ML20113F755

Revised DAEC Emergency Action (EAL) Technical Basis Document

ML20113F755
Person / Time
Site:	Duane Arnold
Issue date:	09/19/1996
From:	Dean Curtland, Hite R, Tirella C IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:
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ML20113F748	List: ML20113F744 ML20113F755
References
NUDOCS 9609250314
Download: ML20113F755 (144)

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l DUANE ARNOLD ENERGY CENTER 1 O

EMERGENCY ACTION LEVEL (EAL)

TECHNICAL BASIS DOCUMENT Revision 1 (ForNRCRevieu)

September 1996 l 4

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

PAGE 1 of 1 APPROVAL SHEET EFFECTIVE DATE: TBD l

l Prepared by: John S. Fuoto, Ogden Environmental and Energy Services

, Reviewed by: / -

Date: f-/6 -% i j Charles P. Tirella, Jr., Prineple Er@cy Planning Specialist l l

l Reviewed by: E Date: 9'#~ b Dean Curtland, Operations Manager l

Reviewed by: 5 /M . // Date: 9'/9 ~9(a Robert Hite, Radiological Protection Manager Approved by: M / Date: 2//9/p/

Kenneth Peveler, Manager, Eme'rgency Planning & Licensing 1

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' Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. I (forNRCreview)

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l TABLE OF CONTENTS EFFECTIVE DATE: TBD i

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TABLE OF CONTENTS l ,

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j INTR O D U CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I- 1 DEFINITIONS....................................................................................................................................D-1 ORGANIZATION OF B ASIS INFORMATION..... .. .. . ... ... . .... .... ........ ............... ..................... O-1 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT CATEGORY  :

, AUI Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment That Exceeds Two Times the .

' _ g) Radiological Technical Specifications For 60 Minutes or Longer - A-1  !

AU2 Unexpected increase in Plant Radiatiot . . . ..A-5 i AAl Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times ,

Radiological Technical Specifications for 15 Minutes or Longer.. -A-8 j l

AA2 Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering ofIrradiated Fuel Outside the Reactor Vessel-- . . . . . . . . . . ..A-12 AA3 Release of Radioactive Material or increases in Radiation Levels Within the Facility That Impedes Operation of i Systems Required to Maintain Safe Operations or to Establish or to Maintain Cold Shutdown. . .. A-15  !

ASI Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mrem TEDE or 500 mrem CDE Thyroid for the Actual or Projected Duration of the Release.. . .. A- 17  ;

AGI Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1,000 mrem TEDE or 5,000 mrem CDE Thyroid for the Actual or Projected Duration of the Release.. . A-21 FISSION PRODUCT BARRIER DEGRADATION CATEGORY 1 i FU1 Any Loss or Any Potential Loss of Primary Containment Barrier. .. . F-1 FA1 Any Loss or Any Potential Loss of Either Fuel Clad Or RCS Barrier.. .. .F-2 FSI Loss Or Potential Loss of Any Two Barriers.. . . F-3 FG1 Loss of Any Two Barriers AND Potential Loss of the Third Barrier. .. F-5

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

A N) PAGE ii of iu TABLE OF CONTENTS EFFECTIVE DATE: TBD FISSION PRODUCT BARRIER DEGRADATION CATEGORY (continued)

FUEL CLAD BARRIER INDICATORS.. =F-6 RCS BARRIER INDICATORS -- ..F-12 PRIMARY CONTAINMENT BARRIER INDICATORS . F-21 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORY HUI Natural and Destructive Phenomena Affecting the Protected Area . . ..H- 1 HU2 Fire Within Safe Shutdown Areas Not Extinguished Within 15 Minutes of Detection . ..H-4 HU3 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant.. ..H-5

) HU4 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant.. ..H-6 HUS Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Declaration of an Unusual Event . ..H-7 HAl Natural and Destructive Phenomena Affecting the Plant Vital Area. . ..H-8  !

HA2 Fire AITecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown . ..H-12 HA3 Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown.. . . H-14 HA4 Security Event in a Plant Protected Area.. . ..H-16 HA5 Control Room Evacuation Has Been Initiat.rd .. . . ..H-17 HA6 Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Declaration of an Alert = ..H-18 HS! Security Event in a Plant Vital Area : . ..H-19 I

l HS2 Control Room Evacuation Has Been initiated and Plant Control Cannot Be Established.. ..H-20 HS3 Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Declaration of a Site Area Emergency H-22 HGI Security Event Resulting in Loss Of Ability to Reach and Maintain Cold Shutdown.. . ..H-23 p HG2 Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Declaration of a General Emergency H-24

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCrevicu) 1 (,h

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I l SYSTEM MALFUNCTION CATEGORY l

SUI Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes - . S-1 '

l SU2 Inability to Reach Required Shutdown Within Technical Specification Limits. .S-2 SU3 Unplanned Loss of All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes

. . . .S-3 SU4 Fuel Clad Degradation. .  ; S-5 SUS RCS Leakage.. . .S-8 SU6 Unplanned Loss of All Onsite or Offsite Communications Capabilities . . . .S-10 SU7 Unplanned Loss of Required DC Povtr During Cold Shutdown or Refuel Mode For Greater Than 15 Minutes... S-12 sal Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Conditions.. .S-14 1

SA2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a l Reactor Protection System Setpoint lias Been Exceeded and Manual Scram Was Successful. .S-IS  ;

SA3 Inability to Maintain Plant in Cold Shutdown . . .S-17 SA4 Unplanned Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a j Significant Transient in Progress, or (2) Compensatory Non-Alarming Indicators are Unavailable.. .S-19 SAS AC Power Capability to Essential Busses Reduced to a Single Power Source for Greater Than 15 Minutes Such That l Any Additional Single Failure Would Result in Station Blackout.- .S-21 SSI Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses. .S-22

SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate on Automatic Reactor Scram Once a

.S-23 Reactor Protection System Setpoint IIas Been Exceeded and Manual Scram Was NOT Successful..

i SS3 Loss of All Vital DC Power.. . .S-24 l

l SS4 Complete Loss of Function Needed to Achieve or Maintain llot Shutdown . . . .S-26 SS5 Loss of Water Level in the Reactor Vessel That lias or Will Uncover Fuel in the Reactor Vessel . .S-28

/~' SS6 Inability to Monitor a Significant Transient in Progress.. . .S-29 V

SG1 Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power.. . .S-31 l

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l- SYSTEM MALFUNCTION CATEGORY (continued) ]

l SG2 Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core . . .. S-33  !

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r l l Duane Arnold Energy Cente-EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) I lO PAGEI-1 of 2 I

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INTRODUCTION IES Utilities has revised the Duane Arnold Energy Center (DAEC) Emergency Plan to incorporate guidance from NUMARC/NESP-007, Revision 2 (January '1992), Methodology for Development of Emergency Action Levels. The NUMARC (now Nuclear Energy Institute or NEI) guidance was developed l to replace Emergency Action Levels (EAL) guidance contained in NUREG-0654/ FEMA-REP-1 (Revision 1), Criteriafor Preparation and Evaluation ofRadiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, November 1980. The NEI-sponsored methodology was used to .

.i develop a set of generic EAL guidelines, together with the basis, so that they could be used and adapted by p each utility in a consistent manner. The NRC has endorsed use of the NEI generic guidance as an I acceptable alternative method to NUREG-0654 for developing plant-specific EALs in Regulatory Guide  :

1.101, " Emergency Planning and Preparedness for Nuclear Power Reactors," Revision 3, August 1992. .

l p This Regulatory Guide further states that: " Licensees may use either NUREG-0654/ FEMA-REP-1 or l l v NUMARC/NESP-007 in developing their EAL scheme but may not use portions of both methodologies." )

1 This EAL basis document was developed to: (1) provide clear documentation of how NEI generic guidance was applied in the development of DAEC upgraded EALs, (2) provide justification of any exceptions or j additions to NEI generic guidance as it is applied to DAEC, and (3) facilitate the regulatory approval of the )

upgraded EALs that is required under 10 CFR 50 Appendix E.  !

i Although there are many similarities, there are some basic differences from the previous EALs based on NUREG-0654 guidance. These include:

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1. Events that are explicitly covered under 10 CFR 50.72 such as one-hour or four-hour reports are no longer classified under the Unusual Event emergency classification. Items such as contaminated injured person transported off-site, partial communications losses, meteorological measurement losses, shutdown within the requirements of technical specifications, and inadvertent actuation of ECCS are no longer treated as emergencies because they are explic.dy defined in 10 CFR 50.72 as "non-emergency" conditions to report.

, 2. Precursor conditions are explicitly included in the Unusual Event emergency classification. This j includes EALs addressing RCS leakage and loss of off-site power.

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3. Conditions such as fire, explosion, gas releases, flooding, low river water level, tornado, or earthquake -

l can be directly escalated only up to the Alert emergency classification. Escalation to Site Area l Emergency or General Emergency is based on degraded system response as would be determined by j ~ fission product barrier, loss of AC power, or projected effluent release EALs.

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4. Core damage sequences are addressed by determining their level of challenge to each of the three primary fission product barriers - fuel clad, reactor coolant system, and the primary containment. The level of challenge is determined in accordance with the Emergency Operating Procedures (EOPs),

l Integrated Plant Operating Instructions (IPOls), Abnormal Operating Procedures (AOPs) and core 1-damage assessment methodology. This allows the operations crew to readily_ recognize the corresponding emergency classification and allows for ready escalation to Site Area Emergency or General Emergency as conditions may worsen.

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5. Offsite radiological releases that can be expected to exceed Environmental Protection Agency (EPA)

Protective Action Guide (PAG) levels for inhalation doses - 1,000 mrem TEDE or 5,000 mrem CDE Thyroid - will result in declaration of a General Emergency.

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

O PAGE D-1 of 10 DEFINITIONS . EFFECTIVE DATE: TBD DEFINITIONS AC- Altemating Current Afecting (in regard to events such as fire, flood, or missiles) - Causing degraded equipment performance as determined by physical observation or by indications in the Control Room or at local control stations.

Alert - Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Prot etive Action Guide (PAG) exposure levels.

All - Initiating Condition applies to all Technical Specification operating modes as well as defueled

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AOP- Abnormal Operating Procedure i

i APRM- Average Power Range Monitor

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j ARM- Area Radiation Monitor j ATWS- Anticipated Transient Without Scram i

. Barrier - Same as " Fission Product Barrier", below, i

Barrier Monitoring Ability - This is a judgment factor in determining whether a fission product barrier is lost or potentially lost. Decreased ability to monitor a barrier results from a loss of/ lack of reliable indicators, including instrumentation operability concerns, readings from portable instrumentation, and

- consideration for offsite monitoring results.

Becquerel- A measurement 'of radioactive decay rate equal to one disintegration per second.

BOP- Balance of Plant BWR - Boiling Water Reactor

. CAM- Continuous Air Monitor CDE - Committed Dose Equivalent as defined in 10 CFR 20.1003

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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

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CEDE- Committed Effective Dose Equivalent as defined in 10 CFR 20.1003 l

' CFM- Cubic Feet per Minute CFS- Cubic Feet per Second

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Cold condition - As defined in Technical Specification 1.0, this refers to the condition where the reactor coolant temperature is less than or equal to 212 F.

j Coldshutdown - As defined in Technical Specification 1.0, the reactor is in the shutdown mode, the reactor l coolant temperature is less than or equal to 212'F, and the reactor is vented to atmosphere.

Compensatory non-alarming indications - Information displayed in the main control room including analog and digital parameter displays, trend recorders, the Safety Parameter Display System (SPDS), and the plant process computer.

Control- As applied to remote shutdown capability, this is the ability to manipulate plant parameters l without reliance on control room devices or instnunentation using components and methods specified by l Abnormal Operating Procedure 915, Shutdown Outside Control Room.

CPS- Counts Per Second CRD- Control Rod Drive 1 l

CSCS- Core Standby Cooling Sywem

! CST- Condensate Storage System Curie (Cl) - A measurement of radioactive oecay rate equal to 3.70E+10 disintegration's per second

' (becquerels).'

[ ' CW- Circulating Water

. DAEC- Duane Amold Energy Center p-DC- Direct Current l

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Duane Arnold Energy Center- l EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1. (forNRC review) l O PAGE D-3 of 10 l

DEFINITIONS EFFECME DAEMD I

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l DEG- Dose Equivalent l l

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- Dominant accident sequences - These will lead to degradation of all fission product barriers. Dominant l accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat ' l removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite l' l power with early HPC1/RCIC failure.

i DW- Drywell  !

i EC- Emergency Coordinator i ECCS- Emergency Core Cooling System EDE- Effective Dose Equivalent as defined in 10 CFR 20.1003 Emergency Action Level (EAL) - A pre-determined, site-specific, observable threshold for a plant Initiating Condition that places the plant in a given Emergency Class. An EAL can be: an instrument reading, an equipment status indicator, a measurable parameter (on-site or offsite), a discrete observable event, results l of analyses, entry into specific emergency operating procedures, or another phenomenon which, if it occurs, indicates entry into a particular Emergency Class.

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Emergency Class - Same as " Emergency Classification Level" below. ,

Emergency Classification Level- These are taken from 10 CFR 50, Appendix E. They are, in escaleung order: (Notification of) Unusual Event (UE), Alert, Site Area Emergency (SAE), and General Emergency l

(GE).

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EOP- Emergency Operating Procedure  !

EPA - Environmental Protection Agency l EPIP-Emergency PlanImplementing Procedure ESF- Engineered Safety Features l l ' ESS- Engineered Safety Systems Establish - Make arrangements for a stated condition, e.g., establish communications with control room.

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PAGE D-4 of 10 DEFINITIONS EFFECTIVE DATE: TBD l

ESW- Emergency Service Water l

l Fission Product Barrier - One of the three principal barriers to uncontrolled release of radionuclides: Fuel l

Clad, Reactor Coolant System (RCS), and the Primary Containment.

i FP - Fuel Pool l

Fuel Clad (Barrier) - The zirconium alloy tubes that contain the fuel pellets.

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General Emergency (GE) .- Events are in process or have occured which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases can reasonably be expected to exceed EPA Protective Action Guide (PAG) exposure levels offsite for more than the immediate site area.

l GPM- Gallons Per Minute GSW- General Service Water i .

l Hot shutdown - As defined in Technical Specification 1.0, the reactor is in shutdown mode and the reactor

! coolant temperature is greater than 212 F.

Hot standby condition - As defined in Technical Specification 1.0, this refers to operation with the reactor coolant temperature greater than 212 F, reactor pressure vessel less than 1055 psig, and the mode switch in

!. Startup position.

HPCI- High Pressure Coolant Injection (system).

Identi/ led Leakage - Identified Leakage shall be:

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or F

i b. Leakage into the containment atmosphere from sources that are both specifically

, located and known not to interfere with the operation of the leakage detection systems.

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t IDLH-Immediately Dangerous to Life and Health

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l l Inadvertent - Accidental or unintentional, e.g., the event occurred because procedures were not strictly I

adhered to.

' Imminent - No tumaround in safety system performance is expected and escalation to a higher emergency classiScation level is expected to occur within two hours.

Implement - Commence a required program or series of procedures.

l l In service - A component or system in the appropriate configuration for normal operation and is considered

operable as defined in the Technical Specifications.

l Indicator - The name for the row on the fission barrier table that is used for convenient grouping of similar symptoms.

Initiate - Take action to begin a process Initiating Condition (IC) - One of a predetermined subset of nuclear power plant conditions where either the potential exists for a radiological emergency or such an emergency has occurred.

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! IPE-Individual Plant Examination >

IPOI- Integrated Plant Operating Instruction

IRM-Intermediate Range Monitor l- Isolate - Remove from service by closing off the flow path l

l kV- Kilovolt (s)

LCO - Limiting Condition for Operation

' LLRPSF- Low Level Radwaste Processing and Storage Facility i

LOCA - Loss ofCoolant Accident LOOP- Loss of OfTsite Power l

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l Loss (of a fission product barrier) - A severe challenge to a fission product barrier exists such that the !

l barrier is considered incapable of performing its safety function.

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- LPCI- Low Pressure Coolant Injection l AICC- Motor Control Center L AICUTL - Maximum Core Uncovery Time Limit I l

l Aficrocurie (pCi) - One millionth of a curie, i.e., 3.7E+4 disintegration's per second (becquerels).

l AfIDAS - Meteorological Information and Dose Assessment System, primary method for detecting and quantifying gaseous releases at the DAEC. I Afillicurie (mci) - One thousandth of a curie, i.e.,3.7E+7 disintegration's per second (becquerels).

3 Afillirem (mrem) - One thousandth of a rem  !

l AIPH- Miles Per Hour mR - milliroentgen, i.e., one thousandth of a roentgen (R) j AISIV- Main Steam Isolation Valve t

- AISL - Main Steam Line NEI- Nuclear Energy Institute (formerly NUMARC) l l Notification of Unusual Event (NOUE) - Same as " Unusual Event", below.

NPSH-Net Positive Suction Head j - NUAMRC - Nuclear Utility Management and Resources Council (now NEI) 1 Dq OBE- Operating Basis Earthquake i ODAAf- Offsite Dose Assessment Manual

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

PAGE D-7 of 10 DEFINITIONS EFFECTIVE DATE: TBD l

l Operating Mode - As defined by Technical Specification Table.1.0-1, Operating Mode describes the I operating status of the unit. Mode designations (and the associated Reactor Mode Switch Positions) used l at DAEC are: RUN/ POWER OPERATION (Run), STARTUP (Startup or Refuel), HOT SHUTDOWN  !

.(Shutdown), COLD SHUTDOWN (Shutdown), and REFUELING (Shutdown or Refuel).  !

' Operable - A system is considered capable of performing its function in accordance with the applicable Technical Specification requirements. Implicit in this definition is the assumption that all auxiliary equipment required for the system is also operable.

OSS- Operations Shift Supervisor PAG- Protective Action Guide l Planned - Loss of a component or system due to expected events such as scheduled maintenance and l

~ testing activities. l Potential Loss (of a fission product barrier) - A challenge to a fission product barrier exists such that the -

! barner is considered degraded in its ability to perform its safety function. .;

L L . Primary Containment (Barrier) - The drywell, the torus, their respective interconnecting paths, and other i connections up to and including the outermost containment isolation valves, t .

PSIG- Pounds per Square Inch Gauge I RB- Reactor Building f RBCCW- Reactor Building Closed Cooling Water (system) l' '

RCIC - Reactor Core Isolation Cooling (system)

RCS- Reactor Coolant System  ;

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RCS Barrier - The reactor coolant system pressure boundary including the reactor pressure vessel and all j reactor coolant system piping up to and including the outermost isolation valves.

Recognition Category - A logical grouping ofInitiating Conditions, e.g., System Malfunctions, i

L r Rem - Unit of radiation dose as defined in 10 CFR 20.1004 l '

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DEFINITIONS EFFECTIVE DATE: TBD r

l l l Required- Action taken (such as entry into emergency operating procedure) is neither optional nor merely L suggested; rather, it is imperative based on existing conditions. ,

t RHR - Residual Heat Removal (system)  !

RHRSW- Residual Heat Removal Service Water (system)  !

l Roentgen (R) - Unit ofionizing radiation energy absorbed in a cubic centimeter of air RPV- Reactor Pressure Vessel i

RFCU- Reactor Water Clean-Up (system)  :

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. U SBDG- Standby Diesel Generator.

SBGT- Standby Gas Treatment (system) l SBLC - Standby Liquid Control (system)  !

SBO- Station Blackout i i

i S/D- Shutdown SDC- Shutdown Cooling i

SDV- Scram Discharge Volume Shutdown - As def'med in Technical Specification 1.0, the reactor is in a shutdown condition when the reactor mode switch is in the Shutdown position. .

l Sigmycant transient - (See also, " Transient", below.) Includes response to automatic or manually initiated i functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or ':

j' thermal power. oscillations of 10% or greater. i Site Area Emergency (SAE) - Events are in process or have occurred which involve actual or likely major i

. failures of plant functions needed for protection of the public. Any releases are not expected to result in

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G' PAGE D-9 of 10 DEFINITIONS EFFECTIVE DATE: TBD l

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( exposure levels which exceed EPA Protective Action Guide (PAG) exposure levels except near the site I boundary.

i 1 SPDS- Safety Parameter Display System SRM- Startup Range Monitor SRO - Senior Reactor Operator i i

SRV- Safety-Relief Valve Sustained windspeed- Baseline wind speed measured by meteorological tower that does not include gusts l l

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TAF- Top of Active Fuel (344.5 inches above bottom of RPV) I TEDE - Total EfTective Dose Equivalent as defined in 10 CFR 20.1003 Total Leakage - Total leakage is the sum ofIdentified Leakage and Unidentified Leakage.

Transient - A condition that: (1) is beyond the expected steady-state fluctuations in temperature, pressure, power level, or water level, and (2) is beyond the nomial manipulations of the Control Room operating crew, and (3) is expected to require actuation of fast-acting automatic control or protection systems to bring the reactor to a new safe, steady-state condition.

TSC- Technical Support Center Uncontrolled - Condition is not the result of planned actions by the plant staff in accordance with procedures.

Unisolable - Actions taken from the Main Control Board or locally are not successful in eliminating the leakage path.

Unidentified Leakage - Unidentified Leakage shall be all leakage which is not identified leakage.

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i Unplanned - Used to preclude the declaration of an emergency where a component or system has been  !

removed intentionally from service (e.g., for maintenance and/or testing activities). As used in the context of radioactive releases, " unplanned" includes any release for which a radioactive discharge permit was not  :

prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, j alarm setpoints, etc.) on the applicable permit. j Unusual Evem (UE) - Events are in process or have occurred which indicate a potential degradation of the l level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are l expected unless further degradation of safety systems occurs. '

l VAC- Volt (s) Altemating Current l

VDC- Volt (s) Direct Current l Valid - Indication is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the  ;

control panels, reports from plant personnel, or radiological survey results.

WEC- Water Effluent Concentration l

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PAGE O-1 of 3 ORGANIZATION OF BASIS INFORMATION EFFECTIVE DATE: TBD 4

d ORGANIZATION OF BASIS INFORMATION  ;

The format of the EAL Basis information was developed to address training needs, to facilitate NRC approval, and to facilitate future revisions and 10 CFR 50.54(q) evaluations. Each EAL Basis is organized in the following manner:

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1. Emergency Action Level (EAL) Basis Information Organized by Initiating Condition (IC) 4 Initiating Condition Identifier For consistency, DAEC has chosen to make its Initiating Condition (IC) identifiers identical to those used in NEI document NUMARC/NESP-007. The EAL Technical Basis information is organized by generic IC

/ NUMARC/NESP-007 organized the generic information into four V] identifier number and name.

Recognition Categories. These are:

A - Abnormal Rad Levels / Radiological Effluent F - Fission Product Barrier Degradation H - Hazards and Other Conditions Affecting Plant Safety S - System Malfunctions For the A, H, and S recognition categories, all EAL basis information is organized by IC identifier in escalating emergeacy class order from Unusual Event through General Emergency. For the F recognition category, the initiating conditions are the combinations of fission product barrier losses and potential losses that correspond to each emergency classification level. The individual indicators used on the fission barrier table are separately discussed below. The generic IC identifiers use two letters followed by one number.

The first letter corresponds to the event category as shown above. The second letter corresponds to the emergency classification level for the IC:

U -(Notification of) Unusual Event A - Alert S -Site Area Emergency G - General Emergency O The number designates whether the IC is the first, second, third, etc., IC for that recognition category under

( that emergency classification. For example, SU2 is the designator for the second System Malfunction recognition category IC in the Unusual Event classification, etc. Generic information is quoted directly from NUMARC/NESP-007, Revision 2, dated January,1992. Changes from the NUMARC/NESP-007

Duane Arnold Energy Center

. EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

PAGE O-2 of 3 ORGANIZATION OF BASIS INFORMATION EFFECTIVE DATE: TBD text are denoted by caret marks (< >). Such changes are based on correction of typographical errors such as those mentioned in the NUMARC Questions and Answers dated June 1993, reflect changes made in 10 CFR Part 20, or to put the information in proper context.  :

Event Type '

This is the label of the applicable row for the EAL Table shown in EPIP-1.1, Determination ofEmergency Action Levels. The event type lists the general area of concem and includes Offsite Rad Conditions, Onsite Rad Conditions, Natural Disasters, Fire, Other Hazards and Failures, Security, Control Room Evacuation, EC/OSS Judgment, Loss of Power, RPS Failure, Inability to Maintain Shutdown Conditions, Instrumentation / Communication, Coolant Activity, and Coolant Leak. This structure was chosen to be consistent with the previous EAL presentation which is already familiar to the Emergency Coordinators and Operations Shift Supervisors. It is also permissible to organize the generic information in this manner

(] based on the response to Question 5 contained in the NUAMRC Afethodologyfor Development of G Emergency Action Levels NUAfARC/NESP-007 Revision 2 Questions and Answers June 1993.

Apolicable Onerating Modes The applicable operating modes for each Initiating Condition / Emergency Action Level is then listed based on NUMARC/NESP-007 mode descriptions. The DAEC EALs use the operating modes defined in  !

Technical Specifications Table 1.0-1. These are:

1 - Run/ Power Operation 4 - Cold Shutdown 2 - Startup 5 - Refueling 3 - Hot Shutdown To conserve space, the EAL displays use "Run" to mean "Run/ Power Operation" and "S/D" as an abbreviation for" Shutdown."

l Operating mode applicability of EALs is based on the operating mode that the plant was in immediately l before the event sequence leading to entry into the emergency classification. For example, 1 events / conditions addressed by EALs applicable to Run mode are expected to lead to reactor trip which should bring the plant to Hot Shutdown. However, the appropriate emergency classification would still be .

based on the applicable EALs for Run/ Power Operation for these events / conditions. If"ALL" operating i

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modes are specifiedfor the EAL, then the EAL applies to all modes identified above plus defueled conditions.

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Duane Amold Energy Center l EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCrevien) i PAGE O-3 of 3 ORGANIZATION OF BASIS INFORMATION EFFECTIVE DATE: TBD l

Generic Examnle EAL(s)

The generic example EALs are then listed. When more than one is provided, logic phrasing is used to describe whether all EALs are suggested or whether at least one EAL should be chosen.

DAEC EAL Information This contains the plant-specific information used to implement the generic EALs. This section will also include the basis, as appropriate, for deviation from generic EALs. For example, DAEC does not have a Independent Spent Fuel Storage Installation for on-site dry storage of spent fuel. Thus, DAEC does not  !

have EALs corresponding to the generic guidance for this item. As appropriate, description of any supporting calculations, their underlying bases and assumptions, and their results are included in this  ;

section.  :

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(/ References The references used to develop the DAEC EAL Information are listed here, as appropriate.

2. Fission Product Barrier Table Indicators The basis information for the fission barrier table indicators is organized similarly to the other basis information described above. For each barrier - fuel clad, RCS, and primary containment - basis i

. information is organized by " Indicator." The indicator is the name for the row on the fission barrier table

[ and is used for convenient grouping of similar symptoms, similar to the " Event Type" used for the A, H, and S EALs described above. Indicators include Radiation / Core Damage, RPV Level, Leakage, Primary Contairunent Atmosphere, and EC/OSS Judgment.

After the DAEC Indicator, the applicable generic BWR fission product barrier indicators are then displayed, showing both the generic loss and potential loss conditions, as applicable. Next displayed is the appropriate DAEC information and references. These are displayed in the same manner as the A, H, and S recognition category basis information described above.

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I I L i EMERGENCY PLAN IMPLEMENTING PROCEDURE No. EPIP - 1.1 Rev.I

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ABNORMAL RAD LEVEIRRADIOACTIVE EFFLUENT EFFECTIVE DATE:TBD l

l l EVENT TYPE UNUSUAL EVENT ALERT StTE AREA EMERGENCY GENERAL EMERGENCY AU1 AA1 AS1 AGf Valid Reactor Buelding or Turbine Building Valid Reactor Building or Turbine Building Valid Reactor Building or Turbine Building Valid Reactor Building or Turbine Building ventilation (Kaman) rad monitor reading ventilation (Kaman) rad monitor reading ventilation (Kaman) rad monitor reading ventilation (Kaman) rad monitor reading above 1 E-3 pCVcc for more than 60 above 3 E-2 pCVcc for more than 15 minutes, above 6 E-2 pCVcc for more than 15 above 6 E-1 pCi/cc for more than 15 minutes. OR minutes. (Dose assessment not available) minutes. (Dose assessment not available)

OR Valid Offgas Stack (Kaman) rad monitor OR OR VaM Oftgas Stack (Kaman) rad monitor reading above 2 E+1 pCVcc for more than 15 Valid Offgas Stack (Kaman) rad monitor Valid Offgas Stack (Kaman) rad morutor reading above 6 E-1 pCVcc for more than minutes, reading above 4 E+1 pCVcc for more than reading above 4 E+2 pCVcc for more than 60 minutes. OR 15 minutes. (Dose assessment not 15 minutes. (Dose assessment not OR Valid LLRPSF (Kaman) rad monitor reading available) available)

Valid LLRPSF (Kaman) rad monitor reading above 9 E-2 pCVcc for more than above 9 E-4 pCVcc for more than 15 minutes.

OFFSITE RAD 60 minutes. OR CONDITIONS OR Valid GSW rad monitor reading above 3E+5 Valid GSW rad monitor readog above 3E+3 CPS for more than 15 minutes.

CPS for more than 60 minutes. OR OR Vand RHRSW & ESW rad morator reading Valid RHRSW & ESW rad monitor reading above 8E+4 CPS for more than 15 minutes, above 8E+2 CPS for more than 60 minutes. OR OR Valid RHRSW & ESW Discharge Canal rad Valid RHRSW & ESW Discharge Canal rad monitor reading above 1E+5 CPS for more monitor reading above 1E+3 CPS for more than 15 minutes.

than 60 minutes. OR OR Confirmed sample analyses for liquid releases Cordirmed sample analyses for liquid releases indicates concentrations in excess of 200 times indicates concentrations in excess of 2 times ODAM limits for greater than 15 minutes.

ODAM limits for greater than 60 minutes. OR Valid field survey reading outside the site boundary > 10 mR/hr or > 50 mR/hr CDE OR OR Thyroid. Valid field survey reading outside the site Valid field survey reading outside the site boundary > 100 mR/hr or above 500 mR/hr boundary > 1.000 mR/hr or > 5.000mR/hr OR CDE Thyroid. CDE Thyroid.

Dose assessment determines hourty dose OR outside the site boundary above 10 mrem OR OR Dose assessment determines hourly dose TEDE. Dose assessment determines integrated Dose assessment determines integrated outside the site boundaiy above 0.1 mrem accident dose projection outside the site accident dose projection outside the site TEDE. Op. Modes: ALL boundary above 100 mrem TEDE or above boundary above 1,000 mrem TEDE or 500 mrem CDE Thyroid. above 5.000 mrem CDE Thyroid.

Op. Modes: ALL Op.Modos: ALL Op.Modos: ALL AU2 AA2 Uncontrolled loss of reactor cavity or fuel Report of ANY of the following:

pool water level with all spent fuel . Valid ARM HI RAD elarm for the assemblies remaining water covered as Refueling Floor North End, Refueling indicated by ANY of the following: Floor South End, New Fuel Storage

. Report to control room Area, or Spent Fuel Storage Area

. Valid fuel pool level indication (LI-3413) = Valid Refueling Floor North End, below 36 feet and lowering Refuehng Floor South End, or New Fuel

. Valid WR GEMAC Floodup indication Storage Area ARM Reading above 10 (LI-4541) coming on scale. mlVhr

. Vaiid Spent Fuel Storage Area ARM Reading above 100 mR/hr ONSITE RAD OR CONDITIONS ..

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t s Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

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l ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY l

l AU1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment That Exceeds Two Times the Radiological Technical Specifications For 60 Minutes or Longer EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4)

1. A valid reading on < site-specific > monitors that < > indicates that the release may have exceeded <2 x l p site-specific technical specifications for 60 minutes or longer >

V 2. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates with a release duration of 60 minutes or longer in excess of two times (site-specific technical specifications).

3. Valid reading on perimeter radiation monitoring system greater than 0.10 mR/hr above normal l background for 60 minutes [for sites having telemetered perimeter monitors]. i

4. Valid indication on automatic real-time dose assessment capability greater than (site-specific value) for 60 minutes or longer (for sites having such capability]. ,

l DAEC EAL INFORMATION: l Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.

The primary methodfor declaration is by means ofdose assessment using the MIDAS computer model.

This is listed as DAEC EAL 4. However, ifthe monitor readings are sustainedfor longer than 60 minutes and the required dose assessments cannot be completed within this period, then the declaration must be l made based on the validreading.

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The approach taken for calculation of gaseous radioactive effluent EAL setpoints includes use of the ODAM Table 3-2 source tenn computed by BWR-GALE for the DAEC Base Case. The release is g assumed to be from a single release point. Multiple release points would be difficult to present as explicit

() EAL threshold values and in any case, are addressed by off-site dose assessment by MIDAS, which is the preferred method for determining this condition. The calculation methods for setpoint determination are from ODAM Section 3.4 and are based on Regulatory Guide 1.109 methodology. The table below lists the  !

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l Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) 1 PAGE A-2 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY  !

results of the gaseous effluent EAL calculations. The Kaman extended range capability is used because the General Electric OfTgas Stack monitor has a limited range, l l

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Gaseous Efiluent EALs Offgas Stack Kaman 9/10 Turbine Bldg (Kaman 1/2) and l Reactor Bldg (Kaman 3/4,5/6,7/8) l Maximum flow (CFM) 10,000 72,000 Release Limits Concentration Release Rate Concentration Release Rate

( Ci/cc) (pCi/sec) (pCi/cc) (pCi/sec) l Tech Spec 3.2E-1 1.5E+6 6.2E-4 2.1 E+4 Unusual Event (2 x TS) 6.4E-1 3.0E+6 1.2E-3 4.2E+4 O Alert (60 x TS) 1.9E+1 8.9E+7 3.7E-2 1.3E+6 LLRPSF Kaman 12  ;

I Maximum flow (CFM) 99,000 Release Limits Concentration Release Rate i (pCi/cc) ( Ci/sec)

Tech Spec 4.5E-4 2.1E+5 Unusual Event (2 x TS) 9.0E-4 4.2E+5 Alert (200 x TS) 9.0E-2 4.2E+7 The off-gas stack is treated as an elevated release and the turbine building and reactor building vents are treated as mixed-mode releases. The ground level setpoints are taken from the default setpoint calculations from the quarterly surveillance tests performed by DAEC Chemistry technicians. Reactor Building, Turbine Building, LLRPSF (Low Level Radwaste Processing and Storage Facility; and Offgas Stack Noble Gas Monitor alarm setpoints nre calculated based on achieving the Tech Spec instantaneous release limit, assuming annual average meteorology as defined in the ODAM. The Tech Spec Limit currently corresponds to a reactor building or turbine building ventilation alarm setpoint of 6.2 E-04 Ci/cc. The monitor alarm setpoint can be periodically adjusted but typically does not vary by much. The DAEC EAL therefore addresses valid radiation levels exceeding 2 times the alarm setpoint for greater than 60 minutes.

l l Rounded off, this corresponds to 1 E-3 pCi/cc. The corresponding ofTgas stack monitor value is 0.64 l Ci/cc, rounded off to 6 E-1 pCi/cc. The Tech Spec Limit currently for the LLRPSF building ventilation i alann setpoint is 4.5 E-04 Ci/cc. The DAEC EAL therefore addresses valid radiation levels exceeding 2 times the alarm setpoint for greater than 60 minutes. This corresponds to 9 E-4 pCi/cc.

f I Technical specification setpoints for radioactive liquid radiation monitors are 10 times the 10 CFR 20 Appendix B, Table 2, Water Effluent Concentration (WEC) limits. It is the policy of DAEC to process all

i Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

PAGE A-3 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY l

l liquid radwaste so that no release of radioactive liquid to the environment is allowed. The radwaste l effluent line which could be used as a batch release mechanism has a trip function that prevents exceeding '

the DAEC release limit, and therefore no EAL limits are provided. The other pathways to the environment 1 (RHRSW - to cooling tower, RHRSW - to discharge canal) have radiation monitors with readouts going to I the Control Room. These systems could become contaminated if heat exchanger leaks develop; however, historically this has not occurred in the service water systems at DAEC. These monitors are displayed on i panels IC02 and IC10. j Reactor water is the likely source of contamination through the service water systems as opposed to floor j l

drain, detergent drain, and chemical waste discharge. The floor drain and detergent drains go to Radwaste i Processing and would be batch released to the Radwaste effluent discharge line (if such a release were to

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occur). The chemical discharge sump is normally a radioactivity clean system and is tested by Chemistry  ;

7 to ensure no contamination prior to discharging to the canal. '

The setpoints for the three service water radiation effluent monitors vary because of differences in detector efliciencies and background. Setpoints based on the same reactor water sample are listed below to show the differences. The rounded off readings will be used for the EALs for ease of reading the monitor scales.

l Monitor TS Limit Reading UE Level Alert Level )

! GSW 1,555 CPS 1.5E+3 CPS 3E+3 CPS 3E+5 CPS RilRSW & ESW to cooling 413 CPS 4E+2 CPS 8E+2 CPS 8E+4 CPS

tower l l RHRSW & ESW to 507 CPS 5E+2 CPS 1E+3 CPS 1E+5 CPS Discharge Canal l

There are no significant deviations from the generic EALs. However, DAEC does not have a telemetered  ;

radiation monitoring system. As an attemative, use of field instruments was considered. It is not practical  !

t to establish an EAL based on field survey readings of 0.1 mR/hr for greater than 60 minutes because field instruments in use for emergency response do not have a threshold of detection to meet such criteria. Thus,  ;

DAEC does not have an EAL corresponding to generic EAL 3.

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) s PAGE A-4 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY I!ourly Whole Body Dose Corresponding to 2 x ODAM Limit for Gaseous Release ODAM limit = 500 mrem / year Whole Body Dose 2 x ODAM limit = [2 x 500 mrem / year]/8760 hours / year = 0.114 mrem Whole Body in one hour Rounded off to 0.1 mrem Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total EfTective Dose Equivalent (TEDE). This is somewhat difTerent from whole body dose from gaseous effluents determined by ODAM methodology which forms the basis for the radiation monitor readings calculated in accordance (y

' with the generic methodology. The gaseous emuent radiation monitors can only detect noble gases. The contribution ofiodine's to TEDE could therefore only be determined either by: (1) utilizing MIDAS, or (2) gaseous effluent sampling. DAEC EAL 4 is written in terms of TEDE and the gaseous effluent radiation monitor readings are determined based on ODAM.

REFERENCES:

1. Offsite Dose Assessment Manual Section 6.1.2 and 7.1.2 Bases

2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action

3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Effluent Radiation Monitors, January 24,1995 i

4. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems

5. EPA 400-R-92-001, Manual ofProtective Action Guides and Protective Actionsfor Nucleue Incidents

6. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/N(SP-007 Revision 2 l Questions andAnswers, June 1993 V

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCrevien)

PAGE A-5 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY AU2 Unexpected Increase in Plant Radiation < >

EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY: All i

EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4)

1. (Site-specific) indication of uncontrolled water level decrease in the reactor refueling cavity with all irradiated fuel assemblies remaining covered by water.  !

2. Uncontrolled water level decrease in the spent fuel pool < > with all irradiated fuel assemblies i i

remaining covered by water.

( 3. (Site-specific) radiation reading for irradiated spent fuel in dry storage.

4. Valid Direct Area Radiation Monitor readings increases by a factor of 1000 over normal

• levels. l

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

DAEC EAL INFORMATION:

There are no significant deviations from the generic EALs. DAEC does not have a spent fuel transfer canal or u-site dry storage of spent fuel.

Uncontrolled means that the condition is not the result of planned actions by the plant staffin accordance l with procedures. Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.  ;

There are three methods to determine water level decreases of concem. The first method is by report to the )

cort o1 room. The other methods include use of the Floodup level indicator and the spent fuel pool level l indicator. These are further described below.

m During pn paration for reactor cavity flood up prior to entry into refuel mode, reactor vessel level

) instrument LI-4541 (WR GEMAC, FLOODUP) on control room panel IC04 is placed in service by I&C l( personnel connecting a compensating air signal after the reference leg is disconnected from the reactor head. Normal refuel water level is above the top of the span of this flood up level indicator. A valid l

i indication (e.g., not due to loss of compensating air signal or other instrument channel failure) of reactor

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

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PAGE A-6 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY cavity level coming on span for this instrument is used at DAEC as an indicator of uncontrolled reactor cavity level decrease.

DAEC Technical Specifications require a minimum of 36 feet of water in the spent fuel pool. During refueling, the gates between the reactor cavity and the refueling cavity are removed and the spent fuel pool level indicator LI 3413 is used to monitor refueling water level. Procedures require that a normal refueling water level be maintained at 37 feet 5 inches. A low level alarm actuates when spent fuel pool level drops below 37 feet 1 inch. Symptoms ofinventory loss at DAEC include visual observation of decreasing water levels in reactor cavity or spent fuel storage pool, Reactor Building (RB) fuel storage pool radiation monitor or refueling area radiation monitor alarms, observation of a decreasing trend on the spent fuel pool water level recorder, and actuation of the spent fuel pool low water level alarm. To eliminate minor level perturbations from concern, DAEC uses LI 3413 indicated water level below 36 feet and lowering.

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Increased radiation levels can be detected by the local refueling floor area radiation monitors, the refueling floor Continuous Air Monitor (CAM) alarm, refueling areas radiation monitors, fuel pool ventilation exhaust monitors, and by Standby Gas Treatment (SGBT) System automatic start. Applicable area radiation monitors include those that are displayed on Panel IC02 and alarmed on Panel IC04B. The DAEC EAL has also been written to reflect the case where an ARM may go offscale high prior to reaching 1,000 times the normal reading.

NOTE: On Annunciator Panel IC04B, the indicators listed below are expected alarms during pre-planned transfers of highly radioactive material through the affected area. If an HP Technician is present, sending an Operator is not required. Radiation levels other than those expected should be promptly investigated.

The indicators are high radiation alarms from the Hot Laboratory or Administrative Building, the new fuel storage area, and the radwaste building.

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i Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) r"g b PAGE A-7 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY i

REFERENCES:

l 1. Alarm Response Procedure (ARP) 1C04B, Reactor Water Cleanup and Isolation

! 2. Technical Specification 3.9C, Spent Fuel Pool Water Level l 3. Emergency Plan Implementing Procedure (EPIP) 3.1, Inplant Radiological Monitoring, Attachment 1,

! ARM Locations

4. Emergency Operating Procedures (EOP) Basis Document, Breakpoints for RC/L & L l 5. Surveillance Test Procedure (STP) 42A-0001, Daily and Shift Instrument Checks

6. Integrated Plant Operating Instruction (IPOI) 8, Outage and Refueling Operations

7. Fuel & Reactor Component Handling Procedure (F&RCHP) 5, Procedure for Moving Core Components Between Reactor Core and Spent Fuel Pool, Within the Reactor Core, or Within the Spent Fuel Pool

8. NUMARC Methodologyfor Development ofEmergency Action Levels NUAfARC/NESP-007 Revision 2 Questions andAnswers, June 1993 l

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, EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

PAGE A-8 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY 1 l

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AA1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All

, EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4)

1. A valid reading on < site-specific > monitors that < > indicates that the release may have exceeded <200 x site-specific technical specifications for 15 minutes or longer.>

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i L 2. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates in excess of(200 x site-specific technical specifications) for 15 minutes or longer.

3. A valid reading on perimeter radiation monitoring system grecter than 10.0 mR/hr sustained for 15 minutes or longer. (for sites having telemetered perimeter monitors]

4. Valid indication on automatic real-time dose assessment capability greater than (200 x site-specific Technical Specifications value) for 15 minutes or longer. [for sites having such capability]

DAEC EAL INFORMATION:

Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.

l The primary methodfor declaration is by means ofdose assessment using the MlDAS computer model. l l This is listed as DAEC EAL 4. However, ifthe monitor readings are sustainedfor longer than 15 minutes and the required dose assessments cannot be completed within this period, then the declaration must be l made basedon the validreading.

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Duane Amold Energy Center i EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

PAGE A-9 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY Gaseous Effluent EALs Offgas Stack Kaman 9/10 Turbine Bldg (Kaman 1/2) and Reactor Bldg (Kaman 3/4,5/6,7/8)

Maximum flow (CFM) 10,000 72,000 Release Limits Concentration Release Rate Concentration Release Rate (pCi/cc) (pCi/sec) ( Ci/cc) ( Ci/sec) l Tech Spec 3.2E-1 1.5E+6 6.2E-4 2.1 E+4 Unusual Event (2 x TS) 6AE-1 3.0E+6 1.2E-3 4.2E+4 l Alert (60 x TS) 1.9E+1 8.9E+7 3.7E-2 1.3E+6 LLRPSF Kaman 12 Maximum flow (CFM) 99,000 Release Limits Concentration Release Rate (pCi/cc) ( Ci/sec)

Tech Spec 4.5E-4 2.1 E+5 Unusual Event (2 x TS) 9.0E-4 4.2E+5 l Alert (200 x TS) 9.0E-2 4.2E+7 The off-gas stack is treated as an elevated release and the turbine building and reactor building vents are treated as mixed-mode releases. The ground level setpoints are taken from the default setpoint calculations from the quarterly surveillance tests performed by DAEC Chemistry technicians. Reactor Building, Turbine Building, LLRPSF (Low Level Radwaste Processing and Storage Facility) and Offgas Stack Noble Gas Monitor alarm setpoints are calculated based on achieving the Tech Spec instantaneous release limit assuming annual average meteorology as defined in the ODAM. The Tech Spec Limit currently corresponds to a reactor building or turbine building ventilation alarm setpoint of 6.2 E-4 pCi/cc. The l

monitor alarm setpoint can be periodically adjusted but typically does not vary by much. For the Offgas Stack, Reactor Building and Turbine building KAMAN monitor readings, DAEC chose to multiply the technical specification concentration by a factor of 60 (instead of 200) in order to allow for a logical step j progression in monitor setpoints from the AU1 through AA1 to ASI. The DAEC EAL therefore addresses i valid radiation levels exceeding 60 times the alarm setpoint for greater than 15 minutes. Rounded off, this corresponds to 3 E-2 pCi/cc. The corresponding offgas stack monitor value is 19.2 pCi/ce, rounded off to 2 E+1 Ci/cc. The Tech Spec Limit currently for the LLRPSF building ventilation alarm setpoint is 4.5 E-04 pCi/cc. The DAEC EAL therefore addresses valid radiation levels exceeding 200 times the alarm setpoint for greater than 15 minutes. This corresponds to 9 E-2 pCi/cc.

Technical specification setpoints for radioactive liquid radiation monitors are 10 times the 10 CFR 20 Appendix B, Table 2, Water Effluent Concentration (WEC) limits. It is the policy of DAEC to process all

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) i PAGE A-10 of 24 l

ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE:TBD CATEGORY I l l  :

! liquid radwaste so that no release of radioactive liquid to the environment is allowed. The radwaste efiluent line which could be used as a batch release mechanism has a trip function that prevents exceeding ,

the DAEC release limit, and therefore no EAL limits are provided. The other pathways to the environment i l (RHRSW - to cooling tower, RHRSW - to discharge canal) have radiation monitors with readouts going to  !

the Control Room. These systems could become contaminated if heat exchanger leaks develop; however, l historically this has not occurred in the service water systems at DAEC. These monitors are displayed on panels IC02 and ICIO.

Reactor water is the likely source of contamination through the service water systems as opposed to floor l drain, detergent drain, and chemical waste discharge. The floor drain and detergent drains go to Radwaste l Processing and would be batch released to the Radwaste effluent discharge line (if such a release were to i occur). The chemical discharge sump is normally a radioactivity clean system and is tested by Chemistry l to ensure no contamination prior to discharging to the canal.

The setpoints for the three service water radiation effluent monitors vary because of differences in detector ]

l efficiencies and background. Setpoints based on the same reactor water sample are listed below to show l the differences. The rounded off readmgs will be used for the EALs for ease of reading the monitor scales. l l

Monitor TS Limit Reading UE Level Alert Level

! GSW 1,555 CPS 1.5E+3 CPS 3E+3 CPS 3E+5 CPS  ;

RHRSW & ESW to cooling 413 CPS 4E+2 CPS 8E+2 CPS 8E+4 CPS l j i tower RHRSW & ESW to 507 CPS 5E+2 CPS 1E+3 CPS 1E+5 CPS Discharge Canal l

DAEC does not have a telemetered radiation monitoring system. As an altemative, DAEC uses valid field l survey readings greater than 10 mR/hr.

Hourly Whole Body Dose Corresponding to 200 x ODAM Limit for Gaseous Release {

i ODAM limit = 500 mrem / year Whole Body )

200 x ODAM limit = [200 x 500 mrem / year]/8760 hours / year = 11.4 mrem Whole Body in one hour Rounded off to 10 mrem  !

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' l ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY i l

Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE). This is somewhat different from whole body dose from gaseous emuents determined by ODAM methodology which forms the basis for the radiation monitor readings calculated in AUl in 1 accordance with the generic methodology. The gaseous effluent radiation monitors can only detect noble gases. The contribution of iodine's to TEDE could therefore only be determined either by: (1) utilizing MIDAS, or (2) gaseous effluent sampling. DAEC EAL 4 is written in terms of TEDE and the gaseous effluent radiation monitor readings are determined based on ODAM.

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REFERENCES:

1. Offsite Dose Assessment Manual Section 6.1.2 and 7.1.2 Bases l

2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action

3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Effluent l Radiation Monitors, January 24,1995 l 4. UFSAR Section 11.5, Process and Eflluent Radiation Monitoring and Sampling Systems l 5. EPA 400-R-92-001, Manual ofProtective Action Guides and Protective Actionsfor Nuclear Incidents l 6. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers,3une 1993 i

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PAGE A-12 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD l CATEGORY AA2 Major Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering ofIrradiated Fuel Outside the Reactor Vessel EVENT TYPE: Onsite Rad Conditions ,

l OPERATING MODE APPLICABILITY: All l

EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4)

1. < Valid < Site-specific > radiation monitor readings for the refuel floor area, fuel handling area, and the l fuel bridge area.> ,

q 2. Report of Visual observation ofirradiated fuel uncovered. j l Q 3. Water Level less than (site-specific) feet for the Reactor Refueling Cavity that will result in Irradiated Fuel Uncovering.

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4. Water Level less than (site-specific) feet for the Spent Fuel Pool < > that will result in Irradiated Fuel l uncovering. ,

1 DAEC EAL INFORMATION:

l l Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed l on the control panels, reports from plant personnel, or radiological survey results. Valid alarms are solely due to damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel.

l There are no significant deviations from the generic EALs. Increased radiation levels can be detected by the local radiation monitors, in-plant radiological surveys, new fuel and spent fuel storage area radiation monitor alarms di.cplayed on panel IC04B, fuel pool ventilation exhaust monitors, and by Standby Gas Treatment (SBGT) System automatic start. Applicable area radiation monitors include RT 9163, RT 9164, RT 9153, and RT 9178. These monitors are located in the north end of the refuel floor, the south end of the refuel floor, the new fuel vault area, and near the spent fuel pool, respectively.

Per ARP 1C04B, the applicable area radiation monitor alarms actuate when radiation levels increase above

' 100 mR/hr in the spent fuel pool area or above 10 mR/hr in the other three areas of concern. If a valid t

actuation of these alarms were to occur, the refueling floor would be immediately evacuated. Thus, a i report of a fuel handling accident with either valid actuation of the fuel area alarms on panel IC04B or with I

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PAGE A-13 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY measured radiation levels in the spent fuel pool or north fuel area are used to address the generic concem consistent with DAEC design and procedures.

During preparation for reactor cavity flood up prior to entry into refuel mode, reac" vessel level instrument LI-4541 (WR GEMAC, FLOODUP) on control room panel IC04 is placed in service by I&C personnel connecting a compensating air signal after the reference leg is disconnected from the reactor head. Normal refuel water level is above the top of the span of this flood up level indicator. A valid on-scale indication (e.g., not due to loss of compensating air signal or other instrument channel failure) from this instrument can be used to determine uncontrolled loss of water level in the reactor cavity.

During refueling, the gates between the reactor cavity and the refueling cavity are removed and the spent 7 fuel pool level indicator LI 3413 is used to monitor refueling water level. This measures the common

() water level in the reactor cavity and the fuel pool. The bottom of the fuel transfer slot between the spent fuel pool and the reactor cavity is 16 feet above the bottom of the spent fuel pool. The top of the active fuel in the spent fuel storage racks is slightly less than 13 feet 9 inches above the bottom of the spent fuel pool.

Therefore, postulated failures which drain the reactor cavity through the reactor vessel cannot uncover fuel in the spent fuel storage racks. However, valid indication of spent fuel pool level less than 13 feet 9 inches would indicate that spent fuel in the storage racks is starting to become uncovered.

F&RCHP 5 requires that upon a loss of water level situation, that the refueling crew on the reiueling floor shall discharge any fuel assembly on the fuel grapple as follows:

. If a fuel assembly is currently being withdrawn from a slot in the core or spent fuel pool, immediately reinsert it into that slot.

• If a fuel assembly is being transferred and is still over or near the core, insert it into the closest available slot in the core.

e if a fuel assembly is being transferred and is over or near the spent fuel pool, insert it into the closest available slot in the spent fuel racks. ,

Following these actions, the refueling floor is to be evacuated of all personnel. The DAEC EAL is written to address the generic concern that a spent fuel assembly was not fully covered by water. This can either be by visual observation of an uncovered spent fuel assembly or by trending fuel pool level in the control room if a spent fuel assembly could not be placed in a safe storage location specified by F&RCHP 5 as

]vdescribed above.

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U PAGE A-14 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY I

i

REFERENCES:

l 1. Alarm Response Procedure (ARP) IC048, Reactor Water Cleanup and Isolation j 2. Technical Specification 3.9C, Spent Fuel Pool Water Level l 3. Emergency Operating Procedures (EOP) Basis Document, Breakpoints for RC/L & L

4. Emergency Plan Implementing Procedure (EPIP) 3.1, Inplant Radiological Monitoring, Attachment 1, ARM Locations

5. Surveillance Test Procedure (STP) 42A-0001, Daily and Shift Instrument Checks

6. Integrated Plant Operating Instruction (IPOI) 8, Outage and Refueling Operations

( 7. Fuel & Reactor Component Handling Procedure (F&RCHP) 5, Procedure for Moving Core Components Between Reactor Core and Spent Fuel Pool, Within the Reactor Core, or Within the Spent Fuel Pool

!gQ 8. Bechtel Drawing C-492, Reactor Building - Reactor Well, Spent Fuel & Dryer-Separator Pool General Arrangement, Rev. 6

9. Bechtel Drawing C-493, Reactor Building - Spent Fuel Liner Plan Elevations and Details, Sheet 1, l Rev.6 l 10. Holtec International Drawing No.1045, Rack Construction - Spent Fuel Storage Racks, Rev. 3 l 11.NUMARC Methodologyfor Development ofEmergency Action Levels NUAfARC/NESP-007 Revision 2 Questions andAnswers, June 1993 i

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PAGE A-15 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY l

AA3 Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe i Operations or to Establish or to Maintain Cold Shutdown EVENT TYPE: Onsite Rad Conditions l l

OPERATING MODE APPLICABILITY: All l I EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2) l l

l 1. Valid (site-specific) radiation monitor readings GREATER THAN (site-specific) values in areas C requiring continuous occupancy to maintain plant safety functions < >

( N)T l 2. Valid (site-specific) radiation monitor readings GREATER THAN (site-specific) values in areas requiring infrequent access to maintain plant safety functions < >

l DAEC EAL INFORMATION:

Valid means that the reading is from instrumentation determined to be operable in accordance with the l Technical Specifications or has been verified by other independent methods such as indications displayed l on the control panels, reports from plant personnel, or radiological survey results. l l There are no significant deviations from the generic EALs. Per the UFSAR, the control room is the only area that is required to be continuously occupied to achieve and maintain safe shutdown following design basis accidents. DAEC EAL 1 is directly applicable to NUMARC EAL 1. However, the capability exists '

for plant shutdown from outside the main control room in the event that the control room becomes  !

uninhabitable using remote shutdown panel IC388. DAEC EAL 2 is directly applicable to NUMARC j EAL2.

The EC/OSS shondd determine the cause of the increase in radiation levels and review other EALsfor applicability. Expected increases in monitor readings due to controlled evolutions (such as lifting the steam dryer during refueling) do not result in emergency declaration. Nor should momentary increases due a

p to events such as resin transfers or controlled movement of radioactive sources result in emergency U declaration. In-plant radiation level increases that would result in emergency declaration, are also unplanned, e.g., outside the limits established by an existing radioactive discharge permit.

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) l 0 Q PAGE A-16 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY l

REFERENCES:

l

1. Alarm Response Procedure (ARP) IC04B, Reactor Water Cleanup and Isolation

2. Abnormal Operating Procedure (AOP) 913, Fire

3. Abnormal Operating Procedure (AOP) 914, Security

4. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room  ;

5. Surveillance Test Procedure (STP) 42A-0001, Daily and Shift Instrument Checks '

l 6. Integrated Plant Operating Instruction (IPOI) 8 , Outage and Refueling Operations i

7. Emergency Plan Implementing Procedure (EPIP) 3.1,Inplant Radiological Monitoring i

8. UFSAR Section 6.4, Habitability Systems l 9. Bechtel Calculation DA-4, Project Number 265-002, Control Room Habitability,9/3/80

, 10. NUMARC Methodologyfor Development ofEmergency Action Levels NUM4RC/NESP-007 Revision 2 l Questions andAnswers, June 1993 I

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PAGE A-17 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT 1 EFFECTIVE DATE: TBD l CATEGORY j 1

ASI Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 < mrem TEDE> or 500 < mrem CDE> ,

Thyroid for the Actual or Projected Duration of the Release l EVENT TYPE: OfTsite Rad Conditions OPERATING MODE APPLICABILITY: All j i

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EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4) i

1. A valid reading on < site-specific > monitors <for greater than 15 minutes which corresponds to an l A offsite dose of 100 mrem or 500 mrem Thyroid in an hour >.

U 2. A valid reading sustained for 15 minutes or longer on perimeter radiation monitoring system greater than 100 mR/hr. (for sites having telemetered perimeter monitors]. l

3. Valid dose assessment capability indicates dose consequences greater than 100 < mrem TEDE> or 500 l

< mrem CDE> thyroid.

4. Field survey results indicate site boundary dose rates exceeding 100 < mrem >/hr expected to continue i for more than one hour; or analyses of field survey samples indicate <CDE> thyroid of 500 < mrem > for one hour ofinhalation.

DAEC EAL INFORMATION:

l Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed l on the control panels, reports from plant personnel, or radiological survey results.

There are no significant deviations from the generic EALs. l DAEC's Meteorological Information and Dose Assessment System (MIDAS) was utilized to determine l

l the KAMAN monitor limits. Eight separate combinations of release point, source term, meteorological  :

. conditions and equipment status were analyzed. Pathways considered were the offgas stack, the turbine building exhaust vent and a single reactor building exhaust vent. Multiple release points were not n considered. In this same vein, it was assumed that only one of the three reactor building vents is on during

() therelease.

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EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) l O PAGE A-18 of 24 1'

ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT -

EFFECTIVE DATE: TBD CATEGORY j The source terms used have been pre-loaded into MIDAS and are the default mixes associated with a loss

, of coolant accident (LOCA) and a control rod drop (CRD). The LOCA mix was used in conjunction with a i release via the offgas stack while the CRD mix was used for releases via the turbine or reactor building l L vents. The source term for a release via the offgas stack is further impacted by the status of the standby gas treatment system. The status of that system was also taken into consideration.

- Based of 1995 data (NG-96-0987), the atmospheric stability was classified as Pascal E 33% of the time.

Consequently, both classifications were evaluated. Based on the same report, the most common wind j ' speeds were:

Pascal Class Altitude Sneed (mnh1 l' D 156' ' 8 - 12' L D 33' 8 - 12 E 156' 8 - 12 E 33' 4-7 Though the temperature setting has no impact on the MIDAS calculations, a value must be entered in order for the program to run Consequently, the temperature was arbitrarily set at 50 F.

The rain estimate was set at zero, to eliminate any on site washout of radioactive material.

For the first MIDAS runs a ICi/cc concentration was assumed. The results of these runs were then  !

normalized to the limits, thus generating a theoretical KAMAN limit. Additional MIDAS runs were made with these theoretical limits as input to verify the normalization process.

In addition to the total integrated dose, MIDAS calculates a peak whole body DDE rate resulting from the 1 plume and a peak thyroid CDE rate resulting from inhalation. Because the ASI and AGI KAMAN limits

. are to be based on a one hour exposure, establishing concentration limits so these peak values match the i NUMARC limits is acceptable.  !

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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) l PAGE A-19 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY l

Site Area Emergency General Emergency Initiating Condition ASI AGl Valid Turbine or Reactor Building ventilation rad ,

monitor (KAMAN) reading for more than 15 0.06 pCi/cc 0.6 pCi/cc l minutes above:  ;

Valid Offgas Stack ventilation rad monitor j (KAMAN) reading for more than 15 minutes 40 pCi/cc 400 Ci/cc l above: 1 The primary methodfor declaration is by means ofdose assessment using the MIDAS computer model.

However, if the monitor readings are sustainedfor longer than 15 minutes and the required dose

&a assessments reading. cannot be completed within thisperiod, then the declaration must be I

DAEC does not have a telemetered radiation monitoring system. As an alternative, DAEC uses valid field survey readings greater than 100 mR/hr.

Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE) and Committed Dose Equivalent (CDE) Thyroid. TEDE is somewhat different from ,

whole body dose from gaseous efiluents determined by ODAM methodology which forms the basis for the  !

radiation monitor readings calculated in AUI. These factors can introduce differences that are at least as large as those mtroduced by using TEDE versus whole body dose. The gaseous effluent radiation monitors l can only detect noble gases. The contribution ofiodine's to TEDE and CDE Thyroid could therefore only  !

be determined either by: (1) utilizing the source term mixture in MIDAS, or (2) gaseous effluent sampling.

Therefore, DAEC EAL 4 is written in terms of TEDE and CDE Thyroid.

REFERENCES:

1. Offsite Dose Assessment Manual, Section 6.1.2 and 7.1.2, Bases

2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action

3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Effluent

! Radiation Monitors, January 24,1995

4. Radiation Engineering Cciculation No. 96-007-A, Determination of DAEC Radioactive Release Initiating Conditions for ASl & AG1 Emergency Classifications, July 3,1996

5. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems

6. EPA 400-R-92-001, Manual ofProtective Action Guides and Protective Actionsfor Nuclear Incidents

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i ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD

- CATEGORY 1

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7. NUMARC Afethodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 l Questions andAnswers, June 1993 i

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PAGE A-21 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE

TBD CATEGORY l

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! AGI Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds <1,000 mrem TEDE> or <5,000 mrem

! CDE> Thyroid for the Actual or Projected Duration of the Release < >

1 EVENT TYPE: Offsite Rad Conditions l

l OPERATING MODE APPLICABILITY: All l EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4) l l 1. A valid reading on < site-specific > monitors <for greater than 15 minutes which corresponds to an l U(] offsite dose of 1,000 mrem or 5,000 mrem Thyroid in an hour >.

2. A valid reading sustained for 15 minutes or longer on perimeter radiation monitoring system greater than 1,000 mR/hr. [for sites having telemetered perimeter monitors].

l 3. Valid dose assessment capability indicates dose consequences greater than 1,000 < mrem TEDE> or 5,000 < mrem CDE> thyroid.

l 4. Field survey results indicate site boundary dose rates exceeding 1,000 < mrem >/hr expected to continue

( for more than one hour; or analyses of field survey samples indicate <CDE thyroid > of 5,000 < mrem >

for one hour ofinhalation.

DAEC EAL INFORMATION:

! Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed

on the control panels, reports from plant personnel, or radiological survey results.

There are no significant deviations from the generic EALs.

DAEC's Meteorological Information and Dose Assessment System (MIDAS) was utilized to determine the KAMAN monitor limits. Eight separate combinations of release point, source term, meteorological conditions and equipment status were analyzed. Pathways considered were the offgas stack, the turbine building exhaust vent and a single reactor building exhaust vent. Multiple release points were not a considered. In this same vein, it was assumed that only one of the three reactor building vents is on during V the release.

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EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

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PAGE A-22 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD CATEGORY l i

I The source terms used have been pre-loaded into MIDAS and are the default mixes associated with a loss l of coolant accident (LOCA) and a control rod drop (CRD). The LOCA mix was used in conjunction with a l release via the offgas stack while the CRD mix was used for releases via the turbine or reactor buildiig l vents. The source term for a release via the offgas stack is further impacted by the status of the standby gas l treatment system. The status of that system was also taken into consideration.

Based of 1995 data (NG-96-0987), the atmospheric stability was classified as Pascal E 33% of the time.

Consequently, both classifications were evaluated. Based on the same report, the most common wind  ;

speeds were: '

Pascal Class Altitude Sneed (mnh)

D 156' 8 - 12

( D 33' 8 - 12  ;

E 156' 8 - 12 E 33' 4-7 l

Though the temperature setting has no impact on the MIDAS calculations, a value must be entered in order for the program to run. Consequently, the temperature was c.rbitrarily set at 50 F. l I

The rain estimate was set at zero, to eliminate any on site washout ofra '.ioactive material.

)

i For the first MIDAS runs a ICi/cc concentration was assumed. The results of these runs were then normalized to the limits, thus generating a theoretical KAMAN limit. Additional MIDAS runs were made with these theoretical limits as input to verify the normalization process.

l In addition to the total integrated dose, MIDAS calculates a peak whole body DDE rate resulting from the l plume and a peak thyroid CDE rate resulting from inhalation. Because the ASI and AGl KAMANlimits I are to be based on a one hour exposure, establishing concentration limits so these peak values match the NUMARC limits is acceptable.

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Duane Amold Energy Center EMERGbNCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

PAGE A-23 of 24 i

ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD ,

CATEGORY Site Area Emergency General Emergency Initiating Condition ASI AGI Valid Turbine or Reactor Building ventilation rad monitor (KAMAN) reading for more than 15 0.06 pCi/cc 0.6 pCi/cc minutes above:

Valid Offgas Stack ventilation rad monitor (KAMAN) reading for more than 15 minutes 40 pCi/cc 400 pCi/cc above:

The preferred methodfor declaration is by means ofdose assessment using the Af1DAS computer model and is therefore is listed as DAEC EAL 4. However, if the monitor readings are sustainedfor longer than 15 minutes and the required dose assessments cannot be completed within this period, then the declaration

(

\ must be made based on the valid reading.

DAEC does not have a telemetered radiation monitoring system. As an attemative, DAEC uses valid field survey readings greater than 1000 mR/hr.

Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE) and Committed Dose Equivalent (CDE) Thyroid. TEDE is somewhat different from whole body dose from gaseous effluents deterrained by ODAM methodology which forms the basis for the radiation monitor readings calculated in AU1. These factors can introduce differences that are at least as large as those introduced by using TEDE versus whole body dose. The gaseous efiluent radiation monitors -

can only detect noble gases. The contribution ofiodine's to TEDE and CDE Thyroid could therefore only be determined either by: (1) utlizing the source term mixture in MIDAS, or (2) gaseous effluent sampling. ,

Therefore, DAEC EAL 4 is written in terms of TEDE and CDE Thyroid.

REFERENCES:

1. Offsite Dose Assessment Manual, Section 6.1.2 and 7.1.2, Bases

2. Emergency Plan implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action l

3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Ellluent l Radiation Monitors, January 24,1995 l - 4. Radiation Engineering Calculation No. 96-007-A, Determination of DAEC Radioactive Release j Initiating Conditions for ASI & AGI Emergency Classifications, July 3,1996 ]

! 5. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems l

6. EPA 400-R-92-001, Afanual ofProtective Action Guides andProtective Actionsfor Nuclear Incidents

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PAGE A-24 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: TBD

• CATEGORY  ;

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m e EMERGENCY PIAN LMPLEMENTING PROCEDURE No. EPIP - 1.1 Rev.I PAGElof1 (ForNRCReview)

FISSION BARRIER TABLE EHEN DATE:TBD INDICATORS FUEL CLAD BARRIER RCS BARRIER l Loss Loss l L Fuel damage assessrnent (PASAP 7.2) determines at L Valid drywell rad monitor reading above 5 R/hr after i least 5% fuel clad damage reactor shutdown ,

OR 1 Fuel damage is indicated by any of the following:

Potentialloss- None l RADIATION / L Valid drywell rad monitor reading above 7E+2 R/hr OR CORE DAMAGE L Valid torus rad monitor reading above 3E+1 R/hr OR L Coolant activity above 300 Ci/gm DOSE EQUIVALENT l-131 PotentialLoss - None j Loss Loss i L RPV Level below -30 inches L RPV Level below 15 inches {

RPV LEVEL PotentialLoss PotentialLoss- None P RPV Levet below 15 inches Loss - None j PotentialLoss P RCS Leakage is above 50 GPM l OR P Unisolable primary system leakage outside the drywg LEAKAGE None as indicated by area temps or ARMS 1

Loss ,

L Drywell cooling operating AND drywell pressure abod l 2 psig l PRIMARY CONTAINMENT None #""# ## ' "#

ATMOSPHERE i

Any condition which in the EC/OSS's Judgment indicates Any condition which in the EC/OSS's judgment indicates EC/OSS ioss or potentiaiioss of tne fuei ciad barrier due to: ioss or potentiai toss of tne RCS barrier due to:

• Imminent barrier degradation

• Imminent barrier degradation

• Degraded fission barrier monitoring capability

• Degraded fission barrier monitoring capability IMMINENT No turnaround in safety system performance is expected and escalation to General Emergency conditions is expected within 2 houn NOTE: Step 1; for allindicators, move from left to right across table, marking allapplicable "L's"and 'P's' for each barrier, based on plantindications. Theq "L's"and 'P's' marked on Barrier Table to the Logic Diagram (at right). "L's* and *P's* should be marked for each affected barrier (working top to bottom) on Step 3, an "L"or *P" marked for each associated barrier will constitute a Logic Iinput. When coincidence is met, then the EAL can be declared.

l L = Loss (of a fission product barrier)- A severe challenge to a fission product barrier exists such that the barrier is considered incapable of performing its C P = Potentialloss (of a fission product barrier)- A challenge to a fission product barrier exists such that the barrier is considered degraded in its ability to $

e- v

PRIMARY CONTAINMENT BARRIER ONE BARRIER AFFECTED l Loss- None l

L P L P L P l PotentialLoss P V . lid drywell rad monitor reading above 3E+3 R/hr CLAD RCS CNTMT l

P VLlid torus rad monitor reading above 1E+7 R/hr E l FUI OR UNUSUAL

)

P Cors damage assessment determines at least 20% o o 1/1 EVENT I fuelclad damage EEEMI '

1/2 l I

)

. FA1 ALERT Loss - None TWO BARRIERS AFFECTED

~

PotentialLoss L P L P L P l P RPV Level below -40 inches CLAD RCS CNTMT Loss

_ L Fcilure of both isolation valves and a downstream pathway to the environment exists gg g' g OR 2/3 L Unisolable primary system leakage outside the drywell Fst cs indicated by area temps or ARMS l I . SITE AREA >

OR '

EMERGENCY L Primary containment venting performed per EOPs THREE BARRIERS AFFECTED "

PotentialLoss- None t P t P L p loss- None CLAD RCS CN WT L Rapid unexplained decrease following initial increase OR l l L L Drywell pressure response not consistent with LOCA conditions EE II PotentialLoss ANSTEC = .

Torus pressure reaches 53 psig OR APERTURE LOSS OF AT NO P Drywell of torus H2 CANNOT be determined to be below 6% AND Drywell or torus 02 CANNOT be CARD = ==

MERS7 determined to be below 5%

AlsO Avattable on YES ro1 Aperture Card  : GENERAL j Any condition which in the EC/OSS's judgment indicates EMERGENCY j loss or potentialloss of the primary containment barrier due to:

o Imrninent barrier degradation o Degraded fission barrier monitoring capability Op. Modes: Run, Startup, Hot S/D vtep 2, trarccribe an e floscturt.

lety function.

form its saf:.ty function.

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O O FISSION PRODUCT BARRIER DEGRADATION CATEGORY O

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PAGE F-1 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY l

s FUI Any Loss or Any Potential Loss of < Primary > Containment < Barrier >

EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown j EXAMPLE EMERGENCY ACTION LEVELS:

See the Fission Barrier Table indicators discussed later in this section.

DAEC INFORMATION:

See the Fission Barrier Table indicators discussed later in this section. The entry conditions for this Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table. )

REFERENCES:

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See the Fission Barrier Table indicators discussed later in this section.

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U,~ l PAGE F-2 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD  !

CATEGORY I FAI Any Loss or Any Potential Loss of Either Fuel Clad Or RCS < Barrier >

EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVELS:

See the Fission Barrier Table indicators discussed later in this section.

DAEC INFORMATION: -

See the Fission Barrier Table indicators discussed later in this section. The entry conditions for this Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table.

REFERENCES:

See the Fission Barrier Table indicators discussed later in this section.

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V PAGE F-3 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD l CATEGORY l l i

FSI < Loss Or Potential Loss of Any Two Barriers >

l EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown l EXAMPLE EMERGENCY ACTION LEVELS:

See the Fission Barrier Table indicators discussed later in this section.

l DAEC INFORMATION:

i The entry conditions for this Initiating Condition are shown by the logic chart located to the right of the i

Fission Barrier Table. DAEC uses " Loss Or Potential Loss of Any Two Barriers." This logic is simplified l from the generic logic based on the following considerations: l 1

l. Human Factors - It is easier to understand and to remember the escalation from Alert to Site Area Emergency to General Emergency using the simpler logic.

l

2. Comprehensiveness - A comparison was made of the combinations of barrier losses and potential losses corresponding to Site Area Emergency between the DAEC logic and the NUMARC/NESP-007 logic. All six generic barrier loss / potential loss combinations are addressed in the DAEC logic that addresses 12 combinations of barrier loss / potential loss. No sequences addressed by the l

NUMARC/NESP-007 logic are significantly affected by the simplified logic when applied to a BWR.

l See the table below.

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REFERENCES:

See the Fission Barrier Table indicators discussed later in this section.

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O PAGE F-4 of 27 l

FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY t

l COMPARISON OF DAEC FS1 BARRIER COMBINATIONS l WITII NUMARC/NESP-007 FS1 BARRIER COMBINATIONS FUEL CLAD B ARRIER RCS BARRIER PRIMARY CONTAINMENT BARRIER LOSS POTENTIAL LOSS POTENTIAL LOSS POTENTIAL LOSS LOSS LOSS 1- D, N D, N

2. D, N D, N -

3. D D 4- D D

5. D, N D, N

6. D, N D, N i 1

7- D, N D, N  ;

8. D D 9- D D 10 D D 11 D, N D, N 12 D D I

D - Barrier status addressed by DAEC simplified logic (Loss Or Potential Loss of Any Two Barriers) l N -Barrier status addressed by NUMARC/NESP-007 generic logic (Loss of BOTII Fuel Clad AND RCS OR Potential Loss of BOTII Fuel Clad AND RCS OR Potential Loss of EITIIER Fuel Clad OR RCS l AND Loss of ANY Additional Barrier) l l

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Rev.1 (forNRCrevicu) t \d

! PAGE F-5 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY

! FG1 Loss of Any Two Barriers AND Potential Loss of <the> Third Barrier EVENT TYPE: See Fission Barrier Table l OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVELS:

, See the Fission Barrier Table indicators discussed later in this section.

l

! DAEC INFORMATION:

3 l (Q See the Fission Barrier Table indicators discussed later in this section. The entry conditions for this j Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table.

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REFERENCES:

See the Fission Barrier Table indicators discussed later in this section.

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! Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

.A l \v) PAGE F-6 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY

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FISSION BARRIER: Fuci Clad DAEC INDICATOR: Rediation/ Core Damage GENERIC INDICATOR:

Drywell Radiation Monitoring LOSS - < Valid > Drywell Rad Monitor Reading GREATER THAN (site-specific) R/hr l POTENTIAL LOSS - Not Applicable i

DAEC INFORMATION- I lg Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, coolant sampling or radiological survey results.

l 1

There is no significant deviation from the generic " loss" indicator. Per NUMARC/NESP-007, the (site- i specific) reading is a value which indicates release into the drywell of reactor coolant with elevated activity l corresponding to about 2% to 5% fuel clad damage. This activity level is well above that expected from iodine spiking. It is intended that determination ofbarrier loss be made whenever the indicator threshold is reached until such time that core damage assessment is performed, at which time direct use of containment rad monitor readings is no longer required.

As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel l damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose l rates at the locations of the containment atmospheric monitoring system radiation detectors in the drywell l and torus locations for assumed releases of gap activity form the core. These calculations were based on

" nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more l appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC i

! Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were l evaluated. In the first case, the released activity was assumed to be contained in the drywell atmosphere.

This case is considered representative of conditions following a line break in which activity is released

[ ( } directly This could into the be applied drywell.

for an In the event which resu'its second in vessel case, isolation and blowdownthe released to the suppression activ v ci:rmber. The results for each case were provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) o b PAGE F-7 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY 3

characteristics of the detector systems. The figures show a drywell reading of about 2.9 x 10 R/hr or a <

2 I torus reading of about 1.1 x 10 R/hr associated with 20% gap release at two hours after shutdown. Scaling this down to 5% gap release:

Calculation of Drywell and Torus Monitor Readings Assuming 5% Gap Release 3

NG-88-0966 value 20% Gap Release at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for drywell = 2.9 x 10 R/lu-3 2 Drywell reading = 2.9 x 10 R/hr x [5 % / 20 %] = 7.25 x 10 R/hr, round off as 7 E+2 R/hr  !

2 NG-88-0966 value 20% Gap Release at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ter torus = 1.1 x 10 R/hr 2

i Torus reading = 1.1 x 10 R/hr x [5 % / 20 %) = 2.75 x 10' R/hr, rocad off as 3 E+1 R/hr l

The results are rounded off for ease of reading the respective radiation monitors' scales. The two hour point was picked because it allows ample time for the Technical Support Center to be operational and core

! damage assessment to begin. These indicators correspond to about 2.5% gap release if they occur immediately after shutdown. Thus, the indicators address the 2%-5% fuel clad damage range of concern described by the generic guidance.

REFERENCES:

1. Office Memo NG-88-0966, G.E. Fuel Damage Documentation / Dose Rate Calculations,03/18/88 i

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PAGE F-8 of 27 l

FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: Fuel Clad l

DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:

Primary Coolant Activity Level LOSS - Coolant activity GREATER THAN (site-specific) value POTENTIAL LOSS -Not Applicable DAEC INFORMATION:

There is no significant deviation from the generic indicator. Consistent with the generic methodology,

) DAEC uses a coolant activity value of 300 Ci/gm Iai equivalent. This value is well above that expected for iodine spikes and would indicate fuel clad damage has occurred.

REFERENCES:

1. Post Accident Sampling and Analysis Procedure (PASAP) 7.2, Fuel Damage Assessment l

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. .C.MERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (for NRCreview) b PAGE F-9 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: Fuel Clad DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:

Other(Site-Specific) Indications LOSS - (Site-specific) as applicable POTENTIAL LOSS -(Site-specific) as applicable DAEC INFORMATION:

As a site-specific loss indicator, DAEC uses determination of at least 5% fuel clad damage, which is (g) consistent with the containment rad monitor reading indicators described previously. This can be determined from the appropriate fuel damage assessment procedures.

No other reliable indications of Fuel Clad " loss" or " potential loss" could be determined.

REFERENCES:

1. Post Accident Sampling and Analysis Procedure (PASAP) 7.2, Fuel Damage Assessment l

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PAGE F-10 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY l

FISSION HARRIER: Fuel Clad DAEC INDICATOR: RPV Level GENERIC INDICATOR:

Reactor Vessel Water Level LOSS - Level LESS THAN (site-specific) value POTENTIAL LOSS - Level LESS THAN (site-specific) value DAEC INFORMATION:

q There are no significant deviations from the generic indicators. The generic loss indicator is based on a site-specific) value that corresponds to the minimum value to assure core cooling without further Q (degradation of the fuel clad. DAEC uses the Minimem Steam Cooling RPV Wa This is defined to be the lowest RPV water level at which the covered portion of the reactor core will generate suflicient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500 F. Consistent with the EOPs, an indicated RPV level below -30 inches that cannot be restored is used.

The potential loss indicator corresponds to the (site-specific) water level at the top of the active fuel (TAF).

Consistent with the EOPs, an indicated RPV level below 15 inches that cannot be restored is used.

REFERENCES:

1. Emergency Operating Procedure (EOP)-1, RPV Control, Sheet 1 of 1

2. ATWS Emergency Operating Procedure (EOP)-RPV Control, Sheet 1 of 1 l 3. Emergency Operating Procedure (EOP) Basis, Curves and Limits, C5, Minimum Steam Cooling RPV Water Level i

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1 1 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD I CATEGORY  !

l FISSION BARRIER: Fuel Clad I

DAEC INDICATOR: EC/OSS Judgment GENERIC INDICATOR:

Emergency Director Judgment Any condition which in the judgment of the Emergency Director that indicates LOSS or POTENTIAL LOSS of the FUEL CLAD barrier l

l DAEC INFORMATION:

There is no significant deviation from the generic indicator. Per EPIP 7.1, Emergency Coordinator Duties, Q the Emergency Coordinator / Operations Shift Supervisor (EC/OSS) performs the emergency director ,

I function at DAEC. EC/OSS considerations for determining whether any barrier " Loss" or " Potential Loss" l include imminent barrier degradation, degraded barrier monitoring capability, and consideration of dominant accident sequences. l l 1 Imminent means that no tumaround in safety system performance is expected and that General Emergency conditions can be expected to occur within two hours. Imminent fission barrier degrajation must be l considered by the EC/OSS to assure timely declaration of a General Emergency and to beiter assure that  !

l offsite protective a::tions can be effectively accomplished. Degraded barrier monitoring capability from l loss of/ lack of reliable indicators must also be considered by the EC/OSS when determining if a fission I barrier loss or potential loss has occurred. This assessment should also include consideration for

]

instrumentation operability, portable instrumentation readings, and offsite monitoring results. Dominant accident sequences can lead to loss of all Fission Barriers. Based on the IPE, the dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite power with l early HPCI/RCIC failure. The EC/OSS should also consult System Malfunction EALs, as appropriate, to i assure timely emergency classification declaration. i l

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties O

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2. Duane Amold Energy Center Individual Plant Examination (IPE) November 1992 l

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, EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

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PAGE F-12 of 27 l FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: RCS DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:

Drywell Radiation Monitoring LOSS - < Valid > Drywell Rad Monitor Reading GREATER THAN (site-specific) R/hr l POTENTIAL LOSS - Not applicable l DAEC INFORMATION:

Valid means that the reading is from instrumentation determined to be operable in accordance with the

] Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, coolant sampling, or radiological survey results.

1 l There is no significant deviation from the generic indicator. This loss indicator is based on conditions after reactor shutdown to assure that it is not misapplied, i.e., to exclude readings due to N-16 effects which are l typically 5 to 8 R/hr at full power conditions.

l l The (site-specific) value for this loss indicator corresponds to instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the drywell atmosphere. The reading will be less than that specified for the loss indicator for Radiation / Core Damage that applies to the Fuel Clad barrier. Thus, this indicator would l be indicative of a RCS leak only. If the radiation monitor reading increased to that value specified by the l Radiation / Core indicator applying to the Fuel Clad barrier, fuel damage would also be indicated.

As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose I rates at the locations of the containment atmosphere monitoring system radiation detectors in the drywell l and torus locations for assumed releases of gap activity form the core. These calculations were based on

" nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were b evaluated. In the first case, the released activity was assumed to be contained in the drywell atmosphere.

! This case is considered representative of conditions following a line break in which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus.

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EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCrevien)

PAGE F-13 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results' for each case were provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response characteristics of the detector systems. The figures show a drywell reading of about 2.1 x 10 R/hr associated with a 100% gap release immediately after shutdown. Assuming 99.99% fuel clad integrity (0.01% gap release) and uniform dispersal of radionuclides into the drywell immediately after shutdown, a i drywell monitor reading is calculated:

l Calculation of Drywell Monitor Reading Assuming 0.01% Gap Release 4

g NG-88-0966 value for 100% Gap Release at 0.01 minutes = 2.1 x 10 R/hr V (2.1 x 10' ) R/hr x [(1 x 10-2 ) percent /100 percent] = (2.1) x 10" R/hr = 2.1 x 10 R/hr = 2 R/hr i

To assure an indicator that is readily discernible on the drywell radiation monitor scale, DAEC uses a valid reading above 5 R/hr after reactor shutdown.

REFERENCES:

1. OfIice Memo NG-88-0966, G.E. Fuel Damage Documentation / Dose Rate Calculations,03/18/88

2. Technical Specification 3.2E, Drywell Leak Detection Instrumentation l

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PAGE F-14 of 27 i FISSION PRODUCT BARRIER DEGRADATION l EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: RCS

! DAEC INDICATOR: RPV Level l ,

i GENERIC INDICATOR:

l Reactor Vessel Water Level

LOSS - Level LESS THAN (site-specific) value 1 j

POTENTIAL LOSS - Not applicable l

l DAEC INFORMATION:

l

'g There is no significant deviation from the generic indicator. This (site-specific) loss indicator corresponds l d to the water level at the top of the active fuel (TAF). Consistent with the EO j i below 15 inches that cannot be restored is used. l l

l

REFERENCES:

i  !

l 1. Emergency Operating Procedures (EOP) Basis, Breakpoints  !

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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

PAGE F-15 of 27 FISSION PRODi>,7 BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY ,

FISSION BARRIER: RCS DAEC INDICATOR: Leakage l

GENERIC INDICATOR:

l RCS Leak Rate LOSS - < Valid > (site-specific) indication of Main Steamline Break L POTENTIAL LOSS - RCS leakage GREATER THAN 50 GPM inside the d.,fwell OR unisolable i primary system leakage outside drywell as indicated by < valid > area temp or area rad monitor alarm DAEC INFORMATION:

A i V Valid means that the reading is from instrumentation datermined to be operable in accordance with the Technical Specifications or has been verified by other ,ndependent methods such as indications displayed

on the control panels, reports from plant personnel, or radiological survey results.

There are no significant deviations from the generic potential loss indicators applying to RCS leakage and indications of unisolable primary system leakage. Please note that RCS leakage inside the drywell excludes Safety-Relief Valve (SRV) discharge through the SRV discharge piping into the torus below the I

l water line. SRV leakage is addressed by SUS, RCS Leakage.

Unisolable primary system leakage outside the drywell includes leakage through portions of the main  ;

steam lines, portions of the Reactor Water Cleanup System (RWCU), and through the Scram Discharge  !

Volumes (SDVs) detected per EOP 3. It is possible to have relatively small amounts ofleakage result in radiation monitor alarms, therefore it is treated as a potential loss of the RCS barrier and loss of the

! Primary Containment barrier (see the discussion under Primary Containment Leakage indicator).

DAEC does not use the generic " loss" indicator for main steam line break. NUMARC Methodologyfor l Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers, June l 1993, discloses that the main steam line break with isolation does not have to be included as a fission barrier table indicator. This event can be appropriately classified in the System Malfunction Recognition Category. This event was classified as a RCS barrier loss indicator in the generic guidance because this event typically results in a puff release with dose consequences greater than 10 millirem whole body, i.e.,

O ofTsite dose consequences consistent with declaration of an Alert in accordance with AAl, Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer. However, UFSAR Section 15.6.6, Table 15.6-1, Steam Line Break - Radiological Effects for Puff Release at 47 Meters, Total Dose, shows a maximum l

i Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

O PAGE F-16 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY l dose of 0.58 mrem.(5.8E-04 rem) passing cloud whole body dose using conservative assumptions.

Therefore, because this event at DAEC has dose consequences similar to those of AUI, Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 2 Times Radiological Technical Specifications for 60 Minutes or Longer, it has been added as an Unusual Event EAL in SUS, RCS Leakage.

REFERENCES:

1. Alarm Response Procedure (ARP) IC04B, Reactor Water Cleanup and Recirculation

2. Alarm Response Procedure (ARP) 1C04C, Reactor Water Cleanup and Recirculation

3. Emergency Operating Procedure (EOP) 3, Secondary Containment Control

4. UFSAR Section 15.6.6, Loss-of-Coolant-Accident

5. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993 l

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FISSION BARRIER: RCS l l

DAEC INDICATOR: Primary Containment Atmosphere

GENERIC INDICATOR

! Drywell Pressure LOSS - < Valid > Pressure < Reading > GREATER THAN (site-specific) psig POTENTIAL LOSS -Not applicable DAEC INFORMATION:

! Valid means that the reading is from instrumentation detennined to be operable in accordance with the 1

[- Technical Specifications or has been verified by. other independent methods such as indications displayed I on the control panels, reports from plant personnel, or radiological survey results. l l

There is no significant deviation from the generic indicator. The (site-specific) value for this loss indicator j corresponds to the drywell high pressure ECCS initiation signal setpoint of 2.0 psig. DAEC also specifies j that drywell cooling is operating to assure that the indicator is not misapplied to conditions that do not j indicate RCS leakage into the drywell, i.e., the drywell pressure increase is not due to loss of drywell l cooling. I l DAEC uses a GE Mark I Containment. During reactor operation, with drywell cooling in operation and the drywell inerted, the normal operating pressure in the drywell is between 0.5 and 1.0 psig. Analysis at the DAEC shows that a 50 gpm RCS leak would result in a 2 to 3 psig pressure rise over a six minute time

, period. Since a 2 psig rise would place DAEC above the ECCS initiation setpoint, ( 2 psig) it is necessary i l to select the DAEC ECCS initiation setpoint of 2 psig to indicate an actual loss of the RCS. Drywell l l cooling is not isolated at the 2 psig ECCS initiation setpoint, therefor further pressure rise would be indicative of a RCS leak.

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REFERENCES:

1. Emergency Operating Procedures (EOP) Bases, Breakpoints

2. Emergency Operating Procedures (EOP)-1, RPV Control

3. Emergency Operating Procedures (EOP) -2, Primary Containment Control

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EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) l(~%0 PAGE F-18 of 27 l FISSION PRODUCT BARRIER DEGRADATION l EFFECTIVE DATE: TBD CATEGORY l FISSION BARRIER: RCS DAEC INDICATOR: Primary Containment Atmosphere GENERIC INDICATOR:

1 Emergency Director Judgment ,

Any condition which in the judgment of the Emergency Director that indicates LOSS or POTENTIAL j l LOSS of the RCS bamer ,

i DAEC INFORMATION:  !

1 l 7)

(

There is no Emergency significant Coordinator deviation

/ Operations from Shift Supervisor the performs (EC/OSS) generic EAL. Per the emergency EPIP director 7.1, Eme function  :

at DAEC. EC/OSS considerations for determining whether any barrier " Loss" or " Potential Loss" include  ;

imminent barrier degradation, degraded barrier monitoring capability, and consideration of dominant l l accident sequences. ,

Imminent means that no tumaround in safety system performance is expected and that General Emergency l conditions will occur within two hours. Imminent fission barrier degradation must be considered by the .

EC/OSS to assure timely declaration of a General Emergency and to better assure that offsite protective

( actions can be effectively accomplished. Degraded barrier monitoring capability from loss of/ lack of l reliable indicators must also be considered by the EC/OSS when determining if a fission barrier loss or l potential loss has occurred. This assessment should also include consideration for instrumentation l operability, portable instrumentation readings, and offsite monitoring results. Dominant accident sequences can lead to loss of all Fission Barriers. Based on the IPE, the dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite power with early l HPCI/RCIC failure. The EC/OSS should also consult System Malfunction EALs, as appropriate, to assure i

timely emergency classification declaration.

l For the RCS barrier, the EC/OSS should also consider safety-reliefvalves (SRVs) open or cycling. If an SRV is stuck open or is cycling and no other emergency conditions exist, an emergency declaration may not be appropriate. However, ifthefuel is damaged and the SR Vis allowingfission products to escape into f primary containment, a loss ofRCS should be determined as having occurred. The EC/OSS should also

\

consult SUS, RCS Leakage, to determine if RCS leakage exceeds the threshold required for declaration of an Unusual Event.

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

PAGE F-19 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties

2. Duane Arnold Energy CenterIndividual Plant Examination (IPE) November 1992

3. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993 1

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) ,

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PAGE F-20 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD l CATEGORY l

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FISSION BARRIER: RCS DAEC INDICATOR: None GENERIC INDICATOR:

Other(Site-Specific) Indications LOSS - (Site-specific) as applicable POTENTIAL LOSS - (Site-specific) as applicable DAEC INFORMATION:

Other indicators were also considered. No other reliable indicators of RCS barrier " loss" or " potential loss" l

( could be determined.

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REFERENCES:

None l

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l l Duane Arnold Energy Center l EMERGENCY ACTION LEVEL BASES DOCUMENT p Rev.1 (forNRCreview)

'C PAGE F-21 of 27 FISSION PRODUCT DARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: Primary Containment DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:

Significant Radioactive Inventory in Containment LOSS - Not applicable POTENTIAL LOSS - Containment Rad Monitor reading GREATER THAN (site-specific) PJhr DAEC INFORMATION:

m There is no significant deviation from the generic indicators. The " potential loss" (site-specific) indicator I value corresponds to at least 20% fuel clad damage with release into the primary containment. This j indicator corresponds to loss of both the Fuel Clad and RCS barriers with Potential Loss of the Primary Containment barrier, and would result in declaration of a General Emergency. The basis for the 20% fuel clad damage threshold is described under the 20% core damage assessment indicator. It is intended that determination ofbarrierpotentialloss be made whenever the indicator threshold is reached until such time that core damage assessment isperformed. at which time direct use ofcontainment rad monitor readings is no longer required.

As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose j rates at the locations of the containment atmospheric monitoring system radiation detectors in the drywell l and torus locations for assumed releases of gap activity form the core. These calculations were based on

" nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more i appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. 'Ihe Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were evaluated. In the first case, the released activity was assumed to be contained in the drywell atmosphere.

This case is considered representative of conditions following a line break in which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus.

This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results for each case were provided for each case in the form of gamma ray dose rate versus p time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate V calculations were canied out independent of any specific information on details of construction or response 3

characteristics of the detector s7 stems. The figures show a drywell reading of about 2.9 x 10 R/hr and a torus reading of about 1.1 x 10 R/hr associated with 20% gap release at two hours after shutdown. These

l Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) l O PAGE F-22 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY l

l l values are rounded to 3 E+3 R/hr and 1 E+2 R/hr , respectively. The two hour point was picked because it allows ample time for the Technical Support Center to be operational and core damage assessment to begin.

l

REFERENCES:

l 1. Office Memo NG-88-0966, G.E. Fuel Damage Documentation / Dose Rate Calculations,03/18/88 l

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (for NRCreview)

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' O PAGE F-23 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: Primary Containment DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:

Other(Site-Specific) Indications LOSS - (Site-specific) as applicable POTENTIAL LOSS -(Site-specific) as applicable DAEC INFORMATION:

)

As a site-specific " potential loss" indicator, DAEC uses determination of at least 20% fuel clad damage, 7

which is consistent with the level of fuel damage indicated by the drywell and torus radiation monitor  ;

readings above. This can be determined using appropriate fuel damage assessment procedures. Regardless  !

ofwhetherprimary containment integrity is challenged, it is possiblefor sigmficant radioactivity within the 1 primary containment to result in EPA PAG plume exposure levels being exceeded even assuming that the primary containment is within technical specification allowable leakage rates. With or without primary containment challenge, however, a major release of radioactivity requiring off-site protective actions from core damage is not possible unless a major failure of the fuel clad barrier allows radioactive material to be l I

released from core into the reactor coolant. NUREG-1228 indicates that such conditions do not exist when the amount of fuel clad damage is less than 20%.

Other indicators were also considered. No other reliable indicators for Primary Containment " loss" or

" potential loss" could be determined.

I

REFERENCES:

l

1. Post Accident Sampling and Analysis Procedure (PASAP) 7.2, Fuel Damage Assessment

2. NUREG-1228, Source Term Estimations During Incident Response to Severe Nuclear Power Plant 1 Accidents, October 1988 l

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EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCrevieu) l PAGE F-24 of 27 i

FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD i CATEGORY l

l l l FISSION BARRIER: Primary Containment l DAEC INDICATOR: RPV Level 1 l

GENERIC INDICATOR:

Reactor Vessel Water Level LOSS -Not applicable l POTENTIAL LOSS - <RPV> level less than (site-specific) value and <no injection source is available>

1 DAEC INFORMATION:

The underlying concem for this indicator is a threshold that represents significant uncovering of the core and imminent core damage. Imminent means that no tumaround in safety system performance would be expected and that General Emergency conditions would be expected within two hours.

Consistent with the underlying concern, the DAEC indicator addresses conditions where the water level is below the Minimum Zero-Injection RPV Water Level of-40 inches with no injection source available.

The Minimum Zero-Injection RPV Water Level is defined to be the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any fuel clad temperature in ,

the uncovered portion of the core from exceeding 1800 F. The Minimum Zero-Injection RPV Water Level is utilized to preclude significant fuel clad damage and hydrogen generation for as long as possible l when no sources of RPV makeup water are available. l Thus, for RPV water level below -40 inches, if no source of injection water was available, water levels would continue to decrease and the fuel clad temperature would be expected to continue to rise. Due to large uncertainties in severe accident progression, it should be assumed that severe core melt is imminent if this condition were to occur. It would not be acceptable to delay the declaration of the General Emergency and issuance of Protective Action Recommendations beyond this point.

i l

l

REFERENCES:

1. Emergency Operating Procedure (EOP) Bases Document, Curves and Limits O

O

2. Emergency Operating Procedure (EOP) RPV/F - RPV Flooding j

3. NUAMRC Afethodologyfor Development ofEmergency Action Levels NUAMRC/NESP-007 Revision 2 l Questions andAnswers, June 1993 l

i l

I Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRC review)

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PAGE F-25 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: Primary Containment

)

I DAEC INDICATOR: Leakage GENERIC INDICATOR:

Containment Isolation Valve Status After Containment Isolation Signal LOSS - Failure of both valves in any one line to close AND downstream pathway to the environment exists M Intentional venting per EOPs M unisolable primary system leakage outside drywell as indicated by < valid > area temp or area rad alarm POTENTIAL LOSS -Not applicable DAEC INFORMATION:

Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.

l The " loss" indicators used at DAEC directly correspond to the generic indicators. Venting of the primary containment can be performed in accordance with EOP 2 irrespective of the offsite radioactivity release rate that will occur and by defeating isolation interlocks es necessary. The consequences of not doing so may be the loss of primary containment integrity, core damage, and an uncontrolled radioactive release much greater than might otherwise occur. Primary containment venting is performed only as necessary to reduce and then maintain torus pressure below the Primary Containment Pressure Limit (PCPL) of 53 psig.

1 Unisolable primary system leakage outside the drywell includes leakage through portions of the main steam lines, portions of the Reactor Water Cleanup System (RWCU), and through the Scram Discharge Volumes (SDV's) detected per EOP 3. It is possible to have relatively small amounts ofleakage result in radiation monitor alarms, therefore it is treated as a " potential loss" of the RCS (see the discussion under RCS Barrier Leakage indicator) and " loss" of the Primary Containment.

REFERENCES:

1. Emergency Operating Procedure (EOP) 2, Primary Containment Control

2. Emergency Operating Procedure (EOP) 3, Secondary Containment Control

- 3. Emergency Operating Procedures (EOP) Bases, Breakpoints

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) l PAGE F-26 of 27 FISSION PRODUCT BARRIER DEGRADATION

! EFFECTIVE DATE: TBD l CATEGORY l

l t

FISSION BARRIER: Primary Containment DAEC INDICATOR: Primary Containment Atmosphere GENERIC INDICATOR:

Drywell Pressure LOSS - Rapid unexplained decrease following initial increase OR Drywell pressure response not consistent with LOCA conditions POTENTIAL LOSS - (site-specific) PSIG OR explosive mixture exists DAEC INFORMATION:

(n) There are no significant deviations from the generic indicators. The " loss" indicators used at DAEC directly correspond to the generic indicators.

The first " potential loss" (site-specific) indicator is torus pressure of 53 psig, which is the Primary Containment Pressure Limit (PCPL) used in the EOPs. The second " potential loss" indicator is based on determination of explosive mixture in accordance with the EOPs. DAEC EOPs require control of drywell and torus atmosphere gas concentrations to less than 6% H2 and less than 5% O2 to assure that an explosive mixture does not exist. This " potential loss" indicator is written to be consistent with the EOPs.

REFERENCES:

1. Emergency Operating Procedure (EOP) 2, Primary Containment Control

2. Emergency Operating Procedure (EOP) PCH - Primary Containment Hydrogen I

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i Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCrevieu)

PAGE F-27 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: TBD CATEGORY FISSION BARRIER: Primary Containment DAEC INDICATOR: EC/OSS Judgrnent GENERIC INDICATOR:

Emergency Director Judgment Any condition which in the judgment of the Emergen'cy Director that indicates LOSS or POTENTIAL LOSS of the RCS barrier DAEC INFORMATION:

There is no significant deviation from the generic indicator. Per EPIP 7.1, Emergency Coordinator Duties, I,) the Emergency Coordinator / Operations Shift Supervisor (EC/OSS) performs the emergency director function at DAEC. EC/OSS considerations for determining whether any barrier " Loss" or " Potential Loss" include imminent barrier degradation, degraded barrier monitoring capability, and consideration of dominant accident sequences.

Imminent means that no turnaround in safety system performance is expected and General Emergency conditions will occur within two hours. Imminent fission barrier degradation must be considered by the EC/OSS to assure timely declaration of a General Emergency and to better assure that offsite protective actions can be effectively accomplished. Degraded barrier monitoring capability from loss of/ lack of reliable indicators must also be considered by the EC/OSS when determining if a fission barrier loss or potential loss has occurred. This assessment should also include consideration for instrumentation operability, portable instrumentation readings, and offsite monitoring results. Dominant accident sequences can trad to loss of all Fission Barriers. Based on the IPE, the dominant accident sequences leading to core damage at DAEC include complete loss of 125 VDC, loss of decay heat removal, ATWS with failure of Standby Liquid Control, prolonged station blackout, and loss of offsite power with early HPCI/RCIC failure. The EC/OSS should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification declaration.

REFERENCES:

l l

1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties

/O 2. Duane Arnold Energy Center Individual Plant Examination (IPE) November 1992 b

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I EMERGENCY PLAN IMPLEMENTING PROCEDURE No. EPIP- 1.i Rev.1 PAGE l of I (ForNRCReview)

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY EFITLTIVE DATE:TBD i

EVENT TYPE UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY l

l hut HA1 Earthquake deteded per AOP 901, Earthquake peak horizontal acceleration l Earthquake. above z 0.06 Gravny.

OR OR Sede Shutdown Areas Report of tomado touchng down within plant Report of tomado strilang plant vital area. Category Ares

protected area or within switchyard. Electncal Power Swichyard,1G31 DG and Day Tank Rooms,1G21 DG OR OR and Day Tank Rooms, i Assessment by the control room that an evert Report to control room of damage affecting has occurred. Battery Rooms, Essential Swtchgear Roorns, Cable l ability to acheve or maintam safe shutdown.

OR Spreading Room OR

! Vehicle crash into plant staJctures or systems Vehicle crash affecting plant vital areas. Heat Smk/ Torus Room, intake Structure, Pumphouse NATURAL within protected area boundary Coolant Supply l DISASTERS OR OR Containment Drywen, Torus

Report of an unanticipated explosion within the Sustasned wind speed above 95 MPH. Lmoi4 NE, NW, SE Comer Rooms. HPCI Room, RCIC Room, protected area boundary resulting in visible Systems RHR Valve Room. North CRD Area. South CRD Area j damage to structures or equipment Other Control Busidmg, Remote Shutdown Panet 1C388 Area, i

OR OR Panel 1C56 Area, SBGT Room

, Turbine failure resultog in casing penetration Missiles affecting abihty to achieve or mamtain

! or damage to turbine or generator seals. safe shutdown.

OR OR \

Riverlevel above 757 feet. Riverlevelabove 767 feet. Water Level Omratog Lmts [

OR OR Max Normal Max Safe Any area watet level above Max Normat Water level above Max Safe Operating Limit Room Ar5a Indicator Operatmg Limit Owating Limit Operating Limet. In 2 or more areas AND Reactor shutdown is (inches) (inches) required. Hpc:Eforn Area LI3768 6 24 Riverlevelbelow725 eet6 inches. Riverlevelbelow724 est6 inches.

Co Room SE 6 Op.Modos: ALL Op. Mart ==:ALL ^

HU2 HA2 8 RHR Comer Room NW L13771 6 23 Fire in buildogs or areas contiguous to any of Fire or explosion in any of the following areas- A'es FIRE the followmg areas nnt extoguished within 15 . Reactor, turbine, control, admirvsecunty Torus Area LI3772 12 24 minutes of control roorn notification or e intake structure venfication of a control room alarm: . Pump house

. Reactor,turt@ onrdrol, admin /secunty AND e intake structure Systems & Equipment of Concem Affected system parameter indications show

. Pump house e Reactivdy Control degraded performance or plant personnei report visible damage to permanent structures . Cun ..non.Morus) or equipment within the speedied area. . RHR/ Core Spray /SRV's Op.Modos: ALL Op. Modes: ALL

• HPCl/RCIC HW HA3

• RHRSW/ River Water /ESW Toxic or flammable gas release affecting Toxic or flammable gas maeng safe shutdown

• Onsite AC Power /EDG's OTHER normaloperation. areas uninhabnable or inaccessible,

• Offsite AC Power HAZARDS AND OR e Inst'umert AC FAILURES Report by local, cxxinty or State official for . DC Power potentiel evacuation of site personnel based on . Remote Shutdown t'apahaty offsite evert Op. Modes: ALL Op. Mart ==:ALL HU4 HA4 HS1 Suspected sabotage device discovered within HGt Intrusson into plant PROTECTED area by a intrusion into plant VfTAL aren by a hoolfie Loss of physicalcorWrolof the Control plant PROTECTED area and odside plant vital hostile force. force. Room.

SECURITY area.

y --- - - - - - -y y Med W deVlCe M irlplert W DEVICE W h the plN1  % deVk:0 h h,10 plart W W d physiCel cortrol Of remote Shi.Adown switchyani protected area. area. capabuy Op. Modes: ALL HS2 Control room esacuston med p Controlrocyn has been evacuatedNO AOP 915, Shutdown Outside W d piant from Fkwnote Shutdown Panel ROOM None 1 NOTestabhshed withn 20 rninutes.

EVACUATION Op. Modes: ALL Op. Modes: ALL NAS HG2 Other conditions exist in thew d Other conditions exist whidi in the ludgnent

~ that syst may eM @

,,e ECoSS --

,,e - a se m,ety ,e --- a ~ and ,,Gt -m. e e i~esa - - core

- we,,,,, -a~- - , -e, i- a ECIOSS -

JUDGMENT . Petertial for uncontrolled radkmchde releases which can reasonath be expected to exceed EPA PAG plume exposure le4 outsde the site boundary Op. Modes: ALL D ****

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, >"0 D 3 Om2 RE 3 CD aa >3 4 oG OCm RE NO "E m k

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6___ _ ._ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ . . _ _ _ _ _ . _ . _ _ _ _ _ _ . . _ _ _ . . . ._ _ _ .._.

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J j HAZARDS AND OTIIER CONDITIONS AFFECTING PLANT SAFETY I CATEGORY

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) b PAGE H-1 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY l

hcl Natural and Destructive Phenomena Affecting the Protected Area EVENT TYPE: Natural Disasters, Other Hazards and Failures OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4 or 5 or 6 or 7)

1. (Site-Specific) method indicates felt earthquake.

2. Report by plant personnel of tomado striking within protected area boundary.

3. Assessment by the control room that an event has occurred.

in 4. Vehicle crash into plant structures or systems within protected area boundary.

U 5. Report by plant personnel of an unanticipated explosion within the protected area boundary resulting in visible damage to permanent structures or equipment.

6. Report of turbine failure resulting in casing penetration or damage to turbine or generator seals. i

7. (Site-Specific) occurrences.

DAEC EAL INFORMATION:

I l There are no significant deviations from the generic EALs. EAL 1 addresses earthquakes that are detected in accordance with AOP 901. For DAEC, a minimum detectable earthquake that is indicated on panel IC35 is an acceleration greater than 0.01 Gravity. DAEC EAL 2 addresses report of a tornado striking j within the protected area or within the plant switchyard. DAEC EAL 3 allows for the control room to determine that and event has occurred and take appropriate action based on personal assessment as opposed l to verification. No attempt is made to assess the actual magnitude of the damage. Such damage can be due to collision, tornadoes, missiles, or any other cause. Damage can be indicated by report to the control l room, physical observation, or by Control Room / local control station instrumentation. Such items as

scorching, cracks, dents, or discoloration of equipment or structures required for safe shutdown are i addressed by this EAL. DAEC EAL 4 addresses a vehicle (automobile, aircraft, forklifl, truck or train) l crash that may potentially damage plant structures containing functions and systems required for safe

! shutdown of the plant. This does not include vehicle crashes with each other or damage to ofF or warehouse structures. Escalation to Alert under HAl would occur if damage was sufficient to a t the ability to achieve or maintain safe shutdown, e.g., damage made required equipment inoperam e or structural damage was observed such as bent supports or pressure boundary leakage.

i l DAEC EAL 5 addresses explosions within the protected area. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts

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pi jgifECTIVE DATE. TilD yl ,l { 1

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[chnwl{ station instrumentation (orl structures required for safe shutdown are

'sectdity aspects of the explosion ,

e.

if applicabl e turbine casing or to the seals of a#

w ad "

elevaji iThese'c6ncerns g-Ute hiat sink. DAEC EAL 7 include extern protecti AOR19028PIMisite' finished grade is at

'~

waterlent MofneaNgradd;would require

$ ZEltR/useia dufdold of flood DAEC EAL activity mishNk hhi5C'l

@@N6@%&Q[J- - [;. . ,

EAL is based onY component failures

~

The Maximum Nhnsis expected to occur during

$la567hereide/thikkh

[hperating)ithih systems functioning prok Containment Control and 'm(ayM fiedlparameterg%

control Suh -

p secondary containment. ExceedihFth&

^* , A degradation in the level of the safety $f- [

emergency classification. The maximte no 1

EOP 3 and are shown in the table 7 below:

d, v

Maximum Operating L AtTected IAcation IIPCI Room Area Indicator M RCIC Room Area ~ LI3768 f6

~

LI3769 A RHR Corner Room SE Area LI 3770 6 B RHR Corner Room NW Area LI 3771 6 inche orus Area 6 inches : < ~

LI 3772 12 inches N jy 1#

EAL 9 addresses the efTects of low supply systems (river water RllR service water and em The intake structurd river water level. _ ); * ~ *'

m ergency service water)is' [

($ - __-_

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

O PAGE H-2 of 25

HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE

TBD PLANT SAFETY CATEGORY l

significant energy to near-by structures or equipment. Damage can be indicated by report to the control room, physical observation, or by Control Room / local control station instrumentation. Such items as scorching, cracks, dents, or discoloration of equipment or structures required for safe shutdown are addressed by this EAL. The EC/OSS needs to consider the security aspects of the explosion, if applicable.

DAEC EAL 6 addresses turbine failure causing observable damage to the turbine casing or to the seals of the generator.

EALs 7 through 9 address site-specific occurrences of concem. These concems include external flood water levels, intemal flooding, and low river water level affecting the ultimate heat sink. DAEC EAL 7 l addresses the observed effects of flooding in accordance with AOP 902. Plant site finished grade is at elevation 757.0 ft. Personnel doors and railroad and truck openings at or near grade would require -

protection in the event of a flood above elevation 757.0 ft. Therefore, EAL 7 uses a threshold of flood waterlevels above 757.0 ft.

t DAEC EAL 8 addresses intemal flooding can be due to system malfunctions, component failures, or repair activity mishaps (such as failed freeze seal) that can threaten safe operation of the plant. Therefore, this EAL is based on a valid indication that the water level is higher than the maximum normal operating limits.  :

The Maximum Normal Operating Limits are defined as the highest values of the identified parameter expected to occur during nonnal plant operating conditions with all directly associated support and control l systems functioning properly. Exceeding these limits is an entry condition into EOF 3, Secondary ,

l Containment Control and may be an indication that water from a primary system is discharging into secondary containment. Exceeding the maximum normal operating limit is interpreted as a potential ,

degradation in the level of the safety of the plant and is appropriately treated as an Unusual Event i emergency classification. The maximum normal operating water level limits are taken from AOP 902 and 1

, EOP 3 and are shown in the table below:

l Maximum Operating Limits - Water Levels

! AfTected Location Indicator Maximum Normal OL Maximurn Safe OL HPCI Room Area LI3768 6 inches 24 inches  ;

RCIC Room Area LI 3769 6 inches 18 inches A RHR Corner Room SE Area LI 3770 6 inches 23 inches B RHR Corner Room NW Area LI3771 6 inches 23 inches Torus Area L13772 12 inches 24 inches O EAL 9 addresses the effects of low river water level. The intake structure for the safety-related water supply systems (river water, RHR service water, and emergency service water) is located on the west bank l

Duane Arnold Energy Center l EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) l l PAGE H-3 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING l

EFFECTIVE DATE

TBD  :

l PLANT SAFETY CATEGORY l

l l of the Cedar River. An overflow-type barrier across the river was designed and constructed in accordance

! with Seismic Category I criteria to intercept the streambed flow and divert it to the intake structure. This makes the entire flow of the river available to the safety-related water supply systems. A minimum flow of 13 cubic feet per second (cfs) from a minimum 1000-year river flow of 60 cfs must be diverted. The top of the barrier wall is at elevation 725 ft. 6 in. River water level below this level represents a potential degradation in the level of safety of the plant and is addressed by EAL 9.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 901, Earthquake l 2. Abnormal Operating Procedure (AOP) 902, Flood l 3. Abnormal Operating Procedure (AOP) 903, Tomado

4. Emergency Operating Procedure (EOP)-3, Secondary Containment Contial

5. EOP Basis Document, EOP-3, Secondary Containment Control

6. UFSAR Chapter 3, Design of Structures, Components, Equipment, and Systems

7. Bechtel Drawing BECH-M017, Equipment Location - Intake Structure Plans at Elevations, Rev. 6 q

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

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PAGE H-4 of 25 l

HAZARDS AND OTHER CONDITIONS AFFECTING l EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY l

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l HU2 Fire Within Protected Area BoundaryNot Extinguished Within 15 Minutes of Detection EVENT TYPE: Fire OPERATING MODE APPLICABILITY: All l EXAMPLE EMERGENCY ACTION LEVEL:

1. Fire in buildings or areas contiguous to any of the following (site-specific areas) areas not extinguished within 15 minutes of control room notification or verification of a control room alarm:

r

()y . (Site-specific) list DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL. The purpose of this EAL is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems.  !

This includes such items as fires within the administration building, and security building (buildings j r 'iguous to the reactor building, turbine building and control building), yet, excludes fires in the

! -house or construction support center, waste-basket fires, and other small fires of no safety ensequence. ,

Per AOP 913, the location of a fire can be determined by observing XL3 alarm messages, Zone Indicating Unit (ZIU) alarms, or fire annunciators on panels IC40 and IC40A. The location of a fire can also be  !

determined by verbal report of the person discovering the fire. Venfication of the alarm in this context  !

means those actions taken to determine that the control room alarm is not spurious.

l l

REFERENCES:

l. Abnormal Operating Procedure (AOP) 913, Fire l 2. Abnormal Operating Procedure (AOP) 914, Security 1

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Duane Arnold Energy Center i EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) I I

PAGE H-5 of 25 4 l

1 IIAZARDS AND OTIIER CONDITIONS AFFECTING 1 EFFECTIVE DATE: TBD i PLANT SAFETY CATEGORY l t

l l HU3 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation l of the Plant j

EVENT TYPE: Other Hazards and Failures I 1

! OPERATING MODE APPLICABILITY: All l

EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)

1. Report or detection of toxic or flammable gases that could enter within the site area boundary in amounts that can afTect normal operation of the plant.

1 p h 2. Report by Local, County or State Officials for potential evacuation of site personnel based on offsite event.

l DAEC EAL INFORMATION:

i There is no sigmficant deviation from the generic EALs. This IC is based on releases in concentrations wit} n the site boundary that will affect the health of plant personnel or affecting the safe operation of the 1 plant with the plant being within the evacuation area of an offsite event (i.e., tanker truck accident releasing l

toxic gases, etc.) The evacuation area is as determined from the DOT Evacuation Tables for Selected Hazardous Materials, in the DOT Emergency Response Guide for Hazardous Materials.

For the purposes of this IC, CO2(such as is discharged by the fire suppression system) is not toxic. CO 2 )

l- can be lethal ifit reduces oxygen to low concentrations that are immediately dangerous to life and health  ;

(IDLH). CO discharge 2 into an area is not basisfor emergency classification under this IC unless: (1) l Access to the affected area is required, and (2) CO2 concentration results in conditions that make the area uninhabitable or inaccessible (i.e., IDLH).

REFERENCES:

1 4

l

1. UFSAR Section 2.2, Nearby Industrial, Transportation, and Military Facilities q 2. UFSAR Section 6.4,IIabitability Systems O

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Duane Arnold Energy Center l EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

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l PAGE H-6 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD l PLANT SAFETY CATEGORY l

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HU4 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant EVENTTYPE: Security OPERATING MODE APPLICABILITY: All l EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)

. i

1. Bomb device discovered within plant Protected Area and outside the plant Vital Area.

lUA 2. Other security events as determined from (site-specific) Safeguards Contingency Plan.

DAEC EAL INFORMATION:

There is no significant deviation from the generic EALs. Security events which do not represent at least a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. The term " suspected sabotage device" is used in place of " bomb device" for consistency with the DAEC Safeguards Contingency Plan.

l i

Other (site-specific) security events of concern at DAEC include discovery of a suspected sabotage device l in the plant switchyard, which is located outside the protected area.

Suspected sabotage devices discovered within the plant Vital Area would result in escalation via other ,

Security Event ICs. l

REFERENCES:

1. Abnormal Operating Procedure (AOP) 914, Security Events O

l

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l Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCrevieu) l PAGE Il-7 of 25 I

IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY '

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' j HUS Other Conditions Existing Which in the Judgment of the <EC/OSS> Warrant Declaration of an Unusual Event EVENT TYPE: EC/OSS Jed ,f ment  ;

l OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:

! 1. Other conditions exist which in the judgment of the Emergency Director indicate a potential

degradation of the level of safety of the plant.

O i V DAEC EALINFORMATION:

There is no significant deviation from the generic EAL.

Per EPIP 7.1, the Emergency Coordinator / Operations Shift Supervisor (EC/OSS) is the title for the emergency director function at DAEC. The EAL addresses conditions that fall under the Notification of Unusual Event emergency classification description contained in NUREG-0654, Appendix 1 that is retained under the generic methodology.

REFERENCES:

i

1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties

2. NUREG-0654/ FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, October 1980, Appendix 1 l

I Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) l PAGE H-8 of 25 HAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY HA1 Natural and Destructive Phenomena Affecting the Plant Vital Area EVENT TYPE: Natural Disasters, Other Hazards and Failures OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4 or 5 or 6 or 7)

1. (Site-Specific) method indicates Seismic Event greater than Operating Basis Earthquake (OBE).

2. Tomado or high winds striking plant vital areas: Tomado or high winds greater than (site-specific) mph strike within protected area boundary.

p 3. Report of any visible structural damage on < site-specific structures >

4. (Site-Specific) indications in the control room.

() 5. Vehicle crash affecting plant vital areas.

6. Turbine failure generated missiles result in any visible structural damage to or penetration of any of the following < site-specific areas >

7. (Site-Specific) occurrences.

DAEC EAL INFORMATION:

There are no significant deviations from the generic EALs. For the events ofconcern here, the key issue is not the wind speed, earthquake intensity, etc., but whether there is resultant damage to equipment or structures required to achieve or maintain safe shutdown, regardless of the cause. Determination of damage affecting the ability to achieve or maintain safe shutdown can be indicated by reports to the control room, physical observation or by Control Room / local control station instrumentation.

EAL 1 addresses OBE events that are detected in accordance with AOP 901. For DAEC, the OBE is associated with a peak horizontal acceleration of + 0.06 Gravity. DAEC EAL 2 addresses report of a tomado striking a plant vital area. DAEC EAL 3 addresses a report to the control room of damage affecting the ability to achieve or maintain safe shutdown. The reported damage can be from tomadoes, high winds, flooding, missiles, collisions, or any other cause.

DAEC EAL 4 addresses vehicle (automobile, aircraft, forklift, truck or train) confirmed crashes affecting

) plant vital areas. This does not include vehicle crashes with each other or damage to office or warehouse structures. DAEC EAL 5 addresses sustained high wind speeds as measured by the 33-Foot or 156-Foot elevations on the Meteorological Tower. Sustained wind speed means the baseline wind speed measured by meteorological tower that does not include gusts. The design basis wind speed is 105 miles per hour.

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PAGE H-9 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD

PLANT SAFETY CATEGORY l

l However, the meteorological instrumentation is only capable of measuring wind speeds up to 100 miles per hour. Thus the alert level for sustained high wind speed,95 miles per hour, is selected to be on-scale for the meteorological instrumentation and to conservatively account for potential measurement errors. DAEC EAL 6 addresses missiles affecting the ability to achieve or maintain safe shutdown. Such missiles can be from any cause, e.g., tomado-generated; turbine, pump or other rotating machinery catastrophic failure; or l generated from an explosion.

Per AOPs 913 and 914, the following areas are identified as safe shutdown areas and are shown on the EAL tables. This table is displayed as an aid to the Emergency Coordinator in determining appropriate l areas ofconcern.

l Safe Shutdown Areas l(n) Category Electrical Area Switchyard,1G31 DG and Day Tank Rooms,1G21 DG and Day Tank Rooms, Power Battery Rooms, Essential Switchgear Rooms Cable Spreading Room Heat Sink / Torus Room, Intake Structure, Pumphouse Coolant Supply Containment Drywell, Torus Emergency NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Valve Room, North Systems CRD Area, South CRD Area Other Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C56 Area, SBGT Room DAEC EALs 7,8, and 9 address site-specific occurrences of concem. These concerns include extemal flood water levels, intemal flooding, and low river water level affecting the ultimate heat sink. DAEC EAL 7 addresses river water levels exceeding design flood water levels. All Seismic Category I structures and non-seismic structures housing Seismic Category I equipment are designed to withstand the hydraulic l head resulting from the " maximum probable flood" to which the site could be subjected. The design flood water is at elevation 767.0 ft Major equipment penetrations in the exterior walls are located above l elevation 767.0 ft. Openings below the flood level are either watertight or are provided with means to control the inflow of water in order to ensure that a safe shutdown can be achieved and maintained.

O Consideration has also been given to providing temporary protection for openings in the exterior walls up

() to flood levels of 769.0 ft All buildings were also checked for uplift (buoyancy) for a flood level at i elevation 767.0 ft, and the minimum factor of safety used was 1.2. Therefore, DAEC EAL 7 uses as its threshold flood water levels above 767 feet.DAEC EAL 8 addresses intemal flooding consistent with the

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PAGE H-10 of 25 i

HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY i

l requirements of EOP 3, Secondary Containment Control. If RPV pressure reduction will have no effect on l leakage into secondary containment, then EOP 3 requires that reactor shutdown be performed in ,

accordance with Integrated Plant Operating Instruction (IPOI) 3,4, or 5 as necessary if the water level l

l exceeds its maximum safe operating limits in two or more areas. If RPV pressure reduction will decrease l

l leakage into secondary containment then this is due to leakage from the primary system, which is  !

addressed by the Fission Barrier Table indicators and System Malfunction EALs, and is not addressed here.

l Therefore, EAL 8 addresses conditions in which water level in two or more areas is above Maximum Safe l Operating Limits and reactor shutdown is required.

l Required means that the reactor shutdown was procedurally mandated by EOP 3 and is not merely performed as a precaution or inadvertently. Maximum Safe Operating Limits are defined as the highest parameter value at which neither (1) equipment necessary for safe shutdown of the plant will fail nor (2)

() personnel access necessary for the safe shutdown of the plant will be precluded. The intemal flooding can j be due to system malfunctious, component failures, or repair activity mishaps (such as failed freeze seal)  !

that can threaten safe operation of the plant. This includes water intrusion on equipment that is not designed to be submerged (e.g., motor control centers).

The maximum safe operating water level limits are taken from EOP 3 and are shown on the table below:

l Maximum Operating Limits - Water Levels  !

l AfTected Location Indicator Maximum Normal OL Maximum Safe OL l HPCI Room Area LI3768 6 inches 24 inches RCIC Room Area LI 3769 6 inches 18 inches l l A RHR Corner Room SE Area LI3770 6 inches 23 inches l B RHR Comer Room NW Area LI3771 6 inches 23 inches Torus Area L13772 12 inches 24 inches l DAEC EAL 9 addresses the effects of low river water level. The intake structure for the safety-related l water supply systems (river water, RHR service water, and emergency service water) is located on the west l bank of the Cedar River. The overflow weir is at elevation 724 feet 6 inches. River level at or below this

elevation will result in all river flow being diverted to the safety related water supply systems. The top of the intake 3,tmeture around the pump wells is at elevation 724 feet. If the river water level dropped to this n level, the purap suction would have no continuous supply. Therefore, this EAL uses a threshold of water V level below 724 feet 6 inches as a potential substantial degradation of the ultimate heat sink capability.

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HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY I

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REFERENCES:

1. Abnormal Operating Procedure (AOP) 901, Earthquake

2. Abnormal Operating Procedure (AOP) 902, Flood

3. Abnormal Operating Procedure (AOP) 903, Tornado

4. Abnormal Operating Procedure (AOP) 913, Fire

5. Abnormal Operating Procedure (AOP) 914, Security Events

6. UFSAR Chapter 3, Design of Structures, Components, Equipment, and Systems

7. Bechtel Drawing BECH-M017, Equipment Location - Intake Structure Plans at Elevations, Rev. 6

8. EOP Basis Document, EOP 3 - Secondary Containment Control

9. Emergency Operating Procedure (EOP) 3, Secondary Containment Control O

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IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY l

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i HA2 Fire Affecting the Operability of Plant Safety Systems Required to Establish

or Maintain Safe Shutdown

! l EVENTTYPE: Fire l

l OPERATING MODE APPLICAHILITY: All l

EXAMPLE EMERGENCY ACTION LEVEL: l l

! 1. The following conditions exist: )

a. Fire or explosion in < site-specific areas > l l AND j lV b. Affected system parameter indications show degraded performance or plant personnel report visible damage to permanent structures or equipment within the specified area.

DAEC EAL INFORMATION:

1 t

There is no significant deviation from the generic EAL. Of particular concem for this EAL are fires that may be detected in the reactor building, control building, turbine building, pumphouse, and intake structure l as shown in Tabs 1 and 3 of AOP 913. Damage from fire or explosion can be indicated by physical l observation, or by Contrc' 'toom/ local control station instrumentation No attempt is made in this EAL to assess the actual magnitude ofthe damage.

l Per AOP 913, the location of a fire can be determined by observing XL3 alarm messages, Zone Indicating j

Unit (ZIU) alarms, or fire annunciators on panels IC40 and 1C40A.

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

PAGE H-13 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY This table is displayed as an aid to the Emergency Coordinator in determining appropriate areas of concern.

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Systems & Equipment of Concem j e Reactivity Control I e Containment (Drywellfforus) e RHR/ Core Spray /SRV's I e HPCI/RCIC l e RHRSW/ River Water /ESW e Onsite AC Power /EDG's

. Offsite AC Power g e Instrument AC V e DC Power

. Remote Shutdown Capability NOTE:  !

Scope of Systems and Equipment of concern established by review of Appendix R Safe Shutdown credited I systems. Only those systems directly affecting safe shutdown or heat removal are listed for consideration, i due to fire damage. Support Systems and equipment such as HVAC and specific instrumentation, while included in Appendix R analysis is not considered an immediate threat to the ability to shutdown the plant and remove decay heat.

With regard to explosions, only those explosions ofsuficientforce to damage permanent structures or identified equipment requiredfor safe operation. should be considered. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to near-by structures and materials. The occurrence of the explosion with reports of evidence of damage (e.g., deformation, scorching) is sufficient for the declaration. The EC/OSS also needs to consider any security aspects ofthe explosions, ifapplicable.

Per the UFSAR, the control room is the only area that is required to be continuously occupied to achieve and maintain safe shutdown following design basis accidents. However, the capability exists for plant shutdown from outside the main control room in the event that the control room becomes uninhabitable. If

) the control room becomes uninhabitable, remote shutdown panel IC388 is utilized in accordance with AOP 915.

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCrevieu)

O PAGE H-14 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY

REFERENCES:

1. Abnormal Operating Procedure (AOP) 913, Fire

2. Abnormal Operating Procedure (AOP) 914, Security Events

3. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room

4. UFSAR Section 6.4, Habitability Systems l

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PAGE H-15 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY HA3 Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown EVENT TYPE: Other Hazards and Failures OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)

1. Report or detection of toxic gases within a Facility Structure in concentrations that will be life threatening to plant personnel.

O)

's 2. Report or detection of flammable gases within a Facility Structure in concentrations that will affect the safe operation of the plant.

D 1EC EAL INFORMATION:

l There is no significant deviation from the generic EALs. This IC, in addition to IC HA5 below, also l addresses entry of toxic gases that may result in control room evacuation in accordance with AOP 915.

For the purposes of this IC, CO2(such as is discharged by the fire suppression system) is not toxic. CO2 l can be lethal ifit reduces oxygen to low concentrations that are immediately dangerous to life and health (IDLH). CO, discharge into an area is not basisfor emergency classification under this IC unless: (1)

Access to the afected area is required, and (2) CO2 concentration results in conditions that make the area uninhabitable or inaccessible (i.e., IDLH).

Per the UFSAR, the control room is the only area that is required to be continuously occupied to achieve and maintain safe shutdown following design basis accidents. However, the capability exists for plant shutdown from outside the main control room in the event that the control room becomes uninhabitable. If l the control room becomes uninhabitable, remote shutdown panel IC388 is utilized to achieve and maintain cold shutdown.

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PAGE H-16 of 25 l HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY l l

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Per AOPs 913 and 914, the following areas are identified as safe shutdown areas. This table is displayed as j l an aid to the Emergency Coordinator in determimng appropriate areas ofconcern. I l l Safe Shutdown Areas )

Category Area j j Electrical Power Switchyard, IG31 DG and Day Tank Rooms, IG21 DG and Day Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Sink / Coolant Supply Torus Room, Intake Structure, Pumphouse Containment Drywell, Torus Emergency Systems NE, NW, SE Corner Rooms, HPCI Room, RCIC Room, RHR Valve Room, North CRD Area, South CRD Area l (-- Other Control Building, Remote Shutdown Panel IC388 Area, Panel IC56 Area, l '( SBGT Room l

l

REFERENCES:

l

1. Abnormal Operating Procedure (AOP) 913, Fire

! 2. Abnormal Operating Procedure (AOP) 914, Security Events

3. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room

4. UFSAR Section 6.4, Habitability Systems l

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IIAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD l PLANT SAFETY CATEGORY l

l HA4 Security Event in a Plant Protected Area EVENTTYPE: Security i

l OPERATING MODE APPLICABILITY: All j i

l EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)

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

l l 2. Other security events as determined from (site-specific) Safeguards Contingency Plan.

i

' hs DAEC EAL INFORMATION:

l

~

There is no significant deviation from generic EALs.

This class of security events represents an escalated threat to plant safety above that contained in the Unusual Event. For the purposes of this EAL a civil disturbance which penetrates that protected area boundary can be considered a hostileforce. Under this EAL, adversaries within the protected area are not yet affecting nuclear safety systems, engineered safety features, or reactor shutdown capability that are l located within the vital area. Intrusion into a vital area by a hostile force will escalate the event to a Site Area Emergency.

REFERENCES:

1 l 1. Abnormal Operating Procedure (AOP) 914, Security Events I

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[V t PAGE H-18 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY l

l HA5 Control Room Evacuation Has Been Initiated l l

EVENT TYPE: Control Room Evacuation OPERATING MODE APPLICABILITY: All 1 l

EXAMPLE EMERGENCY ACTION LEVEL:

1. Entry into (site-specific) procedure for control room evacuation. I l

DAEC EAL INFORMATION:

G i l Q, There is no significant deviation from the generic EAL. The applicable procedure for control room l evacuation at DAEC is AOP 915.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room

2. UFSAR Section 6.4, Habitability Systems l

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l l Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

(G,,.h l PAGE H-19 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY HA6 Other Conditions Existing Which in the Judgment of the <EC/OSS> Warrant

, Declaration of an Alert l

EVENT TYPE: EC/OSS Judgment OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:

1. 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.

O

( ,/ DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL.

Per EPIP 7.1, the Emergency Coordinator / Operations Shift Supervisor (EC/OSS) is the title for the emergency director fimetion at DAEC. The EAL addresses conditions that fall under the Alert emergency classification description contained in NUREG-0654, Appendix 1.

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties

2. NUREG-0654fFEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support ofNuclear Power Plants, Revision 1, October 1980, Appendix 1 l

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PAGE H-20 of 25 t

HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY l

l HS1 Security Event in a Plant Vital Area EVENTTYPE: Security OPERATING MODE APPLICABILITY: All l EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)

1. Intrusion into plant vital area by a hostile force.

t

2. Other security events as determined from (site-specific) Safeguards Contingency Plan.

DAEC EAL INFORMATION:

There is no significant deviation from generic EAL 1.

This class of security events represents an escalated threat to plant safety above that contained in HA4, l l Security Event in a Plant Protected Area, in that a hostile force has progressed from the Protected Area to l the Vital Area. Under the condition ofconcern here, the adversaries are considered to be in a position to directly and negatively affect nuclear safety systems, engineered safetyfeatures, or reactor shutdown capability.

REFERENCES:

t

1. Abnormal Operating Procedure (AOP) 914, Security Events l

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PAGE H-21 of 25 I i

HAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY l

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l HS2 Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established j EVENT TYPE: Control Room Evacuation l OPERATING MODE APPLICABILITY: All j l

l EXAMPLE EMERGENCY ACTION LEVEL:

1. The following conditions exist:

a. Control room evacuation has been initiated.

! O AND (f b. Control of the plant cannot be established per (site-specific) procedure within (site-specific) minutes.

DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL. The applicable procedure for control room j evacuation at DAEC is AOP 915. Based on the results of the analysis described below, DAEC uses 20 l

! minutes as the site-specific time limit for establishing control of the plant. DAEC has satellite panels i

associated with the remote shutdown panel at various locations through out the plant. It physically takes an operator longer than 15 minutes to lincup all the controls at the various panels. Control of the plant from outside the control room is assumed when the controls are transferred to remote shutdown panel IC388 in accordance with AOP 915.

The EC/OSS is expected to make a reasonable, informedjudgment within the 20 minute time limit that l control ofthe plantfrom the remote shutdown panel has been established. The intent of the EAL is that control ofimportant plant equipment and knowledge ofimportant plant parameters has been achieved in a timely manner. Primary emphasis should be placed on those components and instruments that provide protection of and information about safety functions. At a minimum, consistent with the Appendix R safe shutdown analysis described above, these safety functions include reactivity control, maintaining reactor water level, and decay heat removal.

General Electric performed analyses to demonstrate compliance with the requirements of 10 CFR 50 Appendix R for DAEC. The evaluation of Reactor Coolant Inventory was performed using the GE evaluation model (SAFE). The SAFE code determines if the reactor coolant inventory is above the TAF during the safe shutdown operation. If core uncovery occurs, the fuel clad integrity evaluation is performed i

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Duane Amold Ene.gy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

O PAGE H-22 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY by determining the duration of the core uncovery and the resulting peak cladding temperature (PCT). The PCT calculations were performed by incorporating the SAFE output into the Core Heatup Analysis code (CHASTE). The details of these calculations are provided in Section 4 of the final report for DAEC Appendix R analyses (" Safe Shutdown Appendix R Analyses for Duane Amold Energy Center", MDE 036).

The required analyses include evaluation of the safe shutdown capability of the remote shutdown system for various control room fire events assuming: (1) no spurious operation of equipment, (2) spurious operation of a safety-relief valve (SRV) for 20 minutes, (3) spurious operation of a SRV for 10 minutes, and (4) spurious leakage from a one-inch line. The analyses show that the worst case spurious operation of SRV or isolation valves on a one-inch liquid line (high-low pressure interface) will not affect the safe shutdown ability of the remote shutdown system for DAEC in case of a fire requiring control room

evacuation before the identified time limit for the necessary operator actions at the auxiliary shutdown panels. For the limiting cases of worst case spurious leakage from a one-inch line and spurious operation of a SRV, operator control within 20 minutes would not impact the integrity of the fuel clad, the reactor pressure vessel, and the primary containment.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room

2. General Electric Report MDE-44-0386, Safe Shutdown Appendix R Analysisfor DAEC, March 1986

3. UFSAR Section 6.4, Habitability Systems

4. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993 U

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t Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

( PAGE H-23 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY HS3 Other Conditions Existing Which in the Judgment of the <EC/OSS> Warrant l Declaration of Site Area Emergency l EVENT TYPE: EC/OSS Judgment 1

OPERATING MODE APPLICABILITY: All i

EXAMPLE EMERGENCY ACTION LEVEL:

1. Other conditions exist which in the Judgment of the Emergency Director indicate actual or likely major j failures of plant functions needed for protection of the public. l DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL.

Per EPIP 7.1, the Emergency Coordinator / Operations Shift Supervisor (EC/OSS) is the title for the )

emergency director function at DAEC. The EAL addresses conditions that fall under the Site Area Emergency classification description contained in NUREG-0654, Appendix 1.

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties

2. NUREG-0654fFEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency i 1

Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, October 1980, Appendix 1 l

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PAGE H-24 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD ,

PLANT SAFETY CATEGORY '

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HG1 Security Event Resulting in Loss Of Ability to Reach and Maintain Cold Shutdown EVENTTYPE: Security l

OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2) l

1. Loss of physical control of the control room due to security event.

j q 2. Loss of physical control of the remote shutdown capability due to security event.

!V DAEC EAL INFORMATION:

There are no significant deviations from the generic EALs. The EALs encompass conditions under which a hostile force has taken physical control of vital area required to reach and maintain safe shutdown. This also includes areas where any switches that transfer control of safe shutdown equipment to outside the control room are located.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 914, Security Events

2. UFSAR Section 6.4, Habitability Systems O

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.I 6Br NRCreview l

PAGE H-25 of 25 h IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: TBD PLANT SAFETY CATEGORY HG2 Other Conditions Existing Which in the Judgment of the <EC/OSS> Warrant

, Declaration of General Emergency EVENT TYPE: EC/OSS Judgment OPERATING MODE APPLLABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:

1. Other conditions exist which in the Judgment of the Emergency Director indicate: (1) actual or l

imminent substantial core decradation with potential for loss of containment, or (2) potential for

,q uncontrolled radionucNe releases. These releases can reasonably be expected to exceed EPA PAG C' plume exposure levels outside the site boundary.

DAEC EAL INFORMATION:

! There is no significant deviation from the generic EAL.

Per EPIP 7.1, the Emergency Coordinator / Operations Shift Supervisor (EC/OSS) is the title for the emergency director function at DAEC. The EAL addresses conditions that fall under the General Emergency classification description contained in NUREG-0654, Appendix 1 and is consistent with FG1, Loss of Any Two Barriers AND Potential Loss of Third Barrier, and AGl, Site Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1000 mrem TEDE or 5000 mrem CDE Thyroid for the Actual or Projected Duration of the Release.

REFERENCES:

l

1. Emergency Plan Implementing Procedure (EPIP) 7.1, Emergency Coordinator Duties

2. NUREG-0654/ FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision l, October 1980, Appendix 1

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SYSTEM MALFUNCTION EFFECTIVE DATE:TBD EVENT TYPE UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY '

SUt sat SS1 SGI '

Loss of Offsite Power Lasting More Loss of Voltage on Buses 1 A3 and 1 A4 Loss of Voltage on Buses 1 A3 and 1 A4 Loss of Voltage on Buses 1 A3 and 1 A4 and Then 15 Minutes, lasting more than 15 minutes. lasting more than 15 minutes.

ANYof thefollowing [

. Restoration of power to either Bus 1 A3 or l Op. Modes: ALL Op. Modes: Cold S/D, Refuel, Op. Modes: Run.Startup, Hot S/D 1A4 is NOT likely within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Defueled . RPVlevelindeterminate SU7 .

Unplanned Loss of Div 1 and Div 2 RPV Level below-30 inches. i SAS '

LOSS OF POWER 125 VDC to seguired busses based on Only one AC power source ramains i bus voltage less than 105 VDC available to supph Bes 'A3 or Bus  !

Indicated. 1 A4 AND if it is lost, a station Blackout AND will occur.

Failure to restore power to at least one SS3 required 125 VDC bus within 15 minutes Complete Loss of 125 VDC Lasting More from time ofloss. Then 15 Minutes.

Op. Modes: Cold S/D, Refuel Op. Modes: Run, Stortup, Hot S/D Op. Modes: Run, Startup, Hot S/D Op. Modes: Run, Startup, Hot S/D y SA2 SS2 SG2 Failure of automatic scram. l Failure of automatic and manual scram Entry into ATWS EOP- RPV Controlis AND required i Power remains above 5% AND RPS FAILURE None OR RPV level cannot be maintained above -30 Boron injectxxt required. inches.

OR EOP Graph 4 Heat Capacity Limit is exceeded ,

Op. Modes: Run, Startup Op. Modes: Run, Startup SU2 SA3 SS4 Plant NOT brought to regured mode Loss of decay heat removal systems EOP Graph 4 Heat Capacity Umit is within applicable LCO Action Statement required to me!ntain cold shutdown. exceeded .

Time Umits. AND OR '

Temperature rise that exceeds 212* F. Reactor CANNOT be brought OR subcritical.

INABILITY TO MAINTAIN Uncontrolled temperature rise SHUTDOWN CONDITIONS approaching 212'F. Op. Modes: Run, Stortup, Hot S/D, See Fission Barrier Table SSS NO cooling method lined up or available AND RPVLevelbelow 15 inches. j Op. Modes: Run, Startup, Hot S/D Op. Modes: Cold S/D, Refuel Op. Modes: Cold S/D, Refuel SU3 SA4 SS6 Unplanned loss of most annunciators Unplanned loss of most annunciators Significant transient in progress and on panels 1C03,1C04 and 1C05 on penets 1C03,1C04 and 1C05 BOTH of the following:

lasting more than 15 minutes AND lasting more than 15 minutes and . Loss of annunciators on panels w..vis.setory non-alarming indications EITHER: 1C03,1C04 and 1C05 AND  ;

are available. . Significant transient in progress. . Loss of compensatory non-  ;

e Loss of compensatory non- alarming indications.

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l Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) v PAGE S-1 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD ,

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SUI Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes EVENT TYPE: Loss of Power i

OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:

l l 1. The following conditions exist:

a. Loss ofpower to (site-specific) transformers for greater than 15 minutes.

AND

! b. At least (site-specific) emergency generators are supplying power to emergency busses.

DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL. This event is aprecursor ofa more serious Station Blackout condition and is thus considered as a potential degradation ofthe level ofsafety ofthe plant. It is possible to be operating within Technical Specification LCO Action Statement time limits and make a l declaration ofan Umtsual Event in accordance with this E.4L.

l l Under the conditions of concern, entr/ ntoi AOP 301, Loss of Essential Electrical Power, would be made under Tab 3, Loss of Offsite Power. Indications / alarms related to loss of offsite AC power are displayed l on control room panel IC08 and are hsted in the procedure under " Probable Indications." Under these conditions, Essential 4160V Buses l A3 and 1 A4 would indicate zero volts until A diesel generator 1G-31 4kV breaker 1 A311 and B diesel generator IG-214kV breaker l A411, respectively, close for each bus.

REFERENCES:

l

1. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power

2. UFSAR Section 8.2, OITsite Power System

! 3. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993 rh r

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PAGE S-2 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SU2 Inability to Reach Required Shutdown Within Technical Specification Limits l EVENT TYPE: Inability to Maintain Shutdown Conditions OPERATING MODE APPLICAfMLITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:

1. Plant is not brought to required operating mode within (site-specific) Technical Specifications LCO Action Statement Time.

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l DAEC EAL INFORMATION:

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There is no sigmficant deviation from the generic EAL. LCO Action Statement time limits for placing the

' (\ unit in the required OPCON are provided in the DAEC Technical Specifications, j l

An immediate Notification of an Unusual Event is required when the plant is not brought to the required OPCON within the Technical Specifications LCO Action Statement time limits. Declaration of an Unusual Event is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed.

REFERENCES:

1. DAEC Technical Specifications l l

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SYSTEM MALFUNCTION CATEGORY l EFFECTIVE DATE: TBD 1 i

SU3 Unplanned Loss of All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes EVENT TYPE: Instrumentation / Communication OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:

1. The following conditions exist:

a. Loss of most annunciators < > associated with safety systems for greater than 15 minutes.

AND

b. Compensatory non-alarming indications are available.

[_h AND

c. In the opinion of the < Operations > Shin Supervisor, the loss of annuncivors or indicators requires increased surveillance to safely operate the unit < >.

AND

d. Annunciator or indicator loss does not result from planned action.

DAEC EAL INFORMATION:

Control room panels IC03, IC04, and IC05 contain the annunciators associated with safety systems at I DAEC. Therefore, the DAEC EAL addresses unplanned loss of most annunciators on these panels.

Compensatory non-alarming indications includes the plant process computer, SPDS, plant recorders, or plant instrument displays in the control room. Unplanned loss of annunciators or indicators excludes

! scheduled maintenance and testing activities.

Under the conditions of concem, entry into AOP 302.2, Loss of Alarm Panel Power, would be made. The procedure requires alerting operators on shift to the nature of the lost annunciation. It further requires that.

operators be attendant and responsive to abnormal indications that relate to those systems and components that have lost annunciation. Therefore, the generic criterion related to specific opinion of the Operations Shift Supervisor that additional operating personnel will be required to safely operate the unit is not included in the DAEC EAL because the concern is addressed by the AOP.

O Annunciators on IC03, IC04, and IC05 share a common power supply from 125 VDC Division I that is V fed through circuit breaker 1D13.

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l SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE:TBD l

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Indications ofloss of annunciators associated with safety systems include:  !

o 125 VDC charger, battery, or system annunciators on control room panel IC08 o Loss of" sealed in" annunciators at affected panels i o Failure of affected annunciator panels shiflily testing by plant operators  !

o Expected alarms are not received o Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels IC03, 4 1C04, and 1C05)  !

REFERENCES:

l

1. Operating Instruction (OI) No. 317.2 Annunciator System

2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power

3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power )

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PAGE S-5 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SU4 Fuel Clad Degradation l

l EVENT TYPE: Coolant Activity OPERATING MODE APPLICABILITY: Run, Startup, Hot S/D EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)  ;

1. (Site-Specific) < valid > radiation monitor readings indicating fuel clad degradation greater than  !

Technical Specification allowable limits.

2. (Site-Specific) coolant sample activity value indicating fuel clad degradation greater than Technical l Specification allowable limits.

DAEC EAL INFORMATION:

There are no significant deviations from the generic EALs. These EALs are precursors ofmore serious fitel clad degradation and are thus considered as indicating apotential degradation ofthe level ofsafety of l the plant. Thus, it is possible to be operating within Technical Specification LCO Action Statement time ,

limitsfor lodine spikes and make a declaration ofan Unusual Event. DAEC mode applicability for these EALs are consistent with the Tech Specs.

i l EAL 1 addresses valid pretreat rad monitor exceeding (RM-4104) above 4E+3 mR/hr. The calculation

! supporting this value is described below. Valid means that the pretreat rad monitor reading is determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or coolant sampling results. This reading would be displayed on Control Room panels IC-02 and IC-10 on pretreat rad recorder RR-4104.

As specified in the generic methodology, DAEC EAL 2 addresses coolant samples exceeding coolant technical specifications for iodine spike. DAEC Tech Spec 3.6.B.I.b allows a maximum value of 12 Ci/ml of dose equivalent I-131 for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Radiological Engineering Calculation 94-014A and UFSAR Table 15.4-1 were reviewed to determine a j suitable EAL threshold for the pretreat rad monitor reading corresponding to the Tech Spec 3.6.B.I.a j (V] coolant activity limit of 1.2 Ci/ml of dose equivalent 1-131. Using the condenser noble gasi for the control rod drop accident of 2.38 E +06 Curies shown on UFSAR Table 15.4-1 and the condenser free volume of 55,000 cubic feet, an initial noble gas concentration in the condenser offgas line is determined. Because the ofTgas flow rate is very small (about 50 standard cubic feet per minute) compared  !

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PAGE S-6 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD l

to the total condenser free volume, dilution of the condenser noble gas concentration due to offgas flow is

not considered in the calculation shown below. Decrease in the noble gas source term due to decay of short-lived noble gas radioisotopes and offgas flow dilution effects are addressed by rounding down the j value calculated as shown below.

Calculation 94-014A used an exposure rate method based on using a source term consisting of a defined l mixture of noble gases and iodine from the control rod drop accident as described in the DAEC UFSAR, Section 15.4. The calculation assumed that the activity is released instantly and immediately reached in l equilibrium with the reactor coolant inventory. Using this calculation, using dose correction factors l (DCFs) for child thyroid dose from Reg. Guide 1.109, and adjusting for the specific gravity (0.736) of saturated water at 1050 psia (fluid conditions assumed in the calculation) to adjust for standard conditions, the I-131 dose equivalent (in units of Ci/ml assuming I cc equals 1 ml) is determined for this event. This

result is then linearly scaled for rad monitor readings corresponding to the Tech Spec 3.6.B.I.a allowable l primary coolant activity of 1.2 pCi/ml I-131 dose equivalent, i.e., the relative mixture of noble gases and

' iodine is assumed to remain constant.1-129 is ignored because it has no effect on the calculation result.

l

~ Isotope DCF (mrem /pci) Concentration (pCi/cc) Correction Factor [DCF / l-131 DEQ (pCl/cc)

DCFnv] / 0.736

! l-131 4.39 E-03 1.6 E+01 1.4 E+00 2.2 E+01 1-132 5.23 E-05 2.2 E+01 1.6 E-02 3.6 E-01 1-133 1.04 E-03 3.1 E+01 3.2 E-01 1.0 E+01 1-134 1.37 E-05 3.4 E+01 4.2 E-03 1.4 E-01

! 1-135 2.14 E-04 2.9 E+01 6.6 E-02 1.9 E+00 TOTAL -- -- -- 3.4 E+01

Therefore, for this event, a coolant activity of 34 pCi/cc I-131 dose equivalent is calculated. Scaling the l

results for 1.2 Ci/cc I-131 dose equivalent, a suitable condenser source term and corresponding initial concentration in the offgas flow is then determined. This is then converted to a pretreat rad monitor reading by use of the monitor efficiency factor:

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PAGE S-7 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD Pretreat Rad Monitor (RM 4104) Reading l

NG concentrationu% = NG concentration X[1.2 Ci/cc/34 Ci/cc]

= [2.38 E +6 Ci x 1 E+6 pCi /Ciy [5.5 E+4 ft' x 2.83 E+4 cc/ft') X [1.2 pCi/cc /34 pCi/cc ]

= 1529 pCi x 0.0353 = 54.0 pCi/cc Pretreat rad monitor reading = NG concentration X Rad monitor efficiency Rad monitor efficiency = 89.2 mR/hr / pCi/cc, therefore: l Pretreat rad monitor reading = 89.2 X 54.0 = 4800 mR/hr I To account for isotopic decay and dilution effects of offgas flow, round down to 4E+03 mR/hr.

The calculation results were also reviewed to determine if suitable values for the main steam line (MSL) radiation moniton could be developed. As shown above, the rod drop accident corresponds to coolant Q activity of 34 pCi/cc I-131 dose equivalent. As determined by the reference calculation, this corresponds to a MSL radiation monitor reading of about 5.7 R/hr. Scaling the results for 1.2 pCi/ml I-131 dose equivalent:

MSL Reading Correspondirig to 1.2 Cl/ml 1-131 dose equivalent I

((1.2 Ci/cc) / [34 pCi/cc)) X 5.7 R/hr = 0.2 R/hr = 200 mR/hr 200 mR/hr is at the lower end of the normal MSL monitor readings during full power Because this value is not distinguishable and hydrogen water chemistry system malfunctions that result in increased production of N-16 can also result in increased main steam line radiation levels, it is not appropriate at DAEC to use the main steam line monitor readings.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 672.2, Offgas Radiation / Reactor Coolant High Activity

2. Technical Specification 3/4.6.B, Coolant Chemistry

3. Radiological Engineering Calculation No. 94-014A, Main Steam Line Radiation Monitor Setpoint ,

, Calculation, August 29,1994  !

! 4. Surveillance Test Procedure (STP) No. 46B001, Reactor Coolant Gamma and Iodine Activity l 5. Annunciator Response Procedure (ARP) 1C03A, Reactor and Containment Cooling and Isolation O 6. Annunciator Response Procedure (ARP) IC05B, Reactor Control l\

7. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) f%

O PAGE S-8 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SU5 RCS Leakage EVENT TYPE: Coolant Leak OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown, Cold Shutdown EXAMPLE EMERGENCY ACTION LEVELS < >: < (1 or 2 or 3) >

< l .> Unidentified or pressure boundary leakage greater than 10 gpm.

OR

<27 Identified leakage greater than 25 gpm.

<OR>

<3.> < Valid (site-specific) indication of Main Steamline Break >

C)

V DAEC EAL INFORMATION:

EALs 1 and 2 are precursors of more serious RCS barrier challenges and are thus considered as a potential degradation of the level of safety of the plant. Thus, it is possible to be operating within Technical Specification LCO Action Statement time limits and make a declaration ofan Unusual Event in accordance with these EALs. Creditfor the action statement time limit should only be given when leakage exceeds technical specification limits but has notyet exceeded the Unusual Event EAL thresholds described above. In addition, indication of main steam line break ;has been added here as previously discussed in the basis for Fission Barrier Table RCS Barrier EAL 1, RCS Leak Rate, and is further discussed below. This is in accordance with NUMARC Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 2 Guestions and Answers, June 1993, Fission Product Barrier-BWR section, response to question 4 which states that the main steam line break with isolation can be classified under System Malfunctions.

Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.

l The DAEC Tech Spec Section 3.6.C.1 coolant system leakage LCO limits are: (1) 5 gpm unidentified j j leakage, (2) 2 gpm increase in unidentified leakage within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, and (3) 25 gpm total leakage. i

q Total leakage is defined as the sum ofidentified and unidentified leakage.

l .) 1 DAEC EAL 1 uses the generic value of 10 GPM for unidentified leakage or pressure boundary leakage.

The 10 gpm value for the unidentified or pressure boundary leakage was selected as it is observable with

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4 Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

V PAGE S-9 of 33 4

SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD 3

-w i normal control room indications. DAEC EAL 2 uses identified leakage set at a higher value due to the i lesser significance ofidentified leakage in comparison to unidentified or pressure boundary leakage.

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REFERENCES:

4

1. Technical Specification 3.6.C, Coolant Leakage

2. Surveillance Test Procedure No. (STP) 42A001, Reactor Coolant System Leak Rate Calculation

3. Operating Instruction No. (01) 920, Drywell Sump System 1 4. Alarm Response Procedure (ARP) IC04B, Reactor Water Cleanup and Recirculation

5. Alarm Response Procedure (ARP) IC04C, Reactor Water Cleanup and Recirculation l 6. UFSAR Section 5.2.5, Detection of Leakage through Reactor Coolant Pressure Boundary

7. UFSAR Section 15.6.6, Loss-of-Coolant-Accident

8. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993 4

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PAGE S-10 of 33 SYSTEM MALFUNCTION CATEGORY l

EFFECTIVE DATE: TBD 1

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l SU6 Unplanned Loss of All Onsite or Offsite Communications Capabilities EVENT TYPE: Instrumentation / Communication OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:

1. Either of the following conditions exist:

a. Loss of all (site-specific list) onsite communications capability affecting the ability to perform routine operations.

OR

b. Loss of all (site-specific list) offsite communications capability.

DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL. The communications methods used at DAEC are l

described in the Emergency Plan. In-plant and external agency telephone communication methods include PABX lines, direct-ring lines, and NRC telephones which are extensions for the Emergency Notification System. There is also a microwave system to provide backup emergency telephone conununications.

The availability of one method of ordinary offsite communication is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (relaying of informationfrom radio transmissions, individuals being sent to ofsite locations, etc.) are being utilized to make communicationspossible.

The DAEC plant operations radio system is a UHF system with consoles located in the Control Room, Technical Support Center, Operational Support Center, and the Central Alarm Station. Hand-held transceivers are used in this system to provide simplex communications within the plant and onsite. The DAEC Radiological Survey Radio System is an 800 Mllz trunked / conventional repeater system that provides base-to-portable communications throughout the DAEC EPZ. A secondary high-band system provides back-up capability for the 800 Mllz radio. Consoles are located in the Technical Support Center and the Emergency Operations Facility at the IES Tower. The DAEC Security (backup radiological survey) Radio System provides base-to-portable security communication within the plant and with the Linn County Sheriffs Office using a mobile relay (repeater) type base station and two VliF frequencies. Control

(]

(> consoles are located in the Secondary Alarm Station, Central Alarm Station, Security Control Point, Technical Support Center, and Emergency Operations Facility. The DAEC also has a base station licensed for operation in the Police Radio Service on the law enforcement state-wide, point-to-point VliF

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PAGE S-11 of 33 '

SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD frequency.' The transmitter and one control console are located at the Secondary Alarm Station and in the l Central Alarm Station. This station is for communications with lowa Department of Public Safety radio l station, Linn County S heriffs office, and the Benton County Sheriffs office. This point-to-point channel is also used by the Linn County Emergency Management and other public-safety organizations throughout the state ofIowa.

REFERENCES:

1. Emergency Plan, Section F, Emergency Communications

2. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993 e f CJ i l

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EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) l PAGE S-12 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD l

SU7 Unplanned Loss of Required DC Power During Cold Shutdown or Refuel < >

Mode For Greater Than 15 Minutes i

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EVENT TYPE: Loss of Power I

OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel j EXAMPLE EMERGENCY ACTION LEVEL:

1. he following conditions exist:

a. Unplanned Loss of Vital DC power to required DC busses based on (site-specific) bus voltage indications.

l AND l v[] b. Failure to restore power to at least one required DC bus within 15 minutes from time ofloss.

l DAEC EAL INFORMATION:

l There is no significant deviation from the generic EAL. Unplanned loss of Div. I and Div. 2.125 VDC L busses excludes scheduled maintenance and testing activities. Under the conditions of concern, AOP 302.1, Loss of 125 VDC Power, would be entered. The DAEC EAL's address the loss of both divisions of the 125 VDC systems consistent with AOP 302.1.

l The 125 VDC system is divided into two independent divisions - Division I (1D1) and Division II (1D2) - l each with separate AC and DC (battery) power supplies. Loss of both 125 VDC Divisions could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations. These EAL's are intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will be per SA3 "RCS temperature rise that is not allowed by procedures or Technical Specifications that will result in RCS temperature above 212 F".

Bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment I and may be indicated by the illumination of annunciators "125 VDC System 1 Trouble" on IC08A A-9 and/or"125 VDC System 2 Trouble" on IC08B A-4.

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) lO PAGE S-13 of 33 SYSTO4 MALFUNCTION CATEGORY EFFECTIVE DATE: TBD i ,

REFERENCES:

1. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power .

l 2. Abnormal Operating Procedure (AOP) 388, Loss of 250 VDC Power l 3. Technical Specification 3.8.B, DC Power Systems l 4. UFSAR Section 8.3, Onsite Power Systems l 5. UFSAR Table 8.3-6, Plant Battery System - DC Power, Instrumentation, and Control, Principle DC Loads (125V)

6. ARP IC08A A-9

7. ARP IC08B A-4 q

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I SAI Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold < Conditions >

EVENT TYPE: Loss ofPower OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel, Defueled EXAMPLE EMERGENCY ACTION LEVEL:

1. The following conditions exist:

l a. Loss of power to (site-specific) transformers.

AND

b. Failure of(site-specific) emergency generators to supply power to emergency busses.

O l AND I d c. Failure to restore power to at least one emergency bus within 15 minutes from the time ofloss of both offsite and onsite AC power.

DAEC EAL INFORMATION:

l There is no significant deviation from the generic EAL. Under the conditions of concern, entry into AOP 301.1, Station Blackout, would be made under Tab 1. Indications / alarms related to station blackout are displayed on control room panel 1C08 and are listed in the procedure under " Probable Indications."

At DAEC, the Essential Buses of concern are the 4160V Buses I A3 and 1 A4. Each of these buses feed their associated 480V and 120V AC buses through step down transformers.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout

2. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power

3. Technical Specifications Section 3.8, Auxiliary Electrical Systems i

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PAGE S-15 of 33

, SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SA2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Run, Startup i

l EXAMPLE EMERGENCY ACTION LEVEL:

1. (Site-specific) indication (s) exist that indicate that reactor protection system setpoint was exceeded and automatic scram did not occur, and a successful manual scram occurred.

() DAEC EALINFORMATION:

l The DAEC EAL is written in terms of failure of automatic scram. IPOI 5 specifies manual scram insertion immediately following any automatic scram signal, and therefore separately specifying successful manual j scram is not required. The reactor is considered successfully shutdown if either: (1) all control rods are l I

inserted to least position 02, or (2) it has been detemlined that the reactor will remain shutdown under ALL conditions without boron. If these conditions are not achieved, entry into the ATWS - RPV Control EOP will be made where additional manual actions to be performed at panel IC05 are specified to quickly shutdown the reactor. These actions include reducing recirculation pumps to minimum speed and inserting Alternate Rod Insertion (ARI).

Ifthe mode switch is in Startup and the rods arejidly inserted (i.e., the reactor is shutdown) prior to the automatic signalfailure, then declaration ofan Alert would not be required in this case, the event would be reported under 10 CFR 50. 72 (b) (2) (1) as afour hour report.

The condition of concern is failure of the automatic protection system to scram the reactor. This condition ,

is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient and thus plant safety has been compromised and design limits of the fuel may have been exceeded. In the generic EAL, reactor protection system setpoint being exceeded (rather than limiting safety system setpoint being exceeded) is specified to emphasize that failure of the automatic protection system to complete the scram following generation of a scram signal is the tO issue of concern.

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PAGE S-16 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD L

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REFERENCES:

!- 1. Integrated Plant Operating Instruction (IPOI) No. 5, Reactor Scram l 2. ATWS Emergency Operating Procedure (EOP)- RPV Control

3. Emergency Operating Procedure (EOP) 1 - RPV Control

4. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993 O

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PAGE S-17 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD l

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SA3 Inability to Maintain Plant in Cold Shutdown EVENT TYPE: Inability to Maintain Shutdown Conditions l

OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel l EXAMPLE EMERGENCY ACTION LEVEL: i i l i

1. Loss of < decay heat removal systems required > to maintain cold shutdown l l AND l Temperature increase that either: )

l e Exceeds Technical Specification cold shutdown temperature limit

! OR e Results' in uncontrolled temperature rise approaching cold shutdown technical specification limit. l l

DAEC EAL INFORMATION:

1 Under the conditions of concern for EAL 1, AOP 149, Loss of Decay Heat Removal, would be entered .

l under Tab 1, Loss of Shutdown Cooling. Indications / alarms related to loss of shutdown cooling are displayed on control room panels IC03 and IC05 and are listed in the procedure under " Probable Indications." The procedure requires that shutdown cooling be re-established. If this cannot be done, then l the following actions are to be performed:

l

e Re-establish primary and/or secondary containment.

e Increase reactor water level to between 240 inches and 250 inches to improve natural circulation.

l e Start one reactor recirculation pump, if available.

• Monitor reactor temperatures per STP 46A003 noting that some points will lag behind the bulk coolant temperature.

• Notify Health Physics to begin increased monitoring of the Reactor Building.

. Evaluate plant conditions to determine need to achieve or maintain the plant in cold shutdown.

e Initiate alternate means of shutdown cooling which includes feed and bleed to radwaste or condenser, feed and bleed to the torus through the SRVs, using the reactor water cleanup heat exchanger, or reactor cavity flood up and use of fuel pool cooling.

O V The procedure provides curves of maximum water heat up rates which provide an upper bound of the heatup until an estimated time to boil calculation can be completed by Engineering.

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SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD The DAEC EAL is written to imply a RCS temperature rise above 212 F that is not allowed by plant j procedures. This corresponds to the inability to maintain required temperature conditions for Cold l Shutdown. " Uncontrolled" means that system temperature increase is not the result of planned actions by  ;

l the plan: staff. The wording is also intended to eliminate minor cooling interruptions occurring at the  !

transition between Hot Shutdown and Cold Shutdown or temperature changes that are permitted to occur  !'

during establishment of altemate core cooling so that an unnecessary declaration of an Alert does not occur. The uncontrolled temperature rise is necessary to preserve the anticipatory philosophy of NUREG-0654 for events starting from temperatures much lower than the cold shutdown temperature limit.- .

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REFERENCES:

1. Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Removal  :

2. DAECTechnicalSpecifications  !

3. Surveillance Test Procedure (STP) 46A003, Heatup and Cooldown Rate Log  !

4. NUREG 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates, September 1993

5. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARCJNESP-007 Revision 2 Questions andAnswers, June 1993 i

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! O PAGE S-19 of 33 SYSTEM MALFUNCTION CATEGORY l

EFFECTIVE DATE: TBD SA4 Unplanned Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a Significant Transient in Progress, or (2) l Compensatory Non-Alarming Indicators are Unavailable EVENT TYPE: Instrumentation / Communication OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:

1. The following conditions exist:

a. Loss of most< > annunciators associated with safety systems for greater than 15 minutes.

AND O

g b. In the opinion of the < Operations > Shift Supervisor, the loss of all annunciators or indicators requires increased surveillance to safely operate the unit < >.

AND

c. Annunciator or Indicator loss does not result from planned action.

AND

d. Either of the following:

A significant plant transient in progress.

OR Compensatory non-alarming indications are unavailable.

DAEC EAL INFORMATION:

Control room panels IC03, IC04, and IC05 contain the annunciators associated with safety systems at DAEC. Therefore, the DAEC EAL addresses unplanned loss of annunciators on these panels.

Compensatory non-alarming indications includes the plant process computer, SPDS, plant recorders, or plant instrument displays in the control room. Unplanned loss of annunciators or indicators excludes scheduled maintenance and testing activities. Sigmficant transient includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10% or greater.

Under the conditions of concem , entry into AOP 302.2, Loss of Alarm Panel Power, would be made. The fl procedure requires alerting operators on shift to the nature of the lost annunciation. It further requires that

'd operators be attendant and responsive to abnormal indications that relate to those systems and components l that have lost annunciation. Therefore, the generic criterion related to specific opinion of the Operations I

Duane Arnold Energy Center g.s EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview) lU PAGE S-20 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD Shift Supervisor that additional operating personnel will be required to safely operate the unit is not included in the DAEC EAL because the concem is addressed by the AOP.

Annunciators on IC03, IC04, and IC05 share a common power supply from 125 VDC Division I that is fed through circuit breaker ID13. Therefore, DAEC does not specify a loss of"most" annunciators as specified in the generic methodology.

Indications ofloss of annunciators associated with safety systems include:

l e 125 VDC charger, battery, or system annunciators on control room panel 1C08 e Loss of" sealed in" annunciators at affected panels l

• Failure of afTected annunciator panels shiflily testing by plant operators e Expected alarms are not received

/~"N

• Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels IC03, O 1C04, and 1C05)

REFERENCES:

1. Operating Instruction (01) No. 317.2 Annunciator System

2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power

3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power I

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PAGE S-21 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD i

SAS AC Power Capability to Essential Busses Reduced to a Single Power Source for Greater Than 15 Minutes Such That Any Additional Single Failure Would Result in Station Blackout l

EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:

1. The following conditions exist:

a. Loss of power to (site-specific) transformers for greater than 15 minutes.

s AND

b. Onsite power capability has been degraded to one (train of) emergency bus (ses) powered from a I single onsite power source due to loss of:

(Site-specific list)

DAEC EAL INFORMATION:

The DAEC EAL is written to address the underlying concern, i.e., only one AC power source remains and ifit is lost, a Station Blackout will occur. Under the conditions of concern, entry into AOP 301, Loss of Essential Electrical Power, would be made under Tab 1, Loss of One Essential 4160V Bus, and/or under Tab 3, Loss of Offsite Power. Indications / alarms related to degraded AC power are displayed on control .

room panel 1C08 and are listed in the procedure under " Probable Indications." ,

At DAEC, the Essential Buses of concern are 4160V Buses l A3 and 1 A4. Each of these buses feed their '

associated 480V and 120V AC busses through step down transformers. Onsite power sources at DAEC include the A and B Diesel Generators, IG-31 and IG-21, respectively.

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REFERENCES:

l l 1. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power l 2. UFSAR Chapter 8 Electrical Power

3. Technical Specifications Section 3.8. Auxiliary Electrical Systems

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t SYSTEM MALFUNCTION CATEGORY f EFFECTIVE DATE: TBD l

I SSI Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses EVENT TYPE: Loss of Power l

OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown i

EXAMPLE EMERGENCY ACTION LEVEL: 1 i

1. Loss of all offsite and onsite AC power as indicated by:

l a. Loss of power to (site-specific) transformers.

AND

b. Failure of(site-specific) emergency generators to supply power to emergency busses.  ;

AND

, (~ c. Failure to restore power to at least one emergency bus within <l5 minutes > minutes from the time l' ofloss of both offsite and onsite AC power.

DAEC EAL INFORMATION: )

There is no significant deviation from the generic EAL. In accordance with the generic guidance, DAEC is using a threshold of 15 minutes for Station Blackout to exclude transient or momentary power losses.

Under the conditions of concem, entry into AOP 301.1, Station Blackout, would be made under Tab 1.

Indications / alarms related to station blackout are displayed on control room panel IC08 and are listed in the procedure under " Probable Indications."

1 At DAEC, the Essential Buses of concern are the 4160V Buses l A3 and I A4. Each of these buses feed  !

their associated 480V and 120V AC buses through step down transformers. i l

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REFERENCES:

1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout l 2. Technical Specifications Section 3.8, Auxiliary Electrical Systems

3. UFSAR Chapter 8, Electric Power 3

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O PAGE S-23 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Run, Startup EXAMPLE EMERGENCY ACTION LEVEL:

i.

) 1. (Site-specific) indications exist that automatic and manual scram were not successful.

DAEC EAL INFORMATION:

I The DAEC EAL addresses conditions where failure of an automatic scram has occurred and manual actions performed at panel 1C05 to quickly shutdown the reactor do not meet the success criteria ofIPOI 5 and the ATWS - RPV Control EOP.  ;

Under the conditions of concem for this EAL, the reactor may be producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because  !

conditions exist that lead to imminent loss or potential loss of the primary containment and the fuel clad. l In addition, if the SRV's are open, the RCS is no longer capable of retaining fission products and therefore !

l is not acting as a fission product barrier. Although this EAL may be viewed as redundant to the Fission Barrier Table, its inclusion is necessary to better assure timely recognition and emergency response.

REFERENCES:

1. Integrated Plant Operating Instruction (IPOI) No. 5, Reactor Scram

2. ATWS Emergency Operating Procedure (EOP)- RPV Control

3. NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993 t'

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! PAGE S-24 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD 1

SS3 Loss of All Vital DC Power l

EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:

1. Loss of All Vital DC Power based on (site-specific) bus voltage indications for greater than 15 minutes.

l DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL. Under the conditions of concern, AOP 302.1, Loss l

(Vl of 125 VDC Power, would be entered under Tab 3, Complete Loss of 125 VDC. Conseque EAL addresses loss of both divisions of the 125V DC system consistent with AOP.

At DAEC, the 250V/125V DC Systems ensure power is available for the reactor to be shutdown safely and maintained in a safe condition. The 125V System is divided into two independent divisions - Division I l and Division II - with separate DC power supplies. These power supplies consist of two separate 125V batteries and chargers serving systems such as RCIC, RHR, EDGs, and HPCI.

l The 250V system consists of a single 250V battery and two battery chargers which supply power to heavy l motor loads. The 250V battery and DC distribution system are treated as a Division II system. Two 250V l motor control centers are provided, one of which is useci for the HPCI system. The 250-V battery supplies power for the HPCI turbine oil pump. A loss of the 250-V battery would thus prevent operation of the 1 HPCI system. The HPCI system is redundant in its depresmrization function with the Automatic Depressurization System that does not require 250-VDC for operation. All of the 250-VDC motor-operated isolation valves have redundant counterparts that do not rely on DC power. Thus, the ability to l safely shutdown the reactor and maintain it in a safe condition can be maintained without the 250V system.

Ilowever, loss of both 125V DC Divisions could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations.

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! SYSTEM MALFUNCTION CATEGORY

! EFFECTIVE DATE: TBD

REFERENCES:

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1. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power j 2. Abnormal Operating Procedure (AOP) 388, Loss of 250 VDC Power

3. Technical Specification 3.8 B, DC Power Systerns

4. UFSAR Section 8.3, Onsite Power Systems

5. UFSAR Table 8.3-6, Plant Battery System - DC Power, Instrumentation, and Control, Principle DC ;

Loads (125V) i l <

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SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD  !

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SS4 Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown l

EVENT TYPE: Inability to Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown.

I EXAMPLE EMERGENCY ACTION LEVEL:

l. <EOP Graph 4 Heat Capacity Limit is exceeded >

<OR>

<2. Reactor CANNOT be brought subcritical.>

l l

DAEC EAL INFORMATION:

This EAL addresses complete loss of functions, including ultimate heat sink and reactivity control, required j for hot shutdown with the reactor at pressure and temperature. Under these conditions, there is an actual  !

major failure of a system intended for protection of the public. The reactivity condition criteria is addressed by maintenance of required shutdown margin. Ifinadvertent criticality could not be eliminated

! by performing the actions of AOP 255.1, AOP 255.2, or the ATWS EOP, it corresponds to a failure of a system intended for the protection of the public and thus classification as a Site Area Emergency is warranted.

l This EAL represents an escalation from the conditions of concem in SA3, Inability to Maintain Cold l Shutdown, because the reactor is at operating pressure and temperature and decay heat levels are higher.

l Per DAEC Technical Specifications, the following systems are necessary to achieve or maintain Hot Shutdown conditions:

o Reactor Protection System Instrumentation (T.S. 3.1) o Core and Containment Cooling Systems Instmmentation (T.S. 3.2) e Reactivity Control (T.S. 3.3) e Standby Liquid Control System (T.S. 3.4) e Core and Containment Cooling Systems (T.S. 3.5) pe Primary System Boundary (T.S. 3.6)

V e Auxiliary Electrical Systems (T.S. 3.8)

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SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD l

l Loss of instrumentation is addressed by SS6, Inability to Monitor a Significant Transient in Progress, l below. The Auxiliary Electrical System is addressed by SS1, Station Blackout, and SS3, Loss of 125V j DC, above and are therefore not covered here. Failure of the primary system boundary is covered by the l Fission Barrier Table and SUS, RCS Leakage, above.

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REFERENCES:

l l 1. Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Removal l

2. Abnormal Operating Procedure (AOP) 255.1, Control Rod Movement / Indication Abnormal l l

3. Abnormal Operating Procedure (AOP) 255.2, Power / Reactivity Abnormal Change j

4. Emergency Operating Procedure (EOP) 1 - RPV Control l

5. ATWS Emergency Operating Procedure (EOP)- RPV Control '

6. Emergency Operating Procedure ALC - Altemate Level Control i

,q 7. Emergency Operating Procedure (EOP) Basis, EOP Breakpoints  :

l Q 8. NUMARC Afethodologyfor Development ofEmergency Action Levels NUAfARC/NESP-007 Revision 2 l Questions andAnswers, June 1993 l

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PAGE S-28 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD

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SS5 Loss of Water Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reactor Vessel EVENT TYPE: Inability to Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel EXAMPLE EMERGENCY ACTION LEVEL:

j 1. Loss of Reactor Vessel Water Level as indicated by:

a. Loss of all decay heat removal cooling as determined by (site-specific) procedure.

AND

b. (Site-specific) indicators that the core is or will be uncovered.

,-m DAEC EAL INFORMATION:

1 I

There is no significant deviation from the generic EAL. The DAEC EAL is written in terms of the general concem that no cooling water source is lined up or available for injection into the RPV and water level is

! decreasing below the top of the active fuel (TAF). Under the conditions of concern for EAL 1, AOP 149, i Loss of Decay Heat Removal, would be entered under Tab 1, Loss of Shutdown Cooling. I Indications / alarms related to loss of shutdown cooling are displayed on control room panels IC03 and IC05 and are listed in the procedure. Consistent with the value used in the EOPs, the EAL uses an l indicated RPV level of 15 inches for the water level corresponding to TAF. l l

REFERENCES:

4 1

1. Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Removal l 2. Emergency Operating Procedure (EOP)-1, RPV Control, Sheet 1 of 1

3. Emergency Operating Procedure (EOP) Basis, EOP Breakpoints l

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SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SS6 Inability to Monitor a Significant Transient in Progress l

EVENT TYPE: Instrumentation / Communication l

OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown  ;

i l EXAMPLE EMERGENCY ACTION LEVEL: j

1. The following conditions exist:

a. Loss of < > annunciators associated with safety systems.

l AND t b. Compensatory non-alarming indications are unavailable.

AND

/s c. Indications needed to monitor (site-specific) safety functions are unavailable.

O AND l

d. <Significant> transient in progress.

DAEC EAL INFORMATION:

> l l

The DAEC EAL is written in terms of a sigmylcant transient in progress with loss of both safety system annunciators and loss af compensatory non-alarrning instrumentation. The DAEC EAL structure, which addresses all the key points in the generic EAL better assures that the condition of concern for this EAL l will be readily recognized.

l l Sigmficant transient includes response to automatic or manually initiated functions such as scrams,

! runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of10% or greater.

Control room panels IC03, IC04, and IC05 contain the annunciators associated with safety systems at DAEC. Annunciators on IC03, IC04, and IC05 share a common power supply from 125 VDC Division I

! that is fed through circuit breaker ID13. Therefore, DAEC does not specify a loss of"most" annunciators l as specified in the generic methodology.

Compensatory non-cla,unrg indicctions include the plant process computer, SPDS, plant recorders, or plant instrument dispirT; in the control room. These indications are needed to monitor (site-specific) p)g

( safety functions that are of mncem in the generic EAL.

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.1 (forNRCreview)

PAGE S-30 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD Indications ofloss of annunciators associated with safety systems include:

e 125 VDC charger, battery, or system annunciators on control room panel 1C08 e Loss of" sealed in" annunciators at affected panels e Failure of affected annunciator panels shiftily testing by plant operators e Expected alarms are not received e Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, 1C04, and 1C05)

REFERENCES:

! 1. Operating Instruction (OI) No. 317.2, Annunciator System

2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power

3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power l

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l SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD l

l SGI Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:

l l 1. Prolonged loss of all offsite and onsite AC power as indicated by: l

a. Loss of power to (site-specific) transformers.  !

AND l

b. Failure of(site-specific) emergency diesel generators to supply power to emergency busses.

l(,. AND l

V) c. At least one of the following conditions exist:

Restoration of at least one emergency bus within (site-specific) hours is NOTlikely OR l (Site-Specific) Indication of continuing degradation of core cooling based on Fission Product Barrier monitoring.

1 DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL. Under prolonged Station Blackout (SBO) conditions, fission product barrier monitoring capability may be degraded. Although it may be difficult to  !

l l predict when power can be restored, it is necessary to give the EC/OSS a reasonable idea of how quickly a l General Emergency should be declared based on the following considerations:

t Are there any present indications that core cooling is already degraded to the point where a General Emergency is IMMINENT (i.e., loss of two barriers and a potential loss of the third banier)? )

(

• If there are presently no indications of degraded core cooling, how likely is it that power can be restored prior to occurrence of a General Emergency? l I

The first part of this EAL corresponds to the threshold conditions for Initiating Condition SS1, Station Blackout - namely, entry into AOP 301.1, Station Blackout. The second part of the EAL addresses the ,

% conditions that will escalate the SBO to General Emergency. Occurrence of any of the following is j sufficient for escalation: (1) SBO coping capability exceedr', or (2) loss of drywell cooling that continues l

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)

SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD to make RPV water level measurements unreliable, or (3) indications ofinadequate core cooling. Each of these conditions is discussed below:

1. SBO Coping Capability Exceeded DAEC has a SBO coping duration of four hours. The likelihood ofrestoring at least one emergency bus should be based on a realistic appraisal ofthe situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing publicprotective actions.

2. RPV Water Level Measurements Remaining Unreliable Flashing of the reference leg water will result in erroneously high RPV water level readings giving a false indication of actual water inventory and potentially indicating adequate core cooling when it may not exist.

EOP Graph 1, RPV Saturation Temperature, defines the conditions under which RPV level instrument leg boiling may occur.

3. Indications ofInadagnate Core Cooling i

DAEC uses the RPV level that is used for the Fuel Clad EAL 2 " loss" condition. This is RPV level below l

-30 inches, which is consistent with EOP Level / Power Control requirements.  !

REFERENCES:

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1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout

2. . Letter NG-92-0283, John F. Franz, Jr. to Dr. Thomas E. Murley, Response to Safety Evaluation by  ;

NRC-NRR " Station Blackout Evaluation Iowa Electric Light and Power Company Duane Amold Energy Center," February 10,1992

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3. Emergency Operating Procedure (EOP)1 - RPV Control ,

4. Emergency Operating Procedure (EOP) ALC - Altemate Level Control.

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,V PAGE S-33 of 33 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: TBD SG2 Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Run,Startup EXAMPLE EMERGENCY ACTION LEVEL:

1. The following conditions exist:

a. (Site-specific) indications exist that automatic and manual scram were NOT successful.

,, AND

'( '

b. Either of the following:

(Site-specific) indication exists that the core cooling is extremely challenged.

OR (Site-specific) indication exists that heat removal is extremely challenged.

DAEC EAL INFORMATION: l Automatic and manual scram are not considered successful if action away from the reactor control console is required to scram the reactor. Consistent with the EOPs, the ATWS conditions of concern in this EAL are reactor power that is expected to remain above 5% or that is indeterminate.

1 Escalation to the General Emergency classification requires extreme challenge to core or containment cooling, i.e., imminent barrier loss. If the Main Condenser is available for steam release from the reactor, sufficient heat removal capability exists. However, without the main condenser being available as a heat sink, heat removal capability under these conditions is insufficient and a threat to the Fuel Clad barrier exists. In addition, the SRV's will lift and thus the RCS barrier will not retain fission products. Eventually, the torus water will be heated to the point where the containment function will become inefTective. Thus, the resultant combination of barrier conditions warrants a declaration of a General Emergency if ATWS reactor power-level control methods are ineffective in reducing reactor power level. l REFERENCES-

,O d 1. Emergency Operating Procedure ATWS EOP - RPV Control i

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