/yr.), and then multiplying by a factor of 2 times Technical Specifications [ODCM).

CDE = 2x(Tech Spec Limit)/(hours per ycar)

= 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 Dose Calculation Manual NUMARC Questions and Answers, June 1993, " Abnormal Rad Levels / Radiological Effluents

1. 9" I

O 1

I PBAPs EAL Technical Basis Manuti REV o, November 16,1998 l

Pagi 63 of 128 l

5.0 Radioactivity Release

( V(]

l 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 l EAL l

l Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates exceeding TWO TIMES Tech Specs (Liquid Release ODCM. 3.8.B.1 and Gaseous Release ODCM 3.8.C.1.b) for > 60 minutes l

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.

fm The final integrated dose is very low and is not the primary concem. Rather it is the

( ) degradation in plant control implied by the fact that the release was not isolated within 60 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.

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 l

None

/G REFERENCES

,V NUMARC NESP-007 AU1.2 Offsite Dose Calculation Manual T-104, Radioactivity Release Control i

PBAPs EAL Techni:al Basis Manual REv D, Nontnber 16,1998 Page 64 of 128 5.0 Radioactivity Release 1 5.1 Effluent Release and Dose l ALERT - 5.1.2.a IC Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that 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:

Main Stack, Vent Stack, Radwaste Discharge, Service Water Discharge AND Calculated maximum offsite dose rate exceeds 11.4 mRemlhr TPARD OR 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.

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

l 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.

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 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 Monitor indications are calculated based on the methodology of the site Offsite Dose l

Calculation Manual (ODCM). The HiHi alarm setpoints are set conservatively to indicate when a potential release may approach Technical Specification (ODCM) limits assuming rnultiple release points. Use of this conservative setpoint in establishing a monitor reading will not l cause an inappropriate event classification since this EAL requires the magnitude of the I monitor reading to be two hundred times the setpoint, sustained for >15 minutes, and l

assessment by a dose projection indicating an offsite dose rate in excess of two hundred times 1

Technical Specification (ODCM) limits. In the unlikely event that a dose projection cannot be

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

PBAPS EAL Technical Ba*,is Manual REV D. Nonmber 16.1998 Page 65 of 120

-completed within the 15 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) is equal to

. the thyroid exposure due to iodine. The computerized dose model provides projected TPARD 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 per year (8760 hrlyr.), and then multiplying by a factor of 200 times Technical Specifications [ODCM].

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

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

= 11.4 mrem /hr

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

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

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

= 34.2 mrem /hr f) This event will be escalated to a Site Area Emergency when actual or projected doses are

%/ l determined to exceed 10CFR20 annual average population exposure limits.

' DEVIATION None REFERENCES k

.NUMARC NESP-007 AA1.1 Offsite Dose Calculation Manual NUMARC Questions and Answers, June 1993, " Abnormal Rad Levels / Radiological Effluents

1. 9" t

v

PBAPs EAL TC hnical Basis Manual REV D. Nociember 16. 4998 Page 66 of 128 5.0 Radioactivity Release 5.1 Effluent Release and Dose ALERT - 5.1.2.b IC Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that 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 (Liquid Release ODCM.

3.8.B.1 and Gaseous Release ODCM 3.8.C.1.b) for > 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 safety. The primary concern is the final integrated dose (100 times greater than the Unusual 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 minutes. 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 conditions.

DEVIATION

! None i

REFERENCES NUMARC NESP-007 AA1.2 Offsite Dose Calculation Manual T-104, Radioactivity Release Control

_ - ._ _ ._. .._.._ _.__ ___,_ , ._____ --.__.~ ,. -

_ _.....m_ .m - _ ..._m ._ .. __

PBAPS EAL Technical Basis Manual REV D. Novsmt>Ir 16.1998 l Page 67 of 128 5.0 Radioactivity Release O 5.1 Effluent Release and Dose SITE AREA EMERGENCY - 5.1.3 lC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mR Whole Body or 500 mR Child Thyroid for the Actual or 3 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:

Main Stack 5.84 Ci/cc 1 Vent Stack 2.08E-3 Ci/cc j Torus Vent 203 cpm- q AND Projected offsite dose using computer dose model exceeds 100 mrem TPARD QR 500 mrem child thyroid CDE  ;

iO I

Note: If the required dose projections cannot be completed within the 15 minute period, EE 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, QR Analysis of Field Survey results indicate child thyroid dose commitment of 500 mrem for one hour of inhalation l OPCON =-

i 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:

l

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

! 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.

I An actual or projected dose of 100 mrem Total Protective Action Recommendation Dose f ] (TPARD) is based on the 10 CFR 20 annual average population exposure limit. This value l 7 also provides a desirable gradient (one order of magnitude) between the Site Area Emergency l

[-

PBAPS EAL Technical Bast Manuat REV D. November 16.1998 l Page 68 of 128 The 500 mrem integrated child thyroid dose was and General Emergency classifications.

established in consideration of the 1:5 ratio of the EPA Protective Action Guid 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 duration. The inputs are as follows:

Main Stack Vent Stack Torus Vent E E E Stability Class 6.3 mph 11.4 mph 6.3 mph Wind Speed 45 22' 22 Wind Direction LOCA LOCA LOCA Accident 203 cpm Release Rate 5.84 p/cc 2.08E-3 p/cc Child thyroid dose factors, rather than adult thyroid dose factors, are used for consistency with Pennsyhrania Emergency Management Agency (PEMA) / Bureau of Radiation Protection l (BRP). l 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 l DEVIATION lNone l REFERENCES l

l NUMARC NESP-007, AS1.1, AS1.3 and AS1.4 EPA 400 0

PBAPs EAL Technx:al Batia Minual  ;

REV D, November 16,1998 I l Paga 69 of 128 p 5.0 Radioactivity Release b 5.1 Effluent Release and Dose 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:

Main Stack 58.4 Ci/cc Vent Stack 2.08E-2 Ci/cc Torus Vent 2000 cpm AND Projected offsite dose using computer dose model exceeds 1000 mrem TPARD @

5000 mrem child thyroid CDE p)

(

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.

2 Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 1000 mrem /hr expected to continue for more than one hour, @ Analysis of Field Survey results indicate child thyroid dose commitment of 5000 mrem for one hour of inhalation OPCON i~'"

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 Equivalent (T.EDE) + 4 Day Deposition Dose. Committed Dose Equivalent (CDE) is equal to the thyroid egosure due to lodine. 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 General Emergency. This level constitutes the upper level of the desirable gradient for the

- ________ k

PBAPs EAL Technical Basis Manual REV D. November 16,1998 l l Page 70 of 128 Site Area Emergency. Actual meteorology is specifically identified in the initiating condition since it gives the most accurate dose assessment.

Monitorindications 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 talease duration. The inputs are as follows:

l Main Stack Vent Stack Torus Vent E E E Stability Class 11.4 mph 6.3 mph 6.3 Wind Speed 45 22 22 Wind Direction LOCA LOCA LOCA Accident 58.4 /cc 2.08E-2 p/cc 2.026E+3 cpm Release Rate l

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

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

l l

l O

DEVIATION lNone REFERENCES NUMARC NESP-007, AG1.1, AG1.3 and AG1.4 EPA-400 0

PBAPs EAL Technical B:sia Manual REv D. November 16,1998 Page 71 of 128 n 5.0 Radioactivity Release 5.2 In-Plant Radiation UNUSUAL EVENT - 5.2.1 IC Unexpected increase in Plant Radiation or Airborne Concentration EAL Valid Direct Area Radiation Monitor readings increase by a factor of 1000 over normal

• levels Normal levels 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 g correct.

! i An area monitor reading is considered to be valid when it is verified 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 levels 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 n

_/

PBAPs EAL Technical B: sis Manual REV D. November 16,1998 Page 72 of 120 5.0 Radioactivity Release 5.2 In-Plant Radiation ALERT - 5.2.2.a IC Release of Radioactive Material or increases in Radiation Levels Within the Facility 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-10 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 ve:ue 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 i exposure gi.idelines and limits (i.e.,10 CFR 20), and in doing so, will impede necessary l 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-10. 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 I 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 j or cause of the increased radiation levels and determine if any other IC may be involved. For l example, a dose rate of 15 mR/hr in the control room or hi radiation monitor readings may also I 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.

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

l l

. . _ . 2 _ - .. . _.. . . . .._ _ . . . _ ._. . , ._ _.. . . _ _ . . _ . _ . _ _

1 PBAPs EAL Technicti B: sis Manu*l REV D, Novtmber 16,1998

! Page 73 of 128 i~ rm

-(j This EAL is not meant to apply to increases in drywell 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 l

I NUMARC NESP-007, AA3.2 T-103, Secondary Containment Control SE-1, Plant Shutdown from the Remote Shutdown Panel SE-10, Plant Shutdown from the Attemative Shutdown Panels i i i

i l

(

. 'd i

1 PBAPs EAL Technical Basis Manual REV D, November 16,1998 Page 74 of 128 5.0 Radioactivity Release 5.2 in-Plant Radiation ALERT - 5.2.2.b IC Release of Radioactive Material or increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown EAL Valid Control Room OR Central Alarm Station radiation reading > 15 mR/hr OPCON m- m 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 EAL address radiation levels which would impede operation of systems required to maintain safe operations or to establish or maintain cold shutdown. Radiation levels could be indicated by ARM or radiological survey.

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 safety functions since inputs to radwaste will be isolated with a secondary containment isolation and releases can only be performed manually.

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 Action Plan Requirements", provides that the 15 mR/hr value can be averaged over the 30 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 materials cause significant offsite doses.

O L

PBAPS EAL Technical Baus Manual REV D, Novsmber 16,1998 Page 75 of 128 j i i DEVIATION I i

f l

\ i None l

REFERENCES I l

NUMARC NESP-007 AA3.1 l l

l l

l l

1 l

4 l

I

~ - - -. . .... . . . . . . . . . . - . - . ~ . - - . - . - - . . - - .. .

. - . . . . ~ . . - ~

4 PBAPS EAL Technics! Basis Manual l REV D, November 16,1998 '

l Page 76 cf 128 1

i This page intentionally left blank l

d 1

i

.i k

L l

4 a

i O

O

PBAPs EAL Technicil Basis Manual REV o, November 16.1998 Pag) 77 of 128 g3 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:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer for >15 minutes AND 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

,,h through 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer.

(d Loss of offsite power will cause a reactor scram and a containment isolation. 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 unit). Balance of Plant systems that would i I

assist in plant operations (i.e., condensate pumps, etc.) may be unavailable due the loss of power. l 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.

DEVIATION None

! REFERENCES NUMARC NESP-007, SU1 p SE-11, Station Blackout

\)

PBAPs EAL Technical Basis Manual REV D, November 16.1998 Page 78 of 128 6.0 Loss of Power 6.1 Loss of AC or DC Power UNUSUAL EVENT - 6.1.1.b IC Unplanned Loss of Required DC Power During Cold Shutdown or Refueling Mode for Greater than 15 Minutes EAL The following conditions exist:

Unplanned Loss of ALL safety related DC Power indicated by < 107.5 VDC bus l

voltage indications for DC Panels 2(3)0D21,22,23,24 AND Failure to restore power to at least one required DC bus within 15 minutes from the l J

time of the loss l OPCON e i BASIS 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 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 l

l DC Distribution Panels are as follows:

l Unit 2 Unit 3 l 20D21 30D21 l 20D22 30D22 20D23 30D23 l

20D24 30D24 l

l 107.45 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. The value of 107.5 VDC will be used for human factors concems.

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 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. Routinaly, plants will perform maintenance on a Train re..ited 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.

~4 ,- . . ~ ~ . _ . . . , - - -

-. - - - ~. ~ . . . -. ... - -----.- --...-- .. - ..- -.~. ~ - ---..

PBAPS EAL Technical Basis MInual REV D, Novsmbu 16,1998 Pagt 79 of 128

' ~

DEVIATION ~

I

~ None l

REFERENCES NUMARC NESP-007,'SU7 -

SE-13, Loss of a 125/250 VDC Safety Related Bus i

'l ,

l

'i i

l 1

l j l i 4

1 I

l' l

i i-I.

l 5

j

?

i.- - -

r --. _ - _ _ , . . _ . . . _ _ . _ , _ , _ . , , __ _ , ,_ , , _ _

I PBAPs EAL Vechnic01 Batio Manual REV o, November 16,1998 Page 80 of 126 6.0 Loss of Power 6.1 Loss of AC or DC Power ALERT - 6.1.2.a IC AC power capability to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout EAL The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer 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 I OPCON mm 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 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.

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 Torus cooling may be

' lost due to the need to operate the remaining available RHR pump in other than Torus cooling (e.g., LPCI). As such there is a decrease in the systems 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 ability to cool the nuclear fuel, remove decay heat, and control containment parameters is severely limited. Should equipment l

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

i Escalation of this event would be based on the loss of the remaining Emergency Diesel f

Generator.

DEVIATION None REFERENCES l

NUMARC NESP-007, SAS j SE-11, Station Blackout

- _ . _ _ . . . . _ . _ . . _ --_.__.__.-.-----_.-.._.__-__.___._.-.__..m__ _

PBAPS EAL Technical Basis Manual REV D, November 16.1998 Page 81 of 128 i-

,G ,

6.0 Loss of Power U 6.1 Loss of AC or DC Power i

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

EAL l The following conditions exist:

! Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup I Transformer AND l l Failure to restore power to at least One 4 KV emergency bus within 15 minutes L from the time of loss of both offsite and onsite AC power OPCON ""

BASIS i

L 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 f

l REFERENCES l

l_ NUMARC NESP-007, SA1 l SE 11, Station Blackout

p

>G l

I m - ~ _ , - - r,

PBAPs EAL Technical Basis Manual QEV D. November 16,1998 Page 82 of 128 6.0 Loss of Power 6.1 Loss of AC or DC Power SITE AREA EMERGENCY - 6.1.3.a IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses EAL The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transforrner l AND Failure to restore power to at least One 4 KV emergency bus within 15 m/ mites l

from the time of loss of both offsite and onsite AC OPCON mm 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 systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal, High Pressure Service 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 ol this event would be based on the time that the Emergency Diesel Generator are unavailable.

DEVIATION None REFERENCES NUMARC NESP-007, SS1 SE-11, Station Blackout 9

i h

._ m . ._ m - _m . _ _ . .~ .____._m. ._.. - _. . _ _ _ . . _ . . . . _

PBAPs EAL Technical Basis Minual REV D. Noemtsr 16.1998 Pag) 83 of 128

('o 6.0 Loss of Power 6.1 Loss of AC or DC Power SITE AREA EMERGENCY - 6.1.3.b IC Loss of All Vital DC Power EAL Loss of ALL Safety Related DC Power indicated by < 107.5 VDC on DC Panels 2(3)0D21, 22,23,24 for > 15 minutes 1

OPCON +-

BASIS:  :

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

than 15 minutes) critical plant functions such as RPS Logic, Altemate Rod Insertion, (g

'-'j Emergency Service Water Indication, 4KV Breaker Controls, HPCI, RCIC and RHR pump I

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:

Unit 2 Unit 3 20D21 30D21 20D22 30D22 i 20D23 30D23 I 20D24 30D24 i Loss of all DC Power causes the loss of the following equipment:

. Alternate Rod insertion . ADS

. HPCI . RCIC

. Normal EDG Control . Normal Recirculation Pump Trip

. Containment instrument Gas Compressors ,

. 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.

(

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

PBAPs EAL Technmal Basis Manual REU D. November 16,1999 Page 84 of 128 the heat addition to the containment could result in a loss of the primary containment as a fission product release barrier.

107.45 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This EAL uses 107.5 VDC for human factors concerns. 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 minimum voltage selected when battery sizing is performed.

j DEVIATION None

. REFERENCES 4 NUMARC NESP-007, SS3 T-101, RPV Control T-102, Primary Containment Control SE-11, Station Blackout O

9

\

PBAPs EAL Technical Bisis Manuti REV o, Nov;mber 16.1996 Pace 85 of 128 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 Prolonged loss of all offsite and onsite AC power as indicated by:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer 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 4 KV emergency bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is NOTlikely g . Reactor Water Level cannot be maintained > -172 "

U M

= Torus temperature is greater than the Heat Capacity Temperature Limit (HCTL) l OPCON ~w BASIS Loss of all AC power compromises all plant safety systems requiring electric power including l 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 based 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.

I 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 > -172" (TAF)

2. Torus level low enough to prevent HPCI and/or RCIC steam exhaust line flooding

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

4. Containment pressure < 60 psig, design limit

PBAPs EAL Technical Basis Manual REV o, November ia 1993 Page 86 of 128

6. Drywell temperature

<200 degrees F indefinitely l

<250 degrees F 99 days l

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

<340 degrees F 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> l 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 SE-11, Station Blackout T-101, RPV Control T-102, Primary Containment Control T-104, Radioactivity Release Control O

1 l

i PBAPS EAL Technicd Basis Manual -

REV D, November 16,1998 Page 87 of 128 7.0 Internal Events I

' 7.1 Technical Specification & Control Room Evacuation UNUSUA'L EVENT - 7.1.1 l

.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.  ;

i OPCON 'N 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 l on the circumstances, this may or may not be an emergency or precursor to a more severe  ;

condition. In any case, the initiation of plant shutdown required by the site Technical i 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 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 within 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 j i

None REFERENCES NUMARC NESP-007, SU2 Technical Specifications i

O

PBAPs EAL Technical Basis Manual REV o, November 16,1998 Page 88 of 128 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-10 procedure for Control Room evacuation OPCON BASIS Control Room evacuation requires establishment of plant control from outside the control room (e.g., local control and remote shutdown panel) and support from the Technical Support Center and/or other emergency facilities as necessary. Control Room evacuation represents a serious plant situation since the level of control is not as complete as it would be without evacuation. The establishment of system control outside of the Control Room will bypass many protuctive 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 neb! -007, HA5 SE-10, Alternate Shutdown SE-1 Plant Shutdown from the Remote Shutdown Panel O

l 1

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

PBAPS EAL Technical Basis Manual REV o, Nowmber 16,1998 Page 89 of 128

]

'i 7.0 Internal Events 7.1 Technical Specification & Control Room Evacuation SITE AREA EMERGENCY - 7.1.3 IC Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established EAL s

The following conditions exist:

Control room evacuation has been initiated AND Control of the plant cannot be established per SE-1 or SE-10 within 15 minutes J

OPCON t"- -

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 O, for transfer of controlis based on a reasonable time period for personnel to leave the control 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 may be a threat to plant safety.

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

DEVIATION None REFERENCES NUMARC NESP-007, HS2 SE-10, Altemate Shutdown SE-1, Plant Shutdown from the Remote Shutdown Panel O

4

PBAPs EAL To::hnical Basis Manual REV o, November 16,1999 Page 90 of 128 7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability ALERT - 7.2.2 IC Inability to Maintain Plant in Cold Shutdown EAL The following conditions exist:

Loss of all decay heat removal cooling as determined by procedure GP-12 AND Uncontrolled Temperature increase that either: l f

• Exceeds 212 'F 93

• Results in temperature rise approaching 212 *F OPCON annam BASIS This EAL addresses complete loss of functions required for core cooling during refueling and 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-12, " Core Cooling Procedure," directs the methods for establishing decay heat I removal and guidance on amount of time available to prevent excessive heat-up as a function l of time after shutdown.

l l

" Uncontrolled" means that system temperature increase is not the result of planned actions by

! the plant staff.

1 f The EAL guidance related to uncontrolled temperature rise is necessary to preserve the

! anticipatory philosophy of NUREG-0654 for events starting from temperatures much lower l than the cold shutdown temperature limit.

i This EAL is concerned with the ability to keep the reactor core temperature less than 212 F.

l The criteria of uncontrolled Reactor Coolant temperature increase > 212 F is met as soon as it

! becomes known that sufficient cooling cannot be restored in time to maintain the temperature

< 212 F, regardless of the current temperature. The inability to establish alternate methods of decay heat removal indicates that either altemate methods are unavailable to cool the core in the RPV or when the steam is transferred to the Torus, Torus cooling is unavailable. Loss of Torus cooling will result in a continuing, uncontrolled increase in reactor coolant temperature.

l 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.

. .._.....m....__.m_.._____,...... . . . . , _ _ _ _ . . - - . - _ _ . _ _ _ . - - . . . _ _ . . . _ _ _ . _ _ _ . . . . . . . . . . . . _ . . _ . _ . _ _ . . _ . . . . . .

f-PBAPS EAL Technical Basis Minual' REV D, Novsmbst 10,1998 P De 91 of 128 l

- DEVIATION None-i I

l REFERENCES NUMARC NESP-007, SA3 GP-12, Core Cooling Procedure l l

Technical Specifications 1

I i i

I.

I i

i 4 I

i l

l i i l

i t

t l i

i I

l i~

L

j. 8 i ,

l-.

i l

4 i.

l.

PBAPs EAL Technicaf Ba:is Manuaf l REV D. November 16,1998 l 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 TORUS heat sink capabilities as evidenced by T-102 legs requiring an Emergency Blowdown AND l Either of the following conditions:

. RPV level < -172 "

f l 9.R

= Reactor Power > 3%

OPCON rne l BASIS:

This EAL addresses complete loss of functions, including ultimate heat sink and reactivity l

I 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 TORUS utilizing SRVs. The number of SRVs required to reduce pressure will be dependent upon reactor pressure and power. A low TORUS level would result in Heat Capacity Temperature Limit (HCTL) being exceeded if a full power blowdown occurred at water level in the TORUS. A high TORUS

temperature would result in the TORUS being at the HCTL whereby it can no longer function i as a heat sink. If the TORUS Levelis at a high level the TORUS cannot handle a full power l

blowdown. T-102 requires an Emergency Blowdown before these TORUS conditions are reached to ensure the transfer of the energy to the TORUS. 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 212 F.

1

\ O l

PBAPS EAL Technical Basis Manual REV D, November 16,1998 Page 93 of 128

,A DEVIATION

'\

None REFERENCES NUMARC NESP-007, SS4 T-102, Primary Containment Control, SP/L-8

,l) v O

l

PBAPs EAL Technical Basis Manual REV o, Rovember 16.1998 Page 94 of 128 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability O

UNUSUAL EVENT - 7.3.1.a IC Unplanned Loss of Most or All Safety System Annunciation or Indication in The Control Room for Greater Than 15 Minutes t 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).

OPCON mmm l

BASIS l

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 increated 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 i 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 losses. Unplanned loss of annunciators excludes scheduled maintenance and testing activities. Control Room panels with annunciators and direction for response are included in ON-123, Loss of 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

• 7-2 indicates those indications important for monitoring:

Table 7-2 Safety Function Indicators l

Reactor Power Decay Heat Removal Containment Safety Functions 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 9, l safety systems required for operation.

l This event will be escalated to an Alert if a transient is in progress or if compensatory indications become unavailable, i

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

PBAPS EAL Technicsl Basis Manual F.EV D, Novsmber 16,1998 Pag 195cf128 i

i l -

DEVIATION None REFERENCES i

NUMARC NESP-007, SU3 1 ON 123, Loss of Control Room Annunciators

4. AIT A0004447, EP Self Assessment on Salem Loss of Annunciators 3

a M

e 4-t s

8 l

J 4

't 1

s.

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

PBAPs EAL Technical Base Mand REV D. November 16,1998 Page 96 of 128 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability UNUSUAL EVENT - 7.3.1.b IC Unplanned Loss of All Onsite or Offsite Communications Capabilities EAL Loss of ALL Onsite communications (Table 7-3) affecting the ability to perform routine operations OR l Loss ofTLL 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 Unusual Event declaration is warranted. Table 7-2 is identified as follows:

l Table 7-3 Communications Onsite Offsite Site Phones (GTE System) X X OMN1 System X X Plant Public Address X Station Radio X NRC (FTS-2000) X PA State 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 g

PBAPs EAL Tcchnical Bisis Manual REV o, Nonmber 16,1998 Page 97 of 128 rm 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 Unplanned loss of most or all safety system annunciators (Table 7-1) M 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.

OPCON uuw1 BASIS r This EAL recognizes the difficulty associated in monitoring conditions without normal (3,) 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 seventy above 7.3.1.a in that the Plant Monitoring System (PMS) can not provide compensatory indication, or 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 Function 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:

n V

l PBAPs EAL Techrucal Basis Manual REV o. Movember 16,1998 Page 98 of 128 i

l l Table 7-4 Plant Transients SCRAM i

Recire runbacks > 25% thermal power Thermal power oscillations of 10% or greater l Stuck open relief valves ECCS injection f

l 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-123, l Loss of Control Room Annunciators.

l l

Reportability of Technical Specification imposed shutdowns, or the inability to comply with l 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.

l DEVIATION l

None REFERENCES j NUMARC NESP-007, SA4 ON-123, Loss of Control Room Annunciators i

O

PBAPs EAL Technical Basis Manual REv o, NovImber 16.1998 ,

Page 99 of 120 '

q 7.0 Internal Events V 7.3 Loss of Assessment / Communication Capability SITE AREA EMERGENCY - 7.3.3 l

IC Inability to Monitor a Significant Transient in Progress ,

l EAL Loss of safety system annunciators (Table 7-1) l AND indicators (Table 7-2)

AND PMS l l AND a significant plant transient is in progress. (Table 7-4) l OPCON DNm 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

(]

'v that a significant transient is in progress.

l 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 Table 7-4 significant plant transients include response to automatic or manually initiated actions including:

l Table 7-4 Plant Transients SCRAM 73 Recire runbacks >25% thermal power change t

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

PBAPs EAL Technical B: sis Manual REV o November 16,1998 Page 100 of 128 g

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-123, Loss of Control Room Annunciators.

DEVIATION None REFERENCES NUMARC NESP-007, SS6 ON-123, Loss of Control Room Annunciators O

I 1

i i

e

PBAPs EAL Technical BIsis Manual REV D, Nov1mb;r 16,1998 Page 101 of 128 8.0 External Events 8.1 Security Events UNUSUAL EVENT - 8.1.1 IC Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant EAL Credible sabotage or bomb threat within the Protected Area Credible intrusion and attack threat to the Protected Area Attempted intrusion and attack to the Protected Area M

l Attempted sabotage discovered within the Protected Area Hostage / Extortion situation that threatens normal plant operations OPCON (s

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 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 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.

(

PBAPs EAL Technical Basis Manual REV o. November 16 i998 Page 102 of 128 l

l DEVIATION A bomb device discovered within Plant Protected Area and outside the Plant Vital Areas Alert declaration as determined per the site Safeguards Contingency Pian and therefore is not l

included as an Unusual Event in the EAL scheme.

REFERENCES l

NUMARC NESP-007, HU4.1 and HU4.2

Safeguards Contingency Plan Physical Security Plan 1

i l

l 9

i I

O

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

PBAPS EAL Technical Basis Manual L

REV D, Nonmber 16.1998

Page 103 of 128 p 8.0 External Events 8.1 Security Events l

l ALERT - 8.1.2 IC Security Event in a Plant Protected Area l

EAL l Intrusion into plant protected area by a hostile force l

.Q.8 Confirmed bomb, sabotage or sabotage device discovered in the Protected Area l

OPCON' l

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.

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

1. Intrusion into plant protected area by a hostile force

2. Confirmed bomb, sabotage or sabotage device discovered within the Protected Area i This event will be escalated to a Site Area Emergency based upon a hostile intrusion or act in plant Vital Areas. j DEVIATION None ]

REFERENCES I NUMARC NESP-007, HA4.1 and HA4.2 i Safeguards Contingency Plan l Physical Security Plan l

\

)

i V

PBAPs EAL Technical Basis Manual REV D, November 16,1998 Page 104 of 128 8.0 External Events 8.1 Security Events O

SITE AREA EMERGENCY - 8.1.3 IC Security Event in a Plant Vital Area EAL Intrusion into plant Vital area by a hostile force

_O_B Confirmed bomb, sabotage or sabotage device discovered in a Vital Area OPCON om

• 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 encanger 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

PBAPS EAL Technical Basis Manual REV o. Novemter 16.1998 Pag 1105 of 128 n 8.0 External Events k.) 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 93 Loss of physical control of the remote shutdown capability due to security event OPCON tn" -

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.

7 (b Security events which meet the threshold for declaration of a General Emergency are physical 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 v

i

PBAPs EAL Technical Basis Manuol  ;

REV D, November 16.1998 l Page 106 of 128 l

8.0 External Events 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 ON-114 Plant Vital Structures (Table 8-1) which is not extinguished within 15 minutes of control room notification or verification of a control room alarm l

OPCON l

BASIS The purpose 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 vital 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 contiguous or immediately adjacent to plant vital areas. Verification of the alarm in this context means those actions taken in the control room to determine that the control room

! alarm is not spurious.

, This EAL addresses fires in Plant Vital Structures that house safety systems. These fires may l 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 i

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 j respond and control small fires or to verify that no fire exists. Table 8-1 Plant Vital Structures are as follows:

l l Table 8-1 Plant Vital Structures l Power Block l Diesel Generator Building l Emergency Pump Structure J inner Screen Structure l

Emergency Cooling Tower This event will be escalated to an Alert if the fire damages redundant trains of plant safety l systems required for the current operating condition.

l l

1

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

i PBAPS EAL Technical Basis Manual REV D. Novsenbst 16, t 998  ;

Pags 107 of 128  ;

?

DEVIATION None

, 1

REFERENCES l

1 NUMARC NESP-007, HU2 1

r i

i I

s Y

I 4

1 5

P8APs EAL Techocal Basis Manual REV D, November 16.1998 Page 108 of 1N 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases O

UNUSUAL EVENT - 8.2.1.b IC Release of Toxic or Flammable Gasses Deemed Detrimental to Safe Operation of the 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 9.8 Report by Local, County or State Officials for potential evacuation of site personnel based on offsite event m- o OPCON 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 sk.in 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 O

I l

I l

PBAPs EAL Technical Basia Manual REV D, November 16.1998 Page 109 of 128 t'N 8.0 External Events N.

8.2 Fire / Explosion and Toxic / Flammable Gases UNUSUAL EVENT - 8.2.1.c IC Natural and Destructive Phenomena Affecting the Protected Area EAL Report by plant personnel of an unanticipated explosion within protected area boundary resulting in visible damage to permanent structure or equipment OPCON m- "

BASIS The protected area boundary is typicaliy 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 g the protected area should be considered. As used here, an explosion is a rapid, violent, Q 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 of plant safety systems required for the current operating condition.

DEVIATION None REFERENCES NUMARC NESP-007, HU1.5 th

PBAPs EAL Technical Base Manual REV o, November 16,1998 Page 110 of 128 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases O

ALERT - 8.2.2.a IC Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown EAL The following conditions exist:

Fire or explosion which makes inoperable:

Two orMore subsystems of a Safe Shutdown System (Table 8-2) @ Two or More Safe Shutdown Systems @ 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 concem 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 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 Diesel Generators 4KV Safeguard Buses ADS HPCI RCIC RHR (All Modes)

Core Spray HPSW ESW SBGTS ECW CAC/ CAD 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 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 l 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.

PSAPs EAL Techrucal Basia Manual REV D, Novimttr 16.1998 Pegt 111 of 128

- 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.

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 (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 electrica! 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 None REFERENCES NUMARC NESP 007, HA2 PBAPS Safe Shutdown Analysis NUMARC Questions and Answers, June 1993, " Hazards Question #7"

e - .

PBAPs EAL Technical Basis Manual MEV o. November 16,1998 Page 112 of 128 3.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases O

ALERT - 8.2.2.b IC Release of Toxic or Flammable Gases Within a Facility Structure Which Jeoptrdizes 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 (Tabic. 8-1) in concentrations that will be life threatening to plant personne!

_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 BASIS 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 I concentrations is not a concem, but rather that accecs is required to an area and is impeded.

Plant Vital Structures that must be accessed are as (oflows:

Table 8-1 Plant Vital Structures Power Block l

Diesel Generator Building

) Emergency Pump Structure inner Screen Structure Emergency Cooling Tower l

l The intent of this IC is not to include buildings (e.g., warehouses) or other areas that are not l

contiguous or immediately adjacent to plant Vital Areas. It is appropriate that increased i monitoring be done to ascertain whether consequential damage has occurred. This event will l

be escalated to higher classifications based upon damage consequences covered under other various EAL Sections.

DEVIATION l

None REFERENCES NUMARC NESP-007, HA3.1 and HA3.2

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

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

Page 113 of 128

! l 8.0 Extemal Events l D

! 8.3 Man-Made Events l

i UNUSUAL EVENT - 8.3.1.a l IC Natural and Destructive Phenomena Affecting the Protected Area EAL Vehicle crash within protected area boundary that may potentially damage plant structures 1

containing functions and systems required for safe shutdown of the plant. I I

OPCON " " " ' ~ '

BASIS This EAL is intended to address such items as plane, helicopter, or train crash that may I potentially damage plant structures containing functions and systems required for safe j shutdown of the plant. If the crash is confirmed to affect a plant vital area, the event may be i escalated to Alert.

r i

1 DEVIATION l None l l

REFERENCES NUMARC NESP-007, HU1.4 l

t i

'tO

>v l

u l

PBAPs EAL Technical Basis Manual REV o. No" ember 16.1998 Page 114 of 128 8.0 External Events 8.3 Man Made Events O

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 9 classified by the radiologicalICs 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 leakage 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 O

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

PBAPS EAL Technicd Basis Manual REV D, November 16,1998 Page 115 of 128 8.0 External Events 8.3 Man-Made Events i l

ALERT - 8.3.2 IC Destructive Phenomena Affecting the Plant Vital Area EAL i l

Vehicle crash affecting Plant Vital Structures (Table 8-1)

.98 j Turbine failure generated missiles result in any visible structural damage to or penetration of any Plant Vital Structures (Table 8-1)

~~

OPCON i BASIS This EAL tddress 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. Ne attempt should be made to assess the magnitude of damage to Plant Vital l (3 Structures prior to classification. The evidence of damage is sufficient for declaration. A I V vehicle crash includes aircraft and large motor vehicles, such as a crane. Missile impacts j including flying objects from offsite, onsite rotating equipment or turbine failure causing casing l penetration. Table 8-1 Plant Vital Structures are as follows:  ;

\ l Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower i 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 O

V I

i l

l i

PBAPs EAL Technical Basis Manual REV o, November 16,1998 l

Page 116 of 128 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.a IC Natural and Destructive Phenomena Affecting the Protected Area EAL i

Earthquake >.01 g as determined by procedure SO 67.7.A OPCON -

l j BASIS This EAL addresses a sensed earthquake. The magnitude of .01g is the lowest detectable earthquake measured on PBAPS seismic instrumentation per SO 67.7.A. 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 l

l functions.

l This event will be escalated to an Alert if the carthquake reaches an Operating Basis Earthquake.

DEVIATION None REFERENCES NUMARC NESP-007, HU1.1 SE-5, Earthquake and Bases UFSAR, section 1.6 O

._..._.__-.....m _ _ _ . . . . _ , _ _ - - .. _ ... . _ _ _ __ . .___ _ _ ._ _ . _ .._. _ .

t PBAPS EAL Technical Basis Manual -

{ REV D, November 16.1996

. Page 117 of 128 i

! 8,0 External Events iO

8.4 Natural Events 1

UNUSUAL EVENT - 8.4.1.b IC Natural and Destructive Phenomena Affecting the Protected Area  ;

j EAL Report by' plant personnel of tomado striking within protected area l

[ .Q.R Wind speeds > 75 mph as indicated on site Meteorological data for > 15 minutes l d

1 l

-OPCON -m

]

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 '

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

=, O in the level of safety of the plant. Verification of a tomado will be by direct observation and

~

reporting by station personnel. Verification of wind speeds > 75 mph will be via meteorological data in the control room. For purposes of this EAL, sustained is > 15 minutes.

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 tomado or high wind speeds have caused a iuss of shutdown cooling, then escalation will be by EAL IC, Loss of Decay Heat Removal Capability.

DEVIATION None REFERENCES NUMARC NESP-007, HU1.2 and HU1.7 O

PBAF's EAL T:chnical Basis Manual REV o, Nov;rnber 16,1998 Prge 118 of 128 8.0 External Events 8.4 Natural Events UNUSUAL EVENT - 8.4.1.c IC Natural and Destructive Phenomena Affecting the Protected Area EAL l I

t .

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 occurred and take appropriate action based on personal assessment as opposed to verification (e.g., an earthquake is felt but does not register on any plant-specific instrumentation, etc.)

DEVIATION None REFERENCES O

NUMARC NESP-007, HU1.3 O

PBAPs EAL Technical Bans Manual REV D, Nownbu 16,1998 Pige 119 of 128 i

,c) 8.0 External Events

\ ,/

8.4 Natural Events UNUSUAL EVENT - 8.4.1.d IC Natural and Destructive Phenomena Affecting the Protected Area EAL High River level > 112' O.B Low River level < 98.5' OPCON um- -

BASIS High River level of greater than 112 feet on instrument LI-2(3)278A,B,C or Lt-2(3)278A,B,C is indication of the river being in flood. By procedure, the units will be SCRAMMED and be brought to cold shutdown.

/m

( ')

Low River level of less than 98.5 feet is indication of loss of Conowingo Pond and Icss of circulation water pumps. Procedures require the unit to be SCRAMMED and brought to cold shutdown.

This event will be escalated to an Alert classification based continuation of the river situation.

DEVIATION None REFERENCES NUMARC NESP-007, HU1.7 SE-4, Flood SE-3, Loss of Conowingo Pond G

i f

PBAPS EAL Technical Basis Manual REV D. November 16.1998 Page 120 of 120 8.0 External Events '

8.4 Natural Events ALERT - 8.4.2.a IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL Earthquake >.05 g (Operating Basis Earthquake OBE) as determined by procedure SO 67.7.A .

OPCON wm BASIS This EAL addresses an earthquake that exceeds the Operating Basis Earthquake level of .05g and is beyond design basis limits. An earthquake of this magnitude may be sufficient to cause damage to safety related systems and functions.

The Max Credible Earthquake for PBAPS is 0.12g per UFSAR section 1.6, therefore this EAL is conservative and warrants an Alert classification.

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, HA1.1 SE-5, Earthquake and Bases UFSAR section 1.6 O'

l 1

~ , , _ _ - ..m _ _ _ _ _ _ _ . .. .._m-._--, -- - __ - _ __ . . _

PBAPs EAL Technied Bisis M:,nual REV D, Nowmber 16,1998 Page 121 of 128 8.0 External Events 8.4 Natural Events ALERT - 8.4.2.b IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL Tomado or wind speeds > 75 mph striking Plant Vital Structures (Table 8-1)

OPCON m>> -

BASIS This EAL is based on FSAR design basis. Wind loads of this magnitude can cause damage to safety functions. l This EAL addresses events where Plant Vital Structures have been struck with high winds, and i 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 '

Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Stracture Emergency Cooling Tower 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 m

PBAPs EAL Tochnical Basis Manual REV D, November 16,1998 Page 122 of 129 8.0 External Events i

8.4 Natural Events ALERT - 8.4.2.c IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL Report of any visible structural damage to any Plant Vital Structure (Table 8-1)

OPCON ' " > > "

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 81 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower

[

l Other site structures listed in the NUMARC document are not plant vital structures and are not l required for safe shutdown. Those are: RWST, CST.

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, HA1.3 i

i l

l 9

i

-. .-. . . ~ - . .-. .- . .. -

PBAPs EAL Technical Basis Manual REV o, Honmber 16,1998 Page 123 of 128

,q 8.0 External Events G

8.4 Natural Events ALERT - 8.4.2.d IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL High River level > 116'

_OR Low River Level < 92.5' OPCON '=*

BASIS High River level > 116 feet is indication of the river being in flood. This level is capable of causing flooding that can affect Plant Vital Structures. No attempt should be made to determine the magnitude of flooding. This is a long lead time event but this level is ground elevation of the reactor building and intake pump structure so classification as an Alert Event is (qj appropriate. The evidence of flooding is sufficient for declaration.

Low River level < 92.5 feet is indication of loss of Conowingo Pond and loss of circulation water pumps. Procedures require the unit to be SCRAMMED and brought to cold shutdown and utilization of the ECW pump and Emergency Cooling Tower.

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, HA1.7 SE-4, Flood SE-3, Loss of Conowingo Pond i

i t

N

PBAPS EAL Technical Basic Manual REV D.Wouat:Gr 16,1998 Page 124 of 128 This page intentionally left blank O

O

PBAPs EAL Technical Basis Manual REV D. Novemter 16.1998 Page 125 of 128 9.0 Other

(']

%s 9.1 General UNUSUAL EVENT - 9.1.1 IC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant 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 OPCON '-

BASIS 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 q functionality, not operability. A system, subsystem, train, component or device, though (g degraded in equipment condition or configuration, should be considered functional if it is capable of maintaining respective system parameters within acceptable design limits.

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."

DEVIATION None REFERENCES NUMARC NESP-007, HUS

PBAPs EAL Technical Basis Manual REV D. November 16.1998 Page 126 of 120 9.0 Other 9.1 General O ,

l ALERT - 9.1.2  !

(

IC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant Declaration of an Alert EAL I safety Other conditions systems exist may be degraded and thatwhich increasedin the Judgement monitoring of plant functions is of the Emerge warranted.

'm- ~

OPCON BASIS This EAL allows the Shift Management to declare an Alert upon the 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:

Internal flooding afrects 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 I

I 1

O l

l L

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

PBAPS EAL Technical Balis Manua!

REV o. Nov1mber 16,1998 Page 127 of 123 p g- 9.0 Other D 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 EAL l

Other conditions exist which in the Judgement of the Emergency Director indicate actual or likely major f ailures of plant functions needed for protection of the public l

o~+

l OPCON l

BASIS This EAL allows the Shift Management to declare a Site Area Emergency upon the i 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 l Releases."

l DEVIATION None REFERENCES NUMARC NESP-007, HS3 l

l i

l l

PBAPs EAL Technical Basis Manual REV o,ieovarr.ber 16, t998 Page 123 of 129 9.0 Other 9.1 General O

GENERAL EMERGENCY - 9.1.4 IC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant Declaration of General Emergency EAL Other conditions exist which in 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

~"

CPCON BASIS 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 i None REFERENCES l

l NUMARC NESP-007, HG2 I

I O

l

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

l<

i-PBAPs EAL Table REV D, November 16.1998 Page 1 of 2s l

l ('~] PBAPS EAL Table l V. . Table of Contents l I

1.0 ' Reactor Fuel 1.1 Coolant Activity..... .. . . ...... .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .2 1.2 Irradiated Fuel or New Fuel.. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 3 l

2.0' Reactor Pressure Vessel 2.1 Reactor Water Level . ... . . .. .. . .. . . . . . . . . . . . . ....4 2.2 Reactor Power.... . . . ..... . ... . . . . . . . . . . . . . . . . . . . .. . . . . . . .5 1 3.0 Fission Product Barrier l 3.1 initiating Condition Matrix.. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .....6 )

3.2 Fission Product Barrier Table.. . . . . . . . . . . . . . . . . . . . . . . . . . ........7 4.0 Secondary Containment 4.1 Main Steam Line . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . ....9 5.0 Radioactivity Release 5.1 Effluent Release and Dose .. . . . . . . . . . . . . . . . . . . . . . . . . . . .... 10 5.2 - In-Plant Radiation. ... ..... . . . . . . . . . . . . . . . . .. ..........................12 6.0 Loss of Power f-s 6.1 - Loss of AC or DC Power . ...... .... . .. .. .. . , . . . . . . . . . . . . . .. .. . . . . . . . . 13 l(V

. )

7.0 Internal Events-7.1 Technical Specifications & Control Room Evacuation. .. .. .. . .. ... . . . ... 15 7.2 Loss of Decay Heat Removal Capabililty.. .. ... . . . . . . . . .. . . . . 16 7.3 Loss of Assessment / Communications Capabililty. . .. . . . . . . . . . .. .. 17 i 8.0 ' External Events 8.1 S e cu rity Ev e nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 8.2 Fire / Explosion and Toxic / Flammable Gases .. . . . . . .. .............20 8.3 Man-Made Events ...... ........ ..... ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 8.4 Natural Events. . ........ .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 9.0 Other 9.1 General .. ... ... .. ... . . . . ... . . . . . . . . . . . . .......25 OPCON (MODE) MODE SWITCH POSITION

- Run wov Startup

== Shutdown (hct) a-Shutdown (cold) sma Refueling sunmiG N/A (defueled) i p.

(

I PBAPS EAL Table REV D, November 16,1998 Page 2 of 25 1.0 Reactor Fuel

({}

l 1.1 Coolant Activity CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT IC Fuel Clad Degradation 1.1.1.a me '

Reactor Coolant activity > 4 CFgm Dose Equivalent lodine 131 1.1.1.b EUsa SJAE Radiation (Offgas Monitor) > 2.5x10' mR/hr ALERT None SITE AREA None EMERGENCY GENERAL None EMERGENCY r%

'\.)

u v

1 I

I PBAPs EAL TtNe i REV D, Normber 16,1998 I Page 3 of 25 lp l v 1.0 Reactor Fuel 1 l 1.2 Irradiated Fuel or New Fuel i

CLASSIFICATION EMERGENCY ACTION LEVEL i UNUSUAL EVENT IC Unexpected increase in Plant Radiation or Airborne Concentration.

1.2.1.a <*'w l ,

Uncontrolled water level decrease in the spent fuel pool with all irradiated fuel l assemblies remaining covered by water I l l 1.2.1.b i='M Unexpected Skimmer Surge Tank low level alarm AND l

Visual observation of an uncontrolled water level decrease below the fuel pool i skimmer surge tank inlet l

ALERT 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 1.2.2.a twm Unplanned general area radiation > 500 mR/hr on the refuel floor (Table 1-1) 1.2.2.b '">""'w Report of visual observation of irradiated fuel uncovered 1.2.2.c tuuum l Water Level < 458" above RPV instrument zero for the Reactor Refueling Cavity that will result in Irradiated Fuel uncovering 1.2.2.d r~ w l Water Level < 232 ft 3 Inches plant elevatlon for the Spent Fuel Pool that will result in irradiated Fuel uncovering SITE AREA None EMERGENCY GENERAL None EMERGENCY Table 1-1 Refuel Floor ARMS 3-7 (7-9) Steam Separator Pool 3-8 (7-10) Refuel Slot 3-9 (7-11) Fuel Pool O

V 3-10 (7-12) Refueling Bridge

PBAPs EAL Table REV D, November 16,1998 Page 4 of 25

/T 2.0 Reactor Pressure Vessel L.)

2.1 Reactor Water Level CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT IC Reactor Coolant System Leakage 2.1.1 cram The following conditions exist:

Unidentified Primary System Leakage > 10 ppm into the Drywell

_QB Identified Primary System Leakage > 25 gpm into the Drywell ALERT None SITE AREA IC Loss of Water Level in the Reactor Vessel That Has or Will Uncover fuel in the Reactor Vessel EMERGENCY 2.1.3 msm m RPV level < -172 "'

GENERAL None EMERGENCY l

l l

(. l l

l

PBAPs EAL Tibb REV o, November 16,1998 l Page 5 of 25 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 EEEm 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 Emum Automatic RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 3%

IC Failure of the Reactor Protection System to Complete an Automatic GENERAL EMERGENCY S ram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core 2.2.4 Emm Automatic RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 3%

AND Torus Temperature is > 180 degrees F

. ~~ . ...

PBAPs EAL Table REV D, November 16,1998 Page 6 of 25

(')

w/

3.0 Fission Product Barrier Table 3.1 initiating Condition Matrix CLASSIFICATION EMERGENCY ACTION LEVEL 3.1.1 ==

UNUSUAL EVENT ANY Loss or ANY Potential Loss of Containment 3.1.2 mm ALERT ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS SITE AREA 3.1.3 esna 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 314 "'

GENERAL L ss of ANY Two Barriers EMERGENCY AND Potential Loss of Third Barrier l

I NOTES:

1 If a " Loss" condition is satisfied, the " Potential 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.

(~\

,E

3.2 Fission Produi Barrier Fuel Clad Reactor Coolant Syste Indicator Loss Potential Loss Loss Poteni Reactor Coolant Reactor Coolant N/A N/A N/A Activit/ activity > 300 pC//gm Dose Equivalent lodine 131 RPV Level RPV level < -172 " RPV level < -172 " N/A RPV level < -226 "

RPV Level Unknown N/A N/A N/A RPV leve' determine r.

(, '

RCS Leak Rate N/A N/A N/A RCS leaX

>50 gpm O!

Unisolabt system le:

outside di indicated Tempera 3 Level is e ONE are:

SCRAM 91 Unisolabi:

system le outside di indicated Radiation is exceed-area reqM SCRAM S PO

PBAPS EALTatie REV D. Nowmber 16,1998 Barrier Tcble Page 7of 25 i Primary Containment ILeco 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 -226 "

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 ApgRT W' be established per T-116 ,

CARD N/A N/A Also Available on la Ap 3rture Card SITE AREA EMERGENCY prim:ry Loss of BOTH Fuel Clad AND RCS (ag) OR weH as Potential Loss of BOTH Fuel Clad a T-10 AND RCS 9n teed;d in requiring a Potential Loss of EITHER Fuel Clad OR RCS, and Loss of ANY Additional Barrier prim:ry kags twell as iy a T-103 Action L: vel GENERAL EMERGENCY id in ONE ring a Loss of ANY Two Barriers AND Potential Loss of Third Barrier j i

[

c

i l

3.2 Fiscion Prod $

() Barrier Fuel Clad i Reactor Coolant Sys?

Indicator Loss Potential Loss Loss Poten1 Drywell Pressure N/A N/A N/A Drywell Pressure

> 2.0 psig AND Indication of a leak ,

inside drywell l Drywell Radiation - Drywell Rad Monitor N/A N/A Drywell Rad Monitor l reading > 8x10' R/hr reading > 15 R/hr Containment Isolation N/A N/A N/A N/A 1

l

(-

< Emergency Director Any condition in the judgement of the Any condition in the judgement of t Judgement Emergency Director that indicates Loss or Emergency Director that indicates Potential Loss of the FUEL CLAD barrier Potential Loss of the RCS barrier

PBAPS EALTable REV D, Nowmber 16,1998 Barrier Tcble Page 8 of 25 Primary Containment 3 Loss Loss Potential Loss Rapid, unexplained Drywell Pressure decrease in Drywell > 49 psig and Pressure following increasing UNUSUAL EVENT initialincrease OR Drywell7 essure Drywell Hydrogen ANY Loss or ANY Potential Loss of Conta,inment response not > 6% AND Drywell consistent with LOCA Oxygen > 5% l conditions N/A Drywell Rad Monitor reading > 6x10' R/hr N/A

^

Failure of both valves in any one line to close AND ANY Loss or ANY Potential Loss of downstream pathway EITHER Fuel Clad OR RCS to the environment Am -

exists w uRTURE:

g s CARD Intentional venting per A h o ,4 v ejg @ e o n T-200 is required ^1Mure Card SITE AREA EMERGENCY M

Unisolable primary Loss of BOTH Fuel Clad AND RCS system leakage OR outside drywell as Potential Loss of BOTH Fuel Clad indicated by a T-103 AND RCS Temperature Action OR Level is exceeded in Potent:al Loss of EITHER Fuel Clad ONE a ea reqw,n,ng a OR RCS, and Loss of ANY Additional C Sdifier Unisolable primary syste.m leakage outside drywell as indicated by a T-103 GENERAL EMERGENCY Radiation Action Level is exceeded in ONE Loss of ANY Two Barriers area requinng a AND SCRAM Potential Loss of Third Barrier he Any condition in the opinion of the Emergency

. :s or Director that indicates Loss or Potential Loss of the CONTAINMENT barrier g -

o

_ m . . ._. .

PBAPS EAL Table REV 0, NovImb:r 16,1998 Page 9 of 25

('j 4.0 Secondary Containment R.,

4.1 Main Steam Line CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT IC Fuel Clad Degradation 4.1.1 Main Steam Line HiHi Radiation (10xNFPB)

ALERT IC RCS Leak Rate 4.1.2 carmi Indication of a Main Steam Line Break:

Hi Steam Flow Annunciator AND Hi Steam Tunnel Temperature Annunciator 9.R.,

Direct report of steam releaue SITE AREA None EMERGENCY GENERAL None EMERGENCY o

O 1

l

PBAPs EAL Table REV D. Movember 16,1998 Page 10 of 25 O 5.0 Radioactivity Release V

5.1 Effluent Release and 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 S.i.1.a em A valid reading on one or more of the following radiation monitors that exceeds TWO TIMES the HiHi alarm setpoint value for > 60 m!nutes:

Main Stack, Vent Stack, Radwaste Discharge, Service Water Discharge AND Calculated maximum offsite dose rate using computer dose model exceeds 0.114 mrem /hr TPARD OR 0.342 mrem /hr chlid thyrold 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.

5.1.1.b """m Confirmed sample analyses for gaseous orliquid rebases indicates concentrations or release rates exceeding TWO tin.ES Tech Specs (Liquid I I Release ODCM. 3.8.B.1 and Gaseous Release ODCM 3.8.C.1.b) for

> 60 minutes IC Any Unplanned Release cJ Gaseous or Liquid Radic3ctivity to the ALERT Environment that Exceeds 2e') Times Radiologic 4! Technical Specifications for 15 Minutes .ir Longer S.1.2.a E!Em 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:

Main Stack, Vent Stack, Radwaste Discharge, Service Water Discharge AND Calculated maximum offsite dose rate exceeds 11.4 mrem /hr TPARD QR 34.2 mrem /hr child thyroid CDE based on a 15 minute average Note: If the required dose prcjactions cannot be completed within the 15 minute period, then the declarstion must be made based on the valid sustained monitor reading.

S.1.2.b ""iM Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates excecding TWO HUNDRED TIMES Tech Specs (Liquid Release ODCM. 3.8.B.1 and Gaseous Release ODCM 3.8.C.1.b) for

> 15 minutes l I

l l

PBAPs EAL Table l REV o Moucmt:ct 16,1998 l Page 11 of 25 1

ISITE AREA IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mR Whole Body or 500 mR Child EMERGENCY Thyroid for the Actual or Projected Duration of the Release '

5.1.3 - l A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 m/nutest  !

Main Stack 5.84 Ci/cc Vent Stack 2.08E-3 Ci/cc Torus Vent 203 cpm AND Projected offsite dose using computer dose model exceeds 100 mrem TPARD E 500 mrem child thyroid CDE I 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.

l Ok \

Analysis of Field Survey results indicate site boundary whole body dose rate l exceeds 100 mrem /hr expected to continue for more than one hour, @  !

Analysis of Field Survey results indicate child thyroid dose commitment of 500 l mrem for one hour of inhalation GENERAL IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1000 mR Whole Body or 5000 mR EMERGENCY Child Thyroid for the Actual or Projected Duration of the Release Using Actual Meteorology 5.1.4 -w 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:

Main Stack 58.4 Ci/cc  !

Vent Stack 2.08E-2 Ci/cc Torus Vent 2000 cpm AND Projected offsite dose using computer dose model exceeds 1000 mrem TPARD

@ 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.

Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 1000 mrem /hr expected to continue for more than one hour, @

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

l PBAPs EAL Table REV D, Nov:mb:t 16,1998 Page 12 of 25 l

/G 5.0 Radioactivity Release L) 5.2 In-Plant Radiation CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT IC Unexpected increase in Plant Radiation or Airborne Concentration 5.2.1 ' * " "'

Valid Direct Area Radiation Monitor readings increase by a factor of 1000 over normal

• levels i Norma! levels can be considered as the highest reading in the past I twenty-four hours excluding the current peak value.

IC Release of Radioacuve Material or increases in Radiation Levels Within ALERT 4 the Facility That Impedes Operation of Systems Required to Maintain l Safe Operations or to Establish or Maintain Cold Shutdown 5.2.2.a Ha"M Valid radiation level readings > 5000 mR/hrin areas requiring infrequent access to maintain plant safety functions as identified in procedure SE-1 or SE-10 l l AND Access is required for safe plant operation, but is impeded, due to radiation dose rates 5.2.2.b => ><

Valid Control Room 9R Central Alarm Station radiation reading > 15 mR/hr SITE AREA None EMERGENCY GENERAL None EMERGENCY l

(

(

l

t PBAPs EAL Tabb REV D, November 16,1998 l Paga 13 of 25 I O 6.0 Loss of Power

' C) l 6.1 Loss of AC or DC Power  ;

i l

CLASSIFICATION EMERGENCY ACTION LEVEL IC Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes j l UNUSUAL EVENT i

6.1.1.a Gm23 The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer for >15 minutes AND At least Two Diesel Generators are supplying power to their respective l l 4 KV emergency busses  ;

l lC Unplanned Loss of Required DC Power During Cold Shutdown or j Refueling Mode for Greater than 15 Minutes I 6.1.1.b umra The following conditions exist:

Unplanned Loss of ALL safety related DC Power indicated by

< 107.5 VDC bus voltage indications for DC Pane!s 2(3)0D21,22,23,24 AND l Failure to restore power to at least one required DC bus within 15 minutes l l l 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 ama The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer for >15 minutes 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 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 l 6.1.2.b umEB The following conditions exist:

l Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer 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

l

! PBAPS EAL Table l REV t', November 16,1998 l Pae 14 of 25 l

IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential SITE AREA

! Busses EMERGENCY l

l 6.1.3.a -

The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and

! 343 Startup Transformer l AND l l Failure to restore power to at least One 4 KV emergency bus within 15 l minutes from the time of loss of both offsite and onsite AC IC Loss of All Vital DC Power 6.1.3.b EFFJun Loss of ALL Safety Related DC Power indicated by < 107.5 VDC on DC Panels  ;

2(3)0D21,22,23,24 for > 15 minutes i GENERAL IC Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power 1 EMERGENCY 6.1.4 EME l l Prolonged loss of all offsite and onsite AC power as indicated by:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer AND l Failure of ALL Emergency Diesel Generators to supply power to 4 KV l emergency busses AND At least one of the following conditions exist:

. Restoration of at least One emergency bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is NOT likely

. Reactor Water Level cannot be maintained > -172 "

M

. Torus temperature is greater than the Heat Capacity Temperature Limit (HCTL)

/3 i )

s.>

l

PBAPS EAL Table REV D, Nowmber 16,1998 Page 15 of 25 7.0 Internal Events (v~)

7.1 Technical Specification & Control Room Evacuation CLASSIFICATION EMERGENCY ACTION LEVEL lC Inability to Reach Required Shutdown Within Technical Specification UNUSUAL EVENT Limits 7.1.1 EINE inability to reach required shutdown mode within Tech. Spec. LCO required action completion time.

ALERT IC Control Room Evacuation Has Been Initiated 7.1.2 Nm Entry into SE-1 or SE-10 procedure for Control Room evacuation SITE AREA IC Control Room Evacuation Has Been initiated and Plant Control Cannot Be Established EMERGENCY 7.1.3 I I The following conditions exist:

Control room evacuation has been initiated AND Control of the plant cannot be established per SE-1or SE-10 within 15 minutes GENERAL None EMERGENCY l

["i

\

I PBAPS EAL Table '

REV D, November 16.1998 l Page 16 of 25 7.0 Internal Events

~f}

x._ j 7.2 Loss of Decay Heat Removal Capability CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT None 1

ALERT. IC Inability to Maintain Plant in Cold Shutdown j 7.2.2 M'EDE The following conditions exist:

Loss of all decay heat removal cooling as determined by procedure GP-12 l AND Uncontrolled Temperature increase that either:

• Exceeds 212 *F e Results in temperature rise approaching 212 "F

{ IC Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown (ITE AREA EMERGENCY 7.2.3 m>w" l Loss of Main Condenser as a heat sink ,

AND I Loss of TORUS heat sink capabilities as evidenced by T-102 legs requiring an )

Emergency Blowdown AND Either of the following conditions: l

.l

• RPV level < -172 "

M e Reactor Power > 3%

GENERAL None EMERGENCY l

l I l

Ov l

PBAPs EAL Tabb REV D, Nov:mtnf 16,1998 Page 17 of 25

/'N 7.0 Internal Events i )

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 ens Unplanned loss of most or all safety system annunciators (Table 7-1) QB l indicators (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 unum Loss of ALL Onsite communications (Table 7-3) affecting the ability to perform routine operations OR l Loss ofEL 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 NEm Unplanned loss of most or all safety system annunciators (Table 7-1) QB 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) QR the plant monitoring system (PMS) is unavailable.

IC Inability to Monitor a Significant Transient in Progress SITE AREA EMERGENCY 7,3,3 gy, Loss of safety system annunciators (Table 71) l AND indicators (Table 7-2)

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

{ pENERAL None EMERGENCY

. . _ . _ , . __ m . _ . . _ . _ _ .- . _. . _ . . _ _ . 2 PBAPS EAL Table REV D, November 16,1998 Page 18 of 25 -

'able 7-1 Safety System Annunciators 3CS.

' Containment isolation I

R:: actor Trip _ l

. Process Radiation Monitoring l Table 7-2 Safety Function Indicators R: actor Power D: cay Heat Removal Containment Safety Functions l Trble 7-3 Communications  :

Onsite Offsite  !

Sito Phones (GTE System) X X OMNI System X X Pl:nt Public Address X .

St tion Radio X NRC (FTS-2000) X PA State Police Radio X Lord Dispatcher Radio X

. PECO Dial Network X

~[

\

s p/able 7 Plant Transients SCRAM R:cire Runbacks > 25% thermal power Th:rmal power oscillations > 10%

Stuck open relief valve (s)

ECCS injection-r'%

b

PBAPs EAL Table REV D, Nov".mber 10,1998 Page 19 of 25 (7.0 External Events L) 8.1 Security Threats CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT IC Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant 8.1.1 i~ i i Credible sabotage or bomb threat within the Protected Area M

Credible intrusion and attack threat to the Protected Area M

Attempted intrusion and attack to the Protected Area OR l AttempEsabotage discovered within the Protected Area M

Hostage / Extortion situation that threatens normal plant operations IC Security Event in a Plant Protected Area ALERT

{ I 8.1.2 RNT1 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 g,3,3 , _ , ,

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 8=E Loss of physical control of the control room due to security event M

Loss of physical control of the remote shutdown capability due to security event l I

PBAPs EAL Table REV D, November 16,1998 i Page 20 of 25 r~' 8.0 External Events  !

Q. 1 8.2 Fire / Explosion and Toxic / Flammable Gases CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT IC Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes of Detection 8.2.1.a '~N Fire within ON-114 Plant Vital Structures (Table 8-1) which is not extinguished within 15 minutes of control room notification or verification of a control room alarm l

IC Release of Toxic or Flammable Gasses Deemed Detrimental to Safe l Operation of the Plant l 8.2.1.b rmm a Report or detection of toxic or flammable gases that could enter within the site l area boundary in amounts that can affect normal operation of the plant 08 ,

Report by Local, County or State Officials for potential evacuation of site i personnel based on offsite event i I IC Natural and Destructive Phenomena Affecting the Protected Area 8.2.1.c '"

• Report by plant personnel of an unanticipated explosion within protected area I 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 '~~

The following conditions exist:

Fire or explosion which makes inoperable:

Two orMore subsystems of a Safe Shutdown System (Table 8-2) 98 Two or More Safe Shutdown Systems QR Plant Vital Structures containing Safe Shutdown Equipment AND Safe Shutdown System or Plant Vital Structure is required for the present Operational Condition l I

l PBAPs EAL Table REV D, Noember 16,1998 Page 21 of 25 l l

lC 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 8.2.2.b "-

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_ j Report or detection of flammable gases within Plant Vital Structures (Table 8-1) in concentrations affecting the safe operation of the plant i

SITE AREA None EMERGENCY I

GENERAL None l EMERGENCY Table 8-1 Plant Vital Structures I l

C) Power Block C/ Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower Table 8-2 Safe Shutdown Systems Diesel Generators 4KV Safeguard Buses ADS l HPCI RCIC RHR (All Modes)

Core Spray HPSW ESW SBGTS ECW CAC/ CAD PCIS Control Room Ventilation l

l i

N ]'

PBAPs EAL Table REV D, November 16,1998 Page 22 of 25

~LO External Events l (V l

l 8.3 Man-Made Events CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL' EVENT IC Natural and Destructive Phenomena Affecting the Protected Area j 8.3.1.a >> >*

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 EEE E Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

ALERT IC Destructive Phenomena Affecting the Plant Vital Area 8.3.2 ia m Vehicle crash affecting Plant Vital Structures (Table 8-1)

OR l 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 Trble 81 Plant Vital Structures Power Block Di:sel Generator Building Emergency Pump Structure Inn:r Screen Structure Emergency Cooling Tower l

O G

4 i

i PBAPs EAL Table REV D, November 16,1998 Page 23 of 25 i

(~~'T 8.0 External Events LJ l 8.4 Natural Events CLASSIFICATION EMERGENCY ACTION LEVEL l 1

UNUSUAL EVENT IC Natural and D6J.. Wive Phenomena Affecting the Protected Area i 8.4.1.a em Earthquake >.01 g as determined by procedure SO 67.7.A 8.4.1.b t*me Report by plant personnel of tornado striking within protected area M

Wind speeds > 75 mph as indicated on site Meteorological data for > 15 minutes 8.4.1.c Erde Assessment by the control room that an event has occurred. (Natural and Destructive Phenomena Affecting the Protected Area) 8.4.1.d wm

{ g High River level > 112' M

Low River level < 98.5' IC Natural and Destructive Phenomena Affecting the Plant Vital Area ALERT 8.4.2.a Emid w Earthquake >.05 g (Operating Basis Earthquake OBE) as determined by procedure SO 67.7.A 8.4.2.b tuEm l

Tornado or wind speeds > 75 mph striking Plant Vital Structures (Table 8-1) 8.4.2.c Nais Report of any visible structural damage to any Plant Vital Structure (Table 8-1) l 8.4.2.d > >

High River level > 116' l I Low River Level < 92.5'

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

l PDAPS EAL Table f I REV D, November 16,1998 l Page 24 of 25 l m

l slTE AREA None l l EMERGENCY l

l None EMERGENCY IGENERAL Table 81 Plant Vital Structures

. Power Block l . Di:sel Generator Building l Em rgency Pump Structure Innot Screen Structure Errdrgency Cooling Tower t

0 I.

, s ,<

e l

O 1

PBAPs EAL Table REV D, Noumber 16,1998 Page 25 of 25 9.0 Other 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 nm 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 I

l 9.1.2 >">>

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.

dlTE AREA IC Other Conditions Existing Which in the Judgement of the Emergency Director Warrant Declaration of Site Area Emergency EMERGENCY 9.1.3 -m 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 mmu m Other conditions exist which in 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

>v

P8APs EAL NUMARC comparison REV D, Novcmb216,1998 Page 1 of 4 PBAPS EAL NUMARC Comparison NUMARC PBAPS COMMENTS / EXCEPTIONS AU1.1 5.1.1.a AU1.2 5.1.1.b AU1.3 N/A PBAPS does not have telemetered perimeter monitors.

AU1.4 N/A PBAPS does not use automatic initiation of real time dose assessment.

AU2.1 1.2.1.b Unexpected" and " Uncontrolled" added to EAL for clarity.

AU2.2 1.2.1.a Fuel transfer canal is not applicable to PBAPS.

AU2.3 N/A PBAPS 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 PBAPS does not have telemetered perimeter monitors.

!g AA1.4 N/A PBAPS does not use automatic initiation of real time dose I ( assessment AA2.1 1.2.2.a AA2.2 1.2.2.b AA2.3 1.2.2.c AA2.4 1.2.2.d Fuel transfer canal is not applicable to PBAPS.

AA3.1 5.2.2.b AA3.2 5.2.2.a AS1.1 5.1.3 AS1.2 N/A PBAPS does not have telemetered perimeter monitors.

AS1.3 5.1.3 AS1.4 5.1.3 AG1.1 5.1.4 AG1.2 N/A PBAPS does not have telemetered perimeter monitors.

AG1.3 5.1.4 AG1.4 5.1.4 HU1.1 8.4.1.a

l PBAPS EAL NUMARc Comparison REV D. November 16.1998 Page 2 of 4

'O

<v NUMARC PBAPS COMMENTS / EXCEPTIONS l HU1.2 8.4.1.b HU1.3 . 8.4.1.c 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 HU1.7 8.4.1.b l 8.4.1.d l HU2 8.2.1.a HU3.1 8.2.1.b HU3.2 8.2.1.b HU4.1 8.1.1 A b mb device discovered within Plant Protected Area and outside the Plant Vital Areas is an Alert declaration as determined per the site Safeguards Contingency Plan and therefore is not included as an Unusual Event in the EAL scheme.

HU4.2 8.1.1 HUS 9.1.1 HA1.1 8.4.2.a HA1.2 8.4.2.b HA1.3 8.4.2.c HA1.4 N/A PBAPS 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 8.4.2.d HA2 8.2.2.a Wording of EAL changed for clarity. No change in intent.

HA3.1 8.2.2.b l 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 HAS 7.1.2 i

PBAPS EAL NUMARC Comparison REV D. November 16,1998 Page 3 of 4 NUMARC PBAPS COMMENTS / 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 HG12 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 inte?.

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 SUS 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 hVording 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.

SA5 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 E .

PBAPs EAL NUMARC Comp'fison REV D. November 16.1998 Page 4 of 4 g)

/

NUMARC PBAPS COMMENTS / EXCEPTIONS SSS 2.1.3 During EAL review and approval process, it was determined

! that the condition stated in NUMARC NESP-007, SS5,1.a l " 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.

SS6 7.3.3 Wording of EAL simplified. No deviation in intent.

l 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 l

FC.4 3.2-FC.4 FC.5 3.2 FC.5 p RC.1 3.3-RC.1 4.1.2 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) l 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 l 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