ML20199L348

Rev 0 to Duane Arnold Energy Ctr Emergency Action Level (EAL) Technical Basis Document

ML20199L348
Person / Time
Site:	Duane Arnold
Issue date:	01/31/1998
From:	IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:
References
NUDOCS 9802090182
Download: ML20199L348 (135)

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k Manual #91 NRC NRR To:

Document Control Desk Washington, DC Subjer1: Transmittal / Acknowledgement Form -

Emergency Action Level (EAL)

Technical Basis Document Pleaseplace this Emergency A ction Level (EA L) - Technical Basis Document nest toyour Emergency Plan Implementing Procedure (EPIP) notebook.

The attached document has been transmitted to you for your information and/or use.

Please acknowledge receipt by filling cut this form, sign, date, and RETURN iT WITillN 10 DAYS to the address below.

Do you desire to continue rece! a updates to this document?

Yes No d If n t, please return this document to the address below, is your address co Tect?

i Yes No If not, please provide your new address below:

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Return to: Kathryn Dunlap Emergency Management Department (M) (

3313 DAEC Road Palo,IA 52324 ACKNOWLEDGEMENT II llIl Ili t IIII Print Name:

Y Signature:

Date:

9900090182 900131 .

PDR ADOCK OSOOO331c b F PDR

DUANE ARNOLD ENERGY CENTER -

O EMERGENCY ACTION LEVEL (EAL)

TECHNICAL BASIS DOCUMENT

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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT ao i Rev.O 1

v PAGE l of iv TAHLE OF CONTENTS EFFECTIVE DATE: 100/98 TABLE OF CONTENTS INTRODUCTION...................................................................................................................................11 DEFINIT 10NS........................................................................................................................................D.1 ORG ANIZATION OF B A SI S INFO RM AT10N ...................................................... ..... ....................... 0 1 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT CATEGORY h

v AUl Any Unplanned Release of Gaseous or I iquid T. adioactivity to the Environment That Exceeds Two Times the Radiological Technical Specifications For 60 Minutes or Longer... . .. . ., .. .. .. .. ... .... .... ..... ... ............ ......... .......... A 1 AU2 U nexpected Increase in Plant Radisticn ......... ... . .... ... ....... ....... ........... ................. ..... ... . .. .. . ...........A$

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 AA2 Major Damage to irradiated Fuel or Loss of Water Level that lias 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 Systems Required to Maintain Safe Ope ations or to Establish or to Maintain Cold Shutdown..... . . .... ........A l$

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 AGl Site Boundary Dose Resuking 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 thevelease.. . .. ......... . ... . .....A 21 FISSION PRODUCT BARRIER DEGRADATION CATEGORY FUI Any Loss or Any Potential Loss of Primary Containment Banier.... . . .. .., . ...... . . . . . . . . . . . . . . . . . . . . . . _ . .. .. F 1 FA1 Any Loss or Any Potential Loss of Either Fuel Clad Or RCS Barrier.. .....................................F2 Q FSI Loss Or Potential Loss of Any Two Barriers.... . .. . . . .. .. . ... ...... .. . . . . . . . . . . . . . . . .................F3 7

FG1 Loss of Any Two Barriers AND Potential Loss of the Third Barrier... .. . . . .. . .. . . . . .. . .F 5  :

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PAGE li of iv TAHLE OF CONTENTS EFFECTIVE DATE: 1/30/98 FISSION PRODUCT llARRIER DEGRADATION CATEGORY (continued)

FU EL C LA D llA R RI E R INDIC ATORS .................. .... .. .......................... .. . . . . . . . . ...................................F6 RCS 11 A RRIER INDICATORS . ... . . . ...... .. ... ... .. ............ .. . . ...........................................................F12 PRIMARY CONTAINMENT DARRIER INDICATORS ..... ... .. .. ...........................................................F't1 IIAZARDS AND OTilER CONDITIONS AFFECTING PLANT SAFETY CATEGORY 1101 Natural and Destructive Phenomena Affecting the Protected Area .... ..... .... . .. . ... . .. ......... . . . . . . . . .11 1 11U2 I ie Within Safe Shutdown Areas Not Extinguished Within 15 Minutes of Detection . . ... .... .... .. ...... .. . .. ... ...ll 4 1103 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant... . . . ............. . .. . .115 73 U llV4 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant..... . ..... .11-6 11U5 Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Declaration of an Unusual Event ..... .117 il Al Natural and Destructive Phenomena Affecting the Plant Vital Area ... ... ... ... .. . .. ... . .... .. .. ... .......... . ... .. .118 IIA 2 Fire AITecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown ... ... ..11 12 IIA 3 Release of Toxic or Flammable Oases Within a facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Opecations or to Establish or Maintain Cold Shutdown.. . .. . .. ....................................1115 llA4 Security Event in a Plant Protected Area... .. . .. , . .. . . . . . . . . . .. .. . . .. . . .. . ... .. 11 17 IIA 5 Control Room Evacuatiun lias Been initiated . .. . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . ......11 18 IIA 6 Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Declaration of an Alert.. . .... .11 19 IISI Security Event in a Plant Vital Aren. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. ....11 20 11S2 Control Room Evacuation lias Been initiated and Plant Control Cannot Be Established... .. . .. . ... .... . .. . . 11 21 IIS3 Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Declaration of a Site Area Emergency 11 23 1101 Security Event Resulting in Loss Of Ability to Reach and Maintain Cold Shutdown... . ... .. .. . .11 24

-A U 1102 Other Conditions Existing Which in the Judgment of the EC/OSS Warrant Declaration of a General Emergency 11 25

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PAGEill of iv TABLE OF CONTENTS EFFECTIVE DATE: 1/30/98 SYSTEM MALFUNCTION CATEGORY SUI Loss of All Offsite Pow er to Essential Busses for Orester han 15 Minutes ............ . ........ .. S 1 SU2 Inability to Rnch Required Shutdown Within Technical Speci0 cation Limits ............... ................S2 SU3 Unplanned Loss of All Safety System Annunciation or Indication in the Control Room for Greater han 15 Mmutes

..... . ..................... .... ;S 3 SU4 FuelClad Degradation: ......................... .............................S$

SU$ RCS Leakag e................ . ....... .. .......... ...........................................S8 SU6 Unplanned Loss of All Onsite or Offsite Communications Capabilities ... ................ ;S 10 SU7 Unplanned Loss of Required DC Power During Cold Shutdown or Refuel Mode For Orcater Than 15 Minutes... S 12 sal Loss of All OITsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Conditions.., ........ S 14 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......................... S l$

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 Significant Transient in Progress, or (2) Compensatory Non Alarming Indicators are Unavailable................................ S 19 SA$ AC Power Capability to Essential Busses Reduced to a Single Power Source for Greater inan 15 Minutes Such That Any Additional Single Failure Would Result in Station Blackout... ............................... . . . . . .... ... S 21 SSI Loss of All OfTsite Power and Loss of All Onsite AC Power to Essential Bnsses............ .. ... .......... S 2 2 SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate on Automatic Reactor Scram Once a Reactor Protection System Setpoint lias Been Exceeded and Manual Scram Was NOT Successful.- :S-23 SS3 Loss of All Vital DC Power = . . .....................................S24 SS4 Complete Loss of Function Needed to Achieve or Maintain llot Shutdown .. ..... S 25 SS$ Loss of Water Level in the Reactor Vessel nat lias or Will Uncover Fuel in the Reactor Vessel ....... ..... ...........iS-27 ,

. SS6 Inability to Monitor a Sign 10 cant Transient in Progress.... ............................ S 28 sol Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power................ ;S 30

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- SYSTEM MALFUNCTION C.' frFORY (continued) 502 Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and here is Indication of an Extreme Challenge to the Ability to Cool the Core . . . . .. . .. . .. . .. . . . S 3 2 O

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PAGEl 1 of 2 INTRODUCTION EFFECTIVE DATE: 1/30/98 INTRODUCTION IES Utilities has revised the Duane Amold Energy Center (DAEC) Emergency Plan to incorporate guidance from NUhiARC/NESP 007, Revision 2 (January 1992), Methodology for Devckynnent of Emergency Action Levels. The NUhtARC (now Nuclear Energy Institute or NEI) guidance was developed to replace Emergency Action Levels (EAL) guidance contained in NUREG 0654/FEhiA REP 1 (Revision 1), Criteriafor Preparation and Evaluation ofRadiological Emergency Response Plans and Preparedness in Sayy> ort of Nuc/ car Powcr Plants, November 1980. The NEl sponsored methodology was used to develop a set of generic EAL guidelines, together with the basis, so that they could be used and adapted by each utility in a consistent manner, The NRC has endorsed use of the NEl generic guidance as an acceptable attemative 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.

3 (b This Regulatory Guide further states that: " Licensees may use either NUREG 0654/FEhiA REP 1 or NUhiARC/NESP 007 in developing their EAL scheme but may not use portions of both methodologies."

This EAL basis document was developed to: (1) provide clear documentation of how NEl generic guidance was applied in the development of DAEC upgraded EALs,(2) provide justification of any exceptions or additions to NEl 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.

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

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 oft 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 explicitly 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 includes EALs addressing RCS leakage and loss of oft site power.

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INTRODUCTION l EFFECTIVE DATE: 1/30/98

3. Conditions such as fire, explosion, gas releases, flooding, low river water level, tomado, or earthquake can be directly escalated only up to the Alert emergency classification. Escalation to Site Area Emergency or General Emergency is based on degraded system response as would be determined by fission product barrier, loss of AC power, or projected efIluent release EALs.

4. Core damage sequences are addressed by detennining 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),

integrated Plant Operating lastructions (IPOls), Abnormal Operating Procedures (AOPs) and core damage assessment methodology. This allows the operations crew to readily recognize the corresponding emergency chtssification and allows for ready escalation to Site Area Emergency or General Emergency as conditions may worsen.

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

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k Protective Action Guide (PAG) levels for inhalation doses - 1,000 mrem TEDE or 5,000 mrem CDE

'lhyroid will result in declaration of a General Emergency.

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PAGE D-1 of 10 EFFECTIVE DATE: 1/30/98 DEFINITIONS DEFINITIONS AC- Altemating Current Agceting (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 Protective Action Guide (PAG) exposure levels.

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

O AOP AbnonnalOperatingProcedure APRAI- Average Power Range Monitor ARAI- Area Radiation Monitor ATWS- Anticipated Transient Without Scram Barrier - Same as " Fission Product Barrier", below.

Barricr Afonitoring Ability - This is a judgment factor in detennining 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 concems, readings from portable instrumentation, and consideration for ofTsite monitoring results.

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

BOP- Balance of Plant BWR Boiling Water Reactor - -- - = - - -

CAAf Continuous Air Monitor CDE. Committed Dose Equivalent as dr.nned in 10 CFR 20.1003

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PAGE D 2 of 10 A 1 OM DEFINITIONS CEDE - Committed Effective Dose Equivalent as defined in 10 CFR 20.1003 CFM- Cubic Feet per Minute -

CFS- Cubic Feet per Second Cold condition - This refers to the condition where the reactor coolant temperature is less than or equal to 212'F.

Cold shutdown - As defined in Technical Specification Table 1.1 1, the reactor is in the shutdown mode, the reactor coolant temperature is less than or equal to 212'F, and all reactor vessel head closure bolts fully tensioned.

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 without reliance on control room devices or instrumentation using components and methods specified by Abnormal Operating Procedure 915, Shutdown outside Control Room.

Contiguous - Being in actual contact: touching along a boundary or at a point.

CPS Counts Per Second CRD - Control Rod Drive CSCS Core Standby Cooling System CST Condensate Storage System Curie (Cl) - A measurement of radioactive decay rate equal to 3.70E+10 disintegration's per second (becquerels).

Cil'- Circulating Water DAEC- Duane Arnold Energy Center

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PAGE D 3 of 10 E DATE: 1/30M DEFINITIONS DC- Direct Current DEG- Dose Equivalent Dominant accident sequences - These will lead to degradation of all fission product barriers. 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 llPC1/RCic failure.

DW- Drywell EC- Emergency Coordinator V ECCS- Emergency Core Cooling System EDE - Effective Dose Equivalent as defined in 10 CFR 20.1003 Emergency Action Lcvcl(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 of analyses, entry into specific emergency operating procedures, or another phenomenon which, if it occurs, indicates entry into a particular Emergency Class.

Emergency class Same as " Emergency Classification Level" below.

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

EOP - Emergency Operating Procedure EPA - Environmental Protection Agency (V7 EPIP-Emergency PlanImplementing Procedure ESF- Engineered Safety Features

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PAGE D 4 of 10 EUECME DATE: IGON DEFINITIONS ESS Engineered Safety Systems Establish - Make arrangements for a stated condition, e.g., establish communications with control room.

ESil'. Emergency Service Water F/ssion Product Barrier - One of the three principal barriers to uncontrolled release of radionuclides: Fuel Clad, Reactor Coolant System (RCS), and the Primary Containment.

FP- Fuel Pool Fuct Clad (Barricr) - The zirconium alloy tubes that contain the fuel pellets.

General Emergency (GE) - Events are in process or have occurred 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.

GPM Gallons Per Minute GSil' General Service Water llot shutdown - As defined in Technical Specification Table 1.1-1, the reactor mode switch is in the shutdown position and the reactor coolant temperature is greater than 212*F and all reactor vessel head closure bolts fully tensioned.

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l HPCI- lii h3Pressure Coolant injection (system).

f l /dentifiedLeakage Identified Leakage shall be:

a. Leakage into the drywell such as that from pump seals or valve packing that is captured and conducted to a sump or collecting tank, ot p

b. Leakage into the drywell %.iosphere from sources thz.t are both specifically located and known not to interfere with the operation of the leakage detection .

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PAGE D-S of 10 DEFINITIONS E A :1 OM8

/DLH Immediately Dangerous to Life and llealth Inadvertent - Accidental or unintentional, e.g., the event occmTed because procedures were not strictly adhered to.

Imminent - No tumaround in safety system performance is expected and escalation to a highc: emergency classification level is expected to occur within two hours.

Impicment Commence a required program or series of procedures, in service - A component or system in the appropriate configuration for nomial operation and is considered operable as defined in the Technicel Specifications.

(3 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 predetemiined subset of nuclear power plant conditions where either the potential exists for a radiological emergency or such an emergency has occmTed.

IPE-Individual Plant Examination

/ pol Integrated Plant Operating Instmetion IRM Intermediate Range Monitor Isolate - Remove from service by closing off the flow path ki' Kilovolt (s)

LCO - Limiting Condition for Operation

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v LLRPSF- Low Level Radwaste Processing and Storage Facility LOCA - Loss of Coolant Accident

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PAGE D 6 of 10 EFFECTIVE DATE: 1/30/98 DEFlh'ITIONS LOOP Loss of Offsite Power loss (of a fission product barrier) A severe challenge to a fission product barrier exista such that the barrier is considered incapable of performing its safety function. -

LPC/ Low Pressure Coolant injection AICC- hioter Control Center AICUTL - hiaximum Core Uncovery Time 1.imit Aficrocuric ( Cl) One millionth of a curie,i.e.,3.7E+4 disintegration's per second (becquerelet.

(n) AllDAS hieteorological Information and Dose Assessment System, primary method for detecting and quantifying gaseous releases at the DAEC.

Afillicurie (mCD One thousandth of a curie,i.e.,3.7E+7 disintegration's per second (becquerels).

Afillirem (mrem) - One thousandth of a rem AIPH- hiiles Per 11our mR milliroentgen,1.c., one thousandth of a roentgen (R)

A(Sil'- hiain Steam isolation Valve A/SL - hiain Steam Line g NE/ - Nuclear Energy Institute (fonnerly NUhiARC)

Notification of Umtsual Event (NOUE) Same as Unusual Event", below.

NPSil- Net Positive Suction llead NUAIARC - Nuclear Utility hianagement and Resources Council (now NEI)

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OBE- Operating Basis Earthquake i

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 ph PAGE D 7 of 10 EFFECTIVE DATE: 1/30/98 DEFINITIONS ODAM. OfTsite Dase Assessment Manual Opcrating Mode As defined by Technical Specification Table 1.11, Operating Mode describes the operating status of the unit. Mode designations (and the associated Reactor Mode Switch Positions) used at DAEC are: RUN/ POWER OPERATION (Run), STARTUP (Startup/llot Standby or Refuel),110T SilUTDOWN (Shutdown), COLD SilUTDOWN (Shutdown), and REFUFLINO (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 p

PAG- Protective Action Guide Planned Loss of a component or system due to expected events such as scheduled maintenance and testing ativities.

Potential Loss (of a fission product barrier) A challenge to a fission product barrier exists such that the barrier is considered degraded in its ability to perform its safety function.

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

Protected Arca Any area encompassed by physical baniers and to which access is controlled.

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

RC/C Reactor Core Isolation Cooling (system)

( - RCS Reactor Coolant System RCS Barrier - The reactor coolant system pressure boundary including the reactor } ressure vessel and all reactor coolant system piping up to and including the outermost isolation valves. ,

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PAGE D 8 of 10 EFFECTIVE DATE: I/30/98 DEFINITIONS Recognition Category A logical grouping ofInitiating Conditions, c.g., System Malfunctions.

Rem Unit of radiation dose as dermed in 10 CFR 20.1004 -

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

Rl/R - Residual Ilent Removal (system)

RHRSil' Residuallleat Removal Service Water (system)

Rocntgen (R) - Unit ofionizing radiation energy absorbed in a cubic centimeter of air RPI' Reactor Pressure Vessel Ril'CU- Reactor Water Clean-Up (system)

SBDG Standby Diesel Generator SBGT Standby Gas Treatment (system)

SBLC Standby Liquid Control (system)

SBO Station Blackout S/D - Shutdown SDC- Shutdown Cooling SDi'- Scram Discharge Vohune

,_ Sigmylcant transient - (See also, " Transient", below.) includes response to automatic or manually initiated

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V) functions such as scrams, runbacks involving greater than 25% thennal power change, ECCS injections, or undampened thennal power oscillations greater than normal.

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PAGE D 9 of 10 EFFECTIVE DATE: 1/30/98 DEFINITIONS l

Site Arca Emergency (SAE) Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guide (PAO) exposure levels except near the site boundary.

SPDS Safety Parameter Display System SRM- Startup Range Monitor SRO Senior Reactor Operator SRl'. Safety Relief Valve

(} Sustalncd windspecd- Baseline wind speed measured by meteorological tower that does not include gusts TAF. Top of Active Fuel (344.5 inches above bottom of RPV)

TEDE- Total Effective Dose Equivalent as defined in 10 CFR 20.1003 Total Leakage 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 normal 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.

7SC- 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.

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V Unidentified Leakage All leakage into the drywell that is not identified leakage.

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PAGE D 10 of 10 EFFECTIVE DATE: 1/30/98 DEFINITIONS l 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, alarm setroints, etc.) on the applicable permit.

I Unusual Event (UE) - Events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

I l'AC Volt (s) Altemating Current l'Ita/ Arca any area which contains vital equipment.

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l' ital Equipment . Any equipment, system, device or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation.

l'DC. Volt (s) Direct Current Ibild - 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, li'EC . Water Ellluent Concentration O)

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Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 V

PAGE O 1 of 3 ORGANIZATION OF flASIS INFORMATION EFFECTIVE DATE: 1/30/98 ORGANIZATION OF BASIS INFORMATION ne fonnat of the EAL Basis infonnation 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:

1. Emergency Action Level (EAL) Hasis Information Organized by Initiating Condition (IC)

Initiatino Condition Identifier For consistency, DAEC has chosen to make its initiating Condition (IC) identifiers identical to those used in NEl document NUMARC/NESP-007. The EAL Technical Basis information is orgardzed by generic IC p identifier number and name. NUMARC/NESP-007 organized the generic infonnation into four V Recognition Categories. These are:

A - Abnonnal Rad Levels / Radiological Efiluent F - Fission Product Banier Degradation 11 liazards and Other Conditions Affecting Plant Safety S - System Malfunctions For the A,11, and S recognition categories, all EAL basis infonnation is organized by IC identifier in escalating emergency class order from Umtsual Event through General Emergency. For the F recognition category, the initiating conditions are the combinations of fission product banier losses and potential losses that correspond to each emergency classification level. The individual indicators used on the fission banier table are separately discussed below, ne generic IC identifiers use two letters followed by one number.

The first letter corresponds to the event category as shown above, ne 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 The number designates whether the IC is the tirst, 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

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PAGE 0 2 of 3 ORGANIZATION OF BASIS INFORMATION EFFECTIVE DATE: I/30/98 tent are denoted by caret marks (o). 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 Tyne This is the label of the applicable row for the EAL Table shown in EPIP 1,1, Determination ofEmergency 4 Action Levels. The event type lists the general area of concem and includes Offsite Rad Conditions, Onsite Rad Conditions Natural Disasters, Fire, Other llazards 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 A based on the response to Question 5 contained in the NUMARC Afethodologyfor Development of Emergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers June 1993.

Aeolicable Ooeratina 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 1.1 1. These are:

1 - Run/ Power Operation 4 Cold Shutdowd')

2. Startup S . Refueling"

3. Hot Shutdowd')

(*)All reactor vessel head closure bolts fully tensioned.

"One or more reactor vessel head closure bolts less than fully tensioned.

Operating mode applicability of EALs is based on the operating mode that the plant was in immediately before the event- sequence leading to entry into the emergency class {fication. For example, events / conditions addressed by EALs applicable to Run mode are expected to lead to reactor trip which should bring the plant to llot Shutdown (Mode 3). However, the appropriate emergency clusification would still be based on the applicable EALs for Run/ Power Operation (Mode ,1) for these q events / conditions. If"ALL" operating modes are specifiedfor the EAL, then the EAL applies to all mocies identifled abo,v plus defueled conditions.

Duane Amold Energy Center EMEROENCY ACT, N LEVEL BASES DOCUMENT Rev.O PAGEO 3 of 3 ORGANIZATION OF BASIS INFORMATION EFFECTIVE DATE: 1/30/98 Generic Framnle 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. His 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, uus, 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.

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

2. Fission Product Barrier Table ludicators The basis information for the fission barrier table indicators is organized similarly to the other basis infonnation described above. For each barrier fuel clad, RCS, and primary containment - basis 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 Containment 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 8 ormation and references. These are displayed in the same manner as the A, H, and S recognition catego.y basis information described above.

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ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT CATEGORY O

Duane Amold Energy Center q EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 D

PAGE A 1 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/08 CATEGORY AUl 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: OITsite Rad Conditions OPERATING MODE APPLICAHILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4) m 1. A valid reading on < site specific > monitors that < > indicates that the release may have exceeded <2 x II V

site specific technical specifications for 60 minutes or longer.>

2. Confinned 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 background for 60 minutes (for sites having telemetered perimeter monitors).

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),

DAEC EAL INFORMATION:

l'alid means that the reading is from instrumentation detennined 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 of dose 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 made basedon the validreading.

The approach taken for calculation of gaseous radioactive efiluent EAL setpoints includes use of the ODAM Table 3 2 source term computed by BWR GALE for the DAEC Base Case. The release is assumed to be from a single release point. Multiple release points would be difficult to present as explicit (V3 EAL threshold values and in any case, are addressed by off-site dose assessment by MIDAS, which is the preferred method for detennining 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

Duane Amold Energy Center l

EMERGENCY ACTION LEVEL BASES DOCUhiENT Rev.O PAGE A 2 of 24 AHNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY results of the gaseous efIluent EAL calculations. The Kaman extended range capability is used because the General Electric OfTgas Stack monitor has a limited range.

Gaseous Emuent EALs Offgas Stack Kaman 9/10 Turbine Bldg (Kaman 1/2) and Reactor Bldg (Kaman 3/4. 5/6,7/8) hinximum flow (CFhi) 10,000 72,000 Release Limits Concentration Release Rate Concentration Release Rate (pCi/cc) (pCi/sec) (pCl/cc) (pCi/sec)

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+b 1.2E 3 - 4.2E+4 Alert (60 x TS) 1.9E+1 8.9E+7 3.7E-2 1.3E+6 LLRPSF Kaman 12 hinximum flow (CFhi) 99,000 Release Limits Concentration Release Rate (pCl/cc) (pCi/sec)

Tech Spec 4.5E-4 2.1 E+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, LLIU'SF (Low Level Radwaste Processing and Storage Facility) and Oligas 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-04 pCi/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.

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

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

Duane Amold Energy Center EhtERGENCY ACTION 1.EVEL llASES DOCUhtENT Rev.O PAGE A 3 of 24 ANNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY liquid radwaste so that no release of radioactive liquid to the environment is allowed. The radwaste emuent line which could be used as a batch release mechanism has a trip function that prevents exceeding the DAEC release limit, however, an EAL has been provided. 'Ihe other pathways to the environment (RllRSW to cooling tower, RllRSW - to discharge canal) have radiation monitors with readouts going to the Control Room. These systems could become contaminated if heat exchanger leaks develop; however, historically this has not occurred in the senice water systems at DAEC. These monitors ara displayed on panels IC02 and IC10.

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

The setpoints for the three senice water radiation emuent monitors vary because of differences in detector emclencies 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 case of reading the monitor scales.

hionitor TS Limit Reading UE Level Alert Level osW l.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 RiiRSW & ESW to 507 CPS 5E+2 CPS lE+3 CPS lE+5 CPS Discharge Crnal There are no significant deviations from the generic EALs. Ilowever, DAEC does not have a telemetered radiation monitoring system. As an alternative, use of field instruments was considered, it is not practical 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 Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE A 4 of 24 AHNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY llourly Whole Hody 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

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Dose assessment using MIDAS is based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE). This is urnewhat different from whole body dose from gaseous emuents detennined by ODAM methodology which fo.wts the basis for the radiation monitor readings calculated in accordance O' with the generic methodology. The gaseous emuent radiation monitors can only detect noble gases. The contribution oflodine's to TEDE could therefore only be determined either by: (1) utilizing MIDAS, or (2) gaseous emuent sampling. DAEC EAL 4 is written in tenns of TEDE and the gaseous emuent 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 Emuent Radiation Monitors, January 24,1995

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

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

6. NUhfARC Afethodologfor Develojnnent ofEmergency Action Levels NUhiARC/NESP 007 Revision 2 Questions andAnnvers, Jtme 1993

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O J PAGE A 5 of 24 AHNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY AU2 Unexpected Increase in Plant Radiation < >

EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICAHILITY: All 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 s remaining covered by water.

x 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.

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

DAEC EAI. INFORMATION:

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

Uncontrolled means that the condition is not the result of planned actions by the plant staffin accordance with procedures. l'alid means that the reading is from instrumentation detemiined 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 concern. The first method is by report to the control room. The other methods include use of the Floodup level indicator and the spent fuel pool level indicator. These are further described below.

p V During preparation 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 sersice by I&C personnel connecting a compensating air signal afkr 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 indication (e.g., not due to loss of compensating air signal or other instmment channel failure) of reactor

'O Duane Arnold Energy Center EhiERGENCY ACTION LEVEL BASES DOCUhiENT Rev.O lifS t n V

PAGE A-6 of 24 ABNORMAL RAD LEVELS /IIADIOLOGICAL EFFLUENT t EFFECTIVE DATE: 1/30/98 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 alamis, 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 elimhnte minor level rm penurbations from concem, DAEC uses LI 3413 indicated water level below 36 feet and lowering.

V increased radiation levels can be detected by the local refueling floor area radiation monitors, the refueling floor Continuous Air hionitor (CAhi) 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 alenned on Panel IC04B. The DAEC EAL has also been written to reflect the case where an ARhi 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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PAGE A-6 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY I

cavity level coming on span for this insuument 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 re Dicling cavity are removed and the spent fuel pool level indicator LI 3413 is used to monitor rcfueling 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 levets 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 concem, DAEC uses L13413 indicated water level below 36 feet and lowering.

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(V Increased radiatior. 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 inchtS 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 rnching 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 inoicators are high radiatka alarms from the Hot Laboratory or Administrative Building, the new fuel storage area, an6 the radwaste buildin3 Lj 1 .. .

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT I "0 (o) v i

PAGE A-6 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY cavity level coming on span for this instrument is used at DAEC as an indicator of uncontrolled reactor cavity level decrease.

DAEC Tecimical 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 ofinventny loss at DAEC include visual observation of decreasing water levels in reactor cavity or spent fuel storage pool, Reactor Building (RB) fuel storage 901 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 Irel alarm. To eliminate minor level n perturbations from concem, DAEC uses LI 3413 indicated water level below 36 feet and lowering.

(_)

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 norrnal 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 Bailding, the new fuel storage area, and the radwaste building.

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE A 7 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY

REFERENCES:

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

2. Technical Specification 3.7.8, Spent Fuel Pool Water Level

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

4. Emergency Operating Procedures (EOP) Basis Document, Breakpoin? for RC/L & L

5. Surveillance Test Procedure (STP) 3.0.0.0-OlPA, Daily t.nd 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

- Componems Between Reactor Core and Spent Fuel Pool, Within the Reactor Core, or Within the Spent n Fuel Pool u - 8. - NUMARC Melbdologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993 1 O

Duane Arnold Energy Center ,

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PAGE A-8 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY AAl 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 APPLICAHILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4)

1. A valid reading on < site-specific > monitor- that < > indicates that the release may have exceeded <200 (j

x site-speciFc technical specifications for 1.3 minutes or longer.>

2. Confirmed sample analyses for gaseous or liquid releases indicates concentra; ions 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 greater 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.

The primary methodfor declaration is by means ofdose assessment using the MID,fS computer model.

Dois is listed as DAEC EAL 4. However, if the monitor readings are sustainedfor longer than 15 minutes a.: : the required dose assessments cannot be completed within this period, then the declaration must be made basedon the validreading.

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE A-9 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY ,

Gaseous Emuent EALs OfTgas 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 Concentrntion Release Rate (pCi/cc) (pCi/sec) ( Ci/cc) (pCi/sec)

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 Alert (60 x TS) 1.9E+1 8.9E+7 3.7E-2 1.3E+6 LLRPSF Kaman 12

,g

, Maximum flow (CFM) 99,000 V Release Limits Concentration (pCi/cc)

Release Rate (pCi/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 ::urveillance 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 reactm "' ilding or turbine building ventilation alaim setpoint of 6.2 E-4 pCi/cc. The 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 progression in monitor setpoints from the AU1 through AAl to ASI. The DAEC EAL therefore addresses valid radiation levels exceeding 60 times the alarm setpoint for greater than 15 minutes. Rounded off, this corresponds to 3 E-2 Ci/cc. The corresponding offgas stack monitor value is 19.2 Ci/cc, rotmded off to 2 E+1 pCi/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 v 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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PAGE A 10 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY liquid radwaste so that no release of radioactive liquid to the emironment 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 (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, historically this has not occuned in the service water systems at DAEC. These monitors are displayed on panels 1C02 and IC10.

Reactor water is the Mely source of contamination through the service water systems as opposed to floor drain, detergent drain, and chemical waste discharge. The floor drain and detergent drains go to Radwaste Processing and would be batch released to the Radwaste efiluent discharge line (if such a release were to p occur). The chemical discharge sump is normally a radioactivity clean system and is tested by Chemistry

() 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 efficiencies 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 case of reading the monitor scales.

Monitor TS Limit Reading UE Level Alert Level OSW 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 lOWCT RHRSW & ESW to 507 CPS 5E+2 CPS lE+3 CPS 1E+5 CPS Discharge Canal DAEC does not have a telemetered radiation monitoring system. As an attemative, DAEC uses valid field survey readings outside the site boundary greater than 10 mR/hr or greater than 50 mR/hr CDE Thyroid.

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

i v

Rounded off to 10 mrem

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE A 11 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY Dose assessment using MIDAS is based on the EPA-400 methe 'ogy, e.g., use of Total Effective Dose Equivalent (TEDE). This is somewhat different from whole bos .ose from gaseous emuents determined by ODAM methodology which forms the basis for the radiatie nonitor readings calculated in AUl in accordance with the generic methodology. The gaseous emuent radiation monitors can only detect noble gases. The contribution of iodine'r to TEDE could therefore only be determined either by: (1) utilizing MIDAS, or (2) gaseous emuent sa 'ipling. DAEC EAL 4 is written in tenns of TEDE and the gaseous emuu adiation monitor readings are determined based on ODAM,

REFERENCES:

1. Offsite Dose Assessment Manual Section 6.1.2 and 7.1.2 Bases p 2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective Action V 3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Emuent Radiation Monitors, January 24,1995

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

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

6. NUAfARC Afethodologyfor Developm:nt ofEmergency Action Levels NUAIARC/NESP-007 Revision 2 Questions andAnswers, June 1993

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PAGE A-12 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 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 OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4) 1.- < Valid < Site-specific > radiation monitor readings for the refud iloor area, fuel handling area, and the fuel bridge area.>

2. Report of Visual observation ofirradiated fuel uncovered.

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' 3. Water Level less than (site-specific) feet for the Reactor Refueling Cavity that will result in Irradiated Fuel Uncovering.

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

DAEC E3 L INFORMATION:

Valid m su 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. Valid alarms are solely due to damage to irradiated fuel or loss of water level that has or will result in the uncovering ofirradiated fuel.

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 displayed 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.

O V'

Per ARP 1C04B, the applicable area radiation monitor alarms actuate when radiation levels increase above 100 mR/hr in th:: spent fuel pool area or above 10 mR/hr in the other three areas of concem. If a valid actuation of these alarms were to occur, the refueling floor would be immediately evacuated. Thus, a report of a fuel handling accident with either valid actuation of the fuel area alarms on panel 1C04B or with l

l

i Duane Amold Energy Center p EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O V

PAGE A-13 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY me sured radiation levels in the spent fuel pool or nonh fuel area are used to address the generic concem consistent with DAEC design an.1 procedures.

During preparation for reactor cavity flood up prior to entry into refuel mode, reactor vessel level instrument LI-454) (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 p fuel pool level indicator LI 3413 is used to monitor refueling water level. This measures the common V water level in the reactor cavity and the fuel pool. The bottom of the fuel transfer slot between the spent fuel pool and the rea tor 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 16 feet would '

indicate that spent fuel in the storage racks may potentially become uncovered.

F&RCIIP 5 requires that upon a loss of water level situation, that the refueling crew on the refueling 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.

• 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 concem 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

,3 room if a spent fuel assembly could not be placed in a safe storage location specified by F&RCHP 5 as

() described above.

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 PAGE A-14 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY

REFERENCES:

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

- 2. Technical Specification 3.7.8, Spent Fuel Pool Water Level

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) 3.0.0.0-01, Daily and Shift Instrument Checks

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

7. Fuel & Peac'or Component Handling Procedure (F&RCHP) 5, Procedure for Moving Core Compon Between Reactor Core and Spent Fuel Pool, Within the Reactor Core, or Within the Spent Fuel Poe dp 8. - Bechtel Drawing C-492, Reactor Building - Reactor Well, Spent Fuel & Drycr-Separator Pool General

- Arrangement, Rev. 6

9. Bechtel Drawing C-493, Reactor Building - Spent Fuel Liner Plan Elevations and Details, Sheet 1, ,

Rev.6

10. Holtec International Drawing No.1045, Rack Construction - Spent Fuel Storage Racks, Rev. 3 11.NUMARC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993

i l

Duane Ar..old Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O 3

G PAGE A-15 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT IIFFECTIVE DATE: 1/30/98 CATEGORY AA3 Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or to Maintain Cold Shutdown EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2) g 1. Valid (site-specific) radiation monitor readings GREATER THAN (site specific) values in areas requiring continuous occupancy to maintain plant safety functions < >

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

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 EA.Ls. Per the UFSAR, the control room is the only area that is required to be continuously occupied to achi:ve 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 ramote shutdown panel IC388. DAEC EAL 2 is directly applicable to NUMARC EAL2.

The EC/OSS should 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 to events such as resin transfers or controlled movement of radioactive sources result in emergency (V) 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.

Dur.nc Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE A-16 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY

REFERENCES:

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

2. Abnormal Operating Procedure (AOP) 913, Fire

3. AbnormalOperating Procedure (AOP)914, Security

4. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control 'toom

5. St.tveillance Test Procedure (STP) 3.0.0.0-01, Daily and Shift Instrumer.t Checks

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

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

8. UFSAR Section 6.4,liabitability Systems

9. Bechtel Calculation DA-4, Project Number 265-002, Control Room Habitability,9/3/80 Q 10.NUAMRC Methodologyfor Development ofEmergency Action Levels NUAfARC/NESP-007 Revision 2 Nr Questions and Answers, June 1993 O

I l

Duane Amold Energy Center n EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0

( \

V PAGE A-17 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DA's E: 1/30/98 CAT EGORY 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 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 <for greater than 15 minutes which corresponds to an offsite dose of 100 mrem or 500 mrem Thyroid in an hour >.

(]

\ 2 . A valid reading sustained for 15 minutes or longer on perimeter radiation monitoring system greater than 100 mR/hr. [for sites havin;; telemetered perimeter monitors].

3. Valid dose assessment capabil:ty indicates dose consequences greater than 100 < mrem TEDE> or 500

< mrem CDE> thyroid.

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

(]

V considered. In this same vein, it was assumed that only one of the three reactor building vents is on during the release.

i Duane Amold Energy Center EMERGENCY ACMON LEVEL BASES DOCUMENT Rev.O G

PAGE A 18 of 24 AHNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY The source terms used have been p.o- oaded 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 release via the offgas stack while the CRD mix was used for releases via the turbine or reactor building vents. The source temt 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 speeds were:

Pascal Class Altitude Sneed (mnh) p D 156' 8 - 12 V D 33' 8 - 12 E 156' 8 - 12 E 33' 4-7 Though the temperature setting has no impact on the MIDAS calculati oni, 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 1Ci/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 3 plume and a peak thyroid CDE rate resulting from inhalation. Because the ASI and AGl KAMAN limits are to be based on a one hour exposure, establishing concentration limits so these peak values match the NUMARC limits is acceptable.

/N,

/ \ t y '%

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Duane Arnold Energy Center (m

EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 5 i LJ PAGE A-19 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATFGORY Site Area Emergency General Emergency Initiating Condition ASl AG1 Valid Turbine or Reactor Building ventilation rad monitor (KAMAN) reading for more than 15 0.06 Ci/cc 0.6 pCi/cc minutes above:

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

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

However, if the monitor reading., are sustained for longer than 15 minates and the required dose (3

LJ 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 elternat DAEC uses valid field survey readings outside the site boundary greater than 100 mR/hr or greater than > , mR/hr CDE Thyroid.

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 determined by ODAM methodology which forms the basis for the radiation monitor readings calculated in AUl. These factors can introduce differences that are at least as large as those introduced by using TEDE versus whole body dose. The gaseous effluent radiation monitors 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.

REFERENCFS:

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 Engineerin Calculation No. 96-007-A, Determination of DAEC Radioactive Release

() Initiating Conditions for ASI & AGl 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 l

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE A-20 of 24 ABNORMAL RAD LEVELS /RADIOLOG CAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY

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

Duane Amold Energy Center 3q EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O b PAGE A-21 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY 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 Projceted Duration of the Release < >

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 <for greater than 15 minut:s which corresponds to an O 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].

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

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 EALINFORMATION:

Vaud 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 surve/ results.

There are no significant deviations from the generic EALs.

DAEC's Meteorological Information and Dose Assessment System (MIDAS) was utilized to detennine 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 O

-O considered. In this same vein, it was assuned that only one of the three reactor building vents is on during the release.

Duane Amold Energy Center

,f- EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0

(

PAGE A 22 of 24 f

ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY 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 release via the offgas stack while th- CRD mix was used for releases via the turbine or reactor building vents. The source term for a release via the ofTgas 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 speeds were:

Pascal Class Altitude Speed (mnh)

A D 156' 8- 12 V 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 plume and a peak thyroid CDE rate resulting from inhalation. Because the ASl and AGl KAMAN limits am to be based on a one hour exposure, establishing concentration limits so these peak values match the NUMARC limits is acceptable, p,

.Y

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0

>O PAGE A 23 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 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.% 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 Ci/cc above:

The preferred methodfor declaration is by means ofdose assessment using the AflDAS computer model

. and is therefore is listed as DAEC EAL 4. However, ifthe monitor readings are sustainedfor longer than D

V 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 alternative, DAEC uses valid field survey readings outside the site boundary greater than-1,000 mR/hr or greater than 5,000 mR/hr CDE Thyroid.

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 bcdy dose from gaseous effluents determined by ODAM methcdology which forms the basis for the radiation monitor readings calculated in AUl. These factors can introduce differences that are at least as large as those introduced by using TEDE versus whole body dose. The gaseous effluent 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 7.1.1 Bases

. 2. Emergency Plan Implementing Procedure (EPIP) 3.3s 1 Assessment and Protective Action >

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

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

6. EPA 400-R-92-001, AfanualofProtective Action Guides andProtective Actionsfor NuclearIncidents

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rey,0 PAGE A 24 of 24 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT EFFECTIVE DATE: 1/30/98 CATEGORY

7. NUMARC Methodologyfor Development q, Emergency Action Levels NUMARC/NESP-007 Revision 2 Questions andAnswers, June 1993 b

o O

ya pduI iiin .

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O O FISSION PRODUCT BARRIER DEGRADATION CATEGORY O

Duane Arnold Energy Center

!p EMERGENCY ACTION LEVEL BASES DOCUMENT Rev, O LJ PAGE F-1 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY I

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

EV ENT 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:

O

'd 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.

l 1

Duane Arnold Energy Center EhiERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE F 2 of 27 4 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY 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 Shttdown 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 Barrict Table,

REFERENCES:

S:e the Fission Barrier Table indicators discussed later in this section.

(3 V

Duane Arnold Energy Center fm EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 V ~

PAGE F-3 of 27 l

FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FS1 < Loss Or Potential Loss of Any Two Barriers >

EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown 1

EXAMPLE EMERGENCY ACTION LEVELS:

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

DAEC INFORMATION:

(~h O The entry conditions for this Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table. DAEC uses " Loss Or Potential Loss of Any Two Barriers." This logic is simplified from the generic logic based on the following considerations:

1. 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.

2. Comorehensiveness - A compar: son 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 NUMARC/NESP-007 logic are significantly affected by the simplified logic when applied to a BWR.

See the table below.

REFERENCES:

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

O.

i 1

Duane ntnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE F-4 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY

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

2. D, N D, N 3-D D l

• D D 5.

D. N D, N D, N D, N 7- D, N D, N 8-D D 9-D D 10 D D II D, N D, N 12 D D D - Barrier status addressed by DAEC simplified logic (Loss Or Potential Loss of Any Two Barriers) 9 N -Barrier stems addressed by NUMARC/NESP-007 genaic logic (Loss of BOTH Fuel Clad AND RCS OR Potential Loss of BOTH Fuel Clad AND RCS OR Potential Loss of EITHER Fuel Clad OR RCS AND Loss of ANY Additimal Barrier)

Duane Arnold Energy Center ew EMERGENCY ACTION LEVEL BASES DOCUMENT Rey,0 PAGE F-5 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY k

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

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

DAEC INFORMATION:

O)

( See the Fission Barrier Table indicators discussed later in 4his 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 .ndicators discussed later in this section.

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 G "

PAGE F-6 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: Fuel Clad DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:

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

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

There is no significant deviation from the generic " loss" indicator. Per NUMARC/NESP-007, the (site-specific) reading is a value which indicates release into the drywell of reactor coolant with elevated activity 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 damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose rates at the locations of the containment atmospheric monitoring system radiation detectors in the drywell 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 essumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessmene are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases wue 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 p 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 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

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 PAGE F-7 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY i

3 characteristics of the detector systems. The figures show a drywell reading of about 2.9 x 10 R/hr or a 2

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

Calculation of Drywell and Torus Monitor Readings Assuming 5% Gap Release 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/hr 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 /> for torus = 1.1 x 10 R/hr O

V 2 Torus reading = 1.1 x 10 R/hr x (5 % / 20 %] = 2.75 x 10' R/hr, round off as 3 E+1 R/hr 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

[]

o

l Duane Arnold Energy Center  !

ex EMERGENCY ACTION LEVEL 13ASES DOCUMENT Rev.0 4

ty PAGE F 8 of 27 1

FISSION PRODUCT HARRIER DEGRADATION EFFEC TIVE DATE: 1/30/98 CATEGOR's FISSION HARRIER: Fuel Clad DAEC INI)lCATOR: Radiation / Core Damage GENERIC INDICATOR:

Primary Coolant Activity Level LOSS - Coolant activity OREATER TilAN (site specific) value POTENTIAL LOSS Not Applicable DaEC INFORMATION:

There is no significant deviation from the generic indicator. Consistent with the generic methodology, O' DAEC uses a coolant activity value of 300 pCilgm lui 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

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O fO PAGE F 9 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: Fuel Clad DAEC INDICATOR: Radia':.or' Core Damage GENERIC INDICATOR:

Other(Site Specine)lndications S LOSS -(Site specific) as applicable POTENTI AL LOSS - (Site specific) as applicable DAEC INFORMATION:

p As a site specific loss indicator, DAEC uses determination of at least 5% fuel clad damage, which is

\ consistent with the containment rad monitor reading indicatnrs described previously. This can be determined from the appropriate fuel damage assessment procedures.

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

REFERENCES:

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

_ _ _ _ _ _ _ .=

0

Duane Arnoid Energy Center n EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.3 PAGE F 10 of 27 i

FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: Fuel Clad t

DAEC INDICATOR: RPV Level GENERIC INDICATOR:

Reactor Vessel Water Level 1.OSS - Level LESS TilAN (site specific) value POTENTIAL LOSS - Level LESS TilAN (site-specific) value DAEC INFORMATION:

s' 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 te assure core cooling without further degradation of the fuel clad. DAEC uses the Minimum Steam Cooling RPV Water Level of 30 inches.

This 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 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.

'Ihe 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 I of 1

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

3. Emergency Operating Procedure (EOP) Basis, Curves and Limits, CS, Minimum Steam Cooling RPV Water Level w

Duane "enold Energy Center EMERWNCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE F 11 of 27 FISSION PRODUCT HARRIER DEGRAI)ATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: Fuel Clad DAELINDICATOR: 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 DAEC INFORMATION:

There is no significant deviation from the generic indicator. Per EPIP 7.1, Emergeacy Coordinator buties,

{ 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 ba.rier degradation, degraded barrier monitoring capability, and consideration of dominant accident sequences.

Imminent means that no turnamund in safety system performance is expected and that Oeneral Emergency conditions can be expected to 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 ofTsite 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 deter.nining if a fission barrier loss or potential loss has occuned. This assessment should also include consideration for instrumentation operability, portable instrumentation readings, and ofTsite monitoring results. Dominant accident scquences 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 olTsite power with early llPC1/RCIC failure. The EC/OSS should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification declaration.

REFERENCES:

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

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

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE F-12 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: RCS DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:

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

l l'alid means that the reading is from instrumentation determined to be operable in accordance with the

'N 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 sun'ey results.

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, / e., to exclude readings due to N 16 cffects which are typically 5 to 8 R/hr at full power conditions.

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 (l.c., 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 Clt.d ba rier. Thus, this indicator would be indicative of a RCS leak only, if the radiation monitor reading increased to that value spxified by the Radiation / Core indicator applying to the Fuel Clad banier, 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 rates at the locations of the containment atmosphere monitoring system radiation detectors in the drywell 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 detemiining a minimum threshold reading than inventory assumptions found in the NRC c 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 evaluated. In the first case, the released activity was assumed to be contained in the dr>well 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.

t Duane Amold Energy Center EMERGENCY ACTION LEVEL 13ASES DOCUMENT Rev.O C

PAOB F 13 of 27 ,

FISSION PRODUCT HARRILA DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY  !

Tnis 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 fonn of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dos. rate calculations were carried out independent of any specific infomiation 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 unifomi dispersal of radionuclides into the drywell inunediately after shutdown, a drywell monitor reading is calculated:

Calculation of Drywell Monitor Reading Assuming 0.01% Gap Release NO 88-0966 value for 100% Oap Release at 0.01 minutes = 2.1 x 10' R/hr (2.1 x 10' ) R/hr x [(1 x 10 ) percent /100 percent] = (2.1) x 10" R/hr = 2.1 x 10 R/hr = 2 R/hr To assure an indicator that is readily discemible on the drywell rr.diation monitor scale, DAEC uses a valid reading above 5 R/hr aller reactor shutdown.

REFERENCES:

1. OfTice Memo NO 88-0966, O.E. Fuel Damage Documentation / Dose wie Calculations,03/18/88

2. . Technical Specification 3.4.5, Drywell Leak Detection Instmmentation J

Duane Amold Energy Center EMEROENCY ACTION LEVEL BASES DOCUMENT Rev.0

~

PAGE F 14 of 27 FISSIGN PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: RCS DAEC INDICATOR: RPV Level GENERIC INDICATOR:

Reactor Vessel Water Level LOSS Level LESS TilAN (site-specific) value POTENTIAL LOSS -Not applicable DAEC INFORMATION:

There is no significant deviation from the generic indicator. This (site-specific) loss indicator corresponds bq to the 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. Emerg-ncy Operating Procedures (EOP) Basis, Breakpoints

(-

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 O PAGE F 15 of 27

FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE

1/30/98 CATEGORY i

FISSION HARRIER: RCS DAEC INDICATOR: Leakage GENERIC INDICATOR:

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

D

-C l'alid 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 rhnt 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 l'alve (SRI) discharge through the SRI' discharge piping into the torus below the water line. SRV leakage is addressed by SUS, RCS Leakage.

Unisoleble 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, thercSre 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 Methodologefor Development of Emergency Action Levels NUMARC/NESP-007 Revision 2 Questions and Answers, June 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 banier loss indicator in the generic guidance because this n event typically results in a puff release with dose consequences greater than 10 millirem whole body, i.e.,

offsite 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. Ilowever, UFSAR Section 15.6.6, Table 15.6-1, Steam-Line Break - Radiological Eifects for Puff Release at 47 Meters, Total Dose, shows a maximum

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE F 16 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY dos of 0.58 mrem (5.8E-04 rem) passing cloud whole body dose using consenative assumptim.

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 (ARI ) IC04B, Reactor Water Cleanup and Recirculation

2. Alami 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 O 5. NUhIARC hfethodologvfor Ikvelopment ofEmergency Action Levels NUAfARC/NESP-007 Revision 2 Questions and Answers, June 1993 O

Duane Arnold Energy Center p EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O G

PAGE F 17 of 27 FISSION PRODUCT HARRIFR DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: RCS I)AEC INDICATOR: Primary Containment Atmosphere GENERIC INDICATOR:

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

l'aN means that the reading is from instmmentation determined to be operable in s'cardance with the p(/ Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant perr.onnel, or radiological survey rt %.

There is no significant deviation from the generic indicator. The (site specific) value for this loss indicator corresponds to the drywell high pressure ECCS initiation signal setpoint of 2.0 psig. DAEC also specifies that drywell cooling is operating to assure that the indicator is not misapplied to conditions that do not indicate RCS leakage into the drywell, i.e., the drywell pressure increase is not due to loss of drywell cooling.

DAEC uses a GE Mark 1 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,( psig)it is necessary to select the DAEC ECCS initiation setpoint of 2 psig to indicate an actual loss o. the RCS. Dr>well cooling is no' isolated at the 2 psig ECCS initiation setpoint, therefor further pressure rise would be indicative of a RCS leak.

REFERENCES:

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

2. Emergency Operating Procedures (EOP)-1, RPV Control (p) 3. Emergency Operating Procedures (EOP)-2, Primary C-ontainment Control 1

Duane Amold Energy Center p EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 d

PAGE F-18 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: RCS DAEC INDICATOR: Primary Containment Atmosphere GENERIC INDICATOR:

Emergency Director Judgment Any condition which in thejudgment of the Emergency Director that indicates LOSS or POTENTIAL LOSS of the RCS barrier DAEC INFORMATION:

There is no significant deviation from the generic EAL Per EPIP 7.1, Emergency Coordinator Duties, the Cq 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 tumaround in safety system performance is expected and that General Emergency conditions will occur within two hours. Imminent tission barrier degra.'ation must be considered by the EC/OSS to assure timely' declaration of a General Fmergency and to better assure that offsite protective actions can be efTectively accomplished. Degraded barrier monitoring capability from loss of/ lack of reliable indicators must also be considered by the EC/OSS when detennining if a fission banier loss or potential loss has occuned. This assessment should also include consideration for instrumentation operability, portable instrumentation readings, and offsit: 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 llPCl/RCIC failure. The EC/OSS should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification declaration.

For the PCS barrier, the EC/OSS should also consider safety-relief valves (SRVs) open or cycling, lf an SRV is stuck open or is cycling and no other emergency conditions exist, an emergency declaration may p not be appropriate. Ifowever, ifthefuelis damaged and the SRYis allowingfissionproducts to escape into

() 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.

Duane Arnold Energy Center i

EMEROENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE F 19 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY

REFERENCES:

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

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

3. NUAMRC Alethodologfor Development ofEmergency Action Levels NUAMRC/NESP-007 Revision 2 Questions andAnswers, June 1993 0

O

Duane Arnold Energy Center e EMERGENCY ACTION LEVEL BASES DOCUMENT Rey,0

\

PAGE F 20 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: RCS DAEC INI)lCATOR: None GENERIC INDICATOR:

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

Other indicators were also considered. No other reliable incidors of RCS barrier " loss" or " potential loss" could be determined.

REFERENCES:

None l

lO L. , - -. .. .,

, Duane Amold F.nergy Center

! EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 lO PAGE F 21 of 27 ,

FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: Primary Containment DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:

Significant Radioactive Inventory in Containment LOSS Not applicable POTENTIAL LOSS - Containment Rad Monitor reading GREATER TlIAN (site-specific) R/hr ,

DAEC INFORMATION:

There is no significant deviation from the generic indicators. The " potential loss" (site specific) indicator vi value corresponds to at least 20% fuel clad damage with release into the primary containment. This indicator corresponds to loss of both the Fuel Clad and RCS baniers 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 ofbarriu potentialloss 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 rates at the locations of the containment atmospheric monitoring system radiation detectors in the drywell 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 detennining a minimum threshold readmg 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 evaluated. In the first case, the released actisity was cssumed 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 chamberc The results for each case were provided for each cr.se in the form of gamma ray dose rate versus g 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 infonnation on details of construction or response characteristics of the detector systems. 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 aller shutdown. These

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE F 22 of 27 FISSION PRODUCT BARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY 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 Sup; ort Center to be operational and core damage assessment to begin.

REFERENCES:

1. Office Memo NG.88-0066, G.E. Fuel Damage Documentation / Dose Rate Calculations,03/18/88 1

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev, O PAGE F 23 of 27 FISSION PRODUCT HARRIER 1)EGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: Primary Containment DAEC INDICATOR: Radiation / Core Damage GENERIC INDICATOR:

Other(Site Specific) Indications LOSS - (Site-specific) as applicable POTENTIAL LOSS - (Site-specific) as applicable w

DAEC INFORMATION: 8 C As a site-specific " potential loss" indicator, DAEC uses determinstion of at least 20% fuel clad damage, which is consistent with the level of fuel damage indicated by the drywell and torus radiation monitor readings above, his can be determined using appropriate fuel damage assessment procedures. Regard / css ofwhether primary containment integrity is challenged, it is possiblefor significant radioactivity within the primary containment to result in EPA PAG plume exposure levels being exceeded even assuming that the primary containment is within technical specification allmvable 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 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.

REFERENCES:

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

2. NUREG-1228, Source Term Estimations During incident Respcmse to Severe Nuclear Power Plant Accidents, October 1988 O

V

Duane Amold Energy Center p EMERGENCY ACTION LEVEL UASES DOCUMENT Rev,0 d

PAGE F 24 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: Primary Containment DAEC INDICATOR: RPV Level GENERIC INDICATOR:

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

DAEC INFORMATION:

The underlying concem for this indicator is a threshold that represents significant uncovering of the core and imminent core danmge, 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 concem, 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 suflicient 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 p eclude significant fuel clad damage and hydrogen generation for as long as possible when no sources of RPV makeup water are available, Thus, for RPV water level below -40 inches, if no source of injection water was available, water b vels would coatinue 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 meh is imminent if this condition weie to occur. It would not be acceptable to delay the declaration of the Oeneral Emergency and issuance of Protective Action Recommendations beyond this point.

REFERENCES:

1. Eir.tgency Operating Procedure (EOP) Bases Document, Curves and Limits J 2. Emergency Operating Procedure (EOP) RPV/F - RPV Flocding

3. NUAL4RC Methodologyfor Development ofEmergency Action Levels NUMARC/NESP-007 Revision 2 Questlons and Answers, June 1993

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE F.25 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: Primary Containment 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:

l'alid 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 " loss" indicators used at DAEC directly correspond to the generic indicators Venting of the primary containment can be perfonned in accordance with EOP 2 irrespective of the offsite radioactivity release .

rate that will occur and by defeating isolation interlocks as 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.

' 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 of leakage 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 Opemting Procedure (EOP) 2, Primary Containment Control

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

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

.,r- ., ., -, .,. - , m.-... -, y.y- - , - - - - - =,m

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 PAGE F 26 of 27 FISSION PRODUCT llARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION llARRIER: 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 POTENTI AL LOSS - (site-specific) PSIO ( w explosive mixture exists DAEC INFORMATION:

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 toms 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% 11 2 and less than 5% O 2 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) PCil - Primary Containment liydrogen 5

(^ __

1 I

l Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE F 27 of 27 FISSION PRODUCT HARRIER DEGRADATION EFFECTIVE DATE: 1/30/98 CATEGORY FISSION HARRIER: Primary Containment 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 RCS barrier DAEC INFORMATION:

O There is no significant deviation from the generic indicator. Per EPIP 7,1, Emergency Coordinator Duties, the Emergency Coordinator / Operations Shift Supervisor (EC/OSS) performs the emergency director function at DAEC. EC/OSS considerations for determining whether any banier " Loss" or " Potential Loss" include imminent bairier degradation, degraded barrier monitoring capability, and consideration of-dominant acciden acquences.

hnminent means that no tumaround 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 efTectively 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 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 llPC1/RCIC failure. The EC/OSS should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification declaration.

REFERENCES:

1, Emergency Plan implementing Procedure (EPIP) 7.?, Emergency Coordinator Duties O 2. Duane Arnold Enercy Center Individual Plant Examination (IPE) November 1992

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Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O p/

C PAGE111 of 25 l

IIAZARDS AND OTilER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY 1101 Natural and Destructive Phenomena Affecting the Protected Area EVENT TYPE: Natural Disasters, Other llazards and Failures OPERATING MODE APPLICAHILITY: All EXAMPl.E EMERGENCY ACTION LEVELS: (1 or 2 or 3 or 4 or 5 or 6 or 7)  ;

l

1. (Site Specific) method indicates felt earthquake, l

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

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

/ T 4. Vehicle crash into plant structures or systems within protected area boundary. I

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

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

l 7. (Site Specific) occurrences.

DAEC EAL INFORMATION:

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 l

within the protected area or within *he 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 I to verification. No attempt is made to assess the actual magnitude of th : damage. Such damage can be due to collision, tomadoes, missiles, or any other cause. Damage can be indicated by report to the control

room, physical observation, or by Con'.rol 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. DAEC EAL 4 addresses a vehicle (automobile, aircraft forklift truck or train) 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 office or warehouse structures. Escalation to Alert under HAl would occur if damage was sufficient to affect the ability to achieve or maintain safe shutdown, e.g, damage made required equipment inoperable or V structural damage was observed such as bent supports or pressure boundary leakage.

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

Duane Amold Energy Center EhtERGENCY ACTION LEVEL DASES DOCUhiENT Rev.0

,O PAGE112 of 25 HAZARDS AND OTilER CONDITIONS AFFECTING EFFECTIVE DATE: 130/98 PLANT SAFETY CATEGORY rignificant 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 obsenable 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 extemal flood water levels, intemal flooding, and low river water level affecting the ultimate heat sink. DAEC EAL 7 addresses the observed efTects of flooding in accordance with AOP 902. Plant sit: Pnished grade is at elevation 757.0 fl. - Personnel doors and railroad and tmek openings at or near grade would require O

v protectioriin the event of a flood above elevation 757.0 fl. Therefore, EAL 7 uses a threshold of flood water levels above 757.0 ft.

DAEC EAL 8 addresses intemal flooding can be due to system malfunctions, component failures, or rep"r activity mishaps (such as failed freeze sect) 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, ne Maximum Nonnal 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 systems functioning pmperly. Exceeding these limits is an entry condition into EOP 3, Secondary Containment Control and may be an indication that water from a primary system la 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 emergency classification. The maximum normal operating water level limits are taken from AOP 902 and EOP 3 and are shown in the table below:

Maximum Operating Limits - Water Levels Airected Location Indicator Maximum Normal OL staximum Safe OL ,

llPCI Room Area L13768 6 inches 24 inches RCIC Room Area L13769 6 inches 18 inches A RilR Comer Room SE Area LI3770 6 inches 23 inches B RilR Comer Room NW Area L13771 6 inches 23 inches Toms Area LI3772 12 inches 24 inches EAL 9 addresses the effects of low river water level. The intake structure for the safety related water supply systems (river water, RilR service water, and emergency service water) is located on the west bank

Duane Amold Energy Center p EhiERGENCY ACTION LEVEL BASES DOCUMENT Rev.0

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PAGE113 of 25 llAZARDS AND OTilER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY 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 fl. 6 in. River water level below this level represents a potential n degradation in the level of safety of the plant and is addressed by EAL 9.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 901, Earthquake

2. Abnonnal Operating Procedure (AOP) 902, Flood

3. Abnormal Operating Procedure (AOP) 903 Tomado b(~^ 4. Emergency Operating Procedure (EOP)-3, Secondary Containment Control

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

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

7. Ilechtel Drawing BECll M017, Equipment Location Intake Structure Plans at Elevations, Rev 6 f)

N)

Duane Amold Energy Center EMERGENCY ACflON LEVEL DASES DOCUMENT Rev.0 J

PAGE11-4 of 25 liAZARDS AND OTilER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY 1102 Fire Within Protected Area HoundaryNot Extinguished Within 15 Minutes of Detection EVENT *.YPE: Fire OPERATING MODE APPLICAHILITY: All EXAMPLE EMERGENCY ACTION LEVEL:

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

V e (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 t!u administration building, and security building (buildings contiguous to the reactor building, turbine building and control building), yet, excludes fires in the warehouse or construction support center, waste-basket fires, and other small fires of no safety consequence.

Per AOP 913, the location of a fire can be detennined by observing XL3 alann messages, Zone Indicating Unit (ZiU) alanns, or fire annunciators on panels IC40 and IC40A. The location of a fire can also be determined by verbal report of the person discovering th, Bre. Perification of the alann in this context means those actions taken to determine that the contml room alann is not spurious.

REFERENCES:

1. Abnonnal Operating Piocedure (AOP) 913, Fire

2. Abnomial Operating Procedure (AOP) 914, Security

Duane Amold Energy Center EhiERGENCY ACTION LEVEL BASES DOCUhiENT Rev,0 0-PAGE115 of 25 IIAZARDS AND OTilER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY 11U3 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant EVENT TVPE: Other llazards and Failures OPERATING h10DE APPLICAHILITY: All EXAh1PLE Eh1ERGENCY 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 affect normal operation of the plant.

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

DAEC EAL INFORh1ATION:

There is no significant deviation from the generic EALs. This IC is based on releases in concentrations within the site boundary that will affect the health of plant personnel or affecting the safe operation of the plant with the plant being within the evacuation area of an ofTsite event (i.e., tanker truck accident releasing toxic gases, etc.) The evacuation area is as determined from the DOT Evacuation Tables for Selected llazardous Materials, in the DOT Emergency Response Guide for llazardous Materials.

For the purposes of this IC, CO2 (such as is discharged by the fire . /ppression system) is not toxic, CO2 can be lethal if it reduces oxygen to low concentrations that are immediately dangerous to life and health (IDLil). CO discharge 2 into an area is not basisfor emergency classification under this IC unless: (1)

Access to the affected area is required, and (2) CO2 concentration results in conditions that make the area uninhabitable or inaccenible (i.e., IDLil).

REFERENCES:

1. UFSAR Section 2.2, Nearby industrial. Transportation, and hiilitary Faciiitics Q

-V

2. UFSAR Section 6.4,liabitability Systems

Duane Arnold F:nergy Center i

EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE11-6 of 25 IIAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY IIU4 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant EVENTTYPE: Security OPERATING MODE APPLICAHILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)

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

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

D GC 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 Contingent Plan.

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

Suspected sabotage devices discovered within the plant Vital Area would result in escalatic,n via other Security Event ICs.

REFERENCES:

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

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Duane Arnold Energy Center q EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 V

PAGE117 of 25 IIAZARDS AND OTilER CONDITIONS AFFECIING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY I

11U5 Other Conditions Existing Which in the Judgment of the <EC/OSS> Warrant Declaration of an Unusual Event EVENT TYPE: EC/OSS Judgment OPERATING MODE APPLICAHILITY: 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.

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 Notification of Unusual Event emergency classification description contained in NUREG-06S4, Appendix 1 that is retained under the generic methodology.

REFERENCES:

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

2. NUREG-06S4iFEMA REP.l, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, October 1980, Appendix 1 v - -

I Duane Amold Energy Center p EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE H 8 of 25 IIAZARDS AND OTilER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY HA1 Natural and Destructive Phenomena Affecting the Plant Vital Arce EVENT TYPE: Natural Disasters, Other llazards and Failures OPERATING MODE APPLICAHILITY: 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. Tomada or high winds striking plant vital areas: Tomado or high winds greater than (site-specific) mph strike within protected area boundary.

7 3. Report of any visible structural danu:ge on < site-specific structures >

(V 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., bu: whether shere is resultant damage to equipment or s:ructures required to achieve or maintain safe shutdown, regardless of th: cause. Determination of damage atTecting the ability to achieve or maintain safe shutdown c:m % indicated by reports to the control room, physical obs.:rvation or by Control Room / local control station instrumentation.

I y

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 EAI 2 addresses report of a tomado striking a plant vital area. DAEC Ecl. 3 addresses a report to the control room of damage affecting safe shutdown areas. The reported damage can be from tomadoes, high winds, ficoding, missiles, col;isions, or any other cause.

fg DAEC EAL 4 addresses vehicle (automobile, aircraft, forklift, truck or train) confirmed crashes affecting V plant vital areas. This does not include vehicle crashes rith each other or damage to office or w:rehouse 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.

Duane Amold Energy Center n EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O U

PAGE H-9 of 25 IIAZARDS AND OTilER CONDITIONS AFFECflNG EFFECTIVE DATE.1/30/98 PLANT SAFETY CATEGORY llowever, 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 afTecting safe shutdown areas. Such missiles can be from any cause, e.g.,

tornado-generated; turbine, pump or other rotating machinery catastrophic failure; or 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 areas ofconcern.

CT Safe Shutdown Areas

'd Category Area Electrical Switchyard,1G31 DG and Day Tank Rooms,1G21 DG and Day Tank Rooms, Power Battery Rooms, Essutial 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 concems 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 head resulting frc:n 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-elevation 767.0 fl. 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.

[]

'd 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 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

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE H 10 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY requirements of EOP 3, Secondary Containment Control. If RPV pressure reduction will have no effect on leakage into secondary containment, then EOP 3 requires that reactor shutdown be performed in accordance with Integrated Plant Operatinr 'nstruction (IPOI) 3,4, or 5 as necessary if the water level exceeds its maximum safe operating limits two or more areas, if RPV pressure reduction will decrease 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.

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

Required means that the reactor sEidown was procedurally mandated by EOP 3 and is not merely perfonned as a precaution or inadve :ntly. Maimum Safe Operating Limits are defmed as the highest parameter value at which neither (1) equipment necess.c for safe shutdown of the plant will fail nor (2) 9 personnel access necessary for the safe shutdown of the .mt will be precluded. The intemal flooding can be due to system malfunctions, component failures, or repair activity mishaps (such as failed freeze scal) that can threaten safe opere6on of the plant. This includes water intrusion on equipment that is not designed to be submerged (e.g., motor control cc. A.

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

Maximum Operating Limits - Water Levels Affected Location Indicator Maximum Normal OL Maximum Safe OL llPCI Room Area LI3768 2 inches 6 inches RCIC Room Area LI3769 2 inches 6 inches A RHR Comer Room SE Area LI3770 2 inches 10 inches B RHR Comer Room NW Area LI3771 2 inches 10 inches Torus Area LI3772 2 inches 12 inches DAEC EAL 9 addresses the effects oflow 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 of the Cedar River. The overflow weir is at elevation 724 feet 6 inches. River level at or below this elevution will result in all river flow being diverted to the safety related water supply systems. The top of the intake structuie cround the pump wells is at elevation 724 feet. If the river water level dropped to this 9 level, the pump suction would have no continuous supply. Therefore, this EAL uses a threshold of water level below 724 feet 6 inches as a potential substantial degradation of the ultimate heat sink capability.

-s --- - - .. - . .i _ . _ . . _

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGEli ll of 25 IIAZARDS AND OTilER CONDITIONS AFFECTING FFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY

REFERENCES:

! . Abnorms! Operating Procedure (AOP) 901, Earthquake

2. Abnormal Operating Procedure (AOP) 902, Fical

3. Abnormal Operating Procedure (AOP) 903, Tomado

4. Abnormal Operating Procedure (AOP) 913, Fire

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

6. 1 TFS AR Chapter 3, Design of Structures, Components, Equipment, and Systems

7. Bechtel Drawing DECII-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

O

g Duane Arnold Energy Center ut p EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O V

PAGE11-12 of 2' ilAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY IIA 2 Fire Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown EVENTTYPE: Fire g OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:

1. The following conditions exist:

a. Fire or explosion in < site-specific areas >

O AND V I Affected system parameter indications show degraded performance or plant personnel report visible damage to pennanent structures or equipment within the specified area.

DAEC EAL INFORMATION:

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

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 1C40 and 1C40A.

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Duane Arnold Eners;y Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0

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PAGE1113 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY This table is displayed as an aid to the Emergency Coordinator in determining appropriate areas of concern.

Systems & Equipment of Concem e Reactivity Control e Containment (Drywellfforus) e RilR/ Core Spray /SRV's

. IIPCl/RCIC e RilRSW/ River Water /ESW e Onsite AC Power /EDG's

. Offsite AC Power

-Q U

=

e Instrument AC DC Power e Remote Shutdown Capability NOTE:

Scope of Systems and Equipment of concern established by review of Appendix R Safe Shutdown credited system;. Only those systems directly affecting safe shutdown or heat removal are listed for consideration, 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 of suficientforce 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 suilicient for the declaration. The EC/OSS also needs to consider any security aspets ofthe explosions, ifapplicable.

Per the UFSAR, the control room is the only area that is required a 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, remo'e shutdown panel IC388 is utilized in accordance with

' AOP 915.

Duane Arnold Energy Center ,

I EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 0 PAGE H 14 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 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

'v

(

O

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rey,0 PAGE H 15 of 25 llAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY llA3 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 s

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.

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

DAEC EAL INFORMATION:

There is no significant deviation from the generic EALs. This IC, in addition to IC HAS below, also -

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. CO 2 can be lethal if it 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 nmder this IC unless: (1)

Access to the affected 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 the control room becomes uninhabitable, remote shutdown panel IC388 is utilized to achieve and maintain cold shutdown.

O v

i l

9

, Duane Arnold Energy Center q EMERGENCY ACTION LEVEL BASES DOCUMENT Rev. 0 V 1 PAGE 11-16 of 25 HAZARDS AND OTiiER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY Per AOPs 913 and 914, the following areas are identified as safe shutdown areas. This table is displayedas an aid to the Emergency Coordinator in determining appropriate areas ofconcern.

Safe Shutdown Areas Category - Area Electrical Power Switchyard, IG31 DG and Day Tank Rooms,1G21 DG and Day Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room IIcat Sink / Coolant Supply Torus Room, Intake Structure, Pumphouse Containment Drywell, Toms Emergency Systems NE, NW, SE Ccmer Rooms, HPCI Room, RCIC Room, RHR Valve Room, North CRD Area, South CRD Area

,O Other Control Building, Remote Shutdown Panel IC388 Area, Panel IC56 Area, V SBGT Room

REFERENCES:

1. Abnormal Oper-ting 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

.t

Duane Arnold Energy Center '

o EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O e

V)

PAGE H-17 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY llA4 Security Event in a Plant Protected Area EVENTTYPE: Security OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)

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

7

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

(O) DAEC EAL INFORMATION:

There is no significant deviation fiom 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 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. Abnormal Operating Procedure (AOP) 914, Security Events l' h V

Duane Arnold Energy Center ,

EMERGENCY ACTION LEVEL BASES DOCUMLWT Rev.O G

PAGE H-18 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY t

IIA 5 Control Room Evacuation lias Been Initiated EVENT TYPE: Control Room Evacuation OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:

1. Entry into (site specific) procedare for control room evacuation.

DAEC EAL INFORMATION:

There is no ignificant deviation from the generic EAL. The applicable procedure for control room evacuation at DAEC is AOP 915.

REFERENCES:

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

2. UFSAR Section 6.4, Habitability Systems 9

Duane Arnold Energy Cerer >

EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 q'O PAGE1119 of 25 IIAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY IIA 6 Other Conditions Existing Which in the Judgment o'. the <EC/OSS> Warrant Declaration of an Alert 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.

DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL.

Per EPIP 7.1, the Emergency Coordinator /0;aations Shift Supervisor (EC/OSS) is the title for the emergency director ftmetion 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. NUlGG-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

[C 1

1

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 k]J PAGE H 20of 25 IIAZARDS AND 0711ER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY llSi Security Event in a Plant Vital Area EVENTTYPE: Security OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)

1. Intrusion into plant vital area by a hostile fon:e.

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, Security Event in a Plant Protected Area, in that a hostile force has progressed from the Protected Area to 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:

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

V

Duane Arnold Energy Center EMERGENCY ACTION LEVEL llASES DOCUMENT Rev.O PAGE H 21 of 25 HAZAHDS AND OTilER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY HS2 Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established EVENT TYPE: Control Room Evacuation OPERATING MODE APPLICAHILITY: All EXAMPLE EMERGENCY ACTION LEVEL:

(

l. The following conditions exist:

a. Control room evacuation has been initiated.

I AND

b. Control of the plant cannot be established per (site-specific) procedure within (site-specific) ,

, mmutes.

DAEC EAL INFORMATION:

There is no significant d,:viation from the generic EAL. The applicable procedure for control room evacuation at DAEC is AOP 915. Based on the results of the analysis described below, DAEC uses 20 minutes as the site-specific time limit for establishing control of the plant. DAEC has satellite panels associated with the remote shutdown panel at various locations through out the plant. It physically takes an operator longer than 15 minutes to lineup all the controls at the va ious panels. Control of the plant from outside the control room is assumed when the controls are transfern.1 to remote shutdown panel 1C388 in accordance with AOP 915.

. The EC/OSS is expected to make a reasonable, informedjudgment within the 20 r:.'nute time limit that control of the plantfrom the remote shutdown panel has been established. The intent of the EAL is that control of important 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

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE H 22 of 25 HAZARDS AND OTHER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY by determining the duration of the core uncovery and the re::ulting peak cladding temperature (PCT). The_

PCT ralculations were performed by incorporating the SAFE output into the Core Heatu,n 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 scfety-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-p 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 O

l Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O a ~

PAGE1123 of 2f ,

IIA 7ARDS AND OTilER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY HS3 Other Conditions Existing Which in the Judgment of the <EC/OSS> Warrant Declaration of Site Area Emergency EVENT TYPE: EC/OSS Judgment OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:

1. Other conditions exist which in the Judgment of the Emergency Dire: tor indica'e actual or likely major failures of plant functions needed for protection of the public.

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 function at DAEC. The EAL addresses conditions that fall under the Site Area Emergency classification description contained in NUREG-0654, App:ndix 1.

REFERENCES:

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

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

v

Duane Arnold Energy Center ln EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O D

PAGE H 24 of 25 llAZARDS AND OTIIER CONDITIONS AFFECTING EFFECTIVE DATE: 1/30/98 PLANT SAFETY CATEGORY IIG1 Security Event Resulting in Loss Of Ability to Reach and Maintain Cold Shutdown EVENTTYPE: Security OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)

1. Loss of physical control of the control room due to security event.

/~T 2. Loss ofphysical control of the remote shutdown capability due to security event.

%)

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. 'Ihis 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 V

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O O PAGE H 25 of 25 IIAZARDS AND OTilER CONDITIONS AFFECTING EFFECTIVE DATE: 100/98 PLANT SAFETY CATEGORY llG2 Other Conditions Existing Which in the Judgment of the <EC/OSS> Warrant Declaration of General Emergency EVENT TYPE: EC/OSS Judgment OPERATING MODE APPLICAHILITY: All EXAMPLE EMERGENCY ACTION LEVEL:

1. Other conditions exist which in the Judgment of the Emeroency Director indicate: (1) actual or imminent substantial core degradation with potential for loss if containment, or (2) potential for O uncontrolled radionuclide releases. These releases can reasonably be expected to exceed EPA PAG U 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:

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 of Nuclear Power Plants, Revision 1, October 1980, Appendix 1 O

o w

1

O O SYSTEM MALFUNCTION CATEGORY O

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 (S

O PAGE S-1 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 SUI Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes EVENT TYPE: Loss of Power s

OPERATING MODE APPLICABILITY: All EXAMPLE EMERGENCY ACTION LEVEL:

1. The following conditions exist:

a. Loss of pcwer 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 !NFORMATION:

There is no sigr'ficant deviation from the generic EAL. This event is a precursor 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 declaration ofan Umisual Event in accordance with this EAL.

Under the condidons of concem, entry into 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 on control room panel IC08 and are listed 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 IG-31 4kV breaker I A311 and B diesel generator IG-214kV breaker I A411, respectively, close for each bus.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power

2. UFSAR Section 8.2, Offsite Power System

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

V

l Duane Arnold Energy Center e

5 EMERGENCY ACTION LEVEL L\SES DOCUML'NT Rev.0 L

PAGE S-2 of 32 SYSTEM MALFUNCTION CATEGOR'n' EFFECTIVE DATE: 1/30/98 SU2 Inability to Reach Required Shutdown Within Technical Specification Limits EVENT TYPE: Inability to Maintain Shutdown Conditions OPERATING MODE APPLICAHILITY: 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. ,

DAEC EAL INFORMATION:

(~)

(j There is no significant 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.

An immediate Notirication 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 O

x v

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 (nj PAGE S-3 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 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 a b. Compensatory non-alarming indications are available.

V AND

c. In the opinion of the < Operations > Shift Supervisor, the loss of annunciators 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 1C03,1C04, and 1C05 contain the annunciators associated with safety systems at 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 concern, 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 a+tendant and responsive to abnormal indications that relate to those systems and components that have lost annunciation. Thcrefore, the generic criterion related to specific opinion of the Operations f 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.

A

() Annunciators on IC03, IC04, and IC05 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1D13.

Duane Arnold Energy Center

(~~'s EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O N.-)

PAGE S-4 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 Indications ofloss of annunciators associated with safety systems include:

e 125 VDC charger, battery, or system annunciators on control room panel 1C08

• Loss of" scaled in" annunciators at affected panels e Failure of affected annunciator panels shiflily testing by plant operators

. Expected alarms are not received e Computer point ID B350 indicates "NSS ANN DC LOSc TRBL." (Loss of DC power to panels 1C03, 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 (oj 3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power

,s" f I

'\,_,/

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE S-5 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 SU4 Fuel Clad Degradation EVEBT TYPE: Coolant Activity OPERATING MODE APPLICABILITY: All EX AMPLE 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 Specification allowable limits.

O

() DAEC EALINFORMATION:

There are no significant deviations from the generic EALs. These EAls are precursors of more serious fuel clad degradation and are thus considered as indicating a potential degradation ofthe level ofsafety of 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.

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 technical

- specification 3.4.6, coolant activity less than or equal to 1.2 Ci/mi dose equivalent I-131.

Radiological Engineering Calculation 94-014A and UFSAR Table 15.4-1 were reviewed to determine a suitable EAL threshold for the pretreat rad monitor reading corresponding to the Tech Spec 3.4.6 coolant activity limit of 1.2 Ci/ml of dose equivalent 1-131. Using the condenser noble gas source term for the I i 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 offgas flow rate is very small (about 50 standard cubic feet per minute) compared to the total condenser free volume, dilution of the condenser noble gas concentration due to offgas flow is not

i Duane Amold Energy Center n EMERGENCY ACTION LEVEL DASES DOCUMENT Rev.O L]

PAGE S-6 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 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 value calculated as shown below.

Calculation 94-014A used an exposure rate method based on using a source term consisting of a defined 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 equilibrium with the reactor coolant inventory. Using this calculation, using dose correction factors (DCFs) for child thyroid dose from Reg. Guide 1.109, ad 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 1-131 dose equivalent (in units of pCi/ml assuming I cc equals I ml)is determined for this event. This result is then linearly scaled for rad monitor readings corresponding to the Tech Spec 3.4.6 allowable primary coolant activity of 1.2 pCi/ml 1-131 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.

tg Isotope DCF (mrem /pci) Concentration (pCi/cc) Correction Factor [DCF.m / l-131 DEQ ( Ci/cc)

DCFu3,1/ 0.736 l131 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 s Therefore, for this event, a coolant activity of 34 Ci/cc l-131 dose equivalent is calculated. Scaling the results for 1.2 Ci/cc l-131 dose equivalent, a suitable condenser source term and corresponding initial concentration in the ofTgas flow is then determined. This is then converted to a pretreat rad monitor reading by use of the monitor efliciency factor:

Pretreat Rad Monitor (RM-4104) Reading NG concentrationm% = NG concentrMon X [1.2 pCi/cc /34 pCi/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 Ci/cc/34 pCilcc]

= 1529 pCi x 0.0353 = 54.0 pCi/cc n Pretreat rad monitor reading = NG concentration X Rad monitor efficiency Rad monitor efficiency = 89.2 mR/hr / pCi/cc, therefore:

Pretreat rad monitor reading = 89.2 X 54.0 = 4800 mR/hr To account for isotopic decay and dilution effects of offgas flow round down to 4E+03 mR/hr. j

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE S-7 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 The calculation results were also reviewed to determine if suitable values for the main steam line (MSL) radiation monitors could be developed. As shown above, the rod drop accident corresponds to coolant activity of 34 pCi/cc l-131 dose equivalent. As detemiined 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 1131 dose equivalent:

MSL Reading Corresponding to 1.2 pClimi 1131 dose equivalent (11.2 pCl/cc) / (34 pCi/ccl) 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 leve's, it is not appropriatc 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, 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. 3.4.6-01, Reactor Coolant Gamma and Iodine Activity

5. Annunciator Response Pru:edure (ARP) 1C03 A, Reactor and Containment Cooling and Isolation

6. Annunciator Response Procedure (ARP) 1C05B, Reactor Control

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

Duane Amold Energy Center p EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O b

PAGE S 8 of 32 '

SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 SUS 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

<2.> Identified leakage greater than 25 gpm.

<0R>

<3.> < Valid (site specific) indication of Main Steamline Break >

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. Tints, it is possible to be operating within Technical Specification LCO Action Statement time limits and make a declaration ofan Umtsual Event in accordance with these EALs Creditfor the action statemem 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 NUAfARC Afethodology for Development of- Emergency Action Levels NUAIARC/NESP-007 Revision 2 Questions and Answers, June 1993, Fission Product Barrier-BWR section, response to question 4 which states that the main steam line break with isolation ca.. 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.

The DAEC Tech Spec Section 3.4.4 coolant system leakage LCO limits are: (1) s 5 gpm unidentified leakage, (2) s 23 gpm total leakage averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, and (3) s 2 gpm increase in

( unidentified leakage within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in Mode 1. Total leakage is defined as the sum of identified and unidentified leakage.

Duane Arnold Energy Center q EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O D

PAGE S-9 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 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 normal control rcom indications. DAEC EAL 2 uses identified Icakage set at a higher value due to the lesser significance ofidentified leakage in comparison to unidentified or pressure boundary leakage.

REFERENCES:

1. Technical Specification 3.4.4, Coolant Leakage

2. Surveillance Test Procedure No. (STP) 3.0.0.0-01, Reactor Coolant System Leak Rate Calculation

3. Operating Instruction No. (01) 920, Drywell Sump System

4. Alarm Response Procedure (ARP) 1C04B, Reactor Water Cleanup and Recirculation

5. Alarm Response Procedure (ARP) 1C04C, Reactor Water Cleanup and Recirculation O ' 6. UFSAR

7. UFSARSection 5.2.5,Loss-of Section 15.6.6, Detection of Leakage through Reactor Coolant Pressure Boundary Coolant-Accident

8. NUMARC Methodologyfor Development ofEmergency Action Levels NUAIARC/NESi' 007 Revision 2 Questions and Answers, June 1993

Duane Arnold Energy Center O EMERGENCY ACTION LEVELllASES DOCUMENT Rev.0 PAGE S .cof 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 SU6 Unplanned Loss of All Onsite or Offsite Communleations Capabilities EVENT TYPE: Instrumentation / Communication OPERATING MODE APPLICAlllLITY: All EXAMPLE EMERGENCY ACTION LEVELt

1. Either of the following conditions exist t a. Loss of all (site-specific list) onsite communications capability affecting the ability to perform routine operations.

OR g b. Loss of all(site specific list) offsite conununications capability.

V DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL. The communications methods used at DAEC are described in the Emergency Plan. In plant and extemal agency telephone communicm'on methods include l'AllX 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 communications.

The availability of one method of ordinary offsite communication .. sufficient to inform state and local h- authorities of plant problems. This EAL is intended to be used only when extraordinary means (relajing of

. Informationfrom radio transmissions, Individuals being sent to ofzite locations, etc.) are being utili:ed to make communicationspossible.

The DAEC plant operations radio system is a UllF 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 MHz 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 MHz radio. Consoles are located in the Technical Support Center

-and the Fmergency Operations Facility at th: IES Tower. The DAEC Security (backup radiological g survey) Radio System provides base to-portable security communication within the plant and with the Linn County Sheriffs Offlee usica a mobile relay (repeater) type base station and two VHF frequencies. Control consoles are located in the Secondary Alarm Station, Central Alann 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 VHF  !

l

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 PAGE S ll of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 '

frequency. The transmitter and one control console are located at the Secondary Alarm Station and in the Central Alarm Station. This station is for communications with lowa Department of Public Safety radio /

station, Linn County Sheriffs office, and the Benton County Sheriffs office. This point to po'.nt channel is also used by the Linn County Emergency Management and other public safety organizations throughout the state orlowa.

REFERENCES:

1. Emergency Plan, Section F, Emergency Communications

2. NUAMRC Methodologyfor Development ofEmergency Action Levels NUhMRC/NESP-007 Revision 1 Questlons andAnswers, June 1993 0

_=_ _ _ = = -

0

Duane Arnold Energy Center a EMERGENCY ACTION LEVEL DASES DOCUMENT Rev.O V

PAGE S 12 0f 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 SU7 Unplanned Loss of Required DC Power During Cold bautdown or Refuel < >

Mode For Greater Than 15 Minutes

, EVENT TYPE: Loss of Power OPERATING MODE APPLICAHILITY: Cold Shutdown, Refuel EXAMPLE EMERGENCY ACTION LEVEL:

1. <T>he following conditions exist:

a. Unplanned Loss of Vital DC power to required DC busses based on ' site-specific) bus voltage indications.

q AND Q . Failure to restore power to at least one required DC bus within 15 minutes from time ofloss.

DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL. Unplan red loss of Div. I and Div. 2.125 VDC busses excludes scheduled maintenance and testing activities. Under the condi'.lons 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.

The 125 VDC system is divided into two independent divisions - Division 1 (1D1) and Division 11 (1D2) -

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 co'd 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 ths, will result in RCS temperamre above 212 F".

Bus vobgc is based en the minimum bus voltage necessary for the operation of safety related equipment 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.

O

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE S 13 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98

REFERENCES:

1. - Abnormal Op: rating Procedure (AOP) 302.1, Loss of 125 VDC Power

2. Abnonnal Operating Procedure (AOP) 388, Loss of 250 VDC Power

3. Technical Specification 3.8, Electric Power Systems

4. UFSAR Section 8.3, Onsite Power Systems S. 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 O

O

Duane Amctd Energy Center EMEROENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE S 14 of 32 SYSTEM MALFUNCTI ,N CATEGORY EFFECTIVE DATE: 1/30/98 t SAI Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold < Conditions >

EVENT TYPE: Loss of Power OPERATING MODE APPLit. J.HILITY: Cold Shutdown, Refuel, Defueled EXAMPLE EMERGENCY ACTION LEVEL:

1. ' Die following conditions exist:

a. Loss of power to (site specific) transformers.

AND (q,

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 15 minutes from the time of loss of both offsite and onsite AC power.

DAEC EAL INFORM ATION:

There is no significant deviation from the generic EAL 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 1C08 and are listed in the procedure under "Piobable 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 transfomiers.

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, Electrical Power t stems

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev 0 0m .

PAGE S 15 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 I i

SA2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Ilus Been Exceeded and Manual Scram Was Successful EVENT TYPE: RPS Failure i

OPERATING MODE APPLICABILITY: Run,Stanup 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.

("s V DAEC EAL INFORMATION:

The DAEC EAL is written in terms of failure of automatic scram. IPOl 5 specifies manual scram insenion immediately following any automatic scram signal, and therefore separately specifying successful manual scram is not required. The reactor is considered successfully shutdown if either: (1) all control rods are insened to least position 02, or (2) it has been determined that the reactor will remain shutdown under Al L conditiore 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 Altemate Rod Insenion(ARI).

If the mode switch is in Startup an i the rods are1sdly inserted (i.e., the reactor is shutdown) prior to the automatic signalpilure, then declaration ofan Alert would not be required in this case, the event wordd be reported under 10 CFR $0. 72 (b) (2) (1) as afour hour report.

The condition of concem 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 q of the automatic protection system to complete the scram following generation of a scram sit;nal is the V issue of concem.

Duane Amold Energy Center I

EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE S 16of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98

REFERENCES:

?

1. Integrated Plant Operating Instruction (IPOI) No. 5, Reactor Scram

2. ATWS Emergency Operating Procedure (EOP) RPV Control

3. Emergency Operating Procedure (EOP) 1 RPV Control

4. - NWURC Methodologfor Development ofEmergency Action Levels NUAMRC/NESP-007 Revision 2 Questions andAnswers, June 1993 0

=

0

, Duane Amold Energy Center o EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 Q

PAGE S 17 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 SA3 Inability to Maintain Plant in Cold Shutdown EVENT TYPE: Inability to Maintain Shutdown Conditions OPERATING MODE APPLICAHIL!TY: Cold Shutdown, Refuel EXAMPLE EMERGENCY ACTION LEVEL:

1. Loss of < decay heat removal systems required > to maintain cold shutdown AND Temperature increase that either:

+ Exceeds Technical Specification cold shutdown temperature limit O

e OR Results' in uncontrolled temperature rise approaching cold shutdown technical specification limit.

DAEC EAL INFORMATION:

Under the conditions of concem for EAL 1, AOP 149, Loss of Decay 11 eat Removal, would be entered under Tab 1, Loss of Shutdown Cooling. Indications / alarms related to loss of shutdown cooling are displayed on control room panels IC03 and ICOS and are listed in the procedure under " Probable Indications." The procedure requires that shutdown cooling be re-established.

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.

The DAEC EAL is written to imply a RCS temperature rise above 212 'F that is not allowed by plant procedures. This corresponds to the inability to maintain required temperature conditions for Cold 3 Shutdown. " Uncontrolled" means that system temperature increase is not the result of planned actions by '

the plant staff. The wording is also intended to climinate minor cooling interruptions occurring at the transition between Hot Shutdown and Cold Shutdown or temperature changes that are permitted to occur during establishment of attemate 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.

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE S 18 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98

REFERENCES:

1. Abnormal Operating Procedure (AOP) 149, Loss of Decay lleat Removal

2. DAEC Technical Specifications

3. Surveillance Test Procedure (STP) 3.4.9 01,11eatup and Cooldown Rate log

4. NUREG 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the UnitedStates, September 1993

$. NUAfARC Afethodologyfor Development ofEmergency Action Levels NUAfARC/NESP-007 Revision 2 Questions and Answers, June 1993 0

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Duane Arnold Energy Center p EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O O ~

PAGE E 19 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 SA4 Unplanned Loss of Most or All Safety System Annunciation or Indica (icn in Control Room With Either (1) a Significant Transient in Progress, or (2)

Compensatory Non-Alarming Indicators are Unavailable EVENT TYPE: Instrumentation / Communication OPERATING MODE APPLICAHILITY: Run, Startup,110t Shutdown EXAMPLE EMERGENCY ACTION LEVEL:

1. The following conditions exist:

a. Loss of most< > annunciators associated with safety systems for greater than 15 minutes, r] AND V b. In the opinion of the < Operations > Shift Supenisor, the loss of at: 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 1C03,1C04, and 1C05 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. Significant transient includes response to automatic or manually initiated functions such as scrams, runbacks invoMng greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10% or greater, p Under the conditions of concem , entry into AOP 302.2, Loss of Alarra Panel Power, would be made. The h 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

I Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0

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V PAGE S 20 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 Shifl Supenisor 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.

Annunciators on IC03, IC04, and IC05 share a common power supply frem 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:

• 125 VDC charger, battery, or system annunciators on control room panel IC08 e Loss of" scaled in" annunciators at affected panels t Failure of afTected annunciator panels shillily testing by plant operators

, . Expected alarms are not ieceived (m) e Computer pent ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels IC03, 1C04, ar.d 1C05)

REFERENCES:

1. Operating Instmetion (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 D

b

Duane Arnold Energy Center p EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O V

PAGE S 21 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 SAS AC Power Capability to Essential Busses Reduced to a Single Power Source for Greater Than 15 MLutes Such That Any Additional Single Failure Would Result in Station Blackout EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run, Startup,llot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:

1. The following conditions exist:

a. Loss of power to (site specific) transformers for greater than 15 minutes.

AND

(]

(/ b. Onsite oower capability has been degraded to one (ttain of) emergency bus (ses) powered from a single onsite power source due to loss of:

(Site-specific list)

DAEC EAL INFORMATION:

%e DAEC EAL is written to address the underivi:ig concem, l.c., only one AC power source remains and ifit is lost, a Station Blackout will occur. Unoer the conditions of concem, entry into AOP 301, Loss of hsrential Electrical Power, woald 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 liaed in the procedure under " Probable Indications."

At DAEC, the Essential Buses of concem are 4160V Buses l A3 and 1 A4, Each of these buses feed their aesociated 480V and 120V AC busses through step down transformers. Onsite power sources at DAEC include the A a-d B Diesel Generators,1G-31 and 1G 21, respectively,

REFERENCES:

1, Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power

2. UFSAR Chapter 8 Electrical Power

3. Technical Specifications Section 3.8. Electrical Power Systems p'

U

Duane Amold Energy Center

,q EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 V

PAGE S 22 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 SSI Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run,Startup,llot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:

1. Loss of all offsite and onsite AC power as indicated by:

a. Loss of power to (site-specific) transformers.

AND

b. Failure of(si'.e-specific) emergency generators to supply power to emergency busses.

AND

'p g c. Failure to restore power to at least one emergency bus within <l5 minutes > minutes from the time ofloss of both offsite and onsite AC power.

DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL. In accordance wPh 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 1C08 and are listed in

the procedure under " Probable Indications."

l l At DAEC, the Essential Buses of concem are the 4160V Buses I A3 and 1 A4. Each of these buses feed l their associated 480V and 120V AC buses through step down transformers.

L

REFERENCES:

l

1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout

2. Technical Specifications Section 3.8, Electrical Power Systems

3. UFSAR Chapter 8, Electric Power 3

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i

Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O O

PAGE S 23 of 32 SYSTEM MALFUNCTION CATEGOlW EFFECTIVE DATE: 1/30/98 SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint lias Been Exceeded and Manual Scram Was NOT Successful EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Run, Startup EXAMPLE EMERGENCY ACTION LEVEL:

1. (Site specific) indications exist that automatic and manual scram were not successful, p

a DAEC EALINFORMATION:

The DeEC EAL addresses conditions where failure of an automatic scram has occurred and manual actions performed at panel 1C05 to quickly shutdown tbc reactor do not meet the success criteria ofIPOl 5 and the ATWS - RPV Control EOP, where power remains above 5% or baron injection is required.

Under the conditions of concern 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.

In addition, if the SRV's are open, the RCS is no longer capable of retaining fission products and therefore 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 timcly recognition and emergency response. -

REFERENCES:

1. Integrated Plant Operating Instruction (IPOI) No. 5, Reactor Scram

2. ATWS Emergency Operating Procedure (EOPL RPV Control

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

V i

1

l Duane Amold Energy Center

n. EMERGENCY ACTION LEVEL 13ASES DOCUMENT Rev.O ty PAGE S 24 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 SS3 Loss of All Vital DC Power EVENT TYPE: Loss of Power OPERATING MODE APPLICAlllLITY: Run,Startup,llot Shutdown EXAMPLE EMERGEhCY ACTION L" VEL:

1. Loss of All Wal DC Power based on (site specific) bus voltage indications for greater than 15 minutes.

DAEC EAL INFORMAT' TON:

There is no significant deviation from the generic EAL. Under the conditions of concem, AOP 302.1, Loss p/

y of 125 VDC Power, would be entered under Tab 3, Complete Loss of 125 VDC. Consequently, the DAEC EAL addresses loss of both divisions of the 125V DC system consistent with AOP.

At DAEC, the 250V/125V DC Systems ensure power is a tailable for the reactor to be shutdown safely and maintained in a safe condition. The 125V System is divided into two independent divisions - Division I and Division 11 - with separate DC power supplies. These power supplies consist of two separat: 125V batteries and chargers sen ing systems such as RCIC, RliR, EDGs and llPCI.

Complete loss of bota 125V DC Divisions could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refuelmg operations.

REFERENCES:

1. - Abnennal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power

2. Abnormal Operating Procedure (AOP) 388, Loss of 250 VDC Power

3. Technical Specification 3.8, Electrical Power Systems

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)

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 V

PAGE S 25 of 32 i SYSTEM MALFUNCTION CATEGORY EFFECTIVE D.'TE: 1/30/98 SS4 Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown EVENT TYPE: Inability to Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Rtm, Startup, Ho* Shutdown.

EXAMPLE EMERGENCY ACTION LEVEL:

1, <EOP Graph 4 Ileat Capacity Limit is exceeded >

<0R>

<2. Reactor CANNOT be brought suberitical.>

( DAEC EAL INFORMATION:

This EAL addresses complete loss of functions, including ultimate heat sink and reactivity control, required for hot shutdown with the reactor at pressure and temperature. Under these conditions, there is an actual major failure of a system intended for protection of the public. The reactivity condition criteria is addressed by maintenance of required shutdosyn margin. Ifinadvertent c iticality 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.

This EAL represents an escalation from the conditions of concem in SA3, Inability to Maintain Cold Shutdown, because the reactor is at operating pressure and temperature and decay heat levels are higher.

Per DAEC Technical Specifications, the following systems are necessary to achieve or maintain Hot Shutoown conditions:

• Reactor Protection System Instrumentation e Cme and Containment Cooling Systems Instrumentation

• Reactivity Control

• Standby Liquid Control System

. Core and Containment Cooling Systems e Primary System Boundary

"

• Auxiliary Electrical Systems

Duane Amold Energy Center p EhiERGENCY ACTION LEVEL BASES DOCUhiENT Rev.0

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PAGE S 26 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 Loss of instrumentation is addressed by SS6, Inability to hionitor a Significant Transient in Progress, below. The Auxiliary Electrical System is addressed by SS1, Station Blackout, and SS3, Loss of 125V DC, above and are therefore not covered here. Failure of the primary system boundary is covered by the Fission Barrier Table and SUS RCS Leakage, above.

REFERENCES:

1. Abnormal Operating Proce Jure (AOP) 149, Loss of Decay lleat Removal

2. Abnormal Operating Prmedure (AOP) 255.1, Control Rod hiovement/ Indication Abnormal

3. Abnormal Operating Procedure (AOP) 255.2, Power / Reactivity Abnormal Change

4. Emergency Operating Procedure (EOP) 1 RPV Control

5. ATWS Emergency Operating Procedure (EOP)- RPV Control g 6. Emergency Operating Procedure ALC Alternate Level Control

(") 7. Emergency Operating Procedure (EOP) Basis, EOP Breakpoints

8. NUMARC Methodologyfor Developinent ofEn<ergency Action Levels NUMARC/NESP-007 Revision 2 ,

Questions and Answers, June 1993 i'3 V

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE S 27 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 SS5 Loss of Water Level in the Reactor Vescel That lias 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:

l. 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.

DAEC EAL INFORMATION:

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 concem for EAL 1, AOP 149, Loss of Decay lleat Removal, would be entered under Tab 1, Loss of Shutdown Cooling.

Indications / alarms related to loss of shutdnwn cooling are displayed on control reom panels IC03 and IC05 and are listed in the procedure, consistent with the value used in the EOPs, the EAL uses an indicated PPV level of 15 inches for the water level corresponding to TAF.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Removal

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

3. Emergency Operating Procedure (EOP) Basis, EOP Breakpoints

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O PAGE S 28 of 32 SYSTEM h1ALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 SS6 Inability to Monitor a Significant Transient in Progress s.

EVENT TYPE: Instrumentation / Communication OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:

1. He following conditions exist:

a. Loss of < > annunciators associated with safety systems.

AND

b. Compensatory non alarming indications are unavailable, p AND Q c. Indications needed to monitor (site specific) safety functions are unavailable.

AND

d. <Significant> transient in progress.

DAEC EAL INFORMATION:

The DAEC EAL is written in temis of a sign (ficant transient in progress with loss of both safety system annunciators and loss of compensatory non alarming instrumentation. The DAEC EAL structure, which addresses all the key points in the generic EAL, better assures that the condition of concem for this EAL v,.il be readily recognized.

Sigmjicant transient includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or undamped thermal power oscillations greater than normal.

Control room panels IC03, IC04, and IC05 contain the annunciators associated with safety systems at DAEC. Annunciators on 1C03,1C04, and IC05 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1D13. Therefore, DAEC does not specify a loss of"most" annunciators as specified in the generic methodology.

q Compensatory non-alarming indications include the plant process computer, SPDS, plant recorders, or-g I

plant instrument displays in the control room. Dese indications are needed to monitor (site-specific) safety functions that are of concem in the generic EAL.

Duane Amold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev 0 PAGE S 29 0f 32 ,

SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 t

Indications ofloss of annunciators associated with safety systems include:

e 125 VDC charger, battery, or system annunciators on centrol room panel IC08 e Loss of" scaled in" annunciators at afTected panels

• Failure of affected annunciator panels shiftily testing by plant operators e Expected alarms are not received

. Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels IC03, 1C04, and 1C05) l

REFERENCES:

l. Opc atingInstruction(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 1

Duane Arnold Energy Center r EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0

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PAGE S 30of 32 SYSTEM MALFUNCTION CATEGOR)

EFFECTIVE .TTE: 1/30/98 SG1 Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power EVENT TYPE: Loss of Power OPERATING MODE APPLICAHILITY: Run, Startup,llot Shutdown EXAMPLE EMERGENCY ACTION LEVEL:

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

a. Loss of power to (site-specific) transfomiers.

AND n b. Failure of(site specific) emergency diesel generators to supply power to emergency busses.

V c. At least one of the following conditions exist:

AND Restoration of at least one emergency bus within (site-specific) hours is NOT likely OR

• (Site Specific) Indication of continuing degradation of core cooling based on Fission Product Barrier monitoring.

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 degn 3. Although it may be difficult to predict when power can be restored, it is necessary to give the EC/OSS a reasonable idea of how quickly a General Emergency should be declared based on the following considerations:

Are there any present indications that core cooling is already degraded .o the point where a General Emergency is IMMINENT (i.e., loss of two barriers and a potential loss of the third barrier)?

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?

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 sufficient for escalation: (1) SBO coping capability exceeded, or (2) loss of drywell coolint that continues 1

i Duane Arnold Energy Center EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.0 PAGE S 31 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 i

to make RPV water level measurements unreliable, or (3) indications ofinadequate core cooling. Each of  ;

these conditions is discussed below: l

1. SBO Coninn Canability Exceeded i

DAEC has a SBO coping duration of four hours. The likelihood ofrestoring at least one emergency bus l 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 Remaininc 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 ofinadeauate Core Cooline DAEC uses the RPV level that is used for the Fuel Clad EAL 2 " potential loss" condition. This is RPV level below +15 inches.

REFERENCES:

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," Febmary 10,1992

3. Emergency Operating Procedum (EOP)1 RPV Control

4. Emergency Operating Procedure (EOP) ALC Altemate Level Control O

l

Daane Amold Energy Center n EMERGENCY ACTION LEVEL BASES DOCUMENT Rev.O V

PAGE S 32 of 32 SYSTEM MALFUNCTION CATEGORY EFFECTIVE DATE: 1/30/98 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.

/G AND V 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:

Automatic and manual scram are not considered st.ccessful if action away from the reactor control console is required to scram the reactor. Consistent with the EOPs, the ATWS conditions of concem in this EAL are reactor power that is expected to remain above 5% or that is indeterminate.

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. .owever, without the main condenser being available as a heat sink, heat rcmoval capability under these conditions is insufficient and a threat to the Fuel Clad barrier exists. In addition, the SRV's will lif1 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 ineffective. 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.

REFERENCES:

1. Emergency Operating Procedure ATWS EOP - RPV Control