ML051800397

Radiological Emergency Plan Annex for Dresden Station

ML051800397
Person / Time
Site:	Dresden
Issue date:	05/19/2005
From:	Jamie Benjamin, Mccain S Exelon Nuclear
To:	Document Control Desk, NRC/FSME
References
+kBR1SISP20060530, FOIA/PA-2006-0130 EP-AA-1004, Rev 19
Download: ML051800397 (126)

Text

EP-AA-1004 Revision 19 EXELON NUCLEAR RADIOLOGICAL EMERGENCY PLAN ANNEX FOR DRESDEN STATION Submitted: Jeff Benjamin for S. McCain Date: 05/19/05 Midwest Region Emergency Preparedness Manager Authorized: J. A. Benjamin Date: 05/19/05 Vice President - Licensing and Regulatory Affairs

Table of Contents Section Page Section 1: Introduction ....................................................................................................DR 1-1 1.1 Facility Description........................................................................................DR 1-1 1.2 Emergency Planning Zone.............................................................................DR 1-1 Section 2: Organizational Control of Emergencies ........................................................DR 2-1 2.1 Non-Exelon Nuclear Support Groups............................................................DR 2-1 Section 3: Classification of Emergencies ........................................................................DR 3-1 Section 4: Emergency Measures .....................................................................................DR 4-1 4.1 Notification of the Emergency Organization.................................................DR 4-1 4.2 Assessment Actions .......................................................................................DR 4-1 4.3 Protective Actions for the Offsite Public.......................................................DR 4-1 4.4 Protective Actions for Onsite Personnel........................................................DR 4-2 Section 5: Emergency Facilities and Equipment ............................................................DR 5-1 5.1 Emergency Response Facilities .....................................................................DR 5-1 5.2 Assessment Resources ...................................................................................DR 5-1 5.3 Protective Facilities and Equipment ..............................................................DR 5-5 5.4 First Aid and Medical Facilities.....................................................................DR 5-5 APPENDIXES Appendix 1: NUREG-0654 Cross-Reference Appendix 2: Station Letters of Agreement Appendix 3: EAL Wallboard REVISION HISTORY Revision 0; July 1980 Revision 6j; June 1996 Revision 17; August 27, 2003 Revision 1; April 1981 Revision 6k; January 1997 Revision 18, December 2004 Revision 2; June 1982 Revision 6l; February 1997 Revision 3; September 1984 Revision 6m; May 1997 Revision 4; March 1986 Revision 6n; January 5, 1998 Revision 5; February 1987 Revision 6p; August 14, 1998 Revision 6; January 1991 Revision 7; May 13, 1999 Revision 6a; July 1992 Revision 8; February 11, 2000 Revision 6b; August 1992 Revision 9; May 22, 2000 Revision 6c; November 1992 Revision 10; January 8, 2001 Revision 6d; April 1993 Revision 11; May 3, 2001 Revision 6e; December 1993 Revision 12; October 8, 2001 Revision 6f; January 1994 Revision 13; October 31, 2001 Revision 6g; January 1995 Revision 14; January 3, 2002 Revision 6h; December 1995 Revision 15; September 12, 2002 Revision 6i; February 1996 Revision 16; July 31, 2003 May 2005 ii EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Section 1: Introduction As required in the conditions set forth by the Nuclear Regulatory Commission (NRC) for the operating licenses for the Exelon Nuclear Stations, the management of Exelon recognizes its responsibility and authority to operate and maintain the nuclear power stations in such a manner as to provide for the safety of the general public.

The Exelon Emergency Preparedness Program consists of the Exelon Nuclear Standardized Emergency Plan (Emergency Plan), Station Annexes, Emergency Plan Implementing Procedures, and associated program administrative documents. The Emergency Plan outlines the basis for response actions that would be implemented in an emergency. Planning efforts common to all Exelon Nuclear stations are encompassed within the Emergency Plan.

This document serves as the Dresden Station Annex and contains information and guidance that is unique to the station. This includes Emergency Action Levels (EALs), and facility geography and location for a full understanding and representation of the stations emergency response capabilities.

The Station Annex is subject to the same review and audit requirements as the Emergency Plan.

1.1 Facility Description Dresden Station, Units 1, 2 and 3, is located in the Goose Lake Township of Grundy County in northeastern Illinois. Unit 1 is in permanent shutdown (see Figure 1-1).

The plant consists of three Boiling Water Reactor (BWR) Nuclear Steam Supply Systems (NSSS) and turbine generators provided by General Electric Company. Unit 1 is a dual cycle boiling water reactor designed for a power output of 700 MWt and has officially been retired as of August 31, 1984. Units 2 and 3 are equipped with nuclear steam supply systems (NSSS) designed for a power output of 2957 MWt.

The station property consists of a 953 acre tract of land with boundaries generally following the Illinois River to the north, the Kankakee River on the south and east and the Elgin, Joliet and Eastern Railway right-of-way on the west. Exelon is the sole owner of the 953 acre tract subject only to an easement of the U.S. Government for an access road to Dresden Island Lock and Dam maintained and operated by the U.S. Corps. of Engineers. This road traverses the site from north to south ~ 0.8 mile west of the plant.

In addition to ownership of the 953 acre tract, Exelon Nuclear also leases approximately 17 acres in two narrow strips of river frontage located near the northeast corner of the site from the State of Illinois. The terms of the lease provide that these "buffer" strips shall remain idle.

For more specific site location information, refer to the Station UFSAR.

1.2 Emergency Planning Zone The plume exposure Emergency Planning Zone (EPZ) for Dresden Station is an area surrounding the Station with a radius of about ten miles. (Exact boundaries are determined by the State of Illinois). Refer to Figure 1-1.

The ingestion pathway EPZ for Dresden Station is an area surrounding the station with a radius of about 50 miles.

May 2005 DR 1-1 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Figure 1-1: Dresden Station Location and 10 Mile EPZ May 2005 DR 1-2 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Section 2: Organizational Control of Emergencies Initial response to any emergency is by the normal plant organization present at the site. This organization includes positions that are onsite 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day and is described in Section B.1 of the Emergency Plan.

Once an emergency is declared, the Emergency Response Organization is activated according to Section B.4 of the Emergency Plan and Implementing Procedures.

2.1 Non-Exelon Nuclear Support Groups Exelon Nuclear has contractual agreements with several companies whose services would be available in the event of a radiological emergency. These agencies and their available services are listed in Appendix 3 of the Emergency Plan.

Emergency response coordination with governmental agencies and other support organizations is discussed in Section A of the Emergency Plan.

Agreements exist on file at Dresden Station with several support agencies. These agencies and their support roles are listed in Appendix 2, Station Letters of Agreement.

May 2005 DR 2-1 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Section 3: Classification of Emergencies Section D of the Emergency Plan describes the classification of emergencies into five levels. The first four are the Unusual Event, Alert, Site Area Emergency and General Emergency. These classification levels are entered by meeting the criteria of Emergency Action Levels (EALs) provided in this section of the Annex. These classification levels are escalated from least severe to most severe according to relative threat to the health and safety of the public and emergency workers. The fifth level is Recovery. Recovery can be considered as a phase of the emergency and is entered by meeting emergency termination criteria provided in EP-AA-111 Emergency Classification and Protective Action Recommendations.

The Initiating Conditions and the Emergency Action Level Threshold Values are provided in a matrix. An emergency is classified by assessing plant conditions and comparing abnormal conditions to Initiating Conditions and Threshold Values for each Emergency Action Level.

Individuals responsible for the classification of events refer to the Initiating Condition and Threshold Values on the matrix of the appropriate station Emergency Plan Annex (this document). This matrix will contain Initiating Conditions, EAL Threshold Values, Mode Applicability Designators, appropriate EAL numbering system and additional guidance necessary to classify events. It may be provided as a user aid.

The matrix is set up in four Recognition Categories. The first is designated as "R" and relates to Abnormal Radiological Conditions / Abnormal Radiological Effluent Releases. The second is designated as "F" and relates to Fission Product Barrier Degradation. The third is designated as "M" and relates to System Malfunctions. The fourth is designated as "H" and relates to Hazards and Other Conditions.

The matrix is designed to provide an evaluation of the Initiating Conditions from the worst conditions (General Emergencies) on the left to the relatively less severe conditions on the right (Unusual Events). Evaluating conditions from left to right will reduce the possibility that an event will be under classified. All Recognition Categories should be reviewed for applicability prior to classification.

The Initiating Conditions are coded with a two letter and one number code. The first letter is the Recognition Category designator, the second letter is the Classification Level, U for Unusual Event, A for Alert, S for Site Area Emergency and G for General Emergency. The number is a sequential number for that Recognition Category series. All Initiating Conditions that are describing the severity of a common condition (series) will have the same number.

The code may then be used to reference a corresponding Threshold Value page(s), which provides additional information pertaining to the Initiating Condition;

• Threshold Value

• Mode Applicability

• Basis May 2005 DR 3-1 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Emergency Action Levels are the measurable, observable detailed conditions that must be met in order to classify the event. Classification is not to be made without referencing, comparing and satisfying the Threshold Values specified in the Emergency Action Levels.

Mode Applicability provides the unit conditions when the Emergency Action Levels represent a threat. The Basis provides explanations and justification for including the Initiating Condition and Emergency Action Level.

Definitions are provided for terms having specific meaning as they relate to this procedure. Site specific definitions are provided for terms with the intent to be used for that particular Initiating Condition/Threshold Value and may not be applicable to other uses of that term at other sites, the Emergency Plan or procedures.

References are also included to documents that were used to develop the EAL Threshold Values.

References to the Emergency Director means the person in Command and Control as defined in the Emergency Plan. Classification of emergencies is a non-delegable responsibility of Command and Control for the onsite facilities with responsibility assigned to the Shift Emergency Director (Shift Manager) or the Station Emergency Director (TSC). Classification of emergencies remains the responsibility of the applicable onsite facility even after Command and Control is transferred to the Corporate Emergency Director (EOF).

Classifications are based on evaluation of each Unit. All classifications are to be based upon VALID indications, reports or conditions. Indications, reports or conditions are considered VALID when they are verified by (1) an instrument channel check, or (2) indications on related or redundant indications, or (3) by direct observation by plant personnel, such that doubt related to the indications operability, the conditions existence, or the reports accuracy is removed.

Implicit in this definition is the need for timely assessment.

When two or more Emergency Action Levels are determined, declaration will be made on the highest classification level for the Unit. When both units are affected, the highest classification for the Station will be used for notification purposes and both units classification levels will be noted.

May 2005 DR 3-2 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Key Definitions of Terms used in Threshold Values ADEQUATE CORE COOLING - Core submergence. steam cooling or spray cooling as defined in the Emergency Operating Procedures (EOPs).

ALERT - Events are in progress 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 exposure levels.

CONFIRMED - Indication or report is verified to be valid by source outside the Control Room.

• Calls to the National Earthquake Center or reports from university monitoring stations are the primary confirmation source for seismic event.

• Security or the NRC are the confirmation sources for security events.

CONTROL - Placing all local control switches in local control necessary for operation from remote panels and the Shift Manager has determined that the systems for controlling reactivity, core cooling and heat sink functions are established.

ENVIRONMENT - Outside the Containment boundary.

EXPLOSION - A rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to nearby structures or equipment. This includes any High Energy Line Break.

EXTORTION - An attempt to cause an action at the station by threat of force.

GENERAL EMERGENCY - Events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity.

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

IDENTIFIED - Identified leakage is leakage that is pumped from the drywell equipment sump.

This number is determined by the total flow appearing on the drywell equipment drain recorder when the equipment drain sump is pumped.

IMMINENT - Mitigation actions have been ineffective and trended information indicates that the event or condition will occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

MANUAL SCRAM - Any set of actions by the Control Room Operators at the Reactor Control Panels which causes control rods to be rapidly inserted into the core and brings the reactor to a condition where it will remain shutdown under all conditions without the use of Boron injection.

A manual SCRAM may be accomplished by any of the following: (1) placing mode switch in shutdown; (2) using manual SCRAM pushbuttons, or (3) manual ARI initiation using pushbuttons. If the ARI system successfully inserts rods prior to operators using pushbuttons that is considered a successful Manual SCRAM.

MOST - 75% of safety system annunciators or indications are lost OR a significant risk that a degraded plant condition could go undetected exists.

May 2005 DR 3-3 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Key Definitions of Terms used in Threshold Values (cont.)

NORMAL OPERATIONS- Activities at the plant site associated with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative procedures. Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure from NORMAL OPERATIONS.

POTENTIAL- Mitigation actions are not effective and trended information indicates that the parameters are outside desirable bands and not stable or improving.

PPDS - Plant Process Data System, a system that displays plant information on personal computers. The system contains pre-selected computer points organized to allow easy access to critical data.

SIGNIFICANT TRANSIENT - Evolutions such as a SCRAM, a 25% thermal power change, ECCS actuation, or thermal power oscillation 10%.

SITE BOUNDARY - For classification and dose projection purposes, the boundary is a 800-meter (1/2 mile) radius around the plant. This is the nearest distance from potential release points at which Protective Actions would be required for members of the public.

SITE AREA EMERGENCY - Events are in progress or have occurred which involve actual or likely failure of plant functions needed for protection of the public. Any releases are not expected to result in exposure levels that exceed EPA Protective Action Guideline exposure levels except near the site boundary.

SUSTAINED - 15 minutes in duration.

TOXIC - Exposure to the worker in excess of the limits specified in 29 CFR 1910.1000. In practice, this should be considered for concentrations that are capable of incapacitating the worker.

UNIDENTIFIED - Unidentified leakage is leakage that is pumped from the drywell floor drain sump. This number is determined by the total flow appearing on the drywell floor drain recorder when the floor drain sump is pumped.

UNISOLABLE - A breach or leak that cannot be isolated from the Control Room or within 15 minutes by operators in the field.

UNPLANNED - Not the result of an intended evolution and requiring corrective or mitigating actions.

UNUSUAL EVENT - Events are in progress 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 are expected unless further degradation of safety systems occurs.

May 2005 DR 3-4 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Emergency Action Level Technical Basis Page Index General Site Area Alert Unusual Event EAL Pg. EAL Pg. EAL Pg. EAL Pg RG1 3-14 RS1 3-16 RA2 3-18 RU2 3-20 RA3 3-22 RU3 3-24 FG1 3-25 FS1 3-26 FA1 3-27 FU1 3-28 MG1 3-53 MS1 3-55 MA1 3-56 MU1 3-57 MA2 3-58 MG3 3-59 MS3 3-61 MA3 3-62 MS4 3-63 MU4 3-64 MS5 3-65 MS6 3-66 MA6 3-67 MU6 3-68 MS7 3-69 MA7 3-71 MU7 3-72 MU8 3-73 MU9 3-74 MU10 3-75 MA11 3-76 MU11 3-77 MA12 3-78 MU12 3-79 HG1 3-80 HS1 3-81 HA1 3-82 HU1 3-83 HG2 3-85 HS2 3-86 HA2 3-87 HU2 3-88 HS3 3-89 HA3 3-90 HA4 3-91 HU4 3-93 HA5 3-95 HU5 3-97 HA6 3-98 HU6 3-99 HU7 3-100 E-RU1 3-101 E-HU1 3-102 E-HU2 3-103 Containment 1.a 3-29 Fuel Clad 2.a 3-40 RCS 3.a.1 3-45 1.b 3-30 2.b 3-41 3.a.2 3-46 1.c.1 3-31 2.c 3-42 3.b 3-48 1.c.2 3-32 2.d 3-44 3.c 3-49 1.c.3 3-33 3.d 3-51 1.c.4 3-34 3.e 3-52 1.d 3-36 1.e 3-38 1.f 3-39 May 2005 DR 3-5 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ABNORMAL RAD LEVELS / EFFLUENTS RG1 Actual or projected Site Boundary dose: 1 2 3 4 5 D RS1 Actual or projected Site Boundary dose: 1 2 3 4 5 D 1 Rem TEDE. 100 mRem TEDE. Table R3 Table R4 OR OR Areas Req. Continuous Occupancy Areas Requiring Infrequent Access 5 Rem CDE Thyroid 500 mRem CDE Thyroid.

• Main Control Room

• HPCI Cubicle EAL Threshold Value: EAL Threshold Value:

1. Radiological release in excess of Table R1 General

• Central Alarm Station

• East and West LPCI Pump Areas

1. Radiological release in excess of Table R1 Site Area Emergency threshold UNLESS release can be determined Emergency threshold UNLESS release can be determined to

• Secondary Alarm Station

• East and West CRD Module Areas to be below available Table R2 General Emergency be below available Table R2 Site Area Emergency

• Radwaste Control Room

• Vessel Instrument Rack Area Radiological Effluents thresholds within 15 mins. thresholds within 15 mins.

OR

• Gatehouse

• RWCU Area OR

2. Radiological release exceeds ANY Table R2 column 2. Radiological release exceeds ANY Table R2 column Site

• Isolation Condenser Area General Emergency threshold. Area Emergency threshold.

RA2 Release 1 2 3 4 5 D RU2 Release 1 2 3 4 5 D 10 X ODCM limit for 15 mins. 2 X ODCM limit for 60 mins.

EAL Threshold Value: EAL Threshold Value:

1. Unplanned radiological release in excess of Table R1 Alert 1. Unplanned radiological release in excess of Table R1 threshold for 15 mins. UNLESS release can be determined to be Unusual Event threshold for 60 mins. UNLESS release below available Table R2 Alert thresholds within this period.

can be determined to be below available Table R2 Unusual OR Event thresholds within this period.

2. Unplanned Gaseous or Liquid radiological release exceeds ANY Table R2 column Alert threshold for 15 mins. OR

2. Unplanned Gaseous OR Liquid radiological release exceeds ANY Table R2 column Unusual Event threshold for 60 mins.

RA3 In plant radiation levels impede plant 1 2 3 4 5 D RU3 Increase in plant radiation levels by a 1 2 3 4 5 D Abnormal Rad operations. factor of 1000.

EAL Threshold Value: EAL Threshold Value:

1. Radiation readings 15 mR/hr. in ANY area requiring Area Radiation Monitor readings OR survey results indicate an Levels continuous occupancy, Table R3. unplanned increase in plant radiation levels by a factor of 1000.

OR

2. Radiation readings 2500 mR/hr. in ANY area requiring infrequent access for safe plant operations, Table R4.

Table R1 -- Effluent Monitor Thresholds Table R2 Dose Assessment Thresholds Low, mid or high Range General Site Area Alert Unusual Event Method General Emergency Site Area Emergency Alert Unusual Event As determined by DOP 1700-10 or PPDS Emergency Emergency Sample/ Release 10 X ODCM limit for 2 X ODCM limit Total Station Release: N/A N/A Package 15 mins. for 60 mins.

Units 2 / 3 Rx Bldg SPING 1.5 E+07 µCi/sec 1.5 E+06 µCi/sec > 8.5 E+05 µCi/sec 1.7 E+05 µCi/sec Plus for 15 mins. for 60 mins.

Field Team 1 R/hr. for 1 hr.

Units 2 / 3 Chimney SPING 100 mR/hr. for 1 hr.

Monitoring OR 10 mR/hr.

OR N/A (At or beyond Site 5 Rem CDE for 15 mins.

500 mRem CDE Thyroid Boundary) Thyroid Plant Modes (white boxes indicate applicable modes) Dose Projection 1 Rem TEDE 100 mRem TEDE (At or beyond Site OR OR N/A N/A 1 Power Operations 2 Startup 3 Hot Shutdown 4 Cold Shutdown 5 Refueling D Defueled 5 Rem CDE Thyroid 500 mRem CDE Thyroid Boundary)

May 2005 DR 3-6 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear This Page Intentionally Left Blank May 2005 DR 3-7 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT FISSION PRODUCT BARRIER DEGRADATION FG1 Loss of 2 Fission Product Barriers with 1 2 3 FS1 Loss or Potential Loss of 2 Fission 1 2 3 FA1 Loss or Potential Loss of Fuel Clad 1 2 3 FU1 Loss or Potential Loss of Containment 1 2 3 FPB Loss /

Loss or Potential loss of third barrier. Product Barriers or RCS Barrier. barrier.

EAL Threshold Value: EAL Threshold Value: EAL Threshold Value: EAL Threshold Value:

Review of Fission Product Barrier Matrix indicates: Review of Fission Product Barrier Matrix indicates ANY 2 of Review of Fission Product Barrier Matrix indicates EITHER: Review of Fission Product Barrier Matrix indicates:

Potential Loss

• Loss of 2 Fission Product Barriers. the following:

• Loss OR Potential Loss of Fuel Clad.

• Loss OR Potential Loss of Containment.

AND

• Loss OR Potential Loss of Fuel Clad. OR

• Loss OR Potential Loss of the THIRD barrier.

• Loss OR Potential Loss of RCS.

• Loss OR Potential Loss of RCS.

• Loss OR Potential Loss of Containment.

Fission Product Barrier Matrix

1. CONTAINMENT LOSS POTENTIAL LOSS 2. FUEL CLAD LOSS POTENTIAL LOSS RCS LOSS POTENTIAL LOSS Rapid pressure decrease in a. Primary Coolant Coolant Activity Unisolable Main Steam Line None a.1. RCS Leak Unisolable RCS Leakage containment after increase without Activity. 300 uCi/gm I-131 dose equivalent. break.

Rate. 50 gpm.

containment spray. 1. BOTH of the following:

a. Drywell or Torus Containment pressure OR

• Reactor water level -164 (Minimum Unisolable primary Pressure. 62 psig.

Drywell OR Torus pressure Steam Cooling) system leakage outside response NOT consistent with AND Reactor Water level - primary containment

b. Reactor Vessel LOCA conditions during LOCA. Water level.

• NEITHER core spray loop flows 4750 gpm 143 into specified area AND OR (TAF).

Drywell OR Torus Hydrogen EITHER:

concentration 6 %. 2. Reactor water level -191 Area Temperatures

b. Drywell or Torus (2/3 Core Height). a.2. RCS Leak None Table F1 Max.

Hydrogen None AND Rate.

Normal Value.

Concentration. c. Drywell Drywell OR Torus Oxygen Drywell Radiation in the Fuel Cladding Lost OR Radiation None concentration 5 %. range of Graph 2.c. Area Radiation, Monitors Table F2 Max.

d. Steamline MSL radiation levels 2 X the High None Normal Value.

Radiation. Radiation Alarm setpoint.

c Containment 1.c. CONTAINMENT BREACHED / BYPASSED Drywell pressure ECCS Breached / See 1.c. Containment Breached / Bypassed Table LOSS POTENTIAL LOSS setpoint.

Bypassed. b. Drywell Failure of ALL automatic containment AND None c.1. Containment Pressure.

isolation valves in ANY ONE line Pressure increase due to Breached / None reactor coolant leakage.

penetrating the primary containment to Bypassed. c. Drywell Drywell Radiation monitor isolate on isolation actuation signal.

Drywell radiation monitors Radiation reading in the RCS Lost None

d. Drywell c.2. Containment Breach of primary containment, NOT part in the Containment None Monitor range of Graph 3.c.

Radiation None Breached / of Primary Containment Isolation, that is Potential Loss range of d. Reactor Vessel Reactor Vessel level Monitors. Bypassed. unisolable from the Main Control Room. None Graph 1.d. Water Level. -143 (TAF) c.3. Containment Intentional venting of primary containment Entry into Primary Primary system relief valve

e. Reactor Vessel Breached / per DEOPs or SAMG-1 AFTER None e. Primary None Containment Flooding stuck open.

Water Level. Bypassed. containment isolation. System Relief None SAMG-1. AND

f. Drywell Drywell temperature Valves.

None Unisolable primary system leakage outside Torus temperature 110ºF.

Temperature 281ºF.

containment as indicated by EITHER:

c.4. Containment 1. Area Temperatures, Table F1 Max.

Plant Modes (white boxes indicate applicable modes) Breached / Safe Value. None Bypassed. OR 1 Power Operations 2 Startup 3 Hot Shutdown 2. Area Radiation levels, Table F2 Max.

Safe Value.

4 Cold Shutdown 5 Refueling D Defueled May 2005 DR 3-8 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Fission Product Barrier Matrix - Support Tables and Graphs Note: From time after shutdown to 0.5 hrs., use the 0.5 hrs. value. Note: From time after shutdown to 0.5 hrs., use the 0.5 hrs. value.

9000 2500 Graph 1.d Graph 2.c 8000 00 77 62 00 20 7000 70 2000 25 17 19 64 Containment Potential Loss 50 Fuel Cladding Lost Drywell Radiation Reading (R/hr) 17 Drywell Radiation Reading (R/hr) 75 6000 57 04 16 22 44 50 1500 14 5000 55 12 4000 Containment Intact 39 1000 93 1

3000 30 0

Fuel Cladding Intact 76 2000 99 16 500 5 42 1

1000 32 4

62 6 1 8 15 20 12 50 32 0

0 0.5 1 2 3 4 5 6 7 8 9 10 20 30 40 50 60 70 80 90 100 0.5 1 2 3 4 5 6 7 8 9 10 20 30 40 50 60 70 80 90 100 Time After Shutdown (hrs)

Dresden & Quad Cities Time After Shutdown (hrs)

Dresden & Quad Cities Note: From time after shutdown to 0.5 hrs., use the 0.5 hrs. value.

160 Graph 3.c Table F1 -- Area Temperatures Table F2 -- Area Radiation 140 Max. Max.

Max. Max.

Normal Safe 11 8

Normal Safe 120 mR/hr mR/hr 0 Drywell Radiation Reading (R/hr) 11 Reactor Coolant System Lost HPCI Area 150 ºF 210 ºF HPCI Cubicle Unit 2 (3) 150 (100) 2500 10 0

100 Shutdown Cooling Pump Room 150 ºF 180 ºF East LPCI Pump Area 12 2500* 92 Shutdown Cooling Heat Exchanger 83 150 ºF 180 ºF West LPCI Pump Area 9 2500*

Room 80 72 RWCU Demin Rooms 150 ºF 210 ºF East CRD Module Area 30 2500*

RWCU Pump and Heat Exchanger West CRD Module 60 150 ºF 210 ºF 50 2500* 53 Area Area Reactor Coolant System Intact 43 Vessel Instrument Rack Isolation Condenser Area 150 ºF 180 ºF 30 2500* 40 Area RWCU Area 30 2500* 24 20 18 Isolation Condenser 10 2500 9 Area 3 2

• Measured by Local Survey 0 0.5 1 2 3 4 5 6 7 8 9 10 20 30 40 50 60 70 80 90 100 Time After Shutdown (hrs)

Dresden & Quad Cities May 2005 DR 3-9 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTIONS MG1 Prolonged loss (>4 hrs) of all 1 2 3 MS1 Loss of all ECCS bus AC 1 2 3 MA1 Power to ECCS Buses reduced to a 1 2 3 MU1 Unplanned loss of all offsite AC 1 2 3 4 5 D ECCS AC power. power for 15 minutes. single source for > 15 minutes such power to a units ECCS buses.

EAL Threshold Value:

that failure of that single source would EAL Threshold Value:

EAL Threshold Value:

Loss of AC Power

1. Loss of all AC power to ECCS bus 23-1 (33-1). 1. Loss of all AC power to ECCS bus 23-1 (33-1).

result in Unit Blackout. Unplanned loss of ability to power ECCS buses from:

EAL Threshold Value: TR-21(TR-31).

AND AND Available ECCS bus AC power reduced to only one of the following

2. Loss of all AC power to ECCS bus 24-1(34-1). 2. Loss of all AC power to ECCS bus 24-1(34-1). sources for > 15 minutes: AND AND AND

• Reserve auxiliary transformer TR-22 (TR-32).

TR-22 (TR-32).

3. Any of the following: 3. Failure to restore power within 15 min. to at least:

• Unit auxiliary transformer TR-21 (TR-31).

• Restoration of power to EITHER bus 23-1 (33-1) Bus 23-1 (33-1).

• Unit Emergency Diesel Generator.

or 24-1(34-1) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is NOT likely OR

• Shared Emergency Diesel Generator.

OR

• Unit crosstie breakers.

Bus 24-1(34-1).

• Conditions are imminent that a Loss of two fission product

• SBO Diesel Generator.

barriers AND Loss or Potential Loss of the third (FG1) will occur prior to restoration of AC power to the Unit. MA2 Loss of all ECCS bus AC power for 15 min. 4 5 D EAL Threshold Value:

1. Loss of all AC power to ECCS bus 23-1 (33-1).

MG3 Auto and manual SCRAM not AND 1 2 3 successful and loss of core cooling 2. Loss of all AC power to ECCS bus 24-1(34-1).

Failure of Reactor or heat sink. AND

3. Failure to restore power within 15 min. to at least:

EAL Threshold Value: Bus 23-1 (33-1).

1. Failure of BOTH automatic and manual SCRAMs to establish OR shutdown criteria..

Bus 24-1(34-1).

Protection System AND

2. ANY ONE of the following when reactor power is > IRM Range 7: MS3 Auto and Manual SCRAM not 1 2 3 MA3 Auto SCRAM not successful. 1 2 3

• Indications exist that adequate core cooling is NOT assured. successful. EAL Threshold Value:

OR EAL Threshold Value: Failure of the Reactor Protection System to initiate AND complete an

• Indications exist that Torus temperature has OR is predicted Failure of BOTH automatic AND manual SCRAMs to establish automatic reactor SCRAM AND a successful manual SCRAM occurred.

to exceed 110 °F. shutdown criteria.

MS4 Loss of vital 125 VDC power. 1 2 3 MU4 Loss of vital 125 VDC power. 4 5 Loss of EAL Threshold Value: EAL Threshold Value:

125 VDC Battery buses #2 AND #3 bus voltage 110 volts for 15 DC 125 VDC Battery buses #2 AND #3 bus voltage mins.

110 volts for 15 mins.

MS5 Loss of Mode 4 Capability (HCTL 1 2 3 Heat Sink Exceeded).

EAL Threshold Value:

Emergency Depressurization is required due to the inability to maintain Torus bulk temperature below the Heat Capacity Limit curve of DEOP 0200-01.

MS6 Loss of annunciators and indications 1 2 3 MA6 Loss of annunciators or indications 1 2 3 MU6 Unplanned loss of annunciators or 1 2 3 with transient in progress. requiring increased monitoring.

EAL Threshold Values:

indications for 15 mins.

Plant Modes (white boxes indicate applicable modes)

Loss of Annunciators EAL Threshold Values: 1. Unplanned loss of MOST OR ALL Safety System Annunciators on

1. Loss of ALL Safety System Annunciator Panels:

EAL Threshold Value:

Panels (below) for 15 mins.: Unplanned loss of MOST OR ALL Safety System Annunciators 902(3)-3

• 902(3)-3 OR indications for 15 mins. Panels:

1 Power Operations 2 Startup 902(3)-5

• 902(3)-5 902(3)-8

• 902(3)-3

• 902(3)-8 AND

• 902(3)-5 AND 3 Hot Shutdown 4 Cold Shutdown 2. A significant transient is in progress

2. ANY of the following:

• 902(3)-8 AND

• A significant transient is in progress

• 3. Loss of ALL indications needed to monitor ANY 5 Defueled OR Refueling D of the following: Criticality OR Core Heat Removal OR Fission Product Barrier status *

• Loss of ALL indications needed to monitor; Criticality OR Core Heat Removal OR Fission Product Barrier status.

May 2005 DR 3-10 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTIONS (cont.)

MS7 Loss of decay heat removal and 4 5 MA7 Inability to establish or maintain Mode 4. 3 4 5 MU7 Offgas radiation trip or high reactor 1 2 3 4 5 D Decay Heat core uncovery. coolant activity.

EAL Threshold Values: EAL Threshold Values:

1. Reactor core is OR will be uncovered. 1. Inability to establish Mode 4 when required. EAL Threshold Values:

Removal / Coolant AND OR 1. Offgas system isolation due to valid Offgas post-treatment radiation

2. ANY of the Following: 2. Inability to maintain Mode 4 conditions as indicated by EITHER: monitor signal.

• RCS Temperature increase that exceeds 212 °F. OR Activitiy

• Reactor coolant temperature > 212 °F.

OR OR 2. Coolant activity 5 µCi/gm I-131 dose equivalent.

• Uncontrolled temperature increase approaching 212 °F.

• Uncontrolled RCS temperature increase approaching 212 °F.

MU8 RCS leakage 1 2 3 RCS Leakage EAL Threshold Values:

Table M1 Table M2 Onsite Communications Equipment Offsite Communications Equipment 1. Unidentified RCS leakage into the primary containment > 10 gpm.

OR

• Plant Radio System

• All telephone lines (commercial and microwave) 2. Total RCS leakage (identified + unidentified) into primary

• Plant Paging System

• NARS containment > 35 gpm.

• Sound Power Phones

• ENS

• In-plant Telephones

• HPN MU9 Unplanned loss of all onsite or 1 2 3 4 5 D Loss of

• Cellular Phones offsite communications capabilities.

EAL Threshold Values:

Communic.

1. Loss of ALL onsite communications equipment, Table M1.

OR

2. Loss of ALL offsite communications capability, Table M2.

MU10 Inability to reach required mode within 1 2 3 Technical Technical Specification time limits.

EAL Threshold Value:

Specs Plant is not brought to required mode within Tech. Spec. LCO Action Statement time.

MA11 Major fuel damage or fuel uncovery 1 2 3 4 5 D MU11 Unplanned decrease in Spent Fuel Pool 1 2 3 4 5 D Spent Fuel Events in the Spent Fuel Pool. level.

EAL Threshold Values:

EAL Threshold Value:

1. Refuel Floor Radiation Monitors Unit 2 / 3 reading 10 R/hr.

Inability to maintain Unit 2 OR 3 Spent Fuel Pool water level 33.9 ft.

OR above the fuel.

2. Report of visual observation of a rapid decrease of Spent Fuel Pool water level such that irradiated fuel is predicted to become uncovered.

MA12 Unplanned loss of Refueling Cavity level. 5 MU12 Unplanned loss of Refueling Cavity 5 Plant Modes (white boxes indicate applicable modes) EAL Threshold Values: Level.

Refueling

1. Report of visual observation of a rapid decrease of refueling cavity EAL Threshold Value:

1 Power Operations 2 Startup 3 Hot Shutdown water level such that irradiated fuel is predicted to become uncovered.

Inability to maintain Refueling Cavity Level -59.

OR 4 Cold Shutdown 5 Refueling D Defueled 2. Reactor vessel level -143 (TAF).

May 2005 DR 3-11 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS HG1 Security event resulting in loss of 1 2 3 4 5 D HS1 Confirmed Security Event in a 1 2 3 4 5 D HA1 Confirmed Security Event in the 1 2 3 4 5 D HU1 Confirmed Security Event that 1 2 3 4 5 D physical control of the facility. Vital Area. Protected Area. indicates a potential degradation in EAL Threshold Value: EAL Threshold Values: EAL Threshold Values:

level of safety of the plant.

1. Loss of physical control of the Main Control Room due to a 1. Intrusion into plant Vital Area by a hostile force. 1. Intrusion into the Protected Area by a hostile force. EAL Threshold Values:

security event. OR OR 1. A credible threat to the station reported by the NRC.

OR OR

2. A security event that results in the loss of control of ANY Vital 2. Other Security Events of increasing severity that persist for 2. An Actual Threat that meets ALL of the following criteria:

2. Loss of physical control of the facility (remote shutdown Area (other than the Main Control Room).

• A credible threat reported by ANY other outside agency OR 30 mins. (see Table H2 for example events).

Security Events capability) due to a security event. OR security procedures.

3. Imminent loss of physical control of the facility due to a security Table H2 - Examples of Security Events AND event.

• Is specifically directed towards the station.

Credible bomb device. AND OR

• Is imminent (within 2 hrs.).

4. A confirmed bomb device discovered in a Vital Area. Extortion. OR

3. Bomb device discovered within the Protected Area AND outside a Suspicious fire or explosion. plant Vital Area.

OR Significant security system hardware failure. 4. Confirmed tampering with safety related equipment.

OR Loss of guard post contact. 5. A hostage situation that disrupts normal plant operations.

OR

6. A disturbance (civil, internal OR strike) which disrupts normal plant operations.

HG2 Conditions indicate imminent core 1 2 3 4 5 D HS2 Conditions indicate actual or likely 1 2 3 4 5 D HA2 Conditions indicate actual or potential 1 2 3 4 5 D HU2 Conditions indicate a potential 1 2 3 4 5 D damage and release affecting the failure of functions needed for substantial degradation of the level of degradation of the level of safety of the public. public protection. safety of the plant. plant.

Discretionary EAL Threshold Value: EAL Threshold Value: EAL Threshold Value: EAL Threshold Value:

1. Actual OR imminent core degradation AND potential loss of Other conditions exist which in the judgment of the Emergency Other conditions exist which in the judgment of the Emergency Director Other conditions exist which in the judgment of the Emergency Director Containment. Director indicate actual OR likely major failures of plant functions indicate that plant safety systems may be degraded AND that increased indicate a potential degradation in the level of safety of the plant OR needed for protection of the public. monitoring of plant functions is warranted. (See Table H1 for example conditions).

2. Potential uncontrolled radionuclide release which can reasonably be expected to reach or exceed 1 Rem TEDE OR 5 Rem CDE Table H1 - Example Conditions thyroid plume exposure levels at the Site Boundary. An event onsite OR near the site which requires the assistance of the local Police OR Fire Department.

Uncontrolled RCS cooldown.

HS3 Main Control Room evacuated and 1 2 3 4 5 D HA3 Main Control Room evacuation 1 2 3 4 5 D Plant conditions require operation outside analyzed conditions.

control not established in 30 initiated. Approaching ANY Unusual Event Threshold Value as an Control Room mins. EAL Threshold Value: anticipatory judgment call.

EAL Threshold Values: Entry into DSSP 0100-CR for Control Room evacuation. Abnormal operating conditions persist because mitigation activities

1. Main Control Room evacuation initiated. are NOT effective.

Evacuation AND Additional assistance is required to ensure the protection of the health and safety of the workers and the public.

2. Control of the plant CANNOT be established per DSSP 0100-CR in 30 mins. Unusual aircraft activity over the Protected Area.

Plant Modes (white boxes indicate applicable modes) 1 Power Operations 2 Startup 3 Hot Shutdown 4 Cold Shutdown 5 Refueling D Defueled May 2005 DR 3-12 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS (cont.)

ISFSI Malfunctions HA4 Natural or destructive phenomena inside a Vital Area. 1 2 3 4 5 D HU4 Natural or destructive phenomena inside 1 2 3 4 5 D EAL Threshold Values: the Protected Area or Switchyard.

Unusual Event

1. Confirmed seismic event Operating Basis Earthquake (OBE) as indicated by seismic EAL Threshold Values:

E-RU1 Unexpected Increase in ISFSI Radiation. instrumentation registering > 0.1g.

EAL Threshold Values: 1. Confirmed seismic event EITHER:

OR

1. Radiation reading on a HI-STAR spent fuel storage cask EITHER:

• Felt by onsite personnel

2. Tornado striking structures containing systems required to establish OR maintain Mode 4.

• 250 mR/hr. (neutron + gamma) on contact of ANY exterior vertical surface. OR OR Natural Events OR

• Indicated by installed seismic accelograph reading 0.01 g.

3. Sustained high winds 90 mph. on PPDS, OR indicated for 15 mins. by Control Room report, computer point OR chart recorders. OR

• 160 mR/hr. (neutron + gamma) on contact of ANY top horizontal surface.

OR OR 2. Report of a tornado strike within the Protected Area OR Switchyard.

2. ANY of the following radiation readings on a HI-STORM spent fuel storage cask: 4. Report of visible structural damage to a structure containing systems required to establish OR or maintain Mode 4. 3. Vehicle collision into structures containing systems necessary for safe

• 80 mR/hr. (neutron + gamma) on contact of ANY exterior vertical surface.

OR shutdown within the Protected Area OR in the Switchyard.

• 20 mR/hr. (neutron + gamma) on contact of ANY top horizontal surface.

5. Vehicle collision affecting a Vital Area. OR

• 32 mR/hr. (neutron + gamma) at the inlet AND outlet vent ducts. 4. Main Turbine rotating component failure causing visible casing damage OR RU3 Increase in radiation levels by a factor of 1000. 6. Main Turbine rotating component failure that penetrates the casing AND generates OR damage to the generator seals.

EAL Threshold Value: missiles causing damage to structures containing safety related equipment.

Area Radiation Monitor OR survey results indicate an unplanned increase in plant OR radiation levels by a factor of 1000. 7. Flooding in a Vital Area > Max. Safe Water Level for any area listed in DEOP 300-1.

E-HU1 Damaged to loaded cask confinement boundary. HA5 Fire or explosion affecting operability of 1 2 3 4 5 D HU5 Fire in the Protected Area not extinguished 1 2 3 4 5 D EAL Threshold Value: systems required to establish or maintain Mode 4. in 15 mins. or explosion in the Protected

1. Natural OR destructive phenomena affecting a loaded cask confinement boundary. EAL Threshold Values: Area Natural AND destructive phenomena may include: 1. Fire OR explosion in OR near any structure containing systems required to establish OR EAL Threshold Values:

• *

• maintain Mode 4. (Table H3 Example Safe Shutdown Areas)

Fire dropped cask tipped over cask flood 1. Fire in the Protected Area NOT extinguished in 15 mins. of EITHER:

• explosion

• fire

• lightning AND

• Control Room notification

2. ANY ONE of the following:

• seismic event

• tornado

• burial under debris OR

• Affected system parameter indications show degraded performance.

OR

• Verification of alarms.

OR

2. ANY condition in the opinion of the Emergency Director that indicates loss of loaded fuel

• Report received of visible damage to permanent structures OR equipment within the OR storage cask confinement boundary affected area. 2. Explosion in the Protected Area.

Contaminated E-HU2 Confirmed Security Event with potential loss of level of safety of the HA6 Release of toxic or flammable gases within or 1 2 3 4 5 D HU6 Toxic or flammable gas release affecting 1 2 3 4 5 D ISFSI. adjacent to a Vital Area. normal plant operations.

EAL Threshold Value:

EAL Threshold Value: EAL Threshold Value:

Security Event in OR around the ISFSI as determined by the Security Shift Supervisor. 1. Unanticipated report OR detection of toxic gases within OR adjacent to a Vital 1. Report OR detection of toxic OR flammable gases in amounts that Toxic or Flammable Gas Area in concentrations that may be life threatening to personnel. can affect normal plant operations.

OR OR

2. Unanticipated report OR detection of flammable gases within OR adjacent to a 2. Report by Local, County, OR State officials for evacuation OR Vital Area in concentration greater than the lower flammability limit. sheltering of site personnel based on an offsite event involving Injury Table H3 toxic OR flammable gases in amounts that can affect normal plant operations.

Safe Shutdown Areas Reactor Building Cribhouse HU7 Transportation of radiologically contaminated 1 2 3 4 5 D or potentially contaminated person(s) to an Control Room Aux. Electric Room offsite medical facility.

Diesel Generator Electrical Switchyard EAL Threshold Value:

Switchgear and Battery Remote Shutdown Radiologically contaminated OR potentially contaminated injured person(s) transferred to an offsite medical facility for treatment.

Plant Modes (white boxes indicate applicable modes) 1 Power Operations 2 Startup 3 Hot Shutdown 4 Cold Shutdown 5 Refueling D Defueled May 2005 DR 3-13 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / EFFLUENTS RG1 INITIATING CONDITION Actual or projected Site Boundary dose:

• 1 Rem TEDE OR

• 5 Rem CDE Thyroid EAL THRESHOLD VALUES

1. Radiological release in excess of Table R1 General Emergency threshold UNLESS release can be determined to be below available Table R2 General Emergency thresholds within 15 mins.

OR

2. Radiological release exceeds ANY Table R2 column General Emergency threshold.

Table R1 Effluent Monitor Thresholds Low, mid or high Range as determined by DOP 1700-10 or General Emergency PPDS Total Station Release Unit 2 / 3 Rx Bldg SPING 1.5 E+07 µCi/sec Plus Unit 2 / 3 Chimney SPING Table R2 Dose Assessment Thresholds Method General Emergency 1 R/hr. for 1 hr.

Field Team Monitoring OR (At or beyond Site Boundary) 5 Rem CDE Thyroid 1 Rem TEDE Dose Projection OR (At or beyond Site Boundary) 5 Rem CDE Thyroid May 2005 DR 3-14 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / EFFLUENTS RG1 (cont.)

MODE APPLICABILITY ALL BASIS: (References)

Table R1 Effluent Monitors - Classification is based on effluent monitor reading which would meet or exceed an EPA Protective Action Guideline (1 Rem TEDE or 5 Rem CDE) at the Site Boundary.

It is the instantaneous release rate value if NO dose projections can be performed or verified within 15 minutes of meeting or exceeding the total Chimney and Reactor Building Release Rate values. The Effluent Monitor values were determined using methodologies described in the Dresden DAPAR software requirement specifications, with following assumptions:

• Worst Sector (lowest speed)10 year average meteorology specified in the OCDM --

Wind Speed 8.9 mph and Stability Class D

• Quick Assessment Process Reduction Factors (0.012 for Rx Bldg Release).

• 10 % Gap Damage

• 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Release Duration

• Time after reactor shutdown = 1:00 hour

• The same value is used for ground level (Rx Bldg Vent) and elevated (Chimney) release points. An elevated release may not effect offsite areas as close to plant as ground level release, however use of ground level values will provide conservative estimates for exposure (cloud shine) to an overhead plume. (EPA-400, Section 5.6.1)

Table R2 Field Team Monitoring - The values are for surveys or iodine air samples taken at or beyond the SITE BOUNDARY and are the most accurate indicator of the condition. Field data are independent of release elevation and meteorology. The assumed release duration is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Expected post accident source terms would be dominated by noble gases providing the dose rate value. Direct reading iodine monitors are not available. Sampling of radioiodine by adsorption on charcoal or silver ziolite media followed by field analysis are used for determining the iodine value.

Dose Projection - Any calculated dose projections of 1 Rem total effective dose equivalent or 5 Rem committed dose equivalent to the thyroid is classified based on the Environmental Protection Agency Protective Action Guidelines which indicate that public protection is needed at these levels. Source term and release elevation values are options of the program. Actual meteorology provides the most accurate dose assessment and is used whenever possible. The assumed release duration is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

May 2005 DR 3-15 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / EFFLUENTS RS1 INITIATING CONDITION Actual or projected Site Boundary dose:

• 100 mRem TEDE OR

• 500 mRem CDE Thyroid EAL THRESHOLD VALUES

1. Radiological release in excess of Table R1 Site Area Emergency threshold UNLESS release can be determined to be below available Table R2 Site Area Emergency thresholds within 15 mins.

OR

2. Radiological release exceeds ANY Table R2 column Site Area Emergency threshold.

Table R1 Effluent Monitor Thresholds Low, mid or high Range as determined by DOP 1700-10 or Site Area Emergency PPDS Total Station Release Unit 2 / 3 Rx Bldg SPING 1.5 E+06 µCi/sec Plus Unit 2 / 3 Chimney SPING Table R2 Dose Assessment Thresholds Method Site Area Emergency 100 mR/hr. for 1 hr.

Field Team Monitoring OR (At or beyond Site Boundary) 500 mRem CDE Thyroid 100 mRem TEDE Dose Projection OR (At or beyond Site Boundary) 500 mRem CDE Thyroid May 2005 DR 3-16 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / EFFLUENTS RS1 (cont.)

MODE APPLICABILITY ALL BASIS: (References)

Table R1 Effluent Monitors - Classification is based on effluent monitor reading which would meet or exceed an EPA Protective Action Guideline (1 Rem TEDE or 5 Rem CDE) at the Site Boundary.

It is the instantaneous release rate value if NO dose projections can be performed or verified within 15 minutes of meeting or exceeding the total Chimney and Reactor Building Release Rate values. The Effluent Monitor values were determined using methodologies described in the Dresden DAPAR software requirement specifications, with following assumptions:

• Worst Sector (lowest speed)10 year average meteorology specified in the OCDM --

Wind Speed 8.9 mph and Stability Class D

• Quick Assessment Process Reduction Factors (0.012 for Rx Bldg Release).

• 10 % Gap Damage

• 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Release Duration

• Time after reactor shutdown = 1:00 hour

• The same value is used for ground level (Rx Bldg Vent) and elevated (Chimney) release points. An elevated release may not effect offsite areas as close to plant as ground level release, however use of ground level values will provide conservative estimates for exposure (cloud shine) to an overhead plume. (EPA-400, section 5.6.1)

Table R2 Field Team Monitoring - The values are for surveys or iodine air samples taken at or beyond the SITE BOUNDARY and are the most accurate indicator of the condition. Field data are independent of release elevation and meteorology. The assumed release duration is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Expected post accident source terms would be dominated by noble gases providing the dose rate value. Direct reading iodine monitors are not available. Sampling of radioiodine by adsorption on charcoal or silver ziolite media followed by field analysis are used for determining the iodine value.

Dose Projection - Any calculated dose projections of 100 mRem total effective dose equivalent or 500 mRem committed dose equivalent to the thyroid is classified based on 10% of the EPA Protective Action Guidelines for radiological exposures to the public. Source term and release elevation values are options of the program. Actual meteorology provides the most accurate dose assessment and is used whenever possible. The assumed release duration is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

May 2005 DR 3-17 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / EFFLUENTS RA2 INITIATING CONDITION Release 10 X ODCM limit for 15 mins.

EAL THRESHOLD VALUES

1. Unplanned radiological release in excess of Table R1 Alert threshold for 15 mins.

UNLESS release can be determined to be below available Table R2 Alert thresholds within this period.

OR

2. Unplanned Gaseous or Liquid radiological release exceeds ANY Table R2 column Alert threshold for 15 mins..

Table R1 Effluent Monitor Thresholds Low, mid or high Range as determined Alert by DOP 1700-10 or PPDS Total Station Release Unit 2 / 3 Rx Bldg SPING 8.5 E+05 µCi/sec Plus for 15 mins.

Unit 2 / 3 Chimney SPING Table R2 Dose Assessment Thresholds Method Alert Sample / Release Package 10 X ODCM limit for 15 mins.

Field Team Monitoring 10 mR/hr. for 15 mins.

(At or beyond Site Boundary)

MODE APPLICABILITY ALL May 2005 DR 3-18 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / EFFLUENTS RA2 (cont.)

BASIS (References)

Dose Assessment using actual meteorology should be performed a soon as possible to ensure that a Site Area Emergency is not warranted.

A sustained unplanned release of this greater magnitude that cannot be terminated in 15 minutes represents an uncontrolled situation that is an actual or potential substantial degradation of the level of safety of the plant. The degradation in plant control implied by the fact that the release can not terminated in 15 minutes is the primary concern. The Emergency Director should not wait until 15 minutes has elapsed, but should declare an Alert as soon as the release is determined to be uncontrolled or projected to be non-isolable within 15 minutes.

Table R1 Effluent Monitors - The sum of 2/3 Rx Building SPING and 2/3 Chimney SPING will provide a Total Station Release Rate. The Alert value for gaseous effluents was reduced to 10 X ODCM to ensure sequential classifications. The Effluent Monitor value was determined in accordance with Sargent & Lundy calculation ATD-0212, Rev. 0, 1/12/93 and BR-PR-10, Rev. 1, 2/9/93. The values chosen is on a per station basis for a one-hour ground level release using a 10-year Annual Average Meteorology as specified by ODCM. The release rate was determined in the units of a station-generated normal operating mixture for the no clad damage condition.

Table R2 Grab samples - are used to determine release concentrations or rates to confirm meter readings or when the effluent monitors are not in service. The ODCM uses 10 CFR 20 Appendix B Table 2 data to generate maximum instantaneous release rate limits. These are indicated on Release Packages that are approved.

Field Team Monitoring - Prorating the 10 CFR 20 criteria of 500 mR/year for both time and the multiplier, the associated restricted area dose rate would be 10 mR/hr. The release duration was reduced to 15 minutes in recognition of the increased severity and therefore represents a substantial degradation in the level of safety of the plant.

May 2005 DR 3-19 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / EFFLUENTS RU2 INITIATING CONDITION Release 2 X ODCM limit for 60 mins.

EAL THRESHOLD VALUE

1. Unplanned radiological release in excess of Table R1 Unusual Event for 60 mins.

UNLESS release can be determined to be below available Table R2 Unusual Event thresholds within this period.

OR

2. Unplanned Gaseous OR Liquid radiological release exceeds Table R2 column Unusual Event threshold for 60 mins..

Table R1 Effluent Monitor Thresholds Low, mid or high Range as determined by DOP 1700-10 or Unusual Event PPDS Total Station Release Rate Unit 2 / 3 Rx Bldg SPING 1.7 E+05 µCi/sec Plus for 60 mins.

Unit 2 / 3 Chimney SPING Table R2 Dose Assessment Thresholds Method Unusual Event Sample / Release Package 2 X ODCM limit for 60 mins.

MODE APPLICABILITY ALL May 2005 DR 3-20 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / EFFLUENTS RU2 (cont.)

BASIS (References)

Dose Assessment using actual meteorology should be performed as soon as possible to ensure that an Alert is not warranted.

An unplanned release that cannot be terminated in 60 minutes represents an uncontrolled situation that is a potential degradation of the level of safety of the plant. The degradation in plant control implied by the fact that the release can not be terminated in 60 minutes is the primary concern. The Emergency Director should not wait until 60 minutes has elapsed, but should declare an Unusual Event as soon as the release is determined to be uncontrolled or projected to be non-isolable within 60 minutes.

Table R1 Effluent Monitors - The Unusual Event value for gaseous effluents was reduced to 2 X ODCM to ensure sequential classifications. The Effluent Monitor value was determined in accordance with Sargent & Lundy calculation ATD-0223, Rev. 0, 1/12/93. The values chosen is on a per station basis for a one-hour ground level release using a 10-year Annual Average Meteorology as specified by ODCM. The release rate was determined in the units of a station-generated normal operating mixture for the no clad damage condition. The sum of 2/3 Rx Building SPING and 2/3 Chimney SPING will provide a Total Station Release Rate.

Table R2 Release Package - The ODCM uses 10 CFR 20 Appendix B Table 2 data to generate maximum instantaneous release rate limits.

Grab samples - Grab samples are used to determine release concentrations or rates to confirm meter readings or when the effluent monitors are not in service.

May 2005 DR 3-21 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / EFFLUENTS RA3 INITIATING CONDITION In plant radiation levels impede plant operations.

EAL THRESHOLD VALUES

1. Radiation readings 15 mR/hr. in ANY area requiring continuous occupancy, Table R3.

OR

2. Radiation readings 2500 mR/hr. in ANY area requiring infrequent access for safe plant operations, Table R4.

Table R3 Table R4 Areas Requiring Continuous Areas Requiring Infrequent Occupancy Access Main Control Room HPCI Cubicle Central Alarm Station East and West LPCI Pump Areas Secondary Alarm Station East and West CRD Module Areas Radwaste Control Room Vessel Instrument Rack Area Gatehouse RWCU Area Isolation Condenser Area MODE APPLICABILITY ALL BASIS (References)

This EAL addresses increased radiation levels that impede necessary access to operator stations, or other areas containing equipment that must be operated manually in order to maintain safe operation or to perform a safe shutdown. The concern of the EAL is a loss of control of radioactive material causing high radiation levels. The cause of the increase in radiation levels is not the major concern of this EAL. The source of the increased radiation levels should be considered in view of other EALs. The Emergency Director should also evaluate how these conditions will affect the unaffected unit.

May 2005 DR 3-22 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / EFFLUENTS RA3 (cont)

Threshold Value 1 (Table R3) - The value of 15 mR/hr is derived from the general design criteria 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 value can be averaged over the 30 days, the value is used here without averaging, as a 30 day duration implies an event potentially more significant than an Alert.

Threshold Value 2 (Table R4) - Areas requiring infrequent access and dose rate values are based on those specified in the Emergency Operating Procedures and have associated Maximum Safe Values assigned to each area. Dose rates above these values will impede necessary access. When the source of the high radiation condition is the result of the manipulation of a known source and radiological controls are in place to minimize exposure, classification should be considered based on the ability to access the area to safely operate or perform a safe shutdown of the Station.

May 2005 DR 3-23 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ABNORMAL RAD LEVELS / EFFLUENTS RU3 INITIATING CONDITION Increase in plant radiation levels by a factor of 1000.

EAL THRESHOLD VALUE Area Radiation Monitor readings OR survey results indicate an unplanned increase in plant radiation levels by a factor of 1000.

MODE APPLICABILITY ALL BASIS (References)

Classification of an Unusual Event is warranted as a precursor to more serious events. The concern of this EAL is the loss of control of radioactive material representing a potential degradation of the level of safety of the plant. The Threshold Value tends to have a long lead-time relative to a radiological release and thus the threat to public health and safety is very low.

In light of the increased dose rates the Emergency Director should evaluate how these conditions will affect the other unit.

May 2005 DR 3-24 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FG1 INITIATING CONDITION Loss of 2 Fission Product Barriers with Loss or Potential Loss of third barrier.

EAL THRESHOLD VALUE Review of Fission Product Barrier Matrix indicates:

• Loss of 2 Fission Product Barriers.

AND

• Loss OR Potential Loss of THIRD barrier.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

Conditions / events required to cause the loss of 2 Fission Product Barriers with the loss or potential loss of the third could reasonably be expected to cause a release beyond the immediate site area exceeding EPA Protective Action Guidelines.

May 2005 DR 3-25 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FS1 INITIATING CONDITION Loss or Potential Loss of 2 Fission Product Barriers.

EAL THRESHOLD VALUE Review of Fission Product Barrier Matrix indicates ANY 2 of the following:

• Loss OR Potential Loss of Fuel Clad.

• Loss OR Potential Loss of RCS.

• Loss OR Potential Loss of Containment.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

Loss or potential loss of 2 Fission Product Barriers would be a major failure of plant systems needed for protection of the public.

May 2005 DR 3-26 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FA1 INITIATING CONDITION Loss or Potential Loss of Fuel Clad or RCS Barrier.

EAL THRESHOLD VALUE Review of Fission Product Barrier Matrix indicates EITHER:

• Loss OR Potential Loss of Fuel Clad.

OR

• Loss OR Potential Loss of RCS.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

The Fuel Cladding and the Reactor Coolant System are weighted more heavily than the Containment Barrier.

Loss or potential loss of either the Fuel Cladding or the Reactor Coolant System would be a substantial degradation in the level of plant safety.

May 2005 DR 3-27 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FU1 INITIATING CONDITION Loss or Potential Loss of Containment Barrier.

EAL THRESHOLD VALUE Review of Fission Product Barrier Matrix indicates:

• Loss OR Potential Loss of Containment.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

Loss or potential loss of the Containment would be a potential degradation in the level of plant safety.

The Fission Product Barrier EALs place a higher significance on the Fuel Clad barrier and the Reactor Coolant System barrier than the Containment Barrier which is reflected in the classification matrices.

May 2005 DR 3-28 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT 1.a INITIATING CONDITION Drywell or Torus Pressure.

THRESHOLD VALUE LOSS: ....................................Rapid pressure decrease in containment after increase without containment spray.

OR Drywell OR Torus pressure response NOT consistent with LOCA conditions during LOCA.

POTENTIAL LOSS: ...........Containment Pressure 62 psig.

MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS - A rapid unexplained loss of pressure not due to use of containment sprays following an initial pressure increase indicates a loss of containment integrity.

Drywell or Torus pressure should increase as a result of mass and energy release into the containment from a LOCA. Drywell or Torus pressure NOT increasing under these conditions indicates a breach of the primary containment.

POTENTIAL LOSS - A containment pressure 62 psig is based on the containment design pressure. If the containment design pressure is exceeded this represents a challenge to the containment structure because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists. This constitutes a potential loss of the containment barrier even if a breach has NOT occurred.

May 2005 DR 3-29 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT 1.b INITIATING CONDITION Drywell or Torus Hydrogen Concentration.

THRESHOLD VALUE LOSS: ....................................NONE POTENTIAL LOSS: ...........Drywell OR Torus Hydrogen concentration 6%.

AND Drywell OR Torus Oxygen concentration 5%.

MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS - NONE POTENTIAL LOSS - The Emergency Procedure Guidelines and Severe Accident Guidelines identify that deflagration will occur at a minimum Hydrogen concentration of > 6% and Oxygen concentration > 5%. The deflagration of Hydrogen represents a potential loss of the primary containment.

May 2005 DR 3-30 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT 1.c.1 INITIATING CONDITION Containment Breached / Bypassed.

THRESHOLD VALUE LOSS: ....................................Failure of ALL automatic containment isolation valves in ANY ONE line penetrating the primary containment to isolate on isolation actuation signal.

POTENTIAL LOSS: ...........NONE MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS - A failure of all isolation valves in any one line indicates a breach of the primary containment integrity as described in the Primary Containment Technical Specifications bases section or primary containment Limiting Conditions for Operation.

The breach is NOT isolable from the Control Room OR an attempt for isolation from the Control Room has been made and was unsuccessful. An attempt for isolation should be made prior to the accident classification. If isolable upon identification this Initiating Condition is not applicable.

The intent of this Initiating Condition is to address those isolation devices which should close upon receipt of a containment isolation signal whether the signal is valid or not. The concern in this EAL is the open path to the environment.

POTENTIAL LOSS - NONE May 2005 DR 3-31 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT 1.c.2 INITIATING CONDITION Containment Breached / Bypassed.

THRESHOLD VALUE LOSS: ....................................Breach of primary containment, NOT part of Primary Containment Isolation, that is unisolable from the Main Control Room.

POTENTIAL LOSS: ...........NONE MODE APPLICABILITY 1, 2, 3 BASIS: (REFERENCES)

LOSS - The breach or bypass is NOT isolable from the Control Room OR an attempt for isolation from the Control Room has been made and was unsuccessful. An attempt for isolation should be made prior to the accident classification. If isolable upon identification this Initiating Condition is not applicable.

FPB 1.c.1 addresses a breach of primary containment referring to a loss of primary containment integrity as described in the Primary Containment Technical Specifications bases section or primary containment Limiting Conditions for Operation (LCO).

The intent of this Initiating Condition (FPB 1.c.2) is to address all penetrations not included in Primary Containment Isolation (PCI) and are a release path to the environment. Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a path to the environment.

A bypassed containment is the opening of penetration isolations such that a path to the environment exists.

POTENTIAL LOSS - NONE May 2005 DR 3-32 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT 1.c.3 INITIATING CONDITION Containment Breached / Bypassed.

THRESHOLD VALUE LOSS: ....................................Intentional venting of the primary containment per DEOPs or SAMG-1 AFTER containment isolation.

POTENTIAL LOSS: ...........NONE MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS - Venting of the primary containment as required by the Emergency Operating Procedures (DEOPs) or Severe Accident Management Guidelines (SAMGs) is considered to be a breach of the primary containment for the purposes of accident classification.

POTENTIAL LOSS - NONE May 2005 DR 3-33 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT 1.c.4 INITIATING CONDITION Containment Breached / Bypassed.

THRESHOLD VALUES LOSS: ....................................Unisolable primary system leakage outside containment as indicated by EITHER:

1. Area Temperatures, Table F1 Max Safe Value.

OR

2. Area Radiation Levels, Table F2. Max. Safe Value.

Table F1 -- Area Temperatures Table F2 -- Area Radiation Max. Max.

Max. Max.

Normal Safe Normal Safe mR/hr mR/hr HPCI Area 150 ºF 210 ºF HPCI Cubicle Unit 2 (3) 150 (100) 2500 Shutdown Cooling Pump Room 150 ºF 180 ºF East LPCI Pump Area 12 2500*

Shutdown Cooling Heat Exch Room 150 ºF 180 ºF West LPCI Pump Area 9 2500*

RWCU Demin Rooms 150 ºF 210 ºF East CRD Module Area 30 2500*

RWCU Pump and Heat Exchanger Area 150 ºF 210 ºF West CRD Module Area 50 2500*

Isolation Condenser Area 150 ºF 180 ºF Vessel Inst Rack Area 30 2500*

RWCU Area 30 2500*

Isolation Condenser Area 10 2500

• Measured by Local Survey POTENTIAL LOSS: ...........NONE MODE APPLICABILITY 1, 2, 3 May 2005 DR 3-34 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT 1.c.4 (cont.)

BASIS: (References)

LOSS - Primary system leakage that bypasses the primary containment would not cause the primary containment pressure to increase. Area temperatures or radiation levels greater than the maximum safe values listed in the Emergency Operating Procedures (DEOPs) resulting from a primary system discharging into the reactor building is indicative of conditions in which equipment necessary for the safe shutdown of the plant will fail, or personnel access necessary for the safe shutdown of the plant may be precluded. This condition is therefore deemed to be of such a magnitude as to constitute a loss of containment.

The maximum safe temperature or radiation level is the highest above which equipment necessary for the safe shutdown of the plant will fail, or personnel access necessary for the safe shutdown of the plant will be precluded.

POTENTIAL LOSS - NONE May 2005 DR 3-35 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT 1.d INITIATING CONDITION Drywell Radiation Monitors.

THRESHOLD VALUE LOSS: ....................................NONE POTENTIAL LOSS: ...........Drywell radiation monitor readings in the Containment Potential Loss range of Graph 1.d.

Note: From time after shutdown to 0.5 hrs., use the 0.5 hrs. value.

9000 Graph 1.d 8000 00 77 0 0 7000 70 1 7 64 Containment Potential Loss Drywell Radiation Reading (R/hr) 75 6000 57 2 2 5000 50 4000 Containment Intact 39 3000 30 2000 99 16 1000 4 62 1 8 20 12 0

0.5 1 2 3 4 5 6 7 8 9 10 20 30 40 50 60 70 80 90 100 Time After Shutdown (hrs)

Dresden & Quad Cities MODE APPLICABILITY 1, 2, 3 May 2005 DR 3-36 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT 1.d (cont.)

BASIS: (References)

LOSS - NONE POTENTIAL LOSS - A containment radiation monitor reading of 7000 R/hr @ 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reactor shutdown indicates a fuel clad damage of 20% and a failure of the primary system allowing the entire inventory to be disbursed inside the drywell. Regardless of whether containment is challenged this amount of activity, if released, could have such severe consequences that it is prudent to treat this as a potential loss of primary containment.

Clad damage assessment is NOT required to be performed by control room personnel.

May 2005 DR 3-37 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT 1.e INITIATING CONDITION Reactor Vessel Water Level.

THRESHOLD VALUE LOSS: ....................................NONE POTENTIAL LOSS:..............Entry into Primary Containment Flooding. SAMG-1.

MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS- NONE POTENTIAL LOSS - Entry into SAMG-1, Primary Containment Flooding, as required by the DEOPs occurs when the reactor water level is less than the minimum steam cooling level (-164) without spray cooling OR When Reactor Pressure Vessel water level is below the Jet Pump suction (-191). This is indicative of substantial core degradation and represents imminent core melt sequences that if not corrected could lead to reactor vessel failure and increase potential for containment failure.

May 2005 DR 3-38 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT 1.f INITIATING CONDITION Drywell Temperature.

THRESHOLD VALUE LOSS: ....................................NONE POTENTIAL LOSS: ...........Drywell Temperature 281 °F.

MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS - NONE POTENTIAL LOSS - Drywell temperatures greater than the design temperature represent a challenge to the containment structure because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists. This is considered a potential loss of primary containment.

May 2005 DR 3-39 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FUEL CLAD 2.a INITIATING CONDITION Primary Coolant Activity.

THRESHOLD VALUE LOSS: ....................................Coolant Activity 300 µCi/gm I-131 dose equivalent.

POTENTIAL LOSS: ...........NONE MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS - Reactor coolant activity provides for the most accurate assessment of the extent of fuel clad degradation. Classifications based on this criterion may be delayed due to the time involved in obtaining and analyzing a coolant sample. 300 µCi/gm I-131 equivalent indicates a degradation of approximately 0.7% (S&L calculation D-2-91, Rev 0) of the fuel cladding and is indicative of a loss of the fuel clad fission product barrier.

POTENTIAL LOSS - NONE May 2005 DR 3-40 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FUEL CLAD 2.b INITIATING CONDITION Reactor Vessel Water Level.

THRESHOLD VALUE LOSS: ....................................1. BOTH of the following:

Reactor water level -164" (Minimum Steam Cooling).

AND NEITHER core spray loop flows 4750 gpm.

OR

2. Reactor water level -191 (2/3 Core Height).

POTENTIAL LOSS:............Reactor water Level -143" (TAF).

MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS - Reactor vessel water level less than the minimum steam cooling RPV water level (-164) without spray cooling is the minimum value below which further substantial degradation of the cladding will occur and corrective actions as described in the Emergency Operating Procedures (DEOPs) and Severe Accident Management Guidelines (SAMGs).

Entry into SAMG-1, Primary Containment Flooding, as required by the DEOPs occurs when the reactor water level is less than the minimum steam cooling level (-164) without spray cooling or When Reactor Pressure Vessel water level is below the Jet Pump suction (-191). This is indicative of substantial core degradation and represents imminent core melt sequences that if not corrected could lead to reactor vessel failure and increase potential for containment failure.

POTENTIAL LOSS - The top of the active fuel is the point at which adequate core cooling is still assured but is sufficiently low that any further drop in water level could result in the significant degradation of the cladding. Therefore this is considered to be a potential loss of the fuel cladding barrier.

May 2005 DR 3-41 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FUEL CLAD 2.c INITIATING CONDITION Drywell Radiation Monitors.

THRESHOLD VALUE LOSS: ....................................Drywell radiation in the Fuel Cladding Lost range of Graph 2.c.

POTENTIAL LOSS: ...........NONE Note: From time after shutdown to 0.5 hrs., use the 0.5 hrs. value.

2500 Graph 2.c 62 20 2000 25 19 50 Fuel Cladding Lost 17 Drywell Radiation Reading (R/hr) 4 1 60 44 1500 14 55 12 1

1000 93 0

Fuel Cladding Intact 76 500 5 42 1

32 6

15 50 32 0

0.5 1 2 3 4 5 6 7 8 9 10 20 30 40 50 60 70 80 90 100 Time After Shutdown (hrs)

Dresden & Quad Cities MODE APPLICABILITY 1, 2, 3 May 2005 DR 3-42 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FUEL CLAD 2.c (cont.)

BASIS: (References)

LOSS - A drywell radiation monitor reading of 1750 R/hr @ 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reactor shutdown is indicative of both the loss of the reactor coolant system and 5% clad failure and the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory into the drywell atmosphere. Clad damage of 5% assumes reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations allowed by Technical Specifications. Therefore readings of this level are indicative of a loss of the fuel cladding barrier.

Clad damage assessment is NOT required to be performed by control room personnel.

POTENTIAL LOSS - NONE May 2005 DR 3-43 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION FUEL CLAD 2.d INITIATING CONDITION Steamline Radiation.

THRESHOLD VALUE LOSS: ....................................MSL radiation 2 X MSL high radiation alarm setpoint.

POTENTIAL LOSS: ...........NONE MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS - 2 times the Main Steam Line high radiation alarm setpoint provides an indication of gross fuel failures in accordance with UFSAR accident scenarios. This is considered to be a loss of the fuel clad fission product barrier.

This EAL is NOT intended to apply to cases when the isolation is due to spiking caused by resin intrusion, reactor trips, sensing line air intrusion or other known factors. Confirmatory data should be used before relying on this indication.

POTENTIAL - NONE May 2005 DR 3-44 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION REACTOR COOLANT SYSTEM 3.a.1 INITIATING CONDITION RCS Leak Rate.

THRESHOLD VALUE LOSS: ....................................Unisolable Main Steam Line break.

POTENTIAL LOSS: ...........Unisolable RCS leakage 50 gpm.

MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS - A Main Steam Line break represents a loss of the integrity of the reactor coolant system.

The result of a MSL break inside the Drywell with MSIV closure is a loss of the RCS Barrier.

The result of a MSL break on either side of the MSIVs without closure is a Loss of both the RCS and the Containment Barriers. Design basis accident analysis indicates that even if MSIV closure occurs within design limits, dose consequences from a release would be in excess of 10 mR.

The leak is NOT isolable IF an attempt for isolation has been made and was unsuccessful within 15 minutes. An attempt for isolation should be made prior to the accident classification. If isolable upon identification this Initiating Condition is not applicable.

POTENTIAL LOSS - A reactor coolant system leak rate of greater than 50 gallons per minute is at a level indicative of a small breach of the RCS but which is well within makeup capability of normal and emergency high pressure systems. Core uncovery is NOT a significant concern for a 50 gpm leak, however break propagation leading to a significantly larger loss of inventory is possible. A leak of this magnitude would be readily observable on the main control board. A leak of this size is a precursor of the loss of the reactor coolant system integrity and is therefore considered to be a potential loss of this barrier.

May 2005 DR 3-45 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION REACTOR COOLANT SYSTEM 3.a.2 INITIATING CONDITION RCS Leak Rate.

THRESHOLD VALUE LOSS: ....................................NONE POTENTIAL LOSS: ...........Unisolable primary system leakage outside the primary containment into specified area AND EITHER:

• Area Temperatures, Table F1 Max. Normal Value.

OR

• Area Radiation, Table F2 Max. Normal Value.

Table F1 -- Area Temperatures Table F2 -- Area Radiation Max. Max.

Max. Max.

Normal Safe Normal Safe mR/hr mR/hr HPCI Area 150 ºF 210 ºF HPCI Cubicle Unit 2 (3) 150 (100) 2500 Shutdown Cooling Pump Room 150 ºF 180 ºF East LPCI Pump Area 12 2500*

Shutdown Cooling Heat Exchanger 150 ºF 180 ºF West LPCI Pump Area 9 2500*

Room RWCU Demin Rooms 150 ºF 210 ºF East CRD Module Area 30 2500*

RWCU Pump and Heat Exchanger West CRD Module 150 ºF 210 ºF 50 2500*

Area Area Vessel Instrument Rack Isolation Condenser Area 150 ºF 180 ºF 30 2500*

Area RWCU Area 30 2500*

Isolation Condenser 10 2500 Area

• Measured by Local Survey POTENTIAL LOSS: ...........NONE MODE APPLICABILITY 1, 2, 3 May 2005 DR 3-46 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION REACTOR COOLANT SYSTEM 3.a.2 (cont)

BASIS: (References)

LOSS - NONE POTENTIAL LOSS - Area temperatures or radiation levels greater than the maximum normal values listed in the Emergency Operating Procedures (DEOPs) resulting from a primary system discharging into the reactor building is indicative of conditions in which significant RCS inventory is being lost. This is therefore considered to be a potential loss of the reactor coolant system boundary. Maximum normal levels are determined by using historical data from previous years. These values are found in the Emergency Operating Procedures (DEOPs). Since these values are subject to change specific values are not given here and the DEOPs should be consulted.

The leak is NOT isolable if an attempt for isolation has been made and was unsuccessful within 15 minutes. An attempt for isolation should be made prior to the accident classification. If isolable upon identification this Initiating Condition is not applicable.

May 2005 DR 3-47 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION REACTOR COOLANT SYSTEM 3.b INITIATING CONDITION Drywell Pressure.

THRESHOLD VALUE LOSS: ....................................Drywell pressure ECCS Setpoint.

AND Pressure increase due to reactor coolant leakage.

POTENTIAL LOSS: ...........NONE MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS- Drywell pressure in excess of the emergency core cooling system initiation setpoint resulting from primary system leakage into the drywell is of such a magnitude indicating a loss of the reactor coolant system boundary. Cycling of safety relief valves to reduce primary system overpressure when no fuel damage is indicated is NOT considered reactor coolant leakage.

Cycling of safety relief valves to reduce primary system overpressure when fuel damage is indicated is considered reactor coolant leakage. Primary containment pressure increases due solely to loss of containment heat removal capability are also NOT considered to exceed this threshold.

POTENTIAL LOSS- NONE May 2005 DR 3-48 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION REACTOR COOLANT SYSTEM 3.c INITIATING CONDITION Drywell Radiation Monitoring.

THRESHOLD VALUE LOSS: ....................................Drywell radiation monitors readings in the RCS Lost range of Graph 3.c.

POTENTIAL LOSS: ...........NONE Note: From time after shutdown to 0.5 hrs., use the 0.5 hrs. value.

160 Graph 3.c 140 8

120 11 0

Drywell Radiation Reading (R/hr) 11 Reactor Coolant System Lost 0

10 100 92 83 80 72 60 53 Reactor Coolant System Intact 43 40 24 20 18 9

3 2 0

0.5 1 2 3 4 5 6 7 8 9 10 20 30 40 50 60 70 80 90 100 Time After Shutdown (hrs)

Dresden & Quad Cities MODE APPLICABILITY 1, 2, 3 May 2005 DR 3-49 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION REACTOR COOLANT SYSTEM 3.c (cont.)

BASIS: (References)

LOSS - Radiation from an instantaneous release of the entire primary coolant system inventory with Technical Specification allowed radiochemistry into the drywell would not be discernible on the installed drywell radiation monitors.

Therefore a value of 100 R/hr 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reactor shutdown was chosen because it is above the 1Rem/hr level maintained by the radiation monitors, above ambient background radiation and represents a detectable radiation level if coolant is released to the Drywell above allowed Technical Specification radiochemistry limits but less than the 5% Fuel Clad damage used in the Fuel Clad Fission Product Barrier Threshold Value.

POTENTIAL LOSS - NONE May 2005 DR 3-50 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION REACTOR COOLANT SYSTEM 3.d INITIATING CONDITION Reactor Vessel Water Level.

THRESHOLD VALUE LOSS: ....................................Reactor vessel level -143" (TAF).

POTENTIAL LOSS: ...........NONE MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS - In order to provide normal means to cool the fuel, water level must be maintained above the top of active fuel otherwise extraordinary means must be taken to assure that adequate core cooling exists. In certain failure event sequences reactor vessel water level may be procedurally lowered to the top of active fuel and the reactor coolant system depressurized to allow for steam cooling of the core. Even though fuel clad damage is not predicted under these conditions several safety system failures need to have occurred to reach the condition where steam cooling would be procedurally required. Therefore this is indicative of a loss of the reactor coolant system boundary. Water levels below this value indicate a challenge to core cooling which is a precursor to more serious events. If intentional lowering of the vessel level is required during an ATWS event EAL MS3 should also be consulted.

POTENTIAL LOSS - NONE May 2005 DR 3-51 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY FISSION PRODUCT BARRIER DEGRADATION REACTOR COOLANT SYSTEM 3.e INITIATING CONDITION Primary System Relief Valves.

THRESHOLD VALUE LOSS: ....................................Primary system relief valve stuck open.

AND Torus temperature 110 °F.

POTENTIAL LOSS: ...........NONE MODE APPLICABILITY 1, 2, 3 BASIS: (References)

LOSS - Technical Specifications require a SCRAM when Torus bulk temperature levels reach 110°F. Below this point, the reactor is considered in a safe condition even with relief valves stuck open.

POTENTIAL LOSS - NONE May 2005 DR 3-52 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1 INITIATING CONDITION Prolonged (>4hrs) loss of all ECCS AC power.

EAL THRESHOLD VALUES

1. Loss of all AC power to ECCS bus 23-1 (33-1).

AND

2. Loss of all AC power to ECCS bus 24-1(34-1).

AND

3. Any of the following:

• Restoration of power to EITHER bus 23-1 (33-1) OR 24-1(34-1) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is NOT likely.

OR

• Conditions are imminent that a loss of 2 Fission Product Barriers and Loss or Potential Loss of the 3rd (FG1) will occur prior to restoration of AC power to the Unit.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

Conditions are degraded to the point that Loss or Potential Loss of three Fission Product Barriers can be predicted to occur.

Restoration of power cannot be predicted with confidence in a time that will prevent Loss or Potential Loss of three Fission Product Barriers.

The prolonged loss of all AC power compromises all plant safety systems and represents conditions which indicate imminent substantial core degradation or melting with potential for the loss of containment integrity. Resultant releases can be reasonably expected to exceed Environmental Protection Agency Protective Action Guideline exposure levels offsite for more than the immediate site area.

The four (4) hours for restoration is based on the site blackout coping analysis performed in conformance with 10 CFR 50.63 and Regulatory Guide 1.155, Station Blackout, which includes an appropriate allowance for offsite response. This EAL is used as an anticipatory condition for the Fission Product Barrier EALs in that under these conditions, the capability to monitor the Fission Product Barriers may be degraded.

May 2005 DR 3-53 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG1 (cont.)

A General Emergency should be declared when conditions degrade to the point that a General Emergency (FG1) classification can be predicted by trending Fission Product Barrier Threshold Values. Classification of General Emergency under the predictive conditions will allow as much time as possible for protective actions to be implemented.

May 2005 DR 3-54 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS1 INITIATING CONDITION Loss of all ECCS bus AC power for 15 minutes.

EAL THRESHOLD VALUES

1. Loss of all AC power to ECCS bus 23-1 (33-1).

AND

2. Loss of all AC power to ECCS bus 24-1(34-1).

AND

3. Failure to restore power within 15 minutes to at least:

• Bus 23-1 (33-1).

OR

• Bus 24-1(34-1).

MODE APPLICABILITY 1, 2, 3 BASIS (References)

The loss of all onsite and offsite AC power compromises all plant safety systems and represents failures of plant functions required for the protection of the public. Loss of all ECCS AC power is a precursor to events that could result in losses or potential losses of Fission Product Barriers.

For a prolonged loss of all ECCS AC power, upgrade to a General Emergency via MG1 may be appropriate.

The Emergency Director should address how the loss of power to the affected unit may affect the operation of the other unit due to the possible loss of power for common or shared systems.

15 minutes allows adequate time to cross tie or address diesel generator failures and excludes transient or momentary power losses.

May 2005 DR 3-55 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA1 INITIATING CONDITION Power to ECCS Buses reduced to a single source for > 15 minutes such that failure of that single source would result in Unit Blackout.

EAL THRESHOLD VALUE Available ECCS bus AC power reduced to only one of the following sources for > 15 minutes:

• Reserve auxiliary transformer TR-22 (TR-32).

• Unit auxiliary transformer TR-21 (TR-31).

• Unit Emergency Diesel Generator.

• Shared Emergency Diesel Generator.

• Unit crosstie breakers.

• SBO Diesel Generator.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

The reduction of available reliable power sources to a condition where ANY additional single failure will result in a Unit Blackout is a substantial degradation in the level of safety of the plant. That is, the Unit is down to its last source of AC power. Loss of the single power supply would escalate to a Site Area Emergency via Initiating Condition MS1.

The power supplies listed take into account sources and distribution buses that, if unavailable, establish a single failure vulnerability. This EAL allows for the use of unit crosstie breakers if they are the only source of power to the affected unit. The impact of using the crosstie breakers in so far as they affect the other (unaffected) unit must be addressed by the Emergency Director.

15 minutes was selected as a Threshold Value to exclude transient or momentary power losses.

May 2005 DR 3-56 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU1 INITIATING CONDITION Unplanned loss of all offsite AC power to a units ECCS buses.

EAL THRESHOLD VALUE Unplanned loss of ability to power ECCS buses from:

• TR-21 (TR-31).

AND

• TR-22 (TR-32).

MODE APPLICABILITY ALL BASIS (References)

A loss of offsite AC power reduces the 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. In Modes 1, 2, and 3, Unit crosstie breakers are NOT considered an adequate source and an Alert declaration based on EAL MA1 would be appropriate should they be the sole source of AC power. In Modes 4 and 5, Unit crosstie breakers are considered an adequate source due to the less severe threat to the protection of the health and safety of the public because of the much longer time available to restore power to the decay heat removal systems. Backfeed through the unit auxiliary transformer is considered an adequate source of offsite power.

The intent of this EAL is to declare an Unusual Event when offsite power has been lost and at least one of the emergency diesel generators has successfully started and energized at least one ECCS bus.

Declaration of an Unusual Event is not applicable when the loss of the offsite power sources is due to a planned evolution. As stated in the Emergency Plan in Section 2, a planned evolution involves preplanning to address the limitations imposed by the condition, the performance of required surveillance testing, and the implementation of specific controls prior to knowingly entering the condition.

Although the NEI guidance for this initiating condition allows 15 minutes for restoration of power prior to declaration of NOUE, it is not used here because current station procedures for power restoration call for surveillance measures, which will not be completed within 15 minutes.

May 2005 DR 3-57 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA2 INITIATING CONDITION Loss of all ECCS bus AC power for 15 minutes.

EAL THRESHOLD VALUES

1. Loss of all AC power to ECCS bus 23-1 (33-1).

AND

2. Loss of all AC power to ECCS bus 24-1 (34-1).

AND

3. Failure to restore power within 15 minutes to at least:

• Bus 23-1 (33-1).

OR

• Bus 24-1 (34-1).

MODE APPLICABILITY 4, 5, Defueled BASIS (References)

The loss of both offsite and onsite AC power to the ECCS buses when in Mode 4 or 5 compromises safety systems required for decay heat removal and a substantial degradation of the level of safety of the plant. An Alert is declared in Mode 4 or 5 due to the less severe threat to the protection of the health and safety of the public because of the much longer time available to restore power and decay heat removal systems. 15 minutes was selected to exclude transient or momentary power losses. The inability to restore power within 15 minutes is an indicator of a potential substantial degradation in the level of safety of the plant.

Declaration of an Alert is not applicable when ECCS buses are deenergized for a planned evolution. A planned evolution involves significant preplanning and specific controls are in place prior to knowingly entering the condition.

May 2005 DR 3-58 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG3 INITIATING CONDITION Auto and manual SCRAM not successful and loss of core cooling or heat sink.

EAL THRESHOLD VALUES

1. Failure of BOTH automatic AND manual SCRAMs to establish shutdown criteria..

AND

2. ANY ONE of the following when reactor power is > IRM Range 7:

• Indications exist that adequate core cooling is NOT assured.

OR

• Indications exist that Torus temperature has OR is predicted to exceed 110 °F.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

The conditions of the threshold value are precursors to a core melt sequence and represent imminent substantial core degradation or melting with the potential for loss of containment integrity. Resultant releases can be reasonably expected to exceed Environmental Protection Agency Protective Action Guideline exposure levels offsite for more than the immediate site area.

This EAL is used as an anticipatory condition for the Fission Product Barrier EALs due to the rapid core melt sequence associated with heat generation above the decay heat removal capabilities. When either inadequate core cooling or a degraded heat sink exists at a time when the reactor has not been brought below the power associated with the safety systems design criteria, a core melt sequence condition exists. An uncontrolled temperature rise in the Torus after Standby Liquid Control injection indicates that this means of reactor shutdown is not effective. Inadequate core cooling and a degraded heat sink are indicators of the rapid core melt sequence while the reactor is still critical after an ATWS. Core cooling is extremely challenged when adequate core cooling is not assured as used in the EOPs and the heat sink is extremely challenged when the Torus is used as the last resort heat sink for decay heat removal.

May 2005 DR 3-59 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MG3 (cont.)

This EAL is used as an anticipatory condition for the Fission Product Barrier EALs due to the rapid core melt sequence associated with heat generation above the decay heat removal capabilities. When either inadequate core cooling or a degraded heat sink exists at a time when the reactor has not been brought below the power associated with the safety systems design criteria, a core melt sequence condition exists. An uncontrolled temperature increase in the Torus after Standby Liquid Control injection indicates that this means of reactor shutdown is not effective. Inadequate core cooling and a degraded heat sink are indicators of the rapid core melt sequence while the reactor is still critical after an ATWS. Core cooling is extremely challenged when adequate core cooling is not assured as used in the EOPs and the heat sink is extremely challenged when the Torus is used as the last resort heat sink for decay heat removal.

These thresholds should be evaluated until the reactor is made subcritical by any means.

May 2005 DR 3-60 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS3 INITIATING CONDITION Auto and manual SCRAM not successful.

EAL THRESHOLD VALUE Failure of BOTH automatic AND manual SCRAMs to establish shutdown criteria.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

Automatic or manual SCRAMs are not considered successful if action away from the reactor control console was required to cause control rods to be inserted into the core or any other actions taken to establish shutdown criteria.

Under these conditions the reactor may be producing more heat than the maximum decay heat capacity for which the safety systems were designed. Loss of both the automatic and manual SCRAM capability represents a failure of plant functions necessary for the protection of the public.

This Initiating Condition could lead to observable conditions that could be classified under the Fission Product Barriers. This condition would be recognized earlier than the resultant FPB degradation and is therefore included to provide for more timely recognition and emergency response to potential clad damage and loss of RCS integrity conditions.

May 2005 DR 3-61 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA3 INITIATING CONDITION Auto SCRAM not successful.

EAL THRESHOLD VALUE Failure of Reactor Protection System to initiate AND complete an automatic SCRAM AND a successful manual SCRAM occurred.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

The concern of this EAL is the failure of the automatic Reactor Protection System to initiate and complete a SCRAM to establish shutdown criteria.

The ATWS condition represents a substantial degradation in the level of safety of the plant in that a safety system malfunctioned. Conditions resulting from the ATWS may lead to potential loss of fuel clad or RCS.

The minimum classification for an ATWS is an Alert regardless of actions taken subsequent to the event.

May 2005 DR 3-62 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS4 INITIATING CONDITION Loss of vital 125 VDC power.

EAL THRESHOLD VALUE 125 VDC Battery buses #2 AND #3 bus voltage 110 volts for 15 mins.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

The loss of all vital DC power compromises the ability to monitor and control plant functions required for the protection of the public and is considered a loss of those functions. A prolonged loss of control power may result in core uncovering and loss of containment integrity if there is sufficient decay heat generated by the core and sensible heat in the RCS.

The Threshold Value was chosen to recognize a loss of DC power at a voltage level low enough to be indicative of a severe control system problem. The value is high enough to provide reasonable assurance that the 125-volt batteries will last at least 15 minutes prior to reaching a designed minimum voltage of 105 volts.

May 2005 DR 3-63 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU4 INITIATING CONDITION Loss of vital 125 VDC power.

EAL THRESHOLD VALUE 125 VDC Battery buses #2 AND #3 bus voltage 110 volts for 15 mins.

MODE APPLICABILITY 4, 5 BASIS (References)

The Threshold Value was chosen to recognize a loss of DC buses compromising the ability to monitor and control the removal of decay heat during the Modes 4 and 5.

The Threshold Value was chosen to recognize a loss of DC power at a voltage level low enough to be indicative of a severe control system problem. The value is high enough to provide reasonable assurance that the 125-volt batteries will last at least 15 minutes prior to reaching a designed minimum voltage of 105 volts.

May 2005 DR 3-64 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS5 INITIATING CONDITION Loss of Mode 4 capability (HCTL exceeded).

EAL THRESHOLD VALUE Emergency Depressurization is required due to the inability to maintain Torus bulk temperature below the Heat Capacity Limit curve of DEOP 0200-01.

MODE APPLICABILITY:

1, 2, 3 BASIS: (References)

The value chosen is indicative of a loss heat sink available to reject decay and sensible heat available in a shutdown condition. The value allows a sufficient heat capacity in the Torus water for complete condensation and allows time to establish cooling.

May 2005 DR 3-65 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS6 INITIATING CONDITION Loss of annunciators and indications with a transient in progress.

EAL THRESHOLD VALUE

1. Loss of ALL Safety System Annunciator Panels:

902(3)-3 902(3)-5 902(3)-8 AND

2. A significant transient is in progress.

AND

3. Loss of ALL indications needed to monitor ANY of the following; Criticality OR Core Heat Removal OR Fission Product Barrier status.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

The inability of the Control Room Operators to monitor safety functions or a significant transient in progress represents functions needed for the protection of the public. The inability to monitor functions includes criticality, core heat removal, and the status of the Fission Product Barriers.

Planned actions and time limits are excluded from this EAL since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss or length of the loss is not a factor.

May 2005 DR 3-66 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA6 INITIATING CONDITION Loss of annunciators or indications requiring increased monitoring.

EAL THRESHOLD VALUES

1. Unplanned loss of MOST OR ALL Safety System Annunciators on Panels (below) for 15 mins.:

902(3)-3 902(3)-5 902(3)-8 AND

2. ANY of the following:

A significant transient is in progress.

OR Loss of ALL indications needed to monitor; Criticality OR Core Heat Removal OR Fission Product Barrier status MODE APPLICABILITY:

1, 2, 3 BASIS: (References)

These Threshold Values describe the difficulty associated with assessment of plant conditions during a transient which indicate a potential substantial degradation of the level of safety of the plant. The loss of specific, or several, safety system annunciators or indications should remain a function of the system or component operability status and addressed through Technical Specifications.

The use and definition of MOST is not intended to require a detailed count of lost annunciators or indications but should be used as a guide to assess the ability to monitor the transient. The control panels listed include Reactor Controls, ECCS, Radiological, and Electrical.

May 2005 DR 3-67 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU6 INITIATING CONDITION Unplanned loss of annunciators or indications for 15 mins.

EAL THRESHOLD VALUE Unplanned loss of MOST OR ALL Safety System Annunciators OR indications for 15 mins.

Panels:

902(3)-3 902(3)-5 902(3)-8 MODE APPLICABILITY:

1, 2, 3 BASIS: (References)

These Threshold Values describe a loss of Reactor Control, ECCS, and Electrical panels in the Control Room and represents a potential degradation of the level of safety of the plant in that the ability to monitor the plant is impaired. The loss of specific, or several, safety system annunciators or Control Board indications should remain a function of the system or component operability status and addressed through Technical Specifications. A loss of annunciators is more likely than the loss of indications, both are considered equally for classification purposes in that the ability to monitor the unit is the concern.

The use and definition of MOST is not intended to require a detailed count of lost annunciators or indications but should be used as a guide to assess the ability to monitor the operation of the plant.

May 2005 DR 3-68 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS7 INITIATING CONDITION Loss of decay heat removal and core uncovery.

EAL THRESHOLD VALUES

1. Reactor core is OR will be uncovered.

AND

2. ANY of the following:

• Reactor coolant temperature > 212 °F.

OR

• Uncontrolled temperature increase approaching 212 °F.

MODE APPLICABILITY 4, 5 BASIS (References)

The Threshold Values indicate that severe core damage may occur and Reactor Coolant System (RCS) integrity may not be assured and thus indicate failures of functions needed for the protection of the public.

The conditions address concerns raised by the NRC AEOD Report AEOD/EG09, BWR Operating Experience Involving Inadvertent Draining of the Reactor Vessel, dated August 8,1986. This report states:

In broadest terms, the dominant cause of inadvertent reactor vessel draining are related to the operational and design problems associated with the residual heat removal system when it is entering into or exiting from the shutdown cooling mode. During this transitional period water is drawn from the reactor vessel, cooled by Shutdown Cooling heat exchangers (from the cooling provided by the RBCCW system), and returned to the reactor vessel.

First there are piping and valves in the Shutdown Cooling system which are common to both the shutdown cooling mode and other modes of operation such as low pressure coolant injection and suppression pool cooling. These valves, when improperly positioned provide a drain path for the reactor coolant to flow from the reactor vessel to the suppression pool or the radwaste system.

Second, establishing or exiting the shutdown cooling mode of operation is entirely manual making such evolutions vulnerable to personnel and procedural errors.

May 2005 DR 3-69 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MS7 (cont.)

Third, there is no comprehensive valve interlock arrangement for all the Shutdown Cooling system valves that could be activated during shutdown cooling. Collectively, these factors have contributed to the repetitive occurrences of the operational events involving the inadvertent draining of the reactor vessel.

May 2005 DR 3-70 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA7 INITIATING CONDITION Inability to establish or maintain Mode 4.

EAL THRESHOLD VALUE

1. Inability to establish Mode 4 when required.

OR

2. Inability to maintain Mode 4 conditions as indicated by EITHER:

• RCS temperature increase that exceeds 212 °F.

OR

• Uncontrolled RCS temperature increase approaching 212 °F.

MODE APPLICABILITY 3, 4, 5 BASIS (References)

The inability to establish Mode 4 includes instances when decay heat removal capability is lost prior to reaching Mode 4 when the operational intent is to enter Mode 4 using existing procedures.

The Threshold Values indicate a substantial degradation of the level of safety of the plant by indicating a complete loss of the ability to remove decay heat in Modes 3, 4 and 5.

The Threshold Value related to an uncontrolled temperature increase is necessary to preserve the anticipatory philosophy of NUREG-0654 for events starting from temperatures much lower than 212 °F.

May 2005 DR 3-71 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU7 INITIATING CONDITION Offgas radiation trip or high reactor coolant activity.

EAL THRESHOLD VALUES

1. Offgas system isolation due to valid Offgas post-treatment radiation monitor signal.

OR

2. Coolant activity 5 µCi/gm I-131 dose equivalent.

MODE APPLICABILITY:

ALL BASIS: (References)

The isolation of the offgas system (i.e. chimney isolation valve auto closes) on a valid high radiation signal is an indication of fuel clad degradation and a potential degradation in the level of safety of the plant. The coolant activity level selected exceeds the Technical Specification limits for Iodine spikes and is also indicative of fuel clad degradation. These events are precursors of more severe conditions.

May 2005 DR 3-72 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU8 INITIATING CONDITION RCS leakage.

EAL THRESHOLD VALUES

1. Unidentified RCS leakage into the primary containment > 10 gpm.

OR

2. Total RCS leakage (identified + unidentified) into primary containment > 35 gpm.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

RCS leakage is a precursor to more severe conditions and therefore represents a potential degradation in the level of safety of the plant. The basis for the limit is potential crack propagation of the RCS boundary. Escalation of this EAL to the Alert level is via Fission Product Barrier Degradation RCS Leakage 3.a.1.

The unidentified leakage Threshold Value is twice the Technical Specification limit and is observable in the Control Room but does not require a mass balance calculation.

The total leakage Threshold Value exceeds the Technical Specification limit and is observable in the Control Room but does not require a mass balance calculation.

The Identified leakage contribution to total leakage is a higher value because the source is know, has been evaluated and is therefore of lesser significance than unidentified leakage.

May 2005 DR 3-73 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU9 INITIATING CONDITION Unplanned loss of all onsite or offsite communications capabilities.

EAL THRESHOLD VALUES

1. Loss of ALL onsite communications equipment, Table M1.

OR

2. Loss of ALL offsite communications capability, Table M2.

Table M1 Table M2 Onsite Communications Offsite Communications Equipment Equipment Plant radio system All telephone lines (commercial and microwave)

Plant paging system NARS Sound power phones ENS In-plant telephones HPN Cellular Phones MODE APPLICABILITY ALL BASIS (References)

The Threshold Values recognize a loss of communications capability that either inhibits the ability to perform routine tasks necessary for plant operations or the ability to communicate problems to offsite authorities. The loss of offsite communications capabilities Threshold Value is more comprehensive than those reportable in 10 CFR 50.72. The loss of offsite communications capability is applicable when no direct means is available to communicate with or make notifications to the Electric Operations, State and Federal authorities.

May 2005 DR 3-74 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU10 INITIATING CONDITION Inability to reach required mode within Technical Specification time limits.

EAL THRESHOLD VALUE Plant is NOT brought to required mode within Tech. Spec. LCO Action Statement time.

MODE APPLICABILITY 1, 2, 3 BASIS (References)

Limiting Conditions for Operations require the plant to be brought to a specific mode when an LCO has been entered. Depending on the circumstances this may or may not be an emergency or a precursor to a more serious event. In any case when a plant initiates a shutdown due to having entered an LCO action statement a one-hour report must be made under 10 CFR 50.72(b) non-emergency events. The plant is within its safety envelope when being shutdown within the allowable action statement time of a Tech Spec. An immediate classification of Unusual Event should be made when the plant is not brought to the required mode within the allowable action statement time of any Tech Spec LCO. Declaration is based on the time at which the LCO Action Statement specified time period elapses and is NOT related to how long a condition may have existed.

May 2005 DR 3-75 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA11 INITIATING CONDITION Major fuel damage or fuel uncovery in the Spent Fuel Pool.

EAL THRESHOLD VALUES

1. Radiation reading 10 R/hr. Unit 2 / 3 Refuel Floor OR

2. Report of visual observation of a rapid decrease of Spent Fuel Pool water level such that irradiated fuel is predicted to become uncovered.

MODE APPLICABILITY ALL BASIS: (References)

Uncovering spent fuel represents a substantial degradation of the level of safety of the plant and warrants an Alert classification. Time is available to take corrective actions. NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82," (July, 1987) indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and the risk of injury is low.

Rather than classify based on the fuel pool level falling below the top of irradiated fuel, a conservative radiation level has been chosen. Verification of fuel pool level represents a major ALARA concern considering that radiation levels in excess of 10,000 R/hr on the refuel bridge may be reached when fuel uncovery begins. The value of 10 R/hr was conservatively chosen for classification purposes.

All spent fuel has been removed from Unit 1 Fuel Building.

May 2005 DR 3-76 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU11 INITIATING CONDITION Unplanned decrease in Spent Fuel Pool level.

EAL THRESHOLD VALUE Inability to maintain Unit 2 OR 3 Spent Fuel Pool water level 33.9 ft. above the fuel.

MODE APPLICABILITY ALL BASIS (References)

Classification of an Unusual Event for the Threshold Values is warranted as a precursor to more serious events and thus is a potential degradation of the level of safety of the plant. The events described as Threshold Values tend to have a long lead time relative to a radiological release and thus the threat to public health and safety is very low. The levels provided are the minimum allowed by Technical Specifications.

May 2005 DR 3-77 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MA12 INITIATING CONDITION Unplanned loss of Refueling Cavity level.

EAL THRESHOLD VALUES

1. Report of visual observation of a rapid decrease of refueling cavity water level such that irradiated fuel is predicted to become uncovered.

OR

2. Reactor vessel level -143" (TAF).

MODE APPLICABILITY 5

BASIS (References)

Uncovering irradiated fuel represents a substantial degradation of the level of safety of the plant and therefore warrants an Alert classification. Time is available to take corrective actions.

NUREG/CR-4982, "Severe Accidents in Spent Fuel pools in Support of Generic Safety Issue 82," indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and the risk of injury is low.

May 2005 DR 3-78 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY SYSTEM MALFUNCTIONS MU12 INITIATING CONDITION Unplanned loss of Refueling Cavity level.

EAL THRESHOLD VALUE Inability to maintain Refueling Cavity Level - 59.

MODE APPLICABILITY 5

BASIS (References)

Classification of an Unusual Event for the Threshold Values is warranted as a precursor to more serious events and thus are a potential degradation of the level of safety of the plant. The events described as Threshold Values tend to have a long lead time relative to a radiological release and thus the threat to public health and safety is very low.

May 2005 DR 3-79 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HG1 INITIATING CONDITION Security event resulting in loss of physical control of the facility.

EAL THRESHOLD VALUES

1. Loss of physical control of the Main Control Room due to a security event.

OR

2. Loss of physical control of the facility (remote shutdown capability) due to a security event.

MODE APPLICABILITY ALL BASIS (References)

This EAL is an escalation of the Site Area Emergency declaration for a hostile force intrusion into a Vital Area. A hostile force taking physical control of either the Control Room or all remote shutdown capabilities (e.g., control of Remote Shutdown Panels, ability to perform critical plant activities/manipulations) results in a loss of physical control of the facility.

Threshold #2: A security shift supervisor should be consulted to determine which areas of the plant are affected, and the extent of the effects, prior to making the determination of these threshold values.

May 2005 DR 3-80 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HS1 INITIATING CONDITION Confirmed Security Event in a Vital Area.

EAL THRESHOLD VALUES

1. Intrusion into plant Vital Area by a hostile force.

OR

2. A security event that results in the loss of control of ANY Vital Area (other than the Main Control Room).

OR

3. Imminent loss of physical control of the facility due to a security event.

OR

4. A confirmed bomb device discovered in a Vital Area.

MODE APPLICABILITY ALL BASIS (References)

This Threshold Value escalates from the Alert Protected Area intrusion to a Vital Area intrusion of a hostile force.

A security event is as defined in the Security Plan. A security shift supervisor should be consulted when making the determination of these threshold values.

Loss of physical control of the Control Room or remote shutdown capabilities (e.g., control of Remote Shutdown Panels, ability to perform critical plant activities/manipulations) due to a security event is to be classified as a General Emergency via HG1.

A "confirmed bomb device" is a determination made by the security force through the Security Plan, contingency procedures and other guidance documentation, or any other Federal Law Enforcement Agency.

May 2005 DR 3-81 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HA1 INITIATING CONDITION Confirmed Security Event in the Protected Area.

EAL THRESHOLD VALUES

1. Intrusion into the Protected Area by a hostile force.

OR

2. Other Security Events of increasing severity that persists for 30 mins. (see Table H2 for example events).

Table H2 - Examples of Security Events Credible bomb device.

Extortion.

Suspicious fire or explosion.

Significant security system hardware failure.

Loss of Guard Post contact.

MODE APPLICABILITY ALL BASIS (References)

A civil disturbance that penetrates the Protected Area can be considered a hostile force. Intrusion of a hostile force into the Protected Area represents a potential for a substantial degradation in the level of safety of the plant.

A security event is considered of increasing severity if it is discovered within the Protected Area and is NOT under control of the security force within 30 minutes.

A security shift supervisor should be consulted when making the determination of these threshold values.

May 2005 DR 3-82 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HU1 INITIATING CONDITION Confirmed Security Event that indicates a potential degradation in level of safety of the plant.

EAL THRESHOLD VALUES

1. A credible threat to the station reported by the NRC.

OR

2. An Actual Threat that meets ALL of the following criteria:

• A credible threat reported by ANY other outside agency OR security procedures.

AND

• Is specifically directed towards the station.

AND

• Is imminent (within 2 hrs.).

OR

3. Bomb device discovered within the Protected Area AND outside a plant Vital Area.

OR

4. Confirmed tampering with safety related equipment.

OR

5. A hostage situation that disrupts normal plant operations.

OR

6. A disturbance (civil, internal OR strike) which disrupts normal plant operations.

MODE APPLICABILITY ALL May 2005 DR 3-83 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HU1 (cont.)

BASIS (References)

A security shift supervisor should be consulted when making the determination of these threshold values.

A bomb device discovered within the PA but outside an area that contains safety functions or systems is a potential degradation in the level of safety of the plant and therefore an Unusual Event.

Confirmed tampering is adapted from the list of security plan contingencies.

A hostage situation is considered to disrupt normal operations if it results in the inability to perform surveillance activities, alters unit operations, or as described in the security plan.

A disturbance is considered to disrupt normal plant operations if it obstructs ingress or egress to the owner controlled or protected area, or if it requires actions to intervene and control by Local Law Enforcement Agencies (LLEA).

May 2005 DR 3-84 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HG2 INITIATING CONDITION Conditions indicate imminent core damage and release affecting the public.

EAL THRESHOLD VALUES

1. Actual OR imminent core degradation AND potential loss of Containment.

OR

2. Potential uncontrolled radionuclide release which can reasonably be expected to reach or exceed 1 Rem TEDE OR 5 Rem CDE thyroid plume exposure levels at the Site Boundary.

MODE APPLICABILITY ALL BASIS (References)

This Emergency Action Level allows for classification of events which in the judgement of the Emergency Director warrant the General Emergency classification but do not fit into any other General Emergency criteria.

Emergency Director judgement is to be based on known conditions and the expected response to mitigating activities within a short time period arbitrarily set at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Classification of a General Emergency is not to be delayed pending an extended evaluation of possibilities and probabilities. If time allows and the offsite response organizations are active, consultation with the effected state and the NRC is prudent prior to classification.

May 2005 DR 3-85 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HS2 INITIATING CONDITION Conditions indicate actual or likely failure of functions needed for public protection.

EAL THRESHOLD VALUE Other conditions exist which in the judgment of the Emergency Director indicate actual OR likely major failures of plant functions needed for protection of the public.

MODE APPLICABILITY ALL BASIS (References)

The judgment call of the Emergency Director that conditions indicate actual or likely major failures of plant functions needed for protection of the public is intended for unanticipated conditions not explicit in other EALs. Additional assistance, monitoring and direction is made available by activation of the Emergency Operations Facility consistent with declaration of a Site Area Emergency.

Emergency Director judgement is to be based on known conditions and the expected response to mitigating activities within a short time period. Classification of a Site Area Emergency is not to be delayed pending an extended evaluation of possibilities and probabilities.

May 2005 DR 3-86 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HA2 INITIATING CONDITION Conditions indicate actual or potential substantial degradation of the level of safety of the plant.

EAL THRESHOLD VALUE Other conditions exist which in the judgment of the Emergency Director indicate that plant safety systems may be degraded AND that increased monitoring of plant functions is warranted.

MODE APPLICABILITY ALL BASIS (References)

A condition exists which, in the judgement of the Emergency Director, presents an actual or potential substantial degradation in the level of safety of the plant. Emergency Director judgement is to be based on known conditions and the expected response to mitigating activities within a short time period.

The Emergency Director should further consider how these Threshold Values may affect both units at the site due to the effects on common or shared plant systems. These effects may warrant upgrading the Emergency Class based on the nature of the combined effects between the units.

May 2005 DR 3-87 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HU2 INITIATING CONDITION Conditions indicate a potential degradation of the level of safety of the plant.

EAL THRESHOLD VALUE Other conditions exist which in the judgment of the Emergency Director indicate a potential degradation in the level of safety of the plant (see Table H1 for example conditions).

Table H1 - Example Conditions

• An event onsite OR near the site which requires the assistance of the local Police OR Fire Department.

• Uncontrolled RCS cooldown.

• Plant conditions require operation outside analyzed conditions.

• Approaching ANY Unusual Event Threshold Value as an anticipatory judgment call.

• Abnormal operating conditions persist because mitigation activities are NOT effective.

• Additional assistance is required to ensure the protection of the health and safety of the workers and the public.

• Unusual aircraft activity over the Protected Area.

MODE APPLICABILITY ALL BASIS (References)

Any condition not explicitly detailed as an EAL Threshold Value that in the judgment of the Emergency Director is a potential degradation in the level of safety of the plant. Emergency Director judgement is to be based on known conditions and the expected response to mitigating activities within a short time period.

Some examples of actual events that may require emergency director judgement are listed here for consideration. However this list is by no means all-inclusive and is not intended to limit the discretion of the Emergency Director. This list is provided solely as examples for consideration and it is recognized that actual events may not always follow a preconceived description.

The Emergency Director should further consider how these Threshold Values may affect both units at the site due to the effects on common or shared plant systems. These effects may warrant upgrading the Emergency Class based on the nature of the combined effects between the units.

A security shift supervisor should be consulted when making the determination of the security related threshold values.

May 2005 DR 3-88 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HS3 INITIATING CONDITION Main Control Room evacuated and control not established in 30 mins.

EAL THRESHOLD VALUES

1. Main Control Room evacuation initiated.

AND

2. Control of the plant CANNOT be established per DSSP 0100-CR in 30 mins.

MODE APPLICABILITY:

All BASIS: (References)

A station without Remote Shutdown Panels may not be able to meet the 15-minute criteria per NUMARC NESP-007. 30 minute provides a reasonable time period to gain local control and establish communications DSSP 0100-CR .This Threshold Value is an escalation of the Alert for Control Room evacuation and is more serious because control of plant systems has not been established within 30 minutes of the reactor SCRAM.

The primary concern of establishing control in Mode 3 is ensuring the integrity of the Fission product Barriers. The primary concern of establishing control in Modes 4 and 5 is decay heat removal.

The Emergency Director should further consider how these Threshold Values may affect both units at the site due to the effects on common or shared plant systems. These effects may warrant upgrading the emergency class based on the nature of the combined effects between the Units.

May 2005 DR 3-89 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HA3 INITIATING CONDITION Main Control Room evacuation initiated.

EAL THRESHOLD VALUE Entry into DSSP 0100-CR for Control Room evacuation.

MODE APPLICABILITY All BASIS (References)

Evacuation of the Control Room represents a potential for substantial degradation of the level of safety of the plant and therefore requires an Alert declaration. Additional support, monitoring and direction is required and accomplished by activation of the Technical Support Center at the Alert classification level.

The primary concern of establishing control in Mode 3 is ensuring the integrity of the Fission product Barriers. The primary concern of establishing control in Modes 4 and 5 is decay heat removal.

The Emergency Director should further consider how these Threshold Values may affect both units at the site due to the effects on common or shared plant systems. These effects may warrant upgrading the Emergency class based on the nature of the combined effects between the Units.

May 2005 DR 3-90 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HA4 INITIATING CONDITION Natural or destructive phenomena inside a Vital Area.

EAL THRESHOLD VALUE

1. Confirmed seismic event Operating Basis Earthquake (OBE) as indicated by seismic instrumentation registering > 0.1g.

OR

2. Tornado striking structures containing systems required to establish OR maintain Mode 4.

OR

3. Sustained high winds 90 mph on PPDS, OR indicated for 15 mins. by Control Room report, computer point OR chart recorders.

OR

4. Report of visible structural damage to a structure containing systems required to establish or maintain Mode 4.

OR

5. Vehicle collision affecting a Vital Area.

OR

6. Main Turbine rotating component failure that penetrates the casing AND generates missiles causing damage to structures containing safety related equipment.

OR

7. Flooding in a Vital Area > Max. Safe Water Level for ANY area listed in DEOP 300-1.

MODE APPLICABILITY ALL May 2005 DR 3-91 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HA4 (cont.)

BASIS (References)

These Threshold Values are natural or destructive phenomena that represent actual or potential substantial degradation of the level of safety of the plant. The affects of the phenomena should also be evaluated on a system or component basis in relation to Technical Specification and evaluated for further classification via the System Malfunction and Fission Product Barrier Recognition Categories.

The Emergency Director should consider how these Threshold Values may affect both units due to the affects of common or shared plant systems.

Threshold Value 1 - Seismic events at the UFSAR (2.2.6) Operating Basis Earthquake (OBE) level may cause damage to safety systems and represent a potential substantial degradation of the level of safety of the plant. A confirmation call to the National Earthquake Center will only confirm that an earthquake has occurred and will not provide a confirmation of the amplitude of the earthquake.

Threshold Value 2, 4, and 5 - The destructive phenomena must include an affect on systems or components that affects the operability or integrity of the system or structure within a Vital Area.

Threshold Value 3 - Sustained winds present a more severe loading on the buildings than a gust.

The Threshold Value is 90% of the maximum measurable wind speed.

Threshold Value 4 - Should be used in conjunction with investigations of Threshold Values 1 and 2. A detailed description or assessment of damage is not intended to be used to meet the intent of this Threshold Value.

Threshold Value 5 - A collision by any vehicle on land, from the air or on water that affects structures or equipment necessary for safe operation within a Vital Area.

Threshold Value 6 - This Value addresses the threat to safety equipment imposed by missiles generated by main turbine rotating component failures. This includes all areas containing safety related equipment, controls or power supplies.

Threshold Value 7 - Flooding in Vital Areas that affect operability of safety related systems or components. The source of the flooding need not be known.

May 2005 DR 3-92 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HU4 INITIATING CONDITION Natural or destructive phenomena inside the Protected Area or Switchyard.

EAL THRESHOLD VALUE

1. Confirmed Seismic event EITHER:

• Felt by onsite personnel OR

• Indicated by installed seismic accelograph reading 0.01 g.

OR

2. Report of a tornado strike within the Protected Area OR Switchyard.

OR

3. Vehicle collision into structures containing systems necessary for safe shutdown within the Protected Area OR in the Switchyard.

OR

4. Main Turbine rotating component failure causing visible casing damage or damage to the generator seals.

MODE APPLICABILITY:

ALL BASIS: (References)

These Threshold Values are natural or destructive phenomena which represent potential degradation of the level of safety of the plant. The affects of the phenomena should also be evaluated on a system or component basis in relation to Technical Specification and evaluated for further classification via the System Malfunction Recognition Category.

The Emergency Director should consider how these Threshold Values may affect both units due to the affects of common or shared plant systems.

Threshold Value 1 - Seismic events at the lowest instrumentation actuation level may cause damage to systems and represent a potential degradation of the level of safety of the plant.

Following initial indication of a seismic event, a call to the National Earthquake Center will only confirm that an earthquake has occurred and will not provide a value for the amplitude of the earthquake.

May 2005 DR 3-93 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HU4 (cont.)

Threshold Value 2 - Any report that a tornado has touched down in the Protected Area meets this Threshold Value. The switchyard is considered within the Protected Area for classification purposes.

Threshold Value 3 - A collision by any vehicle on land, from the air or on water which affects structures or equipment necessary for safe shutdown within the Protected Area or in the Switchyard.

Threshold Value 4 - Failure of the rotating components has the potential for leakage of flammable fluids (oil and hydrogen) into the Turbine Building.

May 2005 DR 3-94 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HA5 INITIATING CONDITION Fire or explosion affecting operability of systems required to establish or maintain Mode 4.

EAL THRESHOLD VALUES

1. Fire OR explosion in OR near ANY structure containing systems required to establish OR maintain Mode 4. (Table H3 Example Safe Shutdown Areas).

AND

2. ANY ONE of the following:

• Affected system parameter indications show degraded performance.

OR

• Report received of visible damage to permanent structures OR equipment within the affected area.

Table H3 Example Safe Shutdown Areas Reactor Building Cribhouse Control Room Aux. Electric Room Diesel Generator Electrical Switchyard Switchgear and Battery Rooms Remote Shutdown Areas MODE APPLICABILITY ALL BASIS (References)

Only explosions of sufficient force to cause damage (deformation, scorching) to structures or equipment required for safe operation should be considered. This EAL is based on fires or explosions that have damaged plant structures that will affect the safe operation of the plant.

These structures include buildings and areas immediately adjacent or connected to plant Vital Areas and other significant buildings or areas.

May 2005 DR 3-95 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HA5 (cont.)

To meet the Alert criteria, monitoring of systems in the affected area must also indicate a degradation of the system or reports of conditions from the location indicate visible damage to equipment. Damage assessment should not be extended beyond the report of visible damage for classification purposes. A classification should be made based on a report of visible damage and not delayed for assessment activities.

The declaration of an Alert and the activation of the TSC will provide the Emergency Director with the necessary resources to perform these damage assessments.

The Emergency Director should further consider how these Threshold Values may affect both units at the site due to the effects on common or shared plant systems. These effects may warrant upgrading the Emergency class based on the nature of the combined effects between the Units.

May 2005 DR 3-96 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HU5 INITIATING CONDITION Fire in the Protected Area not extinguished in 15 mins. or explosion in the Protected Area.

EAL THRESHOLD VALUE

1. Fire in the Protected Area NOT extinguished in 15 mins. of EITHER:

• Main Control Room notification.

OR

• Verification of alarms.

OR

2. Explosion in the Protected Area.

MODE APPLICABILITY ALL BASIS (References)

The purpose of this EAL is to address only fires that are potentially significant precursors to safety system damage. This excludes such items as fires within administration buildings, wastebasket fires, and other small fires of no consequence. This EAL applies to buildings and areas immediately adjacent or connected to plant Vital Areas or other significant buildings or areas. The intent is not to include buildings (e.g. warehouses) or areas that are not immediately adjacent or connected to plant Vital Areas. Any other areas would not be considered for classification of an Unusual Event.

Only explosions of sufficient force (including high energy line breaks) to cause damage should be considered. Damage assessment should not be beyond the report of visual damage for classification purposes. The Emergency Director also needs to consider any security aspects of the explosion.

The Emergency Director should further consider how these Threshold Values may affect both units at the site due to the effects on common or shared plant systems. These effects may warrant upgrading the Emergency Class based on the nature of the combined effects between the units.

May 2005 DR 3-97 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HA6 INITIATING CONDITION Release of toxic or flammable gases within or adjacent to a Vital Area.

EAL THRESHOLD VALUE

1. Unanticipated report OR detection of toxic gases within OR adjacent to a Vital Area in concentrations that may be life threatening to personnel.

OR

2. Unanticipated report OR detection of flammable gases within OR adjacent to a Vital Area in concentration greater than the lower flammability limit.

MODE APPLICABILITY ALL BASIS (References)

The EAL Threshold values are met if the conditions are not an anticipated result of a planned evolution that includes compensatory measures This IC is based on gases that affect the safe operation of the plant. It applies to buildings and areas immediately adjacent or connected to plant Vital Areas and other significant buildings or areas. The intent is not to include buildings or other areas that are not immediately adjacent or connected to plant Vital Areas.

EAL #1 is met if, by reasonable determination or measurement of toxic gas concentration results in an atmosphere that is Immediately Dangerous to Life and Health (IDLH) and is not an anticipated result of a planned evolution that includes compensatory measures. Exposure to an IDLH atmosphere will result in immediate harm to unprotected personnel, and would preclude access to any such affected areas.

EAL #2 is met when, by reasonable determination or measurement the flammable gas concentration exceeds the Lower Flammability Limit and is not an anticipated results of a planned evolution that includes compensatory measures. This EAL addresses concentrations at which gases can ignite/support combustion. An uncontrolled release of flammable gasses has the potential to affect safe operation of the plant by limiting either operator or equipment operations due to the potential for ignition and resulting equipment damage/personnel injury.

It is appropriate that increased monitoring be done to ascertain whether consequential damage has occurred. Escalation to a higher emergency class, if appropriate, will be based on other ICs.

May 2005 DR 3-98 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HU6 INITIATING CONDITION Toxic or flammable gas release affecting normal plant operations.

EAL THRESHOLD VALUES

1. Report or detection of toxic OR flammable gases in amounts that can affect normal plant operations.

OR

2. Report by Local, County, OR State officials for evacuation OR sheltering of site personnel based on an offsite event involving toxic OR flammable gases in amounts that can affect normal plant operations.

MODE APPLICABILITY ALL BASIS: (References)

This IC is based on the existence of uncontrolled releases of toxic or flammable gas that may enter the site boundary and affect normal plant operations. The IC assumes an uncontrolled process that has the potential to affect plant operations, or personnel safety. The source of the toxic or flammable gas could be from inside or outside the site and is NOT the issue. The event is considered to affect normal operations if it results in the inability to perform surveillance activities, alters unit operations, or as described in the Security Plan.

It is intended that releases of toxic or flammable gases are of sufficient quantity, and the release point of such gases is such that normal plant operations would be affected. This would preclude small or incidental releases, or releases that do not impact structures needed for plant operation.

The EALs are intended to not require significant assessment or quantification.

Escalation of this EAL is via HA6, which involves a quantified release of toxic or flammable gas affecting Vital Areas.

May 2005 DR 3-99 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY HAZARDS AND OTHER CONDITIONS HU7 INITIATING CONDITION Transportation of radiologically contaminated or potentially contaminated person(s) to an offsite medical facility.

EAL THRESHOLD VALUE Radiologically contaminated OR potentially contaminated injured person(s) transferred to an offsite medical facility for treatment.

MODE APPLICABILITY ALL BASIS (References)

This EAL insures that proper authorities are notified when a contaminated or potentially contaminated individual is transported to an offsite medical facility. If it is uncertain that the individual is contaminated an Unusual Event should be conservatively declared until proven otherwise.

May 2005 DR 3-100 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ISFSI MALFUNCTION E-RU1 INITIATING CONDITION Unexpected Increase in ISFSI Radiation.

EAL THRESHOLD VALUES

1. Radiation reading on a HI-STAR spent fuel storage cask EITHER:

• 250 mR/hr. (neutron + gamma) on contact of ANY exterior vertical surface.

OR

• 160 mR/hr. (neutron + gamma) on contact of ANY top horizontal surface.

OR

2. ANY of the following radiation readings on a HI-STORM spent fuel storage cask:

• 80 mR/hr. (neutron + gamma) on contact of ANY exterior vertical surface.

• 20 mR/hr. (neutron + gamma) on contact of ANY top horizontal surface.

• 32 mR/hr. (neutron + gamma) at the inlet AND outlet vent ducts.

MODE APPLICABILITY N/A BASIS (References)

This EAL addresses the degradation of irradiated spent fuel stored onsite in HI-STAR or HI-STORM spent fuel casks. These casks are designed to standards identified in 10 CFR Part 72.

The dry storage casks are routinely monitored by Radiation Protection personnel, such that any degradation would be detected. Increases in radiation levels may indicate a potential criticality event. Readings of 2 times the ISFSI Technical Specification radiation limits are indicative of degradation of the irradiated spent fuel storage cask.

This EAL is applicable for spent fuel storage casks in the Storage Operations mode as defined by the cask certificate of compliance.

May 2005 DR 3-101 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ISFSI MALFUNCTION E-HU1 INITIATING CONDITION Damage to a loaded cask confinement boundary.

EAL THRESHOLD VALUE:

1. Natural or destructive phenomena affecting a loaded cask confinement boundary.

Natural and Destructive Phenomena may include:

dropped cask tipped over cask explosion fire seismic event tornado flood lightning burial under debris OR

2. ANY condition in the opinion of the Emergency Director that indicates loss of loaded fuel storage cask confinement boundary MODE APPLICABILITY N/A BASIS (References)

An Unusual Event in this Initiating Condition is categorized on the basis of the occurrence of an event of sufficient magnitude that a loaded cask confinement boundary is damaged (visible conditions exist that may have caused violation of the boundary) or violated (indication that the containment boundary is no longer intact). This includes classification based on a loaded fuel storage cask confinement boundary loss leading to the degradation of the fuel during storage or posing an operational safety problem with respect to its removal from storage.

For the second Threshold Value, any condition not explicitly detailed as an EAL Threshold Value, which, in the judgment of the Emergency Director, is a potential degradation in the level of safety of the ISFSI. Emergency Director judgment is to be based on known conditions and the expected response to mitigating activities within a short time period.

ISFSI Technical Specifications allow time to complete required actions if cask seal integrity is not maintained; therefore, classification should not be made based on a loss of seal integrity by itself. However, loss of seal integrity coincident with an accident condition or natural phenomena affecting a cask would justify classification.

May 2005 DR 3-102 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear RECOGNITION CATEGORY ISFSI MALFUNCTION E-HU2 INITIATING CONDITION Confirmed Security Event with potential loss of level of safety of the ISFSI.

THRESHOLD VALUE Security Event in OR around the ISFSI as determined by the Security Shift Supervisor.

MODE APPLICABILITY N/A BASIS: (References)

This EAL is based on the Security Plan. Security events that do not represent a potential degradation in the level of safety of the ISFSI, are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72.

Reference is made to the Security Shift Supervisor because these individuals are the designated personnel qualified and trained to confirm that a security event is occurring or has occurred.

May 2005 DR 3-103 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Section 4: Emergency Measures Exelon Nuclear emergency response actions are the same for all nuclear stations and are thus covered by Section E of the Emergency Plan.

4.1 Notification of the Emergency Organization Standard NARS notifications for the Dresden Station are made to the State of Illinois Emergency Management Agency (IEMA). At the Dresden Station, if a General Emergency is the initiating event, the Emergency Director is responsible for notifying the following additional Illinois and local agencies:

• Grundy County EOC

• Grundy County Sheriff's Office

• Kendall County EOC

• Kendall County Sheriff's Office

• Will County EOC

• Will County Sheriff's Office 4.2 Assessment Actions Throughout each emergency situation, continuing assessment will occur. Assessment actions at Dresden Station may include an evaluation of plant conditions; in-plant, onsite, and initial offsite radiological measurements; and initial estimates of offsite doses. Core damage information is used to refine dose assessments and confirm or extend initial protective action recommendations. Dresden Station utilizes NEDC-33045P-A, Revision 0, (2001) as the basis for the methodology for post-accident core damage assessment.

This methodology utilizes real-time plant indications in addition to samples of plant fluids and atmospheres. Core damage is qualitatively evaluated per NRC Core Condition Categories (1-10) as shown in the table below.

Degree of Minor Intermediate Major Degradation (<10%) (10% to 50%) (>50)

No Core Damage 1 1 1 Cladding Failure 2 3 4 Fuel Overheat 5 6 7 Fuel Melt 8 9 10 4.3 Protective Actions for the Offsite Public To aid Control Room personnel during a rapidly developing emergency situation, Figure 4-1, "Protective Action Recommendation (PAR) Determination Flowchart for Dresden Station" has been developed based on Section J.10.m of the Emergency Plan.

May 2005 DR 4-1 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear 4.3.1 Alert and Notification System (ANS) Sirens This ANS consists of a permanently installed outdoor notification system within a ten mile radius around the station. The ten-mile radius around the station is a mix of agriculture and industry with a relatively low population distribution. The ANS, as installed, consists of mechanical and electronic sirens that will cover this entire area with a minimum sound level of 60 db. Additionally, the ANS will cover the heavily populated areas within the ten-mile radius around the station with a minimum sound level of 70 db to ensure complete coverage. The ANS sirens are controlled and monitored on a daily basis, by a computerized telemetry system. The daily monitoring assures early failure detection and therefore maximizes system operability and reliability.

4.3.2 Evacuation Time Estimates The evacuation time estimates were developed per the requirements of NUREG-0654, and to support the Illinois Plan For Radiological Accidents (IPRA) - Dresden Volume II.

The purpose of the evacuation time estimates is to assess the postulated evacuation times for the Dresden Station Emergency Planning Zone (EPZ).

The evacuation time estimate data was updated per a study performed by Earth Tech. Inc.

documented in their report dated December, 2003 entitled Evacuation Time Estimates for the Dresden Station Plume Exposure Pathway Emergency Planning Zone.

The evacuation times are based on a detailed consideration of the EPZ roadway network and population distribution. The information in Table 4-1 presents representative evacuation times for daytime and nighttime scenarios, for summer and winter seasons, and under various weather conditions for the evacuation of various areas around the Dresden Station, once a decision has been made to evacuate. The evacuation times noted include notification, mobilization, and travel time. These times are for the general population which include permanent population and special facilities (schools, nursing homes, hospitals, and recreational areas). Table 4-2 provides information on the permanent population distribution (by compass sector) that was used for this study.

4.4 Protective Actions for Onsite Personnel Dresden Station has a siren system to assemble personnel during emergency conditions.

Upon hearing a continuous two (2) minute siren, all personnel not having emergency assignments have been instructed to assemble in predesignated assembly areas. Refer to Figure 4-2.

Assembly of site personnel, for purposes of accountability and possible evacuation, is initiated per the requirements of Section J of the Emergency Plan. Accountability of site personnel is handled by the Dresden Station security force.

If a site evacuation of non-essential personnel is required by Section J of the Emergency Plan, personnel will be either relocated and monitored at the Dresden Station Relocation Centers or sent home if there is no release or radiological or safety concerns. The designated relocation centers for Dresden Station are:

• LaSalle County Nuclear Power Station; May 2005 DR 4-2 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear

• Braidwood Nuclear Power Station; For evacuation routes, refer to Figure 4-3 and 4-4. Traffic control for onsite areas will be handled by the Dresden Nuclear Power Station security force, if necessary.

Equipment and personnel would be available at LaSalle County Station and Braidwood Station for monitoring, decontamination and bioassay. Other emergency measures common to all nuclear stations are discussed in the Emergency Plan.

May 2005 DR 4-3 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Figure 4-1: Dresden Station PAR Determination Flowchart No No PARs General Emergency Declared?

Required Yes Controlled Direct Containment Shelter 2 Mile Radius &

Vent < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Yes 5 Miles Downwind.

AND Advise Remainder of EPZ to Release in Progress > PAGs Monitor Local Radio Stations.

No 2 Mile Radius, 5 Miles Downwind Loss of FUEL No WD (from) Subareas Fission Product Barrier 002º to 046º 1, 3. 4, 7 047º to 182º 1, 3. 4 Yes 183º to 292º 1, 3, 4, 12 Loss of RCS No 293º to 299º 1, 3, 4 Fission Product Barrier 300º to 338º 1, 3, 4, 9 339º to 001º 1, 3, 4, 7, 9 Yes Loss of CONTAINMENT No Fission Product Barrier Yes Evacuate 5 Mile Radius & Evacuate 2 Mile Radius &

10 Miles Downwind. 5 Miles Downwind.

Advise Remainder of EPZ to Advise Remainder of EPZ to Monitor Local Radio Stations. Monitor Local Radio Stations.

5 Mile Radius, 10 Miles Downwind 2 Mile Radius, 5 Miles Downwind WD (from) Subareas WD (from) Subareas 002º to 026º 1, 3, 4, 7, 9, 10, 11, 12 002º to 046º 1, 3. 4, 7 027º to 044º 1, 3, 4, 7, 8, 9, 11, 12 047º to 182º 1, 3. 4 045º to 068º 1, 3, 4, 7, 8, 9, 12 183º to 292º 1, 3, 4, 12 069º to 083º 1, 2, 3, 4, 5, 7, 8, 9, 12 293º to 299º 1, 3, 4 084º to 092º 1, 2, 3, 4, 5, 7, 9, 12 300º to 338º 1, 3, 4, 9 093º to 112º 1, 2, 3, 4, 7, 9, 12 339º to 001º 1, 3, 4, 7, 9 113º to 143º 1, 2, 3, 4, 6, 7, 9, 12 144º to 177º 1, 3, 4, 6, 7, 9, 12 178º to 199º 1, 3, 4, 6, 7, 9, 12, 13 200º to 225º 1, 3, 4, 7, 9, 12, 13 226º to 249º 1, 3, 4, 7, 9, 12, 13, 14 250º to 266º 1, 3, 4, 7, 9, 12, 14 267º to 286º 1, 3, 4, 7, 9, 12, 14, 15 287º to 321º 1, 3, 4, 7, 9, 12, 15 322º to 344º 1, 3, 4, 7, 9, 12, 15, 16 345º to 353º 1, 3, 4, 7, 9, 10, 12, 15, 16 354º to 001º 1, 3, 4, 7, 9, 10, 12, 16 May 2005 DR 4-4 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Figure 4-2: Dresden Onsite Assembly Areas and Emergency Response Facilities May 2005 DR 4-5 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Figure 4-3: Dresden Evacuation Route to LaSalle Station Dresden Station Dresden Road Pine Bluff Road 2 1/2 miles south 8 miles west LaSalle Station Rte. 47 7 miles south Grand Ridge Road 13 miles west Directions:

Take Pine Bluff Rd west to Rt. 47. Turn south on Rt. 47 to Grand Ridge Rd in Mazon. Turn west on Grand Ridge Rd. and proceed 13 miles to LaSalle Station.

May 2005 DR 4-6 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Figure 4-4: Dresden Evacuation Route to Braidwood Station Dresden Station Dresden Road 2 1/2 miles south Lorenzo Road I-55 4 1/2 miles east 1 1/2 miles south Rte 129 I-55 4 miles south Rte 113 Rte. 53 1 1/2 miles south Braidwood Station Directions:

Take Lorenzo Rd east to I-55. Turn south on I-55 and proceed to the Rt. 129 Braidwood Exit.

Exit left off I-55 and proceed to Rt. 113 in Braidwood. Turn east on Rt. 113 and immediately upon crossing the RR tracks turn south on Rt. 53. Proceed 1 1/2 miles south on Rt. 53 to Braidwood Station.

May 2005 DR 4-7 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Table 4-1: Dresden Evacuation Time Estimates (in minutes) (1)

Summer Daytime Summer Nighttime Winter Daytime Winter Nighttime PAR Evacuation Zone fair adverse fair adverse fair adverse fair adverse 2 Mile Radius & 5 Miles Downwind WD 002 to 046 [1, 3, 4, 7] 190 215 125 130 180 220 120 130 WD 047 to 182 [1, 3, 4] 185 210 120 125 175 215 115 125 WD 183 to 292 [1, 3, 4, 12] 190 215 125 130 180 220 120 130 WD 293 to 299 [1, 3, 4] 185 210 120 125 175 215 115 125 WD 300 to 338 [1, 3, 4, 9] 185 210 120 125 175 215 115 125 WD 339 to 001 [1, 3, 4, 7, 9] 190 215 125 130 180 220 120 130 5 Mile Radius & 10 Miles Downwind WD 002 to 026 [5R, 10, 11] 190 220 130 135 180 225 125 135 WD 027 to 044 [5R, 8, 11] 190 220 130 135 180 225 125 135 WD 045 to 068 [5R, 8] 190 220 130 135 180 225 125 135 WD 069 to 083 [5R, 2, 5, 8] 190 220 130 135 180 230 125 135 WD 084 to 092 [5R, 2, 5] 190 220 130 135 180 230 125 135 WD 093 to 112 [5R, 2] 190 220 130 135 180 225 125 135 WD 113 to 143 [5R, 2, 6] 190 220 130 135 180 225 125 135 WD 144 to 177 [5R, 6] 190 220 130 135 180 225 125 135 WD 178 to 199 [5R, 6, 13] 215 265 140 170 200 280 125 175 WD 200 to 225 [5R, 13] 215 265 140 170 200 280 125 175 WD 226 to 249 [5R, 13, 14] 215 265 140 170 200 280 125 175 WD 250 to 266 [5R, 14] 190 220 130 140 180 225 125 135 WD 267 to 286 [5R, 14, 15] 190 220 130 140 185 230 130 140 WD 287 to 321 [5R, 15] 190 220 130 135 180 225 125 135 WD 322 to 344 [5R, 15, 16] 195 220 130 135 180 225 125 135 WD 345 to 353 [5R, 10, 15, 16] 195 220 130 135 190 235 125 140 WD 354 to 001 [5R, 10, 16] 195 220 130 135 180 225 125 135 Full EPZ 215 265 140 170 200 280 130 175 (1) Times are rounded to the nearest 5 minutes (2) Sub-Areas in brackets. See Figure 1-2 for Sub-Area locations. PAR evacuation zones per EP-AA-111 (3) "5R" designates all Sub-Areas within 5-mile radius (Sub-Areas 1, 3, 4, 7, 9, 12)

(4) WD is the direction (in degrees) from which the wind is blowing (00 or 360 represents a wind from north to south)

May 2005 DR 4-8 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Table 4-2: Dresden Permanent Population Distribution By Compass Sector May 2005 DR 4-9 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Section 5: Emergency Facilities and Equipment 5.1 Emergency Response Facilities Refer to Figure 4-2 for the location of the Dresden Station Control Room, Technical Support Center (TSC), and Operations Support Center (OSC) within the Station's Protected Area boundary.

5.1.1 Station Control Room The Dresden Station Control Room is the initial onsite center of emergency control. The Dresden Station Units 2 and 3 Control Rooms are located at the 534' level at the east end of the Unit 2/3 Turbine Building.

5.1.2 Technical Support Center (TSC)

A Technical Support Center is located on the southwest corner of the Service Building at Elevation 518'. The TSC fully meets the requirements of Section H.1.b of the E-Plan.

5.1.3 Operational Support Center (OSC)

The Operational Support Center is located in the Radiation Protection Area and the Work Control Area. The OSC conforms to the requirements of Section H.1.c of the Emergency Plan and is the location to which operations support personnel will report during an emergency and from which they will be dispatched for assignments in support of emergency operations.

5.1.4 Joint Information Center (JIC)

The Joint Information Center (JIC) is the facility in which media personnel gather to receive information related to the emergency event. The JIC is located West of Chicago at the Exelon Nuclear Cantera Offices, in Warrenville, IL.

5.2 Assessment Resources 5.2.1 Onsite Meteorological Monitoring Instrumentation The meteorological tower, located approximately 3000 ft. west of the reactor building, is 400 ft. high and is instrumented at three levels. The 35 ft., 150 ft. and 300 ft. levels correspond to the elevations of the possible points of airborne effluent release. Wind speed and wind direction are measured at all three elevations. Ambient temperature is measured at the 35 ft. level and differential temperatures referenced to 35 ft. are measured at 150 ft. and 300 ft. Precipitation is also measured at the site.

The onsite meteorological monitoring program is covered in the contract specification and vendor procedures of the meteorological monitoring contractor. These data are used to generate wind roses and to provide estimates of airborne concentrations of gaseous effluents.

May 2005 DR 5-1 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear 5.2.1.1 Instrumentation The meteorological tower is instrumented with equipment that conforms to the recommendations of Regulatory Guide 1.23 and ANSI/ANS 2.5 (1984). The equipment is placed on booms oriented into the generally prevailing wind at the site. Equipment signals are brought to an instrument building with controlled environmental conditions. The building at the base of the tower houses the analog and digital recording equipment, signal conditioners, and other equipment used to process and re-transmit the data to the end point users.

5.2.1.2 Meteorological Measurement Program During a Disaster Cooperation between the corporate office and the meteorological contractor assures that a timely restoration of any outage can be made. Emergency field visits to the site are made as quickly as possible after detection of a failure.

Should a disaster of sufficient magnitude occur to destroy the tower structure, a contract is maintained to have a temporary tower erected within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, weather conditions permitting. Further, the meteorological contractor maintains two levels of sensors (wind speed, wind direction and temperature) in a state of readiness for use on the temporary tower.

Additionally, Exelon Nuclear's existing instrumentated towers at other nuclear sites provide a measurement network with multiple backup opportunities.

Meteorological data is available to the station Control Room, Technical Support Center, and Emergency Operations Facility for use in the Dose Assessment Computer Model for estimating the environmental impact of unplanned releases of radioactivity from the station.

5.2.2 Onsite Radiation Monitoring Equipment 5.2.2.1 Radiation Monitoring System Onsite radiation monitoring systems at Dresden can be categorized into four systems:

• A process radiological monitoring and sample system;

• An effluent radiological monitoring and sampling system;

• An airborne radioactive monitoring system;

• An area radiation monitoring system; and

• A supply of portable survey and counting equipment.

May 2005 DR 5-2 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear 5.2.2.2 Radiological Noble Gas Effluent Monitoring A wide range monitor is installed in the effluent stream that enters the main chimneys and the reactor building vents. These wide range monitors have a range of 1x10-7 µCi/cc to 1x105 µCi/cc.

The method of converting instrument readings to release rates will be determined after the energy responses of the detector are obtained. Due to system design, the monitors give an estimate of a release. Actual releases will be determined by periodically collecting grab samples, counting the samples collected and calculating the releases.

5.2.2.3 Radioiodine and Particulate Effluent Monitoring Effluent sampling media are analyzed in the Station counting room. Silver based cartridges are available to reduce the interference of noble gases.

5.2.2.4 High-Range Containment Radiation Monitors Two high range containment radiation monitors are installed on each of Dresden's units. The range of these monitors is from 1 R/hr to 108 R/hr.

5.2.2.5 In-plant Iodine Instrumentation Dresden Station has the capability to sample and determine iodine concentrations in the plant using charcoal cartridges and gamma ray spectroscopy. Portable monitors may be used to measure increasing levels of iodine during emergency conditions.

5.2.3 Onsite Process Monitors Adequate monitoring capability exists to properly assess the plant status for the modes of operation. The operability of the post-accident instrumentation ensures information is available on selected plant parameters to monitor and assess important variables following an accident. Instrumentation is available to monitor the parameters and ranges given in the Dresden Station Technical Specifications.

Station procedures have been developed which would aid personnel in recognizing inadequate core cooling using applicable instrumentation.

5.2.4 Onsite Fire Detection Instrumentation Dresden Station has a fire protection system that is designed to quickly detect any fires; annunciating locally and in the Control Room. The fire detection system is designed to applicable National Fire Protection Association (NFPA) standards. The majority of the detectors consist of electrically supervised ionization smoke detectors.

May 2005 DR 5-3 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear 5.2.5 Facilities and Equipment for Offsite Monitoring Consult Chapter 11 of the station specific Offsite Dose Calculation Manual (ODCM) for the most current location for fixed continuous air samplers and TLD.

5.2.6 Site Hydrological Characteristics The hydrological characteristics of Dresden Station are described in Section 2.4 of the Dresden UFSAR. The Dresden site at the confluence of the DesPlaines and Kankakee rivers is at the location considered to divide the upper and lower parts of the Illinois River system. The normal river pool elevation controlled at the adjacent Dresden Island Lock and Dam is nominally 505 feet. In December 1982, the Dresden site was subjected to flood waters that exceeded 509 feet establishing a maximum historical flood elevation.

Nominal ground elevation is about 516 feet at the location of the principal structures of Units 2 and 3, and design plant grade is 517 feet. Consequently the probability of flooding critical areas of the site is remote.

Spillway capacity at the Dresden Island Lock and Dam is well in excess of the estimated maximum instantaneous flow of the Illinois River. The site elevation is well above the vast valley storage area upstream from the dam.

River system flow data applicable to the site are more than adequate to meet the cooling water requirements of the two operating units, to assure the availability of sufficient quantities of water for dilution of all radioactive liquid wastes discharged into the Illinois River within the limits in 10 CFR 20, and to reduce concentrations to approximately one one-thousandth of the maximum permissible concentration in the river below the point of discharge from the station.

The closest point downstream of the station where the Illinois River is used as a source of domestic water is Peoria which is 100 miles downstream. At this point the combined effects of dilution, mixing, radioactive decay, and deposition of radioactivity on the river bottom will have rendered the contribution of radioactivity by the station negligible in relation to that present in the Illinois River from other sources.

5.2.6.1 Probable Maximum Flood on Streams and Rivers Since the site probable maximum flood (PMF) elevation of 528'-0" is above the plant grade (elevation 517'0") and above the lowest opening leading to safety-related equipment (elevation 509'-0"), the safe operation of the plant during the PMF is accomplished via implementation of the flood emergency procedures.

May 2005 DR 5-4 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear 5.2.6.2 Potential Dike and Dam Failures, Seismically Induced An earth dam of the type specified usually does not collapse in its entirety. A break occurs and widens as the water washes through the break. This tends to prolong the time it would take to empty the lake; nevertheless, instantaneous dike losses have been considered since the dikes are not constructed to Class I criteria. The Dresden lock and dam are concrete structures that are operated and maintained by the U.S. Army Corps of Engineers. Operations response procedures are in place to deal with loss of the cooling lake and/or the lock and dam.

5.2.6.3 Ice Effects An 8-foot diameter deicing line connects the discharge canal headworks and the crib house forebay. Its high point is in the headworks at elevation 495'-0" and its low point is in the forebay at elevation 489'-0". A slide gate valve is used to isolate the deicing line when not in use.

5.2.6.4 Cooling Lake The purpose of the cooling lake is to provide adequate cooling of the circulating and service water before discharge to the Illinois River. The water discharged to the river must meet state requirements. The lake is connected to the intake and discharge canals for Units 2 and 3 by two canals (the "hot" and "cold" canals).

The level of the lake is maintained by a concrete spillway located adjacent to the lift station and between the cold canal and the north end of the lake. The spillway is equipped with weir gates that can be lowered to block some of the spillover to maintain the level of the lake. The weir gates can be manually operated if necessary. Operations response procedures are in place regarding loss of lake level control and/or loss of the lift station.

A discussion of the groundwater resources and aquifers in the vicinity of Dresden Station is discussed in the Final Environmental Statement.

5.3 Protective Facilities and Equipment The onsite assembly areas for Dresden Station are described in Section 4 of this annex.

The offsite evacuation assembly areas for Dresden Station are discussed in Section 4 of this annex. These areas are outside the plume exposure Emergency Planning Zone and equipped with monitoring, decontamination and bioassay capabilities.

5.4 First Aid and Medical Facilities Dresden Station has an inplant decon room located in a room adjacent to the Radiation Protection Office. This room is provided with a sink, decontamination shower, and a supply cabinet.

First aid kits, stretchers, sinks, eyewashes, and emergency showers have been placed in strategic locations throughout the station.

May 2005 DR 5-5 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Medical treatment given to injured persons at the station is of a "first aid" nature. When more professional care is needed, injured persons are transported to a local hospital or clinic. Morris Hospital in Morris, Illinois is the Dresden Station primary supporting medical facility for radioactively contaminated injured persons. Provena St. Joseph Medical Center in Joliet, Illinois is the backup medical facility for evaluation and treatment of persons suffering from traumatic injury, medical illness, or radiation exposure and uptake.

May 2005 DR 5-6 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Appendix 1: NUREG-0654 Cross-Reference Annex Section NUREG-0654 1.0 Part I, Section A 1.1 Part I, Section C 1.2 Part I, Section D Figure 1-1 Part I, Section D 2.0 Part II, Section A.4 2.1 Part II, Section A.3 3.0 Part II, Section D 4.1 Part II, Section E.1 & J.7 4.2 Part II,Section I.2 & 3 4.3 Part II, Section J.10.m 4.3.1 Part II, Section E.6 4.3.2 Part II, Section J.8 4.4 Part II, Section J.1-5 Figure 4-1 Part II, Section J.10.m Figure 4-2 Part II, Section J.5 Figure 4-3 4-4 Part II, Section J.2 & 3 Table 4-1 Part II, Section J.8 Table 4-2 Part II, Section J.10.b 5.1 Part II, Section H.1 & G.3 5.2.1 Part II, Section H.5.a & 8 5.2.2 Part II, Section H.5.b & I.2 5.2.3 Part II, Section H.5.c 5.2.4 Part II, Section H.5.d 5.2.5 Part II, Section H.6.b & 7 5.2.6 Part II, Section H.5.a & 6.a 5.3 Part II, Section J.1-5 5.4 Part II, Section L.1 & 2 May 2005 Appendix 1-1 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Appendix 2: Station Letters of Agreement

1. US Army Corp of Engineers - provide information regarding failure of or problems with the Dresden Lock and Dam.

2. Will County Sheriff's Office - provides services of law enforcement.

3. Grundy County Sheriff - services of law enforcement.

4. Morris Hospital of Morris, Illinois, acts as the primary supporting medical facility for Dresden Station.

5. General Electric Midwest Fuel Reprocessing Plant - Health Physics support -

instrumentation and limited technical assistance.

6. Morris Fire Protection and Ambulance District - Fire protection.

7. Coal City Fire Protection and District - Fire protection and advanced life support for transportation of accident victims.

May 2005 Appendix 2-1 EP-AA-1004 (Revision 19)

Dresden Annex Exelon Nuclear Appendix 3: EAL Wallboard May 2005 Appendix 3-1 EP-AA-1004 (Revision 19)