ML20209F560

EAL Basis Document, for Jul 1999

ML20209F560
Person / Time
Site:	Crystal River
Issue date:	07/31/1999
From:	FLORIDA POWER CORP.
To:
Shared Package
ML20209F524	List: 3F0799-22, Provides Update & Rev to Submittal Made by Util Ltr with Regard to EAL Classification Methodology for Unit 3.Reponses to NRC Staff Questions Provided as Attachment D to Ltr & Reflects Discussions Held ML20209F560
References
NUDOCS 9907160062
Download: ML20209F560 (98)

Text

p gi i FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACHMENT B TO LETTER 3F0799-22 CRYSTAL RIVER UNIT 3 EMERGENCY ACTION LEVEL BASIS DOCUMENT JULY 1999 f5*IN8!o"soo$3o2 PDR .

EMERGENCY ACTION LEVEL BASIS DOCUMENT TABLE OF CONTENTS

\

ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT......... .. ... ............ ........ 1 Gaseous Effluents Liquid Effluents Unexpected Radiation Levels F

Fuel Handling / Fuel Handling Pool Water Level NATURAI/ MAN-MADE HAZARDS AND EC JUDGMENT .............. ...................... 14 Earthquake Experienced Flood Hurricane

!. Tornado /High Winds i Air:raf1/ Vehicle Crash Toxic or Flammable Gas Explosions / Catastrophic Pressurized Equipment Failure Fire l Control Room Evacuation Security Event Internal Flooding )

Emergency Coordinator J'udgment l SYSTE M MALFUNCTI ON . .. . . . . . . .. . . . . . . . . . .. . . . . . . .. . . . .. . .. . . . . . .. .. . .. . . . . . .. . . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . 4 4 l Loss of Communication Failure of Reactor Protection Inability to Reach Required Mode Within Improved Technical Specification ,

Time Limits l l

Loss ofIndications i Fuel Clad Degradation Turbine Failure RCS Leakage Loss of Function (Hot Shutdown)

Inadvertent Criticality l Inability to Maintain Plant in Cold Shutdown Loss of Water Level in Reactor Vessel that Has Uncovered or Will l Uncover Fuel l

LO S S O F POWE R . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Loss of AC Power l Loss of AC Power (Shutdown)

Loss of Vital DC Power

. . Loss of Vital DC Power (Shutdown)

FISSION PRODUCT BARRIER MATRIX BASIS .................... ......................... ........ 71 D E FINITIO N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 l

Emerg:ncy Action Level Basis Document ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT l l

Gaseous Effluents MODES: ALL Classification:

Unusual Event Initiating Condition:

An UNPLANNED release of gaseous radioactivity to the environment that exceeds 2 times the ODCM noble gas release setpoint for 60 minutes or longer l Emergency Action Level:

l (1 or 2)

I

1. 'A VALID reading on RM-Al or RM-A2 gas channel exceeds the UNUSUAL EVENT threshold value listed on the Radioactive Release Permit for 60 minutes or longer f2B l 2. Sample analysis confirms gaseous effluent being released exceeds 2 times the ODCM noble gas release setpoint for 60 minutes or longer l Basis:

The Unusual Event threshold value listed on the Release Permit represents 2 times the ODCM limit. l Releases in excess of 2 times the ODCM limits continuing for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose is not the primary concern here; it is the degradation in plant control implied by the fact the release was not isolated within 60 minutes. Therefore, it is not intended for the release to be averaged over 60 minutes. For example, a release of 4 times the ODCM for 30 minutes does not exceed this initiating condition. Further, the Emergency Coordinator should not wait until 60 minutes elapses, but declare the event as soon as it is determined the release duration willlikely exceed 60 minuter. This is identified by an increasing trend in monitor readings. This does not include spikes or other erroneous instrument  !

- rerdouts. I CR-3 Matrix Reference Number:1.1 NEI 97-03

Reference:

AU1 -

1

Emerg:ncy Action Lev:1 Baris Document ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Gaseous Effluents MODES:ALL Classification:

Alert Initiating Condition:

An UNPLANNED release of gaseous radioactivity to the environment that exceeds 200 times the ODCM noble gas release setpoint for 15 minutes or longer l j j

Emergency Action Level:

(1 or 2)

1. A VALID reading on RM-Al or RM-A2 exceeds the ALERT threshold value listed on the Radioactive Release Permit for 15 minutes or longer

_O_.R

2. Sample analysis confirms gaseous effluent being released exceeds 200 times the ODCM noble gas release setpoint for 15 minutes or longer l l

l Basis: l The Alert threshold value listed on the Release Permit represents 200 times the ODCM limit. Unplanned l l releases in excess of 200 times the ODCM limits continuing for 15 minutes or longer represent an l uncontrolled situation and hence, a potential substantial degradation in the level of safety. The primary concern for the time factor here is the loss of control of radioactive material allowing the release to continue. The Emergency Coordinator should not wait until 15 minutes elapses, but declare the event as coon as it is determined the release duration willlikely exceed 15 minutes. l CR-3 Matrix Reference Number:1.2 NEI 97-03

Reference:

AA1 2

1 i

Emergency Action Level Basis Document ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT l Gaseous Effluents MODES:ALL Classification:

Site Area Emergency Initiating Condition:

SITE BOUNDARY dose resulting from an actual or projected release of airborne radioactivity exceeding 100 mR TEDE or 500 mR Thyroid CDE l

Emergency Action Level:

. (1 or 2 or 3) l

1. VALID RM-Al or RM-A2 Mid-Range monitor reading exceeds the values on the following table for I the current stability class for 15 minutes or longer: l Stability Class

~

"I*

Monitor Reading (mR/hr) j A, B, or C 80 D or E 20 F or G 5 l

OR l 2. Dose Assessment results indicate SITE BOUNDARY dose >100 mR TEDE pr > BOO mR Thyroid CDE l l for the actual pr projected duration of the release l

O_R

3. Field survey results indicate closed windows dose rates > 100mR/hr expected to continue for more than one hour; l or analyses of field survey samples indicate thyroid CDE of 500mR for one hour of inhalation, at or beyond site boundary ,

l Basis:

A RM-Al or RM-A2 mid-range monitor reading indicated on the Table based on stability class would )

re; ult in a Site Boundary Dose of >100 mR TEDE using conservative meteorology (1 m/see windspeed, G Stability class, no precipitation as outlined on the attached Table).

- Source term for this EAL is based on RCS with 1</< failed fuel per FSAR safety analysis.

3 l

l

Em:rg:ncy Action Level Ba:is Document ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Gaseous Effluents, continued Basis, continued The 100 mR integrated dose in this initiating condition is based on the 10 CFR 20 annual average population exposure. It is deemed exposures less than this are not consistent with the Site Area Emergency class description. These values are 10% of the EPA 400 Protective Action Guidelines (PAG).

l Integrated doses are generally not monitored in real-time. In establishing the emergency action levels, a release of one hour is assumed, and the EALs are based on a site boundary dose of 100 mR TEDE or 500 mR Thyroid CDE, whichever is more limiting.

Classification for items 2 & 3 above result from emergency response team input. For example, the Environmental Survey Team provides actual dose rates which are used to determine dose for the projected duration of the release. The Dose Assessment Team provides projected dose.

CR-3 Matrix Reference Number:1.3 NEI 97-03

Reference:

ASI 4

Em:rg:ncy Action Lcvel Basis Document ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Gaseous Effluents l MODES: ALL Classification:

General Emergency bitiating Condition:

fsITE BOUNDARY doa resulting from an actual or projected release of gaseous radioactivity exceeding 1000 mR TEDE pr 5000 mR Thyroid CDE I Emergency Action Level: 1 (1 or 2 or 3)

1. Valid RM-Al or RM-A2 Mid-Range monitor reading exceeds the values on the table below for the current stability class for 15 minutes or longer:

l Stability Class M to R a (mR/hr)

A, B, or C 800 l 1

D or E 200  !

1 F or G 50 OR

2. Dose Assessment results indicate SITE BOUNDARY dose >1000 mR TEDE pr >5000 mR Thyroid I CDE for the actual or projected duration of the release and core damage is suspected or has occurred l OR

3. Field survey results indicate closed windows dose rates > 1000mR/hr expected to continue for more than one hour; or analyses of field survey samples indicate thyroid CDE of 5000mR for one hour of inhalation, at or beyond site boundary 1

1 Basis:

To achieve the dose for this initiating condition, core damage with a failure of all the fission product barriers is necessary. Based on FSAR isotopic distribution fbr accidents and very conservative worst case" meteorological data, PAG limits cannot be reached without some amount of fuel damage. In classifying this event, verifying that core damage is suspected or has occurred, precludes erroneous protective netion recommendations based on incorrect or default dose assessments when plant conditions clearly do not support the magnitude of the release.

1 1

! 5

Em:rgency Action Level Basis Document ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Gaseous EfHuents, continued Basis, continued A RM-Al or RM-A2 mid-range monitor reading exceeding the values on the above Table would result in a Site Boundary Dose of >1000 mR TEDE using conservative meteorology (1 m/see windspeed, G Stability class, no precipitation as outlined on the attached Table).

The 1000 mR TEDE and the 5000 mR Thyroid CDE are based on the EPA protective action guidance which indicstes that public protective actions are indicated if the dose exceeds 1000 mrem TEDE or 5000 mrem Thyroid CDE. This is consistent with the emergency class description for a General Emergency.

Actual meteorology (including forecasts) should be used whenever possible.

Classification for items 2 & 3 above result from emergency response team input. For example, the Environmental Survey Team provides actual dose rates which are.used to determine doce for the projected duration of the release. The Dose Assessment Team provides projected dose.

CR-3 Matrix Reference Number:1.4 NEI 97-03

Reference:

AG1 1

l l

6

Em:rgency Action Level Basis Document ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Liquid Effluents MODES:ALL Classification:

Unusual Event j 1

Initiating Condition: j l

An UNPLANNED release ofliquid radioactivity to the environment exceeding 2 times the ODCM release  !

setpoint for 60 minutes or longer Emergency Action Level:

(1 or 2)

1. A VALID reading on RM-L2, RM-L7, or sample analysis confirms the release exceeds 2 times the ODCM release setpoint for 60 minutes or longer I og '

2. Release continued for 60 minutes or longer with no dilution flow

. Basis:

This EAL is based on failure of the monitor interlock to perform its function or loss of dilution flow. If the interlock failed, a factor of 2 times the release setpoint as compared to actual readings, can be used to judge if the EAL is exceeded. For other conditions, an evaluation ofliquid effluent radioactivity must be performed and compared against the ODCM release setpoint to determine entry conditions.

Releases in excess of 2 times the ODCM limits continuing for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose is not the primary concern here; it is the degradation in plant control implied by the fact the release was not isolated within 60 minutes. Therefore, it is not intended for the release to be averaged over 60 minutes. For example, a release of 4 times the ODCM limits for 30 minutes does not exceed this initiating condition. Further, the Emergency Coordinator should not wait until 60 minutes elapses, but declare the event as soon as it is determined the release duration will likely exceed 60 minutes. An evaluation is necessary to compare monitor setpoint against the EAL limit.

CR-3 Matrix Reference Number:1.5 NEI 97-03

Reference:

AU1 7

Emerg:ncy Action Level Basis Document ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Liquid Effluents MODES:ALL i

Classification: {

-Alert Initiating Condition:

An UNPLANNED release of liquid radioactivity to the environment exceeding 200 times the ODCM release setpoint for 15 minutes or longer l !

Emergency Action Level:

A VALID reading on RM-L2, RM-L7, or sample analysis confirms the release exceeds 200 times the ODCM release setpont for 15 minutes or longer l

Basis:

This EAL is based on loss of dilution flow or failure of the monitor interlock to perform its function. If the interlock failed, a factor of 200 times the release setpoint as compared to actual readings, can be used to judge if the EAL is exceeded. For other conditions, an evaluation ofliquid effluent radioactivity must be performed and compared against the ODCM release setpoint to determine entry conditions. For example, a c~omparison would be necessary if dilution flow was lost.

CR-3 Matrix Reference Number:1.6 NEI 97-03

Reference:

AA1 8

Emergency Action Level Basis Document ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Unexpected Radiation Levels MODES:ALL Classification:

Unusual Event Initiating Condition:

An unexpected increase in radiation levels within the plant Emergency Action Level:

One or more VALID Area Radiation Monitor readings unexpectedly exceeds the values listed below for 15 minutes or longer:

RM-G3 = 400 mR/hr RM-G4 = 600 mR/hr RM-G5 = 3000 mR/hr ,

l RM-GS = 100 mR/hr RM-G10 = 800 mR/hr RM-G14 = 1000 mR/hr l RM-G17 = 800 mR/hr Basis:

This EAL addresses unexpected increases in in-plant radiation levels representing a degradation in the control of radioactive material, and a potential degradation in the level of safety of the plant.

The values above represent approximately 1000 times normal monitor levels based on nominal historical dita of the monitors during normal plant operation. Portable surveys may be substituted for in plant radiation monitors. The specific area radiation monitors were chosen as they represent potential release tu eas within the plant and/or access corridors to the plant.

Assessment should be completed such that after the 15 minutes elapsed time, a classification decision thould be made.

Monitor Locations:

RM-G3 (Primary Sample Room)

RM-G4 (Auxiliary Building entrance corridor)

RM-05 (Waste Gas Deyky Tank Area)

RM-G9 (Intermediate ' Building outside Reactor Building (RB) personnel airlock)

RM-G10(Makeup Pump area)

RM-G14(Spent Fuel Pool Storage Area - 143' elev. Aux. Bldg. general area)

RM-G17(inside RB at personnel hatch)

CR-3 Matrix Reference Number: 1.7 NEI 97-03

Reference:

AU2 9

l

Em:rg:ncy Acti:n Level Baris Docum:nt ABNOEMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Unexpected Radiation Levels MODES: ALL Classification:

Alert Initiating Condition:

An unexpected increase in radiation levels within the plant impeding operation of systems required to maintain safe operations or to establish or maintain cold shutdown Emergency Action Level:

(1 or 2)

1. VALID radiation reading greater than 15 mR/hr for 15 minutes or longer in the Control Room (RM-G1) or the Central Alarm Station (CAS)

OR

2. VALID Area Radiation Monitor reading unexpectedly exceed one or more of the values listed below for 15 minutes or longer:

RM-G3 = 5,000 mR/hr RM-G4 = 5,000 mR/hr RM-G9 = 5,000 mR/hr RM-G10 = 5,000 mR/hr ,

RM-G17 = 5,000 mR/hr Basis:

This addresses increased radiation levels ir" , necessary access to operating stations, or other areas containing equipment operated manuall;

• 1er to maintain safe operation or perform a safe shutdown. The specific area radiation mon -

were chosen as they represent access corridors to the pl nt. These monitors cover general areas tnat would require access to maintain safe operations or to establish and maintain safe shutdown. It is this impaired ability to operate the plant that results in the c.ctual or potential substantial degradation of the level of safety of the plant. The cause and/or m7gnitude of the increase in radiation levels is not a concern of this initiating condition. The Emergency Coordinator must consider the source or cause of the increased radiation levels and determine if any other Initiating Condition is involved. For example, a dose rate of 15 mR/hr in the control room may be a problem in itself. However, the increase may also be indicative of high dose rates in the containment due to a LOCA. In this latter case, a Site Area Emergency or General Emergency may be indicated by the Fizsion Product Barrier Matrix Initiating Conditions.

10

Em:rg:ncy Action Level Basis Document j

i ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Unexpected Radiation Levels, continued Basis, continued Portable surveys may be substituted for in plant radiation monitors. A generic emergency action level at gr:ater than 5,000 mR/hr has been chosen for those areas in the plant that would need to be accessed for cr.fe operation or safe shutdown of the unit.

l <

M:nitor Locations:

RM-G3 (Primary Sample Room)

RM-G4 (Auxiliary Building entrance corridor)

RM-G9 (Intermediate Building outside RB personnel airlock)

RM-G10 (Makeup Pump area)

RM-G17 (inside RB at personnel hatch)

- Assessment should be completed such that after the 15 minutes elapsed time, a classification decision thould be made. This Initiating Condition is not intended to apply to anticipated temporary increases due to planned events (e.g., incore detector movement, radwaste container movement, depleted resin tre.nsfers, etc.)

i CAS dose rates are determined by portable monitors.

Areas requiring continuous occupancy includes the control room and any other control stations that are manned continuously, such as the Central Alarm Station. 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 III.D.3 of NUREG-0737, " Clarification of TMI Action Plan Requirements," provides that the 15 mR/hr v:lue can be averaged over the 30 days, the value is used here without averaging, as a 30 day duration implies an ever; potentially more significant than an Alert.

CR-3 Matrix Reference Number:1.8 NEI 97-03

Reference:

AA3 11

I Em:rg:ncy Action Level Basis Document l- ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT I Fuel Handling / Fuel Handling Pool Water Level MODES:ALL 3 Classification:

Unusual Event Initiating Condition:

An uncontrolled water level decrease in spent fuel pool o_r transfer canal with fuel remaining covered Emergency Action Level:

(1 and 2) l

1. (a or b)

a. Uncontrolled level decrease resulting in indications of-2.5 feet in spent fuel pool og

b. Plant personnel report water level drop in spent fuel pool pr fuel transfer canal l

l AND l

i 2. Fuel remains covered with water Basis:

The "-2.5 feet" indication is relative to the normal "zero" reading for spent fuel pool level and represents the minimum 23 feet of water (156 feet plant datum) over the top of the fuel as described in Improved Technical Specifications.

CR-3 Matrix Reference Number:1.9 l NEI 97-03

Reference:

AU2 l

l 12

Emerg:ncy Action Level Basis Document ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT Fuel Handling / fuel Handling Pool Water Level MODES: ALL Classification:

Alert Initiating condition:

Major damage to irradiated fuel n loss of water level has or will uncover irradiated fuel outside the rzactor vessel Emergency Action Level:

(1 or 2)

1. (a and b)

a. Plant personnel report damage ofirradiated fuel AND

b. VALID high alarm as indicated on RM-015 or RM-G16 i

OR l

2. Plant personnel report spent fuel pool or transfer canal water level drop has g will exceed makeup l capacity such that irradiated fuel will be uncovered l

Basis:

There is time available to take corrective actions, and there is little potential for substantial fuel damage.

Thus, an Alert classification for this event is appropriate. Escalation, if appropriate, would occur via Abnormal Rad Levels or Emergency Coordinatorjudgment.

Monitor Locations:

RM-G15 (Auxiliary Building Fuel IIandling Bridge)

RM-G16 (Reactor Building Fuel Handling Bridge)

CR-3 Matrix Reference Number:1.10 NEI 97-03

Reference:

AA2 13

Emergency Action Lovel Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT 1

Earthquake Experienced MODES: ALL 1

Classificationj Unusual Event Initiating Condition:

! Emrthquake detected by seismic instrumentation l

l Emergency Action Level:

(1 and 2) i

1. Ground motion sensed by plant personnel AND

2. Confirmed earthquake causing Annunciator C-3-14 " Seismic System Trouble" alarm Basis:

Damage may be caused to some portions of the site, but should not affect ability of safe shutdown equipment to operate. Method of detection is based on instrumentation, validated by a reliable source, or operator assessment. As defined in the EPRI-sponsor: d, " Guidelines for Nuclear Plant Response to an Earthquake," dated October 1989, a " felt earthquake"is:

"An earthquake of sufficient intensity such that: (a) the ground motion is felt at the I nuclear plant site and recognized as an earthquake based on a consensus of control room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated."

CR-3 Matrix Reference Number:2.1 I NEI 97-03

Reference:

HU1 14

Emcrgency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Earthquake Experienced MODES: ALL Classification:

Alert Earthquake detected by seismic instrumentation (1 and 2)

1. Ground motion sensed by plant personnel or confirmed Annunciator C-3-14 " Seismic System Trouble" clarm AND

2. (a or b)

a. Analysis confirms the earthquake at >0.05g

!LR

b. Indications show degraded SAFE SHUTDOWN EQUIPMENT performance due to the earthquake Basis: j Seismic events of this magnitude can cause damage to safety functions.

- Analysis of earthquakes is completed using AP-961 and its supporting procedures. The analysis to determine the magnitude of an earthquake may take an extended period of time. Ifit is determined even after several hours that the earthquake was >0.05g, the event should be classified.

This EAL is intended to address an earthquake resulting in a plant vital area being subjected to forces beyond design limits, and thus damage is assumed to have occurred to plant safe shutdown equipment.

The initial report is not interpreted as mandating a lengthy damage assessment before classification and no attempt is made to assess the actual magnitude of the damage, l

Additional information on the earthquake (confirmation and magnitude) can be obtained from the U. S. l Geological Survey - Golden, Colorado at (303) 273-8500.  !

If damage from the earthquake is clearly contained and localized to one train, then safe shutdown equipment is not affbeted and item 2b of the EAL is not met. If the extent of the damage is uncertain in terms ofloss of safe shutdown equipment, then entry into this EAL is required.

CR 3 Matrix Reference Number:2.2 NEI 97-03

Reference:

HA1 15

Em:rgency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Flood MODES:ALL Classification:

Unusual Event Initiating Condition:

Flood being experienced Emergency Action Level:

Intake canal level or visual observation indicates flood water level 2 98 feet Basis:

This EAL covers flooding due to natural phenomena. This EAL can be a precursor of more serious events. In particular, since CR-3 may be subject to severe weather as defined in the NUMARC station blackout initiatives, this includes action based on activation of the severe weather mitigation procedures for flooding (e.g., precautionary shutdowns, diesel testing, staff call-outs, etc.).

Ninety-eight (98) feet is contained within the discharge and intake canals banks. The top of the concrete wall at the intake structure is 99 feet.

The highest water level recorded at CR-3 was 99.5 feet during the 03/13/93 "No Name Storm."

At 98 feet, there is no immediate impact on plant equipment but heightened awareness is appropriate chould the level increase.

CR-3 Matrix Reference Number:2.3 NEI 97-03

Reference:

HU1 i

1 16

Emergency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Flood MODES:ALL Classification:

Alert Initiating Condition:

Flood being experienced Emergency Action Level:

(1 and 2)

1. Intake canal level or visual observation indicates flood water level 2 98 feet AND

2. Indications show degraded SAFE SHUTDOWN EQUIPMENT performance due to the flooding Basis:

This EAL covers flooding due to natural phenomena.

This EAL is intended to address flooding that may have resulted in a plant vital area being subjected to forces beyor.d design limits, and thus damage may be assumed to have occurred to plant safety systems. I The initial report is not interpreted as mandating a lengthy damage assessment prior to classification I and no a4empt is made to assess the actual magnitude of the damage. I If damage from the flooding is clearly contained and localized to one train, then safe shutdown equipment is not affected and item 2 of the EAL is not met. If the extent of the damage is uncertain in terms ofloss of safe shutdown equipment, then entry into this EAL is required.

CR-3 Matrix Reference Number:2.4 NEI 97-03

Reference:

HA1 17

Emergency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Hurricane MODES:ALL ClassiScation:

Unusual Event Initiating Condition:

Hurricane Warning Emergency Action Level:

The plant is within a Hurricane Warning area Basis:

This EAL can be a precursor of more serious events. In particular, since CR-3 may be subject to severe weather as defined in the NUMARC station blackout initiatives.

This should include a notification from the National Hurricane Center via the State Warning Point.

CR-3 Matdx Reference Number:2.5 NEI 97-03

Reference:

HU1 18

Emergency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT To0nado MODES:ALL Classification:

Unusual Event Initiating Condition:

Tornado within the Protected Area Emergency Action Level:

Report by plant personnel of a Tornado striking within the Protected Area Basis:

This EAL is based on the assumption a tornado strikes (touches down) within the protected area boundary and may have potentially damaged plant structur containing functions or systems required l for safe shutdown of the plant. If such damage is confirmed vu, tily or by other in-plant indications, the event may be escalated to un Alert. l Waterspouts remaining intact after coming onshore / land are classified as tornadoes.

I CR-3 Matrix Reference Number:2.6 NEI 97-03 Reference HU1 l

19

Em rgency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Tcrnado/High Winds i

MODES:ALL Classi8 cation:

A1:rt I

l Initiating Condition:

Tornado pr High Winds or windborne object (s) strike structures and results in significant visible damage Emergency Action I4 vel:

(1 and 2)

1. Tornado pr High Winds or windborne object (s) strike one of the following structures:

. Auxiliary Building,

.BWST, e Control Complex, i

. EFT-2 Building,

. Diesel Generator Building,

. Intermediate Building,

. Reactor Building l AND

2. (a or b)

a. Confirmed report of significant visible damage to buildings listed above l

OR 1

b. Indications show degraded SAFE SHUTDOWN EQUIPMENT performance due to Tornado, high winds, or windborne objects.

i Basis.

This EAL addresses events that may have resulted in a plant vital area being subjected to forces beyond de:ign limits, and thus damage may be assumed to have occurred to plant safety systems. The initial report is not interpreted as mandating a lengthy damage assessment before classification and no attempt is made to assess the actual magnitude of the damage.

The highest recorded sustained windspeed at CR-3 during the 03/13/93 "No Name Storm" was 56 mph.

20 l

s Emergency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Tornado /High Winds, continued Basis, continued Waterspouts remaining intact after coming onshore / land are classified as tornadoes.

If damage from the tornado /high winds is clearly contained and localized to one train, then safe shutdown equipment is not affected and item 2b of the EAL is not met. If the extent of the damage is uncertain in terms ofloss of safe shutdown equipment, then entry into this EAL is required.

CR-3 Matrix Reference Number:2.7 NEI 97-03

Reference:

HA1 l

1 21

Emcrg:ncy Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Aircraft / Vehicle Craah MOLUS:ALL Classification:

Unusual Event Initiating Condition:

Aircraft pr Vehicle crash within the Protected Area potentially damaging plant structures containing functions and systems required for safe shutdown of the plant Emergency Action Level:

Report by plant personnel of Aircraft pr Vehicle Crash involving any of the' following permanent ctructures within the Protected Area:

l

. Auxiliary Buildmg, j

.BWST, )

. Control Complex,

• EFT-2 Building, .

. Diesel Generator Building,

. Intermediate Building,

. Reactor Building Basis:

This EAL is intended to address such items as a plane, helicopter or vehicle crash, potentially damaging plant structures containing functions and systems required for safe shutdown of the plant. Automobiles, trucks, and forklifts are vehicles within the context of this EAL. The intent is to address any vehicle that can cause significant damage to plant structures.

CR-3 Matrix Reference Number:2.8 NEI 97-03

Reference:

HU1 i

22

7 Emerg:ncy Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Aircraft / Vehicle Crash MODES:ALL Classification:

Alert Initiating Condition:

Aircraft o_r Vehicle strikes vital structures and results in significant visible damage Emergency Action Level:

(1 and 2)

1. Confirmed report of significant visible damage to any of buildings listed below:

. Auxiliary Building,

.BWST,

. Control Complex,

. EFT-2 Building,

• Diesel Generator Building, I

• Intermediate Building,

. Reactor Building AND

2. Indications show degraded SAFE SHUTDOWN EQUIPMENT performance due to Aircraft pr Vehicle Crash Basis:

This EAL is intended to address such items as a plane or helicopter or vehicle crash damaging plant ctructures containing functions and systems required for safe shutdown of the plant. Automobiles, trucks, and forklifts are also vehicles within the context of this EAL. Significant damage refers to ctructural damage that is beyond cosmetic dama7e.

This EAL is intended to address events that may have resulted in a plant vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems.

The initial report is not interpreted as mandating a lengthy damage assessment prior to classification cnd no attempt is made to asses the actual magnitude of the damage.

If damage from the vehicle or aircraft crash is clearly contained and localized to one train. then safe shutdown equipment is not affected and the EAL is not met. If the extent of the damage is uncertain in  !

terms ofloss of safe shutdown equipment, then entry into this EAL is required.

CR-3 Matrix Reference Number:2.9 NEI 97-03

Reference:

HA1 t

23

Emergency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Toxic or Flammable Gas MODES:ALL Classification:

Unusual Event Initiating Condition:

Release of Toxic or Flammable Gas within, or potentially affecting the Protected Area Emergency Action Level:

)

(1 or 2)

1. Report or detection of toxic or flammable gases within the SITE BOUNDARY that could er.ter the Protected Area in amounts that can affect normal operation of the plant OR

2. Confirmed notification by FPC, County, or State personnel of a potential evacuation or sheltering of site personnel based on an offsite event Basis: l i

This Initiating Condition is based on releases in concentrations within the Site Boundary that could; (1) l l

affect the within put the plant health and safety an evacuation of plant or sheltering personnel; area (2)event.

due to an offsite affect the safe operation of th Gases within the Site Boundary that are below life-threatening R IDLH) or flammable concentrations l are not applicable to this Initiating Condition. Concentrations at these levels would not affect plant personnel or the safe operation of the plant. Gases at the site boundary that are above life-threatening or flammable concentrations, yet have not exceeded those concentrations within a facility structure, w:uld satisfy the first EAL and would require the declaration of an Unusual Event.

Toxic County,or Flammable gases which arepotential released offsite (e.g., transportation accident) of the confi Local, or State personnel have the for requiring the evacuation or sheltering Owner Controlled Area (Site Boundary).

A localized /small-scale event within the Site Boundary that may involve gases at life threatening or fkmmable concentrations do not meet the intent of this Initiating Condition.

CR-3 Matrix Reference Number:2.10 NEI 97-03

Reference:

HUS 24

Emergency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Toxic or Flammable Gas MODES:ALL Classification:

Alert Initiating Condition:

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 Emergency Action Level:

(1 or 2)

L Flammable Gas levels >25% Lower Explosive Limits 9E l

2. Toxic gas levels 2 IDLH levels within the Protected Area such that plant personnel are unable to l perform actions necessary to maintain safe operations or establish and maintain cold shutdown using personal protective equipment Basis:

This ' Initiating Condition is based on gases that have entered a plant structure affecting the safe operation oflthe plant. This Initiating Condition applies to buildings and areas contiguous to plant vital creas or other significant buildings or areas.

Concentrations at these amounts will restrict or prevent normal actions from being taken to operate the  ;

plent. This EAL is not intended to include precautionary general evacuation of personnel. j I

For toxic gases, this initiating condition only applies to areas that do not require continuous occupancy.

If personnel can safely enter areas necessary to establish and maintain safe shutdown using protective  !

equipment, this Initiating Condition /EAL is not met.

IDLH - Immediately Dangerous to Life or Health CR-3 Matrix Reference Number:2.11 NEI 97-03

Reference:

HA3 25

Emergency Action Lcvel Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Explosions / Catastrophic Pressurized Equipment Failure MODES:ALL Classification:

Unusual Event i

Initiating Condition: I UNPLANNED EXPLOSION within the Protected Area Emergency Action I4 vel: 1 Report by plant personnel of VISIBLE DAMAGE to permanent structures pr equipment within the PROTECTED AREA due to an EXPLOSION or catastrophic failure of pressurized equipment Basis:

For this EAL, only those explosions of sufficient force to damage permanent structures or equipment within the protected area should be considered. 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.,

d formation, scorching)is sufficient for declaration.

This EAL is not intended to cover small steam line breaks, small pipe cracks, or small steam / feed leaks.

The Emergency Coordinator also needs to consider security aspects of the explosion and, if applicable, refer to the security EALs.

CR-3 Matrix Reference Number:2.12 NEI 97-03

Reference:

HU1 i

1 26

l Emergency Action Level Basis Document l

l NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT l

Explosions / Catastrophic Pressurized Equipment Failure l

MODES:ALL Classification:

Alert Initiating Condition:

EXPLOSION within the plant affecting the operability of plant safety systems required to establish or I maintain safe shutdown Emergency Action Level:

(1 and 2) 1

1. EXPLOSION or catastrophic failure of pressurized equipment in any of the following structures:

. Auxiliary Building,

.BWST,

• Control Complex,

. EFT-2 Building,

• Diesel Generator Building,

. Intermediate Building,

• Reactor Building

2. (a or b) l a. Report by plant personnel of EXPLOSION or catastrophic failure of pressurized equipment l causing VISIBLE DAMAGE to SAFE SHUTDOWN EQUIPMENT l

l OR l

l l

b. Indications show degraded SAFE SHUTDOWN EQUIPMENT performance due to the Explosion

, or pressurized equipment failure 1

Basis:

i This EAL is intended to address events that may have resulted in a plant vital area being subjected to j forces beyond design limits, and thus damage may be assumed to have occurred to plant safe shutdown equipment. The initial report 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 the damage.

The observation of damage to a structure is sufficient to make a declaration.

27 i

(

Emergency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Explosions / Catastrophic Pressurized System Failure, continued Basis, continued If damage from the explosion or catastrophic failure of pressurized equipment is clearly contained and localized to one train, then safe shutdown equipment is not afrected and the EAL is not met. If the extent of the damage is uncertain in terms ofloss of safe shutdown equipment, then entry into this EAL is required.

CR-3 Matrix Reference Number:2.13 NEI 97-03

Reference:

HA2 l

1 l

I l

28

Emergency Action Level Basia Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Fire MODES:ALL Classification:

Unusual Event Initiating Condition FIRE within the PROTECTED AREA that could affect SAFE SHUTDOWN EQUIPMENT Emergency Action Level:

(1 and 2)

1. FIRE in or threatening one of the following structures:

• Auxiliary Building,

• BWST,

. Control Complex,

• EFT-2 Building,

• Diesel Generator Building,

. Intermediate Building,

. Reactor Building AND

2. FIRE not extinguished within 15 minutes from Control Room notification o_r receipt of a verified Control Room fire alarm Basis:

This EAL 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-b:sket fires, and other small fires of no safety consequence. This Initiating Condition applies to buildings and areas contiguous to plant vital areas or other significant buildings or areas.

Verification of the alarm in this context means those actions taken in the control room or other location to determine the control room alarm is not spurious.

Fire in other areas adjacent to vital areas may warrant classification if the fire is of a magnitude that threatens vital areas.

29 ,

Emerg:ncy Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Fire, continued Basis, continued The 15 minute time period begins with the time when a credible notification that a fire is occurring or a verified fire detection system alarm is received. The intent of the 15 minute duration is to discriminate against small fires that are readily extinguished.

CR-3 Matrix Reference Number:2.14 NEI 97-03

Reference:

HU2 l

l l

l l

l l

1 l ,

l '

l l

30

Emcrgency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT l

Fire MODES:ALL Classification:

A1:rt Initiating Condition:

' FIRE affecting the operability of SAFE SHUTDOWN EQUIPMENT i

Emergency Action Level: l (1 or 2)

1. Report by plant personnel of VISIBLE DAMAGE to SAFE SHUTDOWN EQUIPMENT due to the FIRE I

I OR

2. Indications show degraded SAFE SHUTDOWN EQUIPMENT performance due to the FIRE 1 i

Basis:

The key to classifying fires as an Alert is the damage as a result of the incident. The fact that the equipment required for safe shutdown of the unit has been affected or damaged as a result of the fire is thm driving force for declaring the Alert.

If damage from the fire is clearly contained and localized to one train, then safe shutdown equipment is not affected and the EAL is not met. If the extent of the damage is uncertain in terms of loss of safe shutdown equipment, then entry into this EAL is required.

l l

CR-3 Matrix Reference Number:2.15 NEI 97-03

Reference:

HA2  !

l l

l l

31 u

r 1 1

Emergency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Control Room Evacuation

, MODES: ALL ClassiBeation:

A1:rt Initiating Condition:

Evacuation of Control Room is Required Emergency Action Level:

Control Room evacuation is required per AP-990," Shutdown Outside of the Control Room" l

I

_ Basis:

With the control room evacuated, additional support, monitoring and direction through the Technical Support Center and/or the Emergency Operations Facility is necessary.

l Declaration of an Alert may be delayed until the transfer to remote shutdown is completed. This is appropriate since establishing control of the plant takes precedence.

CR-3 Matrix Reference Number:2.16 NEI 97-03

Reference:

HA5 i

32

Em::rgency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Control Roorn Evacuation MODES: ALL Classification:

Site Area Emergency Initiating Condition:

Evacuation of Control Room is Initiated and Plant Control cannot be established Emergency Action Level:

(1 and 2)

1. Control Room evacuation is required per AP-990," Shutdown Outside of the Control Room" 1 AND j j

2. Control of the necessary equipment nqt established per AP-990 within 15 minutes Basis:

The 15 minutes begins at the first attempt to turn the transfer switch to transfer control from the Main Control Room to the Remote Shutdown Panel.

The timely transfer of control to alternate control areas has not been accomplished. The failure to transfer control would be evidenced by deteriorating reactor coolant system or steam generator parameters.

The determination of whether or not control is established at the Remote Shutdown Panel is based upon the judgment of the Nuclear Shift Manager. The Nuclear Shift Manager is expected to make a reasonable, informed judgment within fifteen minutes of the transfer from the Control Room that the l operating crew has control of the plant from the Remote Shutdown Panel.

CR-3 Matrix Reference Number:2.17 NEI 97-03

Reference:

HS2 I

33  !

Emcrgency Action Level Ensis Docum;nt NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Security Event MODES: ALL Classification:

Unusual Event Initiating Conditions:

Confirmed Security event which indicates a potential degradation in the level of safety of the plant f

Emergency Action Level:

l (a or b or c)

Report by Security Shift Supervisor of one or more of the following events:

J

a. Occurrence of SABOTAGE / INTRUSION O_g

b. HOSTAGE / EXTORTION situation or hostile strike action threatening to interrupt plant operations QB

c. A violent CIVIL DISTURBANCE ongoing outside the PROTECTED AREA but within the Owner Controlled Area (SITE BOUNDARY)

Basis:

This EAL is based on CR-3 Physical Security Plan. Security events which do not represent at least a potential degradation in the level of safety of the plant, are reported under 10 CFR 73.71 or in some cases ,

under 10 CFR 50.72. i CR-3 Matrix Reference Number:2.18 l

NEI 97-03

Reference:

HU4 j 34

Emergency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Security Event MODES:ALL Classincation:

Alert Initiating Condition:

Confirmed Security Event in a plant Protected Area.

Emergency Action Level:

(1 or 2)

1. Discovery of BOMB within the PROTECTED AREA  ;

l OR j i

2. INTRUDER (S) penetrates the PROTECTED AREA Basis:

This class of security events represents an escalated threat to plant safety above that contained in the Unusual Event. For the purposes of this Initiating Condition, a civil disturbance which penetrates the  !

protected area boundary can be considered an intrusion.

CR-3 Matrix Reference Number:2.19 NEI 97-03

Reference:

HA4 35

Emergency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Security Event MODES: ALL ClassiBeation:

Site Area Emergency Initiating Condition:

Confirmed Security Event in a plant Vital Area.

Emergency Action Level:

1. INTRUDER (S) penetrates or _.IB is discovered in any of the areas listed below:

. Auxiliary Building,

.BWST,

. Control Complex,

. EFT-2 Building,

. Diesel Generator Building,

. Intermediate Building, eReactor Building Basta:

This class of security events represents an escalated threat to plant safety above that contained in the Alert IC in that an intruder has progressed from the Protected Area to a Vital Area.

If an intruder or intruders penetrates or a bomb is detonated in the Control Room or Remote Shutdown Room, a General Emergency would be declared.

CR-3 Matrix Reference Number:2.20 NEI 97-03

Reference:

HS1 l

l 36

Emerg:ncy Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Security Event MODES: ALL ClassiBcatian:

General Emergency Initiating Condition:

Security Event resulting in loss of physical control of the facility Emergency Action Level:

INTRUDER (S) has taken control of the Control Room, or Remote Shutdown Room or plant equipment such that plant personnel are unable to operate equipment required to establish and maintain safe shutdown conditions Basis:

This Initiating Condition encompasses conditions under which a hostile force has taken physical control of a vital area or equipment required to reach and maintain safe shutdown. If control of the plant equipment necessary to maintain safety functions can be transferred to another location. then the above limiting condition is not met. Loss of physical control of the Control Room or remote shutdown capability alone may not prevent the ability to maintain safety functions per se.

CR-3 Matrix Reference Number:2.21 NEI 97-03

Reference:

HG1 l

37

En gncy Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Internal Flooding MODES: ALL Classification:

Unusual Event Initiating Condition:

Internal flooding affecting areas containing safe shutdown equipment Emergency Action Level:

(1 and 2)

1. Indication of uncontrolled flooding in the Auxiliary Building or Intermediate Building AN_D

2. Water level / flooding has the potential to affect or immerse SAFE SHUTDOWN EQUIPMENT Basis:

This addresses the possible effects of flooding from system malfunctions, component failures, or repair activity mishaps that could threaten the safe operation of the plant. The flooding could affect equipment not designed to be submerged.

CR-3 Matrix Reference Number:2.22 NEI 97-03

Reference:

HU1 l

1 38

l Emergency Action Level Baris Document l

NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT l

Internal Flooding MODES:ALL Classification:

A1:rt Initiating Condition:

Internal flooding affecting SAFE SHUTDOWN EQUIPMENT Emergency Action Level:

(1 and 2)

1. Water level exceeds 1.5 feet in the Auxiliary Building or latermediate Building AND J

l

2. (a or b) st. Indications show degraded SAFE SHUTDOWN EQUIPMENT due to the flooding OR

b. Electrical hazards prevent plant personnel normal access to areas of plant containing SAFE SHUTDOWN EQUIPMENT Basis:

This addresses the possible eheets of flooding from system malfunctions, component failures, or repair cctivity mishaps that has either threatened the safe operation of the plant or resulted in a complete loss of function required for cold shutdown.

The water level was selected based on a level when Motor Control Centers would experience water  !

intrusion. 1 If damage from the internal flooding is clearly contained and loc lized to one train, then safe shutdown ]

equipment is not affected and item 2 of the EAL is not met. If the extent of the damage is uncertain in i terms ofloss of safe shutdown equipment, then entry into this EAL is required.

CR-3 Matdr Referer.r- Number: ?.23 NEI 97-03

Reference:

HA1 39

l Emergency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Emergency Coordinator Judgment MODES:ALL Classification:

Unusual Event ,

Initiating Conditions:

Other conditions existing, which in the judgment of the Emergency Coordinator, warrant declaration of an Unusual Event Emergency Action Level:

Other conditions exist which indicate a potential degradation in the level of safety of the plant s Basis:

This EAL addressea unanticipated conditions not addressed explicitly elsewhere but warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the Unusual Event emergency class.

This EAL should also be referenced if, in th( judgment of the Emergency Coordinator, an Unusual Event should be classified if plant symptoms are less than the threshold of an existing EAL.

CR-3 Matrix Reference Number:2.24 NEI 97-03

Reference:

HU5 l 1

i 40

Emergency Action Level Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Emergency Coordinator Judgment MODES:ALL Class 18 cation:

A1:rt Initiating Conditions:

Other conditions exist, which in the judgment of the Emergency Coordinator, warrant declaration of an Alert Emergency Action Level:

Other conditions exist which indicate that events are in process or have occurred which involve potential or actual substantial degradation of the level of safety of the plant Basis:

This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the Alert emergency class.

Any release is expected to be limited to small fractions of the EPA plume Protective Action Guideline Exposure Levels.

This EAL should also be referenced if,in the judgment of the Emergency Coordinator, an Alert should be classified if plant symptoms are less than the threshold of an existing EAL.

CR-3 Matrix Reference Number:2.25 NEI-97-03

Reference:

HA6 i

41

1

)

Em:rgency Action Levd Bania Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT I

l Emergency Coordinator Judgment MODES:ALL Classification:

Site Area Emergency Initiating Conditions:

Other conditions exist, which in the judgment of the Emergency Coordinator, warrant declaration of a Site Area Emergency Emergency Action Level:

Other conditions exist. which indicate actual or likely major failures of plant functions needed for the protection of the public Basis:

This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that i wtrrant declaration of an emergency because conditions exist which are believed by the Emergency i Coordinator to fall under the emergency class description for Site Area Emergency.

A release is not expected to result in exposure levels exceeding EPA plume Protective Action Guideline Exposure Levels beyond the Site Boundary (1 Rem TEDE or 5 Rem Thyroid CDE).

The Site Boundary is defined as 0.83 miles radially from the center of the Reactor Building.

This EAL should also be referenced if, in the judgment of the Emergency Coordinator, a Site Area Emergency should be classified if plant symptoms are less than the threshold of an existing EAL.

CR-3 Matrix Reference Number:2.26 NEI 97-03

Reference:

HS3 l

42

Emergency Action Lcvel Basis Document NATURAL / MAN-MADE HAZARDS AND EC JUDGMENT Emergency Coordinator Judgment MODES: ALL Classification:

G:neral Emergency Initiating Condition:

Other conditions exist, which in the judgment of the Emergency Coordinator, warrant declaration of a l G:neral Emergency Emergency Action Level:

(1 or 2)

I Other conditions exist which indicate:

1. Actual or imminent substantial core degradation with potential for loss of containment integrity O__R

2. The potential for uncontrolled radionuclide releases that can be expected to exceed EPA Protective Action Guidelines Plume Exposure Levels beyond the SITE BOUNDARY Basis: l This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency l Coordinator to fall under the General Emergency class. I Relea ses can reasonably be expected to exceed EPA Protective Action Guidelines Plume Exposure Levels beyond the Site Boundary (1 Rem TEDE or 5 Rem Thyroid CDE).

CR-3 Matrix Reference Number:2.27 NEI 97-03

Reference:

HG2 l

l 1

43

e Em:rg:ncy Action Lev:1 Ba:i2 Document SYSTEM MALFUNCTION Loss of Communication MODES:ALL J Classification:  ;

Unusual Event Initiating Condition:

Unplanned loss of all In-Plant or all offsite Communication capability l Emergency Action Level:

(1 or 2)

1. Loss ord the following in-plant communications capability:

• FPC Internal Telephone System

. PAX e Portable UHF Radios OR

2. Loss of d the following offsite communications capability: .

. FPC Telephone System l

. State Hot Ring Down (SHRD)

. State-Wide Emergency Satellite Communication (ESATCOM) System j

. ALL FTS 2000 NRC Phones (Emergency Notification System (ENS), Health Physics Network (HPN), Counterpart Links)

. Control Room Cellular Phone Basis:

The purpose of this Initiating Condition and its associated EALs is to recognize a loss of communications capability either defeating the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72.

The onsite or offsite communications loss must encompass the loss of all means of routine direct communications with intended parties. This includes the ENS, Commercial lines, Microwave, and FAX transmissions. This EAL is used only when extraordinary means are used to make communications possible (relaying of information from radio transmissions, individuals being sent to offsite locations, etc.). Credit is not taken for portable satellite phones due to the time it takes to establish a communications link. Once a link is established with a portable satellite phone, the event may be terminated.

CR-3 Matrix Reference Number:3.1 NEI 97-03

Reference:

SU6 44

Emtrg:ncy Action Level Basis Document SYSTEM MALFUNCTION Failure of Reactor Protection MODES: 1,2,3 Classification:

Alert Initiating Condition:

Frilure of Reactor Protection System (RPS) instrumentation to complete or initiate an automatic reactor trip once an RPS setpoint has been exceeded and manual trip was successful Emergency Action Level:

(1 and 2)

1. RPS Trip actpoint exceeded and no Reactor trip occurred AND

2. Manual Reactor trip from Control Room was successful and the reactor is shutdown Basis: l This condition indicates failure of the Reactor Protection System to trip the reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did .

n:t function in response to a plant transient and thus the plant safety has been compromised, and design j limits of the fuel may have been exceeded.

l An Alert is indicated because conditions exist that lead to potential loss of fuel clad or RCS. Reactor i protection system setpoint being exceeded (rather than limiting safety system setpoint being exceeded) is epecified here because failure of the automatic protection system is the issue. A manual trip is any set of cctions by the reactor operator (s) in the Control Room which causes sufficient coni rol rods to be rapidly inserted into the core and brings the reactor suberitical (e.g., reactor trip bu' mn, de-energizing control rod power from the control room). Operator actions to drive rods or other actions taken or occurring  ;

cutside the control room does not constitute a reactor trip because it does not meet the rapid insertion criterion.

An automatic reactor trip is considered as the RPS tripping the reactor.

CR-3 Matrix Reference Number:3.2 NEI 97-03

Reference:

SA2 l

45

y  ;

3 Em rg:ncy Acti:n Level Bari Docum:nt SYSTEM MALFUNCTION Failure of Reactor Protection MODES: 1,2 Classification:

Site Area Enzergency initiating Condition:

F ilure of Reactor Pa,tection System (RPS) instrumentation to complete or initiate an automatic reactor trip once an RPS setpoint has been exceeded and manual trip was NOT successful Emergency Action Level:

(1 and 2)

1. RPS Trip setpoint exceeded and no Reactor trip occurred AND

2. Manual Reactor trip from Control Room was not successful in shutting down the reactor Basis:

Automatic and manual trips are not considered successful if action away from the Control Room was required to trip the reactor. Manual trip is successful if the trip push button or de-energizing control rod power in the Control Room results in shutting down the reactor. l An automatic reactor trip is considered as the RPS tripping the reactor.

The trip is considered unsuccessful when enough control rods have not inserted to cau a a. reactor power to fall below that percent power associated with the ability of the safety systems to .emove heat end continue to decrease. Subsequent actions necessary for the reactor to be prepared for a cooldown and d: pressurization are not to be considered.

Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because conditions exist that lead to imminent loss or potential loss of both fuel clad and RCS. Although this Initiating Condition may be viewed as redundant to the Fission Product Barrier Matrix, its inclusion is necessary to better assure timely recognition and emergency response.

CR-3 Matrix Reference Number:3.3 NEI 97-03

Reference:

SS2 i 46

Em:rg:ncy Action Level Basis Document SYSTEM MALFUNCTION Failure of Reactor Protection MODESj l.2 Classification:

General Emergency Initiating Condition:

Fcilure of the Reactor Protection System to complete an automatic trip and manual trip was NOT cuccessful and there is indication of extreme challenge to the ability to cool the core Emergency Action Level:

(1 and 2)

1. (a and b)

a. RPS Trip setpoint exceeded and no Reactor trip occurred AND

b. Manual Reactor trip from Control Room was not o successful in shutting down the reactor AND

2. (a or b)

a. Core exit thrmocouple temperatures > 700*F, as indicated on SPDS O_R_

b. '

'e Secondary Cooling is not available Basis:

I Under the conditions of this Initiating Condition and its associated EALs, the efforts to bring the reactor suberitical have been unsuccessful and, as a result, the reactor is producing more heat than the maximum decay heat load for which the safety systems were designed. Although there are capabilities cway from the reactor control console, such as emergency boration, the continuing temperature rise indicates that these capabilities are not effective. This situation could be a precursor for a core melt sequence.

700$F is a good indicator of an extreme challenge to the ability to cool the core and is consistent with the

" potential loss" factor in the Fission Product Barrier Matrix. 1 Another consideration is the inability to initially remove heat during the early stages of this sequence. If j cmergency feedwater flow is insuflicient to remove the amount of heat required by design from at least cne steam generator, an extreme challenge should be considered to exist.

In the event either of these challenges exist at a time the reactor has not been brought below the power cssociated with the safety system design a core melt sequence exists. In this situation, core degradation

. c:n occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the Fission Product Barrier Matrix declaration to permit maximum offsite intervention time.

CR-3 Matrix Reference Number:3.4 I NEI 97-03

Reference:

SG2 l

47  !

I I

Emergency Action Level Basis Document SYSTEM MALFUNCTION Inability to Reach Required Mode Within Improved Technical Epecification Time Limits MODES: 1,2,3,4 ClassiBeation:

Unusual Event Initiating Condition:

Inability to reach required operating mode within Improved Technical Specification limits Emergency Action Level:

(1 and 2)

1. Entry into an Improved Technical Specification LCO statement requiring a mode reduction l AND

2. The plant is no_t brought to the required operating mode within the time prescribed by tre LCO

. required action Basis: l Limiting Conditions for Operation (LCOs) require the plant to be brought to a required shutdown mode when the Improved Technical Specification required configuration cannot be restored. The plant is within its safety envelope when being shut down within the allowable required action time in the  !

Improved Technical Specifications. An immediate Notification of an Unusual Event is required when the i plant is not brought to the required operating mode within the allowable required action time in the Improved Technical Specifications.

Declaration of an Unusual Event is based on the time at which the LCO-specified required action time period elapses under the Improved Technical Specifications and is not related to how long a condition may have existed.

CR-3 Matrix Reference Number:3.5 NEI 97-03

Reference:

SU2 48

Em:rgency Action Lev:1 Ba::is Document SYSTEM MALFUNCTION Loss ofIndications MODES: 1,2,3,4 Classification:

Unusual Event Initiating Condition:

UNPLANNED loss of most p_r all Control Room Annunciators for 15 minutes or longer Emergency Action Level:

(1 or 2) i I

1. UNPLANNED loss of Annunciator panels A-G and Annunciator printer for 15 minutes or longer OR

2. UNPLANNED loss of NNI-X and NNI-Y for 15 minutes or longer Basis:

This Initiating Condition and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment. Recognition of the availability of computer based indication equipment is considered (SPDS, i plant computer, etc.). The Annunciator printer includes the Annunciator CRT display as the display mimics the printer. A loss of both is required to meet the 10.

A loss of Annunciators is considered to be a loss of the visual, as opposed to a loss of the audible portion of the Annunciator. Annunciator panels A-G contain the major control systems (RPS, ES, ICS, etc.).

Loss on NNI-X and NNI-Y will cause the loss of most or all safety system indication.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Due to the limited number of safety systems in operation during cold shutdown, refueling, and defueled modes, no IC is indicated during these modes of operation.

CR-3 Matrix Reference Number:3.6 NEI 97-03

Reference:

SU3 l

l 49

)

Em:rg:ncy Action Level Basis Document l SYSTEM MALFUNCTION I l

1 i

Loss ofIndications MODES: 1,2,3,4 Classification:

Alert J

Initiating Condition.

UNPLANNED loss of most gr all Control Room Annunciators for 15 minutes or longer with either a SIGNIFICANT TRANSIENT in progress gr Plant Computer and SPDS unavailable Emergency Action Level:

(1 and 2)

1. (a or b)

a. UNPLANNED loss of. Annunciator panels A-G and Annunciator printer for 15 minutes or longer OB

b. UNPLANNED loss of NNI-X and NNI-Y for 15 minutes or longer AND .

2. (a or b)

a. SIGNIFICANT TRANSIENT in progress i O_R

b. Loss of Plant Computer and SPDS Basis:

This Initiating Condition and its associated EAL are intended to recognize the difficulty associated with m:nitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a transient. Recognition of the availability of computer based indication equipment is considered (SPDS, plant computer, etc.).

The Annunciator printer includes the Annunciator CRT display as the display mimics the printer. A loss 4 of both is required to meet the IC.

A loss of Annunciators is considered to be a loss of the visual, as opposed to a loss of the audible portion of the Annunciator.

Annunciator panels A-G contain the major control systems (RPS, ES, ICS, etc.)

50

Emergency Action Level Basis Document SYSTEM MALFUNCTION Loss of Indications, continued Basis, continued Due to the limited number of safety systems in operation during cold shutdown, refueling and defueled modes no IC is indicated during these modes of operation.

CR-3 Matrix Reference Number:3.7 NEI 97-03

Reference:

SA4

(

- }.

51

Em:rg:ncy Acti:n Level Ba:is Document SYSTEM MALFUNCTION Loss ofIndications MODES: 1,2,3,4 Classification:

Site Area Emergency Initiating Condition:

Inability to monitor a SIGNIFICANT TRANSIENT in progress  ;

l Emergency Action Level:

(1 and 2 and 3 and 4)

1. (a or b)

a. Loss of Annunciator panels A-G and Annunciator printer for 15 minutes or longer OR

b. Loss of NNI-X and NNI-Y for 15 minutes or longer AND

2. SIGNIFICANT TRANSIENT in progress I l

AND '

i

3. Ioss of Plant Computer and SPDS AND j

4. Inability to directly monitor any one of the following:

Suberiticality; Core Cooling '

Containment Conditions RCS Inventor J

. Basis:

This Initiating Condition and its associated EAL are intended to recognize the inability of the control room staff to monitor the plant response to a transient.

The Annunciator printer includes the Annunciator CRT display as the display mimics the printer. A loss i cf both is required to meet the IC, A loss of Annunciators is considered to be a bss of the visual, as opposed to a loss of the audible portion of the Annunciator.

52

Em:rg:ncy Acti:n Level Basis Document SYSTEM MALFUNCTION Loss ofIndications, continued Basis, continued

{

Indications needed to monitor safety functions necessary for protection of the public must include control I room indications, computer generated indications and dedicated annunciation capability. The specific i indications should be those used to determine such functions as the ability to shut down the reactor, miintain the core cooled and in a coolable geometry, to remove heat from the core, to maintain the r actor coolant system intact, and to maintain containment intact. l Pl nned and unplanned actions are not differentiated in this EAL since the loss ofinstrumentation of this magnitude is of such significance during a transient, that the cause of the loss does not make the condition more tolerable.

CR-3 Matrix Reference Number:3.8 NEI 97-03

Reference:

SS6 l

l l

1 l

l l 53 1

L

Em:rg:ncy Action Level Basis Document SYSTEM MALFUNCTION Fuel Clad Degradation MODES:ALL Classi8 cation:

Unusual Event Initiating Condition:

' RCS specific activity exceeds LCO Emergency Action Level:

Radiochemistry analysis indicates:

(a or b)

c. Dose Equivalent Iodine (I-131) >1 pCi/gm for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer I

_O_R

b. Specific activity >100/E-bar for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer i

Basis:

This Initiating Condition is included as an Unusual Event because it is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This EAL addresses RCS samples exceeding Improved Technical Specifications for radioactivity levels in the ,

RCS.

l RCS purification will provide for Iodine and crud cleanup in the reactor coolant system and reduce activity to < 1.0 pCi/gm within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

The EAL values are based on Improved Technical Specification Limits.

E-bar is the weighted average energy of RCS isotopes.

CR-3 Matrix Reference Number:3.9 NEI 97-03

Reference:

SU4 ,

54 t

I i

i Emergency Action Level Basis Documcnt SYSTEM MALFUNCTION Turbine Failure l MODES: 1,2,3 Classi6 cation:

Unusual Event j

Initiating Condition:

Turbine failure results in casing penetration Emergency Action Level:

Report by plant personnel of turbine failure causing penetration of the turbine casing qr damage to main g:nerator seals l Basis:

This EAL is intended to address main turbine rotating component failures of sufficient magnitude to )

c use observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustilJe fluids (lubricating oils) and gases (hydrogen) to the plant cnvirons. Actual fires and flammable gas build up are appropriately classified via Fire and Flammable Gis' EALs This EAL is consistent with the defimition of an Unusual Event while maintaining the enticipatory nature desired and recognizing the risk to non-safety related equipment.

Escalation of the emergency classification is based on the potential damage done by missiles generated ,

by the failure. It is not the intent of this Initiating Condition to declare an event based on damage  !

discovered in a maintenance evolution. Generator seal damage observed after generator purge does not meet the intent of this EAL because it did not impact normal operation of the plant.

CR-3 Matrix Reference Number:3.10 NE197-03

Reference:

HU1 55

I Em:rg:ncy Action Lev:1 Ba:is Document I

r L SYSTEM MALFUNCTION l

l Turbine Failure MODES: 1,2,3 Classification:

l l A1:rt initiating Condition:

Turbine failure generated projectiles cause significant visible damage to SAFE SHUTDOWN EQUIPMENT Emergency Action Level:

(1 or 2) l

1. Report by plant personnel of projectiles generated by a turbine failure causing significant visible damage to any of the following structures:

. Auxiliary Building,

. BWST, '

. Control Complex,

. Diesel Generator Building,

. EFT-2 Building,

. Intermediate Building, e Reactor Building OR

2. Indications show degraded SAFE SHUTDOWN EQUIPMENT perfortnance due to turbine generated projectiles Basis:

This EAL is intended to address the threat to safe shutdown equipment imposed by missiles generated

~by main turbine rotating component failures. The list of areas includes all areas containing safe shutdown equipment, their controls, and their power supplies. This EAL is, therefore, consistent with tha definition of an Alert in that if missiles have damaged or penetrated areas containing safety-related equipment the potential exists for substantial degradation of the level of safety of the plant.

This EAL is intended to address events that may have resulted in a plant vital area being subjected to l f rces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems.

The initial report is not interpreted as mandating a lengthy damage assessment prior to classification i cnd no attempt is made to asses the actual magnitude of the damage. This EAL is not intended to be used for temporary loss of control room habitability where timely repairs can be affected.

56

Emergency Action Level Basis Document SYSTEM MALFUNCTION Turbine Failure, continued Basis. continued If damage from the turbine failure is clearly contained and localized to one train, then safe shutdown equipment is not affected and the EAL is not met. if the extent of the damage is uncertain in terms of loss of safe shutdown equipment, then entry into this EAL is required.

CR-3 Matrix Reference Number:3.11 NEI 97-03

Reference:

HA1 l

I l

l l

l 8 I 57

. Em:rg:ncy Action Lev:1 Basis Docum:nt SYSTEM MALFUNCTION RCS Leakage MODES: 1,2,3,4 Classi8 cation:

Unusual Event Initiating Condition:

RCS leakage Emergency Action Level:

(1 or 2)

1. Unidentified Leakage or Pressure Boundary Leakage 10 gpm OR

2. Identified Leakage 25 gpm J Basis:  !

l The terms " identified," " unidentified," and " pressure boundary" leakage are as defined in Improved i Technical Specifications.

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

The 10 gpm value for the unidentified and pressure boundary leakage was selected as it is observable with normal control room indications. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances). The EAL for identified leakage is set at a higher v:lue due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.

CR-3 Matrix Reference Number:3.12 l NEI 97-03 '

Reference:

SU5 58

Em:rgency Action Level . Basis Document I SYSTEM MALFUNCTION Loss of Function (Hot Shutdown) i MODES: 1,2,3,4 Classification:

Site Area Emergency Initiating Condition:

C:mplete loss of core heat removal capability Emergency Action Level:

(1 and 2)

1. Complete loss of Main, Emergency, and Auxiliary Feedwater and unable to establish HPI cooling AND

2. Loss of Subcooling Margin (SCM)

Basis:

This EAL addresses complete loss of functions, including loss of heat removal capability, required for hot chutdown. 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. The EALs will allow for HPI/PORV cooling upon loss of all feedwater. Subcooling margin may be lost momentarily during this evolution but the ability to establish HPI cooling will not have been lost.

CR-3 Matric Reference Number:3.13 1 NEI 97-03

Reference:

SS4 59

Em:rg:ncy Action Level Basis Document SYSTEM MALFUNCTION Inadvertent Criticality MODES: 2, 3, 4, 5, 6 Classification:

Unusual Event Initiating Condition:

Inadvertent criticality during refueling or shutdown Emergency Action Level:

An extended or unplanned sustained positive startup rate monitored by nuclear instrumentation Basis:

This condition can be identified using the startup rate monitor. The term " extended"is used to allow for exclusion of expected short term positive startup rates from planned fuel bundle or control rod movements during core alterations. The short term startup rates are the result of the increase in neutron population due to suberitical multiplication.

This Initiating Condition /EAL is not intended to classify an early criticality during reactor startup. This type event is indicative of errors in reactivity data / calculations and/or mis-operation. The loss of the required shutdown margin can be quickly restored by manual actions or automatic reactor trip.

CR-3 Matrix Reference Number:3.14 l NEI 97-03

Reference:

SUS {

l 60

Em:rg:ncy Action Level Ba:is Document SYSTEM MALFUNCTION Inability To Maintain Plant In Cold Shutdown MODES: 5, 6 ClasdScation:

Alert (

Initiating Condition:

C:mplete loss of functions required for core cooling during refueling and cold shutdown modes Emergency Action Level:

(1 or 2)

1. Inability to maintain reactor coolant temperature below 200*F OR

2. Uncontrolled reactor coolant temperature approaching 200"F Basis:

Fcr PWRs, this Initiating Condition and its associated EAL are based on concerns raised by Generic Letter 88-17 " Loss Of Decay Heat Removal." A number of phenomena, such as pressurization, vortexing, steam generator U-tube draining, RCS level differences when operating at a mid-loop condition, decay h:at removal system design and level instrumentation problems can lead to conditions where decay heat removalis lost and core uncovery can occur. NRC analyses show that sequences can cause core uncovery in 15 to 20 minutes, and severe core damage within an hour after decay heat removal is lost. Under these conditions, RCS integrity is lost and fuel clad integrity is lost or permanently lost, which is consistent with a Site Area Emergency. Site-specific indicators for these EALs are those methods used by the plant in response to Generic Letter 88-17, which include core exit temperature monitoring and RCS water level monitoring. In addition radiation monitor readings may also be appropriate as an indicator of this condition.

" Uncontrolled" means that system temperature increase is not the result of planned actions by the plant ctaff. 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 than the cold chutdown temperature limit.

A momentary UNPLANNED excursion above 200"F when the heat removal function is available is not intended to constitute an alert. Separate statements (1 and 2) are included to recognize additional plant l capability to maintain cooling of the reactor.

CR-3 Matrix Reference Number:3.15 NEI 97-03

Reference:

NEI-SAS 61

Emerg:ncy Action Level Basis Document SYSTEM MALFUNCTION Loss of Water Level in Reactor Vessel That Has or Will Uncover Fuel MODES: 5, 6 Classification:

Site Area Emergency I

Initiating Condition:

Loss of water level in the reactor vessel that has or will uncover fuel in the reactor vessel Emergency Action Level:

(1 and 2 )

1. Loss of decay heat removal per AP-404 AND

2. (a or b)

a. Incores indicating superheated conditions OR

b. Incores unavailable and time to uncovery exceeded as specified in OP-301 Basis:

Under the conditions specified by this Initiating Condition, severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured. OP-301, " Operation of .e Reactor Coolant System," contains time to core uncovery without decay heat removal curves. l For CR-3, this Initiating Condition covers sequences such as prolonged boiling following loss of decay heat removal. Thus, declaration of a Site Area Emergency is warranted under the conditions specified by the Initiating Condition.

CR-3 Matrix Reference Number:3.16 NEI 9703

Reference:

SS5 62

Emerg:ncy Action Level Basis Document LOSS OF POWER l Loss of AC Power MODES: ALL Classification:

Unusual Event Initiating Condition:

' Loss of All Offsite Power for 15 minutes or longer Emergency Action Level:

(1 and 2)

1. Offsite Power Transformer (OPT) and Backup ES Transformer (BEST) and Auxiliary Transformer not available for 15 minutes or longer AND

2. EDGs supplying power to required 4160V ES Bus (ses) l Basis:

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 is used as a threshold to exclude transient or momentary power losses.

l Available indicates transformers are capable of energizing required busses.

'In Modes 1-4, both ES busses are required to be available.

In Modes 5-6, one ES Bus is required to be available.

CR-3 Matrix Reference Number:4.1 NEI 97-03

Reference:

SUI 63

Em:rg:ncy Action Level Basis Document -

LOSS OF POWER Loss of AC Power MODES: 1,2,3,4 Classification:

Alert Initiating Condition:

AC power capability to required buses reduced to a single source for 15 minutes or longer such that an )

additional failure would result in station blackout J l

Emergency Action Level:

AC power capability to the ES 4160V busses reduced to a single power source for 15 minutes or longer such that only one of the following is available:

)

1

. "A" EDG

. ' "B" EDG l

. Offsite Power Transformer (OPT)

. Backup ES Transformer (BEST)

. Aux Transformer )

Basis: ,

]

This Initiating Condition and the associated EALs are intended to provide an escalation from " Loss of Offsite Power for Greater Than 15 Minutes." The condition indicated by this Initiating Condition is the d gradation of the offsite and onsite power systems such that any additional single failure would result j in a station blackout.

Available indicates transformers are capable of energizing required busses.

EDG = Emergency Diesel Generator CR-3 Matrix Reference Number:4.2 NEI 97-03

Reference:

SA5 64

Em:rgency Action Level B: sis Document LOSS OF POWER Loss of AC Power MODES: 1,2,3,4 Classification:

Site Area Emergency Initiating Condition:

Loss of All Offsite and required Onsite AC Power for 15 minutes or longer Emergency Action Level:

Neither 4160 ES bus is capable of being energized within 15 minutes Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will cause core uncovering and may challenge containment integrity. The fifteen minute time duration is to exclude transient or momentary power losses.

NOTE:In Modes 5 and 6, the same initiating condition /EAL is an Alert classification.

CR-3 Matrix Reference Number:4.3 NEI 97-03

Reference:

SSI i

l l

65

Em:rg:ncy / tion Lev:1 B=is Dot.ument LOSS OF POWER Loss of AC Power MODES: 1,2,3,4 Classification:

General Emergency Initiating Condition:

Prolonged Loss of All Offsite and Onsite AC power Emergency Action Level:

(1 and 2)

1. Neither 4100 ES bus is capable of being energized AND

2. (a or b)

a. Restoration of 4160V ES Bus A pr 4160V ES Bus B is not likely within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> OR

b. Core exit thermocouples >700*F as indicated on SPDS Basis:

700 F is a good indicator of an extreme challenge to the ability to cool the core and is consistent with the

" potential loss" factor in the Fission Product Barrier Matrix.

Loss of all AC power compromises all plant safety systems requiring electric power including ECCS and th3 Ultimate Heat Sink. Prolonged loss of all AC power will lead to loss of fuel clad, RCS, and may challenge containment integrity. The four hours to restore AC power is based on the CR-3 station bl:ckout coping analysis performed in conformance with 10 CFR 50.63 and Regulatory Guide 1.155,

" Station Blackout." Although this Initiating Condition may be viewed as redundant to the Fission Product Barrier Matrix, its inclusion is necessary to better assure timely recognition and emergency re ponse.

This Initiating Condition is specified to assure that in the unlikely event of a prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.

66

Em:rg:ncy Action Level Basis Document LOSS OF POWER Loss of AC Power, continued Basis, continued The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the eituation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions.

CR-3 Matrix Reference Number:4.4 NEI 97-03

Reference:

SGI j i

)

1

)

67 L

Em:rgency Action Level Basis Document LOSS OF POWER Loss of AC Power (Shutdown)

MODES: 5, 6, No Mode (defueled)

Classification:

Alert Initiating Condition:

Loss of All Offsite and Onsite AC Power to Required Busses During Cold Shutdown or Refueling Mode l for 15 minutes or longer Emergency Action Level:

N:ither 4160 ES bus is capable of being energized within 15 minutes Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including ECCS, Containment 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. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

CR-3 Matrix Reference Number:4.5 NEI 97-03

Reference:

SA1 l j

I 68

Emcrgency Action Lcvel Basis Document LOSS OF POWER Loss of Vital DC Power MODES: 1,2,3,4

)

ClassiScation:

Site Area Emergency I

Initiating Condition:

l Loss of all Vital DC Power for 15 minutes or longer Emergency Action Level:

Standby Power Status Lights for BUS A1, A2, and BUS B1, B2 on the Main Control Board (SSF Panel) are out I

Basis:

Loss of all DC power compromises ability to monitor and control plant safety functions. Prolonged loss of j cll DC power could cause core uncovery and loss of containment integrity when there is significant decay i heat and sensible heat in the reactor system. Fifteen minutes is used to exclude transient or momentary I power losses.

CR-3 Matrix Reference Number:4.6 NEI 97-03

Reference:

SS3 I

l l

1 1

69

Em:rgency Action Lcvel Basis Document

-LOSS OF POWER

~ Loss of Vital DC Power (Shutdown)

MODES: 5, 6 I

Classification:

Unusual Event l

Initiating Condition:

I Loss of all Vital DC Power for 15 minutes or longer Emergency Action Level:

Standby Power Status Lights for BUS A1, A2 and BUS B1, B2 on the Main Control Board (SSF Paned are out Basis:

Loss of required DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power could cause core uncovery and loss of containment integrity when there is significant {

decay heat and sensible heat in the reactor system. Fifteen minutes is ured to exclude transient or momentary power losses.

CR-3 Matrix Reference Number:4.7 NEI 97-03

Reference:

SU7 l

i l

70

Emergency Action Level Basis Document FISSION PRODUCT BARRIER MATRIX BASIS FUEL CLAD LOSS FACTORS

1. CORE CONDITIONS IN REGION THREE OR SEVERE ACCIDENT REGION OF INADEQUATE CORE COOE1NG CURVES (EOP-07)

The initial core damage assessment curve is used to relate the observable parameters of incore temperature and RCS pressure to clad temperature. In region three or the severe accident region, elevated clad temperatures may exceed temperatures that will lead to zirc/ water reactions and rapid failure of the clad will occur if not halted.

4 l.{..{. ; gf i - - -

/- + -t),-

y_ _

.f _p _ f. _

4 /p 4 _

_/

4 . f _._p _ }.f4 .. p f.

+ - _

+ fr _ _

y _

q f

.p .

i . . .. . . . i j j_ _ _

/p _

af g "- 4 _ _

4.. .... . qy + p .

g "" ++

Q "" -;.

_;j

. j.

_. - lj a

l ,.*.7 .. , l 4 g af f,_ gj _

4 q ..f_ _ j .. 4 _

q_ .._

.g __

4 f.............i .

y .j _j- - _

af. .

y. 4 - -

y j_A ,

-.. p y .. - -

_g 4 4 .A _ p. ...

.p 4 _

1... en

2. RCS ACTIVITY >300 pCi/gm I*

This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2% to 5% fuel clad damage. This amount of clad damage indicates significant clad heating and thus the Fuel Clad Barrier is considered lost.

3. RM-G29 OR 30 >100 R/hrfor 15 minutes or longer Monitor readings have increased and are sustained, not spikes. Readings of >100 R/hr on these monitors indicate activity in the Reactor Building above what would be expected for normal reactor coolant. The 15 minutes will aid in accounting fbr spikes and uneven mixing that occurs in the initial phases of an ItCS leak in the RB. High initial concentrations that accumulate in the upper portion of the RB may lead to erroneous fuel damage assumptions.

4. EC DEEMS FUEL CIAD BARRIER 15 LOST Based on Emergency Coordinatorjudgment.

71

Emergency Action Level Basis Document FISSION PRODUOT BARRIER BASIS FUEL CLAD POTENTIAL LOSS FACTORS

1. RCS CONDITIONS WARRANT ENTRYINTO EOP-07 EOP-07 is the " Inadequate Core Cooling" precedure which indicates that there are superheated conditions in the core which may lead to clad degradation.

2. CORE EXIT THERMOCOUPLES >700"F 700*F is a good indicator of an extreme challenge to the ability to cool the core. Temperatures are determined using guidance in EOP-07.  ;

l

3. EC DEEMS FUEL CLAD BARRIER IN JEOPARDY Based on Emergency Coordinatorjudgment.

I 72

l Emergency Action Level Basis Document FISSION PRODUCT BARRIER BASIS RCS LOSS FACTOR

1. RCS LEAK RESULTING IN LOSS OF ADEQUATE SUBCOOLING MARGIN (SCM)

A loss of adequate SCM resulting from RCS leakage would indicate that the rate ofleakage from the l RCS is exceeding the rate of addition from the injection system. Therefore, the RCS boundary should be considered lost any time adequate SCM is lost due to leakage.

2. OTSG TUBE IRAK RESULTING IN LOSS OF ADEQUATE SUBCOOLING MARGIN (SCM)

A loss of adequate SCM resulting from RCS leakage would indicate that the rate ofleakage from the RCS is exceeding the rate of addition from the injection system. In addition, with a loss of SCM, accurate RCS inventory cannot be determined. Therefore, the RCS boundary should be considered lost any time adequate SCM is lost due to leakage.

3. RM-G29 OR 30 > 10 R/hr The reading of > 10 R/hr is a value which indicates the release of reactor coolant to the containment.

The reading is based on RCS activity in normal operation concentrations.

4. EC DEEMS RCS BARRIER IS LOST l

Based on Emergency Coordinatorjudgment.

l

\

I 73

)

l l

Emergency Action Level Basis Document FISSION PRODUCT BARRIER BASIS RCS POTENTIAL LOSS FACTORS l

I. RCS LEAK REQUIRING ONE OR MORE INJECTION VALVES By procedure, the HPI injection valves will be used to increase RCS inventory if pressurizer level cannot be maintained greater than 50 inches with letdown isolated. Thus, the use of one or j capability and therefore a potential loss of the RCS barrier. )

2. OTSG 'ITJBE LEAK REQUIRING ONE OR MORE INJECTION VALVES By procedure (EOP-06), the HPI injection valves will be used to increase RCS inventory if pressurizer level cannot be maintained at 220 inches during a tube leak event. Thus, the use of one or more injection valves would indicate leakage in excess of the normal makeup capability and therefore a potentialloss of the RCS barrier.

3. RCS LEAK RESULTS IN ES ACTUATION ON LOW RCS PRESSURE Should the injection system fail or the operator fail to open the injection valves upon a failure of the Makeup system to maintain RCS inventory, RCS pressure will decrease to the ES actuation setpoint.

This potential loss factor in addition to number one (above) will ensure that the RCS barrier will be considered potentially lost for any inability of the makeup system to maintain adequate inventory during a loss of coolant event.

4. OTSG TUBE LEAK RESULTS IN ES ACTUATION ON LOW RCS PRESSURE Should the injection system fail or the operator fail to open the injection valves upon a failure of the Makeup system to maintain RCS inventory, RCS pressure will decrease to the ES actuation setpoint.

This potential loss factor in addition to number one (above) will ensure that the RCS barrier will be considered potentially lost for any inability of the Makeup system to maintain adequate inventory during an OTSG tube leak event.

I

5. RCS PRESSURE TEMI'ERATURE RELtTIONSHIP VIObtTES NDT LIMITS.

RCS conditions of high pressure accompanied by low temperature increase the potential for Reactor Coolant System brittle failure. This potential loss factor will ensure that the RCS harrier is considered potentially lost wheaever the system is at risk of a non-ductile failure.

6. EC DEEMS RCS BARRIER IN JEOPARDY Based on Emergency Coordinatorjudgment.

74

Emergency Action Level Basis Document FISSION PRODUCT BARRIER BASIS CONTAINMENT LOSS FACTORS

1. RAPID UNEKPLAENED RB PRESSURE DECREASE FOLLOWING INITIAL INCREASE During a loss of coolant event, RB pressure should rise to some value determined by the size of the leak and the response of the RB cooling systems. Following the initial peak, RB pressure should exhibit a steady decreasing trend. Any deviation from this should be the result of a known change in plant status. A rapid decrease of unknown cause is therefore indicative of possible containment failure.

2. CONTAINMENT PRESSURE OR SUMP LEVEL NOT CONSISTENT WITH LOCA CONDITIONS Sump level or containment pressure Not increasing indicates containment bypass and a loss of containment integrity.

3. AN OTSG HAS TUBE LEAK AND UNISOLABLE STEAM LEAK OUTSIDE RB PER EOP-06, l "GTEAM GENERA 1DR TUBE RUPTURE" An unisolable steam leak outside the RB on an OTSG with a tube leak would provide a direct leakage path to the environment from the RCS. This " loss" EAL recognizes that steam generator tube leakage can represent a bypass of the containment barrier as well as a loss of the RCS barrier.

The first " loss" EAL addresses the condition in which a ruptured steam generator is also faulted.

This condition represents a bypass of the RCS and containment barriers. In conjunction with RCS Barrier "Potentialloss" EAL #2, this would always result in a Site Area Emergency.

4. CONTAINMENT ISOLATION IS INCOMPLETE AND BRTRASE PATH TO THE ENVIRONMENT l EXISTS This factor should be used any time an incomplete RB isolation results in a direct path from the RB atmosphere to the environment. The conditions expected for this EAL would be a known path or a visual indication of the failure or path. Confirmation may be from elevated radiation readings in areas e.djacent to the RB. j

5. EC DEEMS CONTAINMENT BARRIER IS LOST Based on Emergency Coordinatorjudgment.

i 1

75 I

1 Em:rgency Action Lcvel Basis Document FISSION PRODUCT BARRIER BASIS CONTAINMENT POTENTIAL LOSS FACTORS

1. RB PRESSURE >54 psig RB design pressure is 54.4 psig. Internal pressure greater than this value has the potential to exceed design leakage values.

2. RB HYDROGEN CONCENTRATION >4%

Hydrogen concentrations > 4% are above the lower explosive limit.

1

3. RB PRESSURE >30 psig WITH NO BUILDING SPRAY AVAILABLE I

The RB spray actuation setpoint is 30 psig. With RB pressure above this value and no spray available, the potential exists to exceed the RB design values.

4. RMG-29 OR 30 READINGS >25,000 R/hr This monitor reading is indicative of severe core damage conditions and is consistent with the monitor indication listed on the Protective Action Recommendations Table for " core damage indications." Monitor readings have increased and are sustained, not spikes. Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted.

5. CORE CONDITIONS IN SEVERE ACC1 DENT REGION OF ICC CURVES FOR GREATER THAN 15 MINUTES Core conditions in the Severe Accident Region represent imminent melt sequence which, if not corrected within 15 minutes, could lead to vessel failure and an increased potential for containment failure. The Emergency Coordinator should make the declaration as soon as it is determined that the restoration procedures have been, or will be ineffective. .

G. EC DEEMS CONTAINMENT BARRIER IN JEOPARDY Based on Emergency Coordinatorjudgment.

I 76

Emergency Action Level Basis Document DEFINITIONS ALERT: Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline evosure levels.

BOMB: An explosive device (See EXPLOSION)

CIVIL DISTURBANCE: A group of ten (10) or more people violently protesting station operations or activities at the site. A civil disturbance is considered violent when force has been used in an attempt to injure site personnel or damage plant property.

EXPLOSION: A rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment that imparts energy of sufficient force to potentially damage permanent structures, systems, or components.

EXTORTION: An attempt to cause an action at CR-3 by threat of force. Bomb threats that are unsubstantiated are not included in this definition.

FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required iflarge quantities of smoke and heat are observed.

GENERAL EMERGENCY: Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential loss of containment integrity. Releases can be )

reasonably be expected t.o exceed EPA Protective Action Guidelines exposure levels at the SITE j BOUNDARY.

HOSTAGE: A person or object held as leverage against the station to ensure that demands will be met by CR-3.

INTRUSION / INTRUDER: Suspected hostile individual present in a protected area without l suthorization.

PROTECTED AREA: All areas within the CR-3 security perimeter fence that require badged authorization for entry.

SABOTAGE: Deliberate damage, mis-alignment, or mis-operation of safe shutdown equipment with the intent to render the equipment unavailable.

SAFE SHUTDOWN EQUIPMENT: Equipment necessary to achieve and maintain the reactor suberitical with controlled decay heat removal. l 77

Em:rg:ncy Action Levtl Basis Document DEFINITIONS, continued SIGNIFICANT TRANSIENT: An UNPLANNED event involving one or more of the following:

(1) Automatic turbine trip at >25% reactor thermal power (2) Electricalload rejection >25% full electricalload (3) Plant runback (4) Reactor trip (5) Safety injection system actuation (6) >10% thermal power oscillations (7) Loss of decay heat removal in Modes 5 or 6 SITE AREA EMERGENCY: Events are in process or have occurred which involve actual or likely failures cf plant functions needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels at the SITE BOUNDARY.

SITE BOUNDARY: That area, including the PROTECTED AREA, that extends 4400 ft. or 0.83 miles in a circle around the Reactor Building. Also referred to as the Owner Controlled Area.

UNPLANNED: An event or action is UNPLANNED ifit is not the expected result of normal operations, te ting, or maintenance. Events that result in corrective or mitigative actions being taken in accordance with abnormal or emergency procedures are UNPLANNED.

NOTE: With specific regard to radioactive releases, a release of radioactivity is UNPLANNED if the release is not authorized by a Release Permit or exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.

UNUSUAL EVENT: Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring cre expected unless further degradation of safety occurs.

VALID: An indication or report or condition is considered VALID when it is conclusively verified by (1) en instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely c:sessment (e.g., within 15 minutes).

VISIBLE DAMAGE: Damage to equipment or structure that is readily observable without measurements, testing, or analyses. Damage is sufficient to cause concern regarding the continued cperability or reliability of affected safety structure, system, or component. Example damage includes:

d: formation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface bhmishing (e.g., paint chipping, scratches) should not be included.

78 a

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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 t

ATTACHMENT C '1'O LETTER 3F0799-22 DEVIATIONS FROM NEI 97-03 DRAFT FINAL REVISION 3 OCTOBER 1998 (FORMERLY NUMARC/NESP-007)

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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACHMENT D TO LETTER 3F0799-22 i

i ISSUES ON CR-3 EAL SUBMITTAL DISCUSSED WITH NRC STAFF DURING TELECONFERENCES HELD APRIL 12 AND 14, AND JUNE 17,1999 l

l

Responses to NRC Issues Regarding Adoption of New EAL Methodology Page 1 of 9 The following issues were discussed during teleconferences held April 12 and 14, and June 17,1999.

FPC's emergency action level (EAL) is identified in bold typeface followed by a brief summary of the U. S. Nuclear Regulatory Conunission's (NRC's) question or concern. FPC's response is in italic typeface.

EAL 1.1-1 (NUMARC/NESP-007 AUI)

Please verify that all postulated gaseous releases will be monitored by RM-Al and RM-A2.

All anticipated gaseous releases will be monitored by RM-Al and RM-A2 as noted in FSAR Section 11.4.2.1.2. If a steam generator tube rupture (SGTR) occurs and a secondary safety hfts )

however, there is no monitorfor this type of release.

Please discuss how the thresholds for radioactive releases are set on the radioactive release permit. An example of a radioactive release permit will facilitate discussions.

Existing procedures will be revised to identify the Unusual Event (UE) and Alert thresholds on the radioactive release permits.

EAL 1.1-2 (NUMARC/NESP-007 AU1)

Please discuss why this EAL refers only to the noble gas instantaneous release limit.

Tids is the only species which is monitored real-uime. Lodines andparticulates are not monitored as a release rate but rather are collected on afilter and evaluated in terms of the cumulative release. Note that the words " noble gas instantaneous release limit" have been replaced with

" noble gas release setpoint"in EALs 1.1,1.2,1.5, and 1.6.

EAL 1.3-1 (NUMARC/NESP-007 AS1-1)

Please discuss the following issues- '

i Justification for deviation from NUMARC/NESP-007 guidance (i.e., accuracy, timeliness).

Additionalinformation was included in Attachment C, Deviation Matrixfor this EAL explaining how initial radiation monitor readings in the Control Room ensure timely and accurate dose assessment.

Potential numbering error in basis.

The numbering error was corrected.

Responses to NRC Issues Regarding Adoption of New EAL Methodology Page 2 of 9 Does a shift dose assessment capability exist and if it does, why does the emergency response

' team'need to provide input.

Additionalinformation uns included in Attachment C, Deviation Matrixfor this EAL explaining that ship dose assessment capability exists in the ControlRoom. If the release

. is unmonitored, control room stafis instructed to base prctective actions on plunt conditions, not dose assessmentfrom the Control Room. TSCformal dose assessment would be used to adjust early PARS that were taken based on plant conditions.

Why a note which provides for classifying on real time dose assessment (when available) instead

- of radiation monitor setpoints was not included.

Additionalinformation was included in Attachment C, Deviation Matrixfor this EAL explaining that the Note in NEl 97-03 regarding delay of the declaration was not necessary since real-time meteorological data has been incorporated into the monitor readings in this EAL.

What was the source term which was used in establishing the radiation monitor setpoints.

The Basis section of this EAL uns revised to state " source termfor this EAL is based on RCS with 1 % failedfuelper FSAR safety analysis. "

EAL 1.4-2 (NUMARC/NESP-007 AGI)

Please. discuss why indication of core damage is necessary to classify this event.

De Basis section of this EAL was revised to explain the statement that core damage is necessary to classify this event because it precludes erroneous actions taken to recommend evacuation of the

. public based on incorrect or default dose assessment results when plant conditions clearly do not support the magnitude of the release.

EAL 1.7-1,1.8-2 (NUMARC/NESP-007 AU2-3,AA3-2)

. Please discuss why this EAL does not include all area radiation monitors. In particular, the staff would like to discuss whether local radiation levels could limit access to safety equipment without causing the

. listed radiation monitors to exceed the 'specified setpoints.

. nis EAL was revised to include two additional radiation monitors, RM-G5 located in the Waste Gas Decay Tank area, and RM-G14 located in an access corridor adjacent to the Spent Fuel Storage walls. These and other monitors were chosen because they monitorpotential release areas and/or access corridors to the plant.

Responses to NRC lssues Regarding Adoption of New EAL Methodology Page 3 of 9 EAL 1.9 (also EAL 1.10.2) (NUMARC/NESP-007 AU2-1, AA2-2)

Please discuss why this EAL does not specify a water level in the refueling cavity.

Additionalinformation was included in Attachment C, Desiation Matrixfor this EAL explaining that FPC does not have water levelindication in the refueling cavity. Spent Fuelpoollevel indication is used when the transfer tubes are open duringfuel movement.

EAL 1.10-1b (NUMARC/NESP-007 AA2-1)

Please discuss why this EAL refers to increasing trend instead of providing a setpoint. Further, the staff would like to discuss how this event would be classified if the increasing trend stabilized.

Tids EAL was revised to replace " increasing trend' with "high alarm. "

EAL 2.1 (NUMARC/NESP-007 HUI-1)

Please discuss why this EAL did not specify ground motion sensed as a stand-alone condition. In addition, please discuss the setpoint used (Revision 18 had the annunciated earthquake at the Alert level).

Sensing ground motion may not be a result of an earthquake but may be due to other causes such as turbine vibration. Furthermore, the seismic alarm is set at 0.01g which is below the operating basis earthquake (OBE) level and is consistent with the NUMARC/NESP-007 guidance.

EAL 2.6 (NUMARC/NESP-007 HUI-2) ,

Please discuss why high winds were not included in this EAL.

)

l i

As described in the Attachment C Deviation Matrixfor this EAL, the primary sources of high winds are hurricanes or tornadoes that would result in an Unusual Event classification. The Deviation Matrixfor this EAL wasfurther clarfled to explain that the design basis wind speedfor missiles per FSAR Section 5.2.1.2.6 is 300 mph which cannot be reliably measured on site.

EAL 2.7 (NUMARC/NESP-007 HAl-2)

Please discuss an apparent editorial mistake in this EAL. ,

1 Attachment A was revised to correct the numerical error in this EAL. l EAL 2.10 (NUMARC/NESP-007 HU3-2)

Please discuss why " sheltering" was not included in this EAL. I Tids EAL was revised to address sheltering in addition to evacuation ofsite personnel, i

Responses to NRC Issues Regarding Adoption of New EAL Methodology Page 4 of 9 EAL 2.11 (NUMARC/NESP-007 AU1)

Plearc discuss why this EAL would not be applicable if personnel protective equipment was used.

(Th wif would like to discuss whether there are other EALs which include statements in its basis which limit their applicability.)

This EAL was separated into two items. One item addressesflammable gas levels and explosive limits that could result in an explosion. The remaining item addresses toxic gas levels. The Basis for this EAL nm expanded to clanfy thatfor toxic gas, the initiating condition only applies to areas that do not require continuous occupancy.

EAL 2.15 (NUMARC/NESP-007 HAl-4)

Please discuss the statements limiting the applicability of this EAL contained in the basis for the EAL.

FPC recognizes that NEl 97-03 does not adequately accountfor afire that may afect safety-related equipment not neededfor safe shutdonn in a spectfc event. This inadequacy uns discussed with members of the NEl Task Force. FPCproposed, and the Task Force agreed that thefolloning statement should be included in the Basisfor this EAL: Ifdamagefrom thefire is clearly contained and localized to one train, then safety shutdoun equipment is not afected and the EAL is not met. If the extent of the damage is uncertain in terms ofloss ofsafe shutdown equipment, then entry into this EAL is required. " Similar changes were made to EALs 2.2, 2.4,

2. 7, 2.9, 2.13, 2.23, and 3.11.

EAL 2.16 (NUMARC/NESP-007 HA5-1)

Please discuss why it is not appropriate to declare this event as soon as possible.

The Basisfor this EAL was revised to state " Declaration of an Alert may by delayed until transfer to remote shutdoun is completed. " "Tids is appropriate since establishing control of theplant takes precedence. "

EAL 2.17 (NUMARC/NESP-007 HS2)

Please discuss the initiating condition for this EAL and its relationship to the NUMARC/NESP-007 guidanc.e. Please discuss the ease of classifying this event using the referenced procedure.

FPC maintains that this EAL does mt deviatefrom the NUMARC/NESP-007 guidance and the NRC agreed as rwtedin Reference '. To resolve this issue, FPChas venfed that AP-990, "Shutdoun Outside of the Control Room"provides suficient guidance to support declaration of a Site Area Emergency in the event plant control cannot be established after evaluationfrom the ControlRoom has been initiated.

l 1

1

Responses to NRC Issues Regarding Adoption of New EAL Methodology Page 5 of 9 EAL 2.19,20 (NUMARC/NESP-007 HA1, HSI)

Please discuss how this EAL corresponds to its NUMARC/NESP-007 counterpart and whether there are any other safeguards events which should be included.

The staff would like to discuss deviations from NUMARC/NESP-007 in the initiating conditions specified for the security event EALs.

The initiating conditionfor EAL 2.18 was revised to state, "Conynned Security Event which indicates a potential degradation in the level ofsafety of theplant. " The initiating conditionfor EAL 2.19 was revised to state, ' Confirmed Security Event in a plant Protected Area. " The initiating conditionfor EAL 2.20 was revised to state, "Confnned Security Event in a plant Vital Area. " These changes meet the NUMARC/NESP-007 criteria. CR-3 security representatives reviewed the EALs to ensure they capture all safeguard events of concern.

EAL 2.21 (NUMARC/NESP-007 HG1)

Please discuss the justification for this deviation from the NUMARC/NESP-007 guidance.

The Basis for this EAL was expanded to incorporate the following statements from the corresponding EALfrom NEl 97-03, Draft Revision 3: "If control of the plant equipment necessary to maintain safety functions can be transferred to another location, then the above limiting condition is not met. Loss ofphysical control of the Control Room or remote shutdown capability alone may not prevent the ability to maintain safetyfunctions per se. "

EAL 2.23 (no corresponding NUMARC/NESP-007 EAL)

Please discuss the statement in the basis regarding the applicability of the EAL.

The Basisfor this EAL was clanfied to add the statements, "If damagefrom the internalflooding is clearly contained and localized to one train, then safe shutdown equipment is not afected and item 2 of the EAL is not met. " "If the extent of the damage is uncertain in terms ofloss of safe shutionia equipment, then entry into this EAL is required. "

EAL 3.2 (NUMARC/NESP-007 SA4-1)

Please discuss why including the condition " reactor power 5% and decreasing" is not a deviation from the NUMARC/NESP-007 guidance.

EAL 3.2 ans revised to replace "<5% and decreasing" with " reactor is shutdown"for consistency with terminology used in Emergency Operating Procedures (EOPs)for thefailure to trip the reactor.

Responses to NRC lssues Regarding Adoption of New EAL Methodology Page 6 of 9 EAL 3.3 (NUMARC/NESP-007 SS2-1)

Please discuss whether reactor power less than 5% power corresponds to the power level for which the l

safety systems are designed.

EAL 3.3 was revised to replace "<5% and decreasing" with " shutting down the reactor"for consistency with terminology used in Emergency Operating Procedures (EOPs)for thefailure to trip the reactor.

EAL 3.4 (NUMARC/NESP-007 SG2-1)

Please discuss how this EAL relates to the NUMARC/NESP-007 guidance for including indications that heat removal is extremely challenged.

EAL 3.4 was revised to replace "<5% and decreasing" with " shutting down the reactor"for consistency with tenninology used in Emergency Operating Procedures (EOPs)for thefailure to trip the reactor. In addition, this EAL was revised to add an indicator ofloss ofsecondary heat transfer to address conditions that pose an extreme challenge to the ability to cool the core.

EAL 3.6 (NUMARC/NESP-007 SU3-1)

Please discuss the purpose of the annunciator printer.

Note that this response also applies to EAL 3.7. CR-3 has an alarm system that is basically divided into two parts. There are annunciator windows that come into alarm above the main control board. These windows may have multiple alarm points thatfeed the window. For example, a radiation monitor going to high alann will annunciate the same window regardless of which radiation monitor is actually in alarm. The secondpan of the system is the events recorder / printer. A CRTis located above the Main Control Board and a printer is located behind the main control board. These two devices print or show which alann point has come in on the annunciator window. Using the example above, the alarmprinter/CRT willindicate which individual radiation monitor has come into alann. Both parts of the annunciator system work independently but have the same input. The loss of either system degrades but does notpreclude the operatorfrom properly monitoring plant alarms.

' EAL 3.15 (NUMARC/NESP-007 SA3-1)

Please discuss how this EAL relates to the guidance provided in NUMARC/NESP-007.

The classificationfor this EAL was changedfrom an Unusual Event to an Alert to meet the NUMARCINESP-007 guidance. In addition, two typographical errors were corrected.

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Responses to NRC Issues Regarding Adoption of New EAL Methodology Page 7 of 9 EAL 3.16 (NUMARC/NESP-007 SS5-1)

Please discuss how this EAL relates to the guidance provided in NUMARC/NESP-007.

This EAL was revised to match more clearly the NEI guidance. Presious item 3 of the EAL regarding a direct pathfrom the containment uns deleted as this is not addressed in either the NUMARC or NEl 97-03 guidance. Item 2a of the EAL was revised to change " saturated" to "superheated. " Item 2b of the EAL was revised to change " time to boil" to " time to uncovery.

Corresponding changes were made to the initiating conditions.

EAL 4.4 (NUMARC/NESP-007 SGI-1)

Please discuss an apparent editorial error in this EAL.

This EAL was revised to change the word " incapable" to "is capable. "

EAL Fission Product Barrier (FPB) - Loss of Fuel Clad based on Radiation Monitor Reading Please discuss the basis for the setpoint used (e.g. RM-G29 > 100 R/hr).

FPC has deternined that the 100 R/hr serpoint is appropriatefor CR-3 based on actual experience which resulted in a 65 R/hr spike. As noted in the NRC's Response Technical Manual (RTM) RTM-93 Training Overview Manual, the CR-3 event that occurred in February,1980 was an example and states thatfuel (clad) damage would be indicated at 100 R/hr. In addition to the serpoint question, the Basis Documentfor this EAL was revised to explain the use of the 15 minute timeframe in the EAL. The 15 minutes will help accountfor spiking and uneven mixing that occurs in the initialphases of the RCS leak in the reactor building. High initial concentrations l

1 may lead to erroneousfuel damage assumptions.

EAL Fission Product Barrier (FPB) - Potential Loss of Fuel Clad based on CSFS Core Cooling-Orange OR Heat Sink-Red l Please discuss whether there is any equivalent site-specific conditions which can be used to develop an l EAL which corresponds to this NUMARC/NESP 007 EAL.

The Deviation Matrh in Attachment C ns revisedfor this EAL to n~te that EAL 3.13 addresses the ~11 eat Sink Red" condition as a potential loss status. The initiating conditionfor EAL 3.13 is a complete loss of core heat removal capability.

EAL Fission Product Barrier (FPB) -- Loss of Reactor Coolant System based on Steam Generator Leak Plea::, discuss:

How the loss of adequate subcooling margin is determined and how this setpoint relates to the l NUMARC/NESP-007 and NEI EAL guidance.

Subcooling margin is determined by observing either of two Safety Parameters Display j System (SPDS) screens located on the main control board. As subcooling margin lowers, j the SPDS screen willflash red when adequate subcooling margin is lost. There is also an audible alarm on the SPDS monitor indicating a loss of adequate subcooling margin in ,

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Responses to NRC lssues Regarding Adoption of New EAL Methodology Page 8 of 9 the event of a loss ofboth SPDS systems, operators can easily plot RCSpressure and temperature on curves supplied in procedures.

Ilow the use of the setpoint "one or more injection valves" relates to the NUMARC/NESP-007 and NEl EAL guidance.

A note was included in the Deviation Matrixfor this EAL to explain how the setpoint "one or more injection valves" relates to the NEI guidance. The note explains that nonnal makeup is supplied to the RCS via normally operating, centnfugal makeup pumps and a makeup control valve. In the event operators are unable to maintain pressurizer level, procedures direct the operator to manually open additional high pressure injection (HPI) valves as required.

EAL Fission Product Barrier (FPB) - Potential Loss of Reactor Coolant System based on CSFS RCS Integrity - Red OR IIcat Sink-Red Please discuss whether there are any equivalent site-specific conditions which can be used to develop an EAL which corresponds to this NUMARC/NESP-007 EAL.

This EAL was revised to add an item on NDT Limits corresponding with the RCS integrity aspect of the NEI guidance. The additionalpotentiallossfactor will ensure that the RCS barrier is consideredpotentially lost whenever the system is at risk of a non-ductilefailure. Reference to this additionalitem and EAL 3.13 (for Heat Sink Red") has been included in the Deviation Matrh, Attachment C,for this EAL.

EAL Fission Product Barrier (FPB) - Loss of Containment based on Steam Generator Leak Please discuss how the CR EAL relates to the NUMARC/NESP-007 and NEI EAL guidance.

Ihe Basisfor this EAL was expanded to include thefollowing statementfrom NEl 97-03, Draft Revision 3: "This ' loss' EAL recognizes that SG tube leakage can represent a bypass of11 e containment barrier as well as a loss of the RCS barrier. Thefirst ' loss' EAL addresses the condition in which a ruptured steam generator is alsofaulted. This condition represents a bypass of the RCS and containment barriers. In conjunction with RCS Barrier 'Potentialloss' EAL M 2, this would always result in a Site Area Emergency. "

EAL Fission Product Barrier (FPB) - Loss of Containment based upon ICC curves Please discuss why the CR EAL did not include the 15-minute criteria provided in the NUMARC/NESP-007 and NEl EAL guidance. In addition the staff would like to discuss why a setpoint based upon reactor vessel level was not included.

This EAL was revised to include the 15-minute criteria relative to core exit thermocouples to address the NEl recommended time period during which it would be reasonable to determine if restoration procedures have been efective.

Responses to NRC Issues Regarding Adoption of New EAL Methodology Page 9 of 9 EAL Fission Product Barrier (FPB) - Loss of Containment based upon area Radiation Levels Please discuss whether this EAL might misclassify an event due to shine from the containment in the event of a severe accident.

This EAL was deleted.

EAL Fission Product Barrier (FPB) - Loss of Containment based upon Sump Indications Please discuss why CR did not include an EAL equivalent to the NUMARC/NESP-007 EAL

" Containment pressure or sump level response not consistent with LOCA conditions."

A new item to address containment pressure or sump level was added to the Containment Loss Factor EAL and its Basis, it states that " sump level or containment pressure not increasing indicates containment bypass and a loss of containment integrity. "

NUMARC/NESP-007 EAL SU4 Please discuss whether there are any monitors which may be used in this EAL.

FPC's EAL 3.9 addresses NEI EAL SU4 andfuel clad degradation. The Attachment C Deviation Matrix corresponding with this EAL states that FPC does not have afailedfuel monitoring system, but samples RCS on a daily basis. This statement was expanded to refer to the liquid RCS monitor (RM-L1) which monitors RCS letdown. An alarm on this monitor would trigger an RCS sample to be takenfor use in classifying the event.

NUMARC/NESP-007 EAL HUS Please discuss whether rapid depressurization of the secondary side should be included as an EAL.

Revision 18 included this condition as an EAL, classified at the Unusual Event level.

FPC stated in the referenced conference calls that this type of event would be classified as an Unusual Event under EAL 2.12for catastrophicfailure ofpressurized equipment or as an Alert under EAL 2.13. During the conference call on June 17,1999, the NRC stafindicated this was reasonable.

GENERAL In addition to the above noted changes, FPC made editorial changes to EALs 1.5, 2.11, 3.7, and 3.9.

Other minor editorial changes were made throughout.