NG-04-0529, Revision to Emergency Action Levels

Revision to Emergency Action Levels

ML043240346
Person / Time
Site:	Duane Arnold
Issue date:	10/22/2004
From:	Peifer M Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
+sispjld120050824, NG-04-0529
Download: ML043240346 (383)

Text

NMC Committed to Nuclear Excellence Duane Arnold Energy Center Operated by Nuclear Management Company, LLC October 22, 2004 NG-04-0529 10 CFR 50 Appendix E (IV)(B)

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Duane Arnold Energy Center Docket 50-331 License No. DPR-49 Revision To Emergency Action Levels

References:

1) Letter from NRC, to L. Liu, IES Utilities Inc., "EVALUATION OF IES UTILITIES' PROPOSED EMERGENCY ACTION LEVEL CHANGES FOR THE DUANE ARNOLD ENERGY CENTER (TAC# M93692)"

dated May 19,1997

2) Letter from NRC, to M. Peifer, Nuclear Management Company, LLC, "DUANE ARNOLD ENERGY CENTER - NEW EMERGENCY ACTION LEVELS RELATED TO INDEPENDENT SPENT FUEL STORAGE INSTALLATION FOR THE EMERGENCY PLAN (TAC NO. MB7981)'

dated May 12, 2003

3) NRC REGULATORY ISSUE

SUMMARY

2003-18, "USE OF NEI 99-01, "METHODOLOGY FOR DEVELOPMENT OF EMERGENCY ACTION LEVELS," REVISION 4, DATED JANUARY 2003," dated October 8, 2003

4) NRC REGULATORY ISSUE

SUMMARY

2003-18, SUPPLEMENT 1, "USE OF NUCLEAR ENERGY INSTITUTE (NEI) 99-01,

'METHODOLOGY FOR DEVELOPMENT OF EMERGENCY ACTION LEVELS,' REVISION 4, DATED JANUARY 2003," dated July 13, 2004 Nuclear Management Company, LLC, (NMC) is transmitting for NRC review and approval, revisions to the Duane Arnold Energy Center (DAEC) emergency action levels (EALs). This revision implements new emergency action levels based on NEI 99-01, Revision 4, "Methodology for Development of Emergency Action Levels," using the guidance of Reference 4.

The enclosed security EALs are in compliance with NEI 99-01, Revision 4, but are not aligned with the "Order For Interim Safeguards and Security Compensatory Measures for - Duane Arnold Energy Center," dated February 25, 2002. The DAEC response to that Order is dated August 28, 2002. The NRC has indicated that additional security 3277 DAEC Road

• Palo, Iowa 52324-9785 Telephone: 319.851.7611 A

Document Control Desk Page 2 EALs are being developed for threat advisories. If revised security EALs are issued before the enclosed EALs are approved, NMC will provide a supplement to reflect the updated security EALs. NMC intends to remain in compliance with the security EALs in the August 2002 security Order response, as modified by the threat advisories, until the NRC approves new security EALs for DAEC.

EALs approved in Reference 1 implemented the guidance in NUMARC/NESP-007, "Methodology for Development of Emergency Action Levels," Revision 2 with some variations. The EALs related to DAEC's Independent Spent Fuel Storage Installation (ISFSI) which reflect the guidance in NEI 99-01, Revision 4, were previously approved in Reference 2 by the Staff.

This revision of the DAEC EALs, updates the EALs to NEI 99-01, Revision 4, for consistency of EALs within the NMC fleet. Reference 3 allows for licensees currently using NEI EAL's to upgrade to NEI 99-01 Revision 4 EALs per the IOCFR 50.54(q) process. However, these EAL changes are being submitted for review and approval to ensure compliance with existing regulations and Staff expectations.

This submittal follows the guidance provided in Reference 4. An additional column has been added to the Justification Matrix, Enclosure 4, Attachment 3, which shows the last SER approved DAEC EAL version for comparison with the current NEI wording, as well as DAEC's proposed wording. This submittal does not include supporting documentation for setpoints found in EALs which were previously approved per Reference 1 and 2.

This request contains the following enclosures: : Table of Contents : Summary Explanation : State/Local Government Official Agreement Documentation : Detailed Supporting Information : Compact Disk of Enclosures, References, and Supporting Documentation These proposed EALs have been discussed and agreed to by applicable state and local government officials, and have been reviewed and approved by the DAEC Operations Committee.

It is requested that the enclosed EAL revision be approved by November 1, 2005.

Summary of Commitments This letter makes the following new commitment:

If revised security EALs are issued before the enclosed EALs are approved, NMC will provide a supplement to reflect the updated security EALs.

Document Control Desk Page 3 Should you have any questions regarding this submittal, please contact Mr. Paul Sullivan, DAEC Emergency Planning Manager, at (319) 851-7191.

Mark A. Peifer Site Vice President, Duane Arnold Energy Center Nuclear Management Company, LLC

Enclosures:

(5)

CC Administrator, Region ll, USNRC (2 copies)

Project Manager, DAEC, USNRC Resident Inspector, DAEC, USNRC Spent Fuel Project Office, USNRC

ENCLOSURE I TABLE OF CONTENTS contains the Summary Explanation, or Executive Summary. contains the State and Local Government Official Agreement Documentation. contains the Detailed Supporting Information and is divided into three attachments:

Attachment 1: Red-line of the Technical Basis Document Attachment 2: Clean copy of the Technical Basis Document Attachment 3: Justification Matrix contains a compact disk of enclosures, references and supporting documentation.

Page I of I

ENCLOSURE 2

SUMMARY

EXPLANATION

References:

(1) Letter from NRC, to L. Liu, IES Utilities Inc., "EVALUATION OF IES UTILITIES' PROPOSED EMERGENCY ACTION LEVEL CHANGES FOR THE DUANE ARNOLD ENERGY CENTER (TAC# M93692)"

dated May 19,1997 (2) Letter from NRC, to M. Peifer, Nuclear Management Company, LLC, "DUANE ARNOLD ENERGY CENTER - NEW EMERGENCY ACTION LEVELS RELATED TO INDEPENDENT SPENT FUEL STORAGE INSTALLATION FOR THE EMERGENCY PLAN (TAC NO. MB7981)"

dated May 12, 2003 (3) NRC REGULATORY ISSUE

SUMMARY

2003-18, "USE OF NEI 99-01, 'METHODOLOGY FOR DEVELOPMENT OF EMERGENCY ACTION LEVELS,' REVISION 4, DATED JANUARY 2003," dated October 8, 2003 This submittal includes the transmittal letter, and five enclosures. The enclosures include a table of contents (Enclosure 1), this executive summary (Enclosure 2),

documentation of state and local government officials' agreements (Enclosure 3),

detailed supporting information for each emergency action level (EAL) (Enclosure 4),

and additional supporting information (Enclosure 5).

EALs approved in Reference 1 implemented the guidance in NUMARC/NESP-007, "Methodology for Development of Emergency Action Levels," Revision 2 with some variations. The EALs related to DAEC's Independent Spent Fuel Storage Installation (ISFSI) which reflect the guidance in NEI 99-01, Revision 4, were previously approved in Reference 2 by the Staff.

Nuclear Management Company, LLC (NMC) requests approval to change the existing scheme for DAEC to that described in NEI 99-01, Revision 4, "Methodology for Development of Emergency Action Levels," January 2003, as endorsed by the NRC in Regulatory Guide 1.101, " Emergency Planning and Preparedness for Nuclear Power Reactors," Revision 4, July 2003. This revision is for consistency within the (NMC) fleet.

Reference 3 allows for licensees currently using NUMARC-007 EALs to upgrade to NEI 99-01 Revision 4 EALs per the 10 CFR 50.54(q) process, however the proposed EAL changes are being submitted for review and approval to ensure compliance with existing regulations and Staff expectations.

Page l of3

The following state and local government officials' agreement is contained in Enclosure 3:

Don Flater, Iowa Department of Public Health Ned Wright, Linn County Emergency Management Agency Scott Hansen, Benton County Emergency Management Agency David Miller, Iowa Homeland Security & Emergency Management Division contains detailed supporting information for any differences/deviations from the NEI 99-01, Rev. 4. Attachment 1 contains a red-line copy of the Technical Basis Document. This document includes the pertinent information to describe each EAL (category, description, modes, basis, etc.). The red-line areas indicate the specific changes necessary. Attachment 2 contains a clean copy of the Technical Basis Document. Attachment 3 contains the detailed justification matrix. This matrix provides the cross-reference comparing the last SER approved EAL version for DAEC, the NEI 99-01 Rev. 4 version of the EAL, and the proposed DAEC version of the EAL. An additional column contains an explanation as well as justification for any differences/deviations between the proposed EAL version and the NEI 99-01 Rev. 4 version.

The Technical Basis Document, justification matrix, and supporting reference material are contained on compact disk in Enclosure 5. Supporting documentation for setpoints found in EALs which were previously approved per Reference I and 2 are not included.

Following is a description of the operational modes for DAEC:

MODE TITLE REACTOR MODE SWITCH AVERAGE REACTOR POSITION COOLANT TEMPERATURE (OF)

I Power Operation Run NA 2 Startup Refuel (a) or Startup/Hot NA

-Standby 3 Hot Shutdown (a) Shutdown > 212 4 Cold Shutdown 'a Shutdown < 212 5 Refueling b Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bolts less than fully tensioned.

It is recommended the following strategy be used during review of this submittal:

• Review the wording in Enclosure 4, Attachment 3, Justification Matrix, for a given EAL. The EAL may be flagged as a "DIFFERENCE" and an explanation for the "DIFFERENCE" is given. (Note: DAEC does not intend to deviate from NEI 99-01, Revision 4.)

Page 2 of 3

• Review the wording for the associated EAL in the corresponding red-line Basis document in Enclosure 4, Attachment 1. The EAL Basis Document contains all supporting guidance and justification for the EAL.

Defueled EALs are not included in this submittal since DAEC is not a Permanently Defueled Station.

In summary, this submittal provides the basis and justification for changing the DAEC EAL scheme to the NEI 99-01 requirements and demonstrates compliance with 10 CFR 50.54 (q).

Page 3 of 3

ENCLOSURE 3 STATE/LOCAL GOVERNMENT OFFICIAL AGREEMENT DOCUMENTATION 5 pages follow

NMc Co&mW to NudearExcelence' To: David L. Miller, Iowa Homeland Security & Emergency Management Division Don Flater, Iowa Department of Public Health Scott Hansen, Benton County Emergency Management Agency Ned Wright, Linn County Emergency Management Agency From: Don A. Johnson Date: 9/8/2004 Re: Revision to the EAL Basis Document and EAL Tables File: A-221b NEP-2004-0030 The Duane Arnold Energy Center (DAEC) proposes modifying areas within the Emergency Action Level (EAL) Bases Document (EBD) and EAL Tables (EPIP Forms).

The revisions are in order to incorporate industry 'lessons learned' given in revision 4 of NEI 99-01 and as endorsed by the NRC via RIS 2003-18. These changes are not philosophically different than those already In place. These changes merely serve to ensure DAEC Staff have clear, concise, and regulatory endorsed guidance when determining EAL's at DAEC. A Training Packet has been drafted detailing the existing EAL and the new proposed EAL.

DAEC respectfully asks for State & County approval of these EAL's prior to submitting these EAL's to the NRC as DAEC desires to ensure that the key stakeholders in the Iowa Emergency Preparedness Program have an understanding of these EAL's and approve their use at DAEC.

If you have any questions or desire additional information regarding this matter, please contact Paul Sullivan, Manager, Emergency Planning at (319) 851-7191 or Don A. Johnson at (319) 851-7872.

Sincerely,1 (

Don A. Johnson (319)851-7872

Commuted to NudearExceJvt To: Paul R. Sullivan From: Ned Wright, Unn County Emergency Management Agency Re: EAL Changes as Referenced in NEP 2004-0030 File: A-221a I concur with the changes to the EAL's in order to incorporate NEI 99-01 Revision 4 EAL's. The Emergency Preparedness staff at DAEC has effectively answered any question I had regarding these changes.

(-

Oa k44-- ___ ___ ___ __

Name Agency It C&4' 0 Date

Covined Nudetarxc To: Paul R. Sullivan From: Scott Hansen, Benton County Emergency Management Agency Re: EAL Changes as Referenced in NEP 2004-0030 File: A-221a I concur with the changes to the EAL's in order to incorporate NEI 99-01 Revision 4 EAL's. The Emergency Preparedness staff at DAEC has effectively answered any question I had regarding these changes.

Name Agency Date

.N ConnIOd to NudearExce/l To: Paul R. Sullivan From: Don Flater, Iowa Department of Public Health Re: EAL Changes as Referenced in NEP 2004-0030 File: A-221a I concur with the changes to the EAL's in order to incorporate NEI 99-01 Revision 4 EAL's. The Emergency Prepatedness staff at DAEC has effectively answered any question I had regarding these changes.

4".1 6 ,UA,,"t v.,-/Z Name Agency Date

Coamsd to NudiaExee To: Paul R. Sullivan From: David L Miller, Iowa Homeland Security &Emergency Management Division Re: EAL Changes as Referenced in NEP 2004-0030 File: A-221a I concur with the changes to the EAL's in order to incorporate NEI 99-01 Revision 4 EAIs. The Emnergency Preparedness staff at DAEC has effectively answered any question I had regarding these changes.

Agency Date

ENCLOSURE 4 DETAILED SUPPORTING INFORMATION 405 pages follow

ATTACHMENT 1 RED-LINE TECHNICAL BASIS DOCUMENT 183 pages follow

EAL BASES DOCUMENT Rev. 20 INDEX Page 1 of 1 PROCEDURE TITLE REV# REV. DATE EBD-REF EAL Supporting Reference Information 0 SUBMITTED EBD-REG Regulatory Context 0 SUBMITTED EBD-C Cold Shutdown/Refueling 0 SUBMITTED EBD-E ISFSI Abnormal Events Category 1 SUBMITTED EBD-F Fission Product Barrier Degradation Category 6 SUBMITTED EBD-H Hazards & Other Conditions Affecting Plant Safety 9 SUBMITTED EBD-R Abnormal Rad Levels/Radiological Effluent Category 9 SUBMITTED EBD-S System Malfunction Category 7 SUBMITTED

EAL BASES DOCUMENT EBD-REF Rev. 0 EAL SUPPORTING REFERENCE INFORMATION Page 1lof 19l I Usage Level l Information Use l Effective Date:

TECHNICAL REVIEW Prepared by: _ Date:

Reviewed by: Date:

Independent Reviewer Reviewed by: _ Date:

Operations Reviewer PROCEDURE APPROVAL I am responsible for the technical content of this procedure and for obtaining the necessary approval from the State and County Emergency Management officials prior to implementation.

Approved by: Date:

Manager, Emergency Planning

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 2 of 19 Table of Contents Section Page EXECUTIVE

SUMMARY

................ 3 ACRONYMS .................. 4

1. PURPOSE .6

2. REFERENCES .6

3. DISCUSSION .6

4. TECHNICAL BASES INFORMATION .14

5. DEFINITIONS .16

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 3 of 19 EXECUTIVE

SUMMARY

DAEC implemented NEI EALs via NESP-007 in January 1998. NMC Management directed that all NMC Plants submit for NRC approval an EAL scheme based upon NEI 99-01 Revision 4.

Using NEI 99-01 Rev. 4, DAEC conducted an EAL implementation upgrade project that produced the EALs discussed herein. While the upgraded EALs are site-specific, an objective of the project was to ensure to the extent possible EAL conformity and consistency between the NMC plant sites. The vast majority of the EALs are functionally similar to what is already in place at DAEC. However, many EALs were revised to better align with the intent and wording of NEI 99-01 Revision 4, or to align with the NMC Fleet. It is important to note that DAEC does not deviate from the intent of NEI 99-01 Revision 4 guidance, however the following conditions should be noted:

• DAEC is a BWR and therefore cannot implement PWR specific guidance.

• DAEC does not have a perimeter rad monitoring system or an automatic dose assessment system and therefore cannot implement guidance for these systems.

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 4 of 19 ACRONYMS AC Alternating Current ATWS Anticipated Transient Without Scram BWR Boiling Water Reactor CCW Component Cooling Water CDE Committed Dose Equivalent CFR Code of Federal Regulations CMT Containment DC Direct Current DHR Decay Heat Removal DOT Department of Transportation EAL Emergency Action Level ECCS Emergency Core Cooling System ECL Emergency Classification Level EOF Emergency Operations Facility EOP Emergency Operating Procedure EPA Environmental Protection Agency EPG Emergency Procedure Guideline EPIP Emergency Plan Implementing Procedure EPRI Electric Power Research Institute ESF Engineered Safeguards Feature ESW Emergency Service Water GE General Emergency HPCI High Pressure Coolant Injection IC Initiating Condition IDLH Immediately Dangerous to Life and Health IPEEE Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI Independent Spent Fuel Storage Installation LCO Limiting Condition of Operation LER Licensee Event Report

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 5 of 19 LFL Lower Flammability Limit LOCA Loss of Coolant Accident MSIV Main Steam Isolation Valve mR milliRem Mw Megawatt NEI Nuclear Energy Institute NESP National Environmental Studies Project NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NUMARC Nuclear Management and Resources Council OBE Operating Basis Earthquake ODCM Offsite Dose Calculation Manual PRA/PSA Probabilistic Risk Assessment / Probabilistic Safety Assessment PSIG Pounds per Square Inch Gauge R Rem RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System RPS Reactor Protection System RPV Reactor Pressure Vessel SAE Site Area Emergency SBGT Stand-By Gas Treatment SPDS Safety Parameter Display System SRO Senior Reactor Operator SSE Safe Shutdown Earthquake TEDE Total Effective Dose Equivalent TOF Top of Active Fuel TSC Technical Support Center UE Notification Of Unusual Event USAR Updated Safety Analysis Report

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 6 of 19

1. PURPOSE This document provides the detailed set of Emergency Action Levels (EALs) applicable to the Duane Arnold Energy Center (DAEC) and the associated Technical Bases using the EAL development methodology found in NEI 99-01 Revision 4 [Ref. 2.1]. Personnel responsible for the classification of emergencies may use this document as a technical reference and an aid in EAL interpretation.

The primary tool for determining the emergency classification level is the Emergency Action Level Matrix. The user of the Emergency Action Level Matrix may (but is not required) to consult the EAL Technical Basis Document in order to obtain additional information concerning the EALs under classification consideration.

2. REFERENCES 2.1 NEI 99-01 Revision 4, Methodology for Development of Emergency Action Levels 2.2 Emergency Action Level Matrix 2.3 DAEC Technical Specifications 2.4 DAEC Emergency Plan & implementing Procedures

3. DISCUSSION 3.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the DAEC Emergency Plan.

In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG 0654 EAL guidance.

NEI 99-01 (NUMARC/NESP-007) Revision 4 represents the most recently NRC endorsed methodology per RG 1.101 Rev 4, "Emergency Planning and Preparedness for Nuclear Power Reactors." Enhancements over earlier revisions included:

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 7 of 19

• Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.

• Addressing initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations.

• Simplifying the fission product barrier EAL threshold for a Site Area Emergency.

Using NEI 99-01 Rev. 4, DAEC conducted an EAL implementation upgrade project that produced the EALs discussed herein. While the upgraded EALs are site-specific, an objective of the project was to ensure to the extent possible EAL conformity and consistency between the NMC plant sites.

3.2 Key Definitions in EAL Methodology The following definitions apply to the generic EAL methodology:

EMERGENCY CLASS: One of a minimum set of names or titles, established by the Nuclear Regulatory Commission (NRC), for grouping of normal nuclear power plant conditions according to (1) their relative radiological seriousness, and (2) the time sensitive onsite and off site radiological emergency preparedness actions necessary to respond to such conditions. The existing radiological emergency classes, in ascending order of seriousness, are called:

• Notification of Unusual Event (NOUE or UE)

• Alert

• Site Area Emergency (SAE)

• General Emergency (GE)

Section 3.3 provides further discussion of the emergency classes.

INITIATING CONDITION (IC): One of a predetermined subset of nuclear power plant conditions when either the potential exists for a radiological emergency, or such an emergency has occurred.

• An IC is an emergency condition which sets it apart from the broad class of conditions that may or may not have the potential to escalate into a radiological emergency.

• It can be a continuous, measurable function that is outside technical specifications, such as elevated RCS temperature or falling reactor coolant level (a symptom).

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 8 of 19

• It also encompasses occurrences such as FIRE (an event) or reactor coolant pipe failure (an event or a barrier breach).

EMERGENCY ACTION LEVEL (EAL): A pre determined, site-specific, observable threshold for a plant Initiating Condition that places the plant in a given emergency class. An EAL can be: an instrument reading; an equipment status indicator; a measurable parameter (onsite 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.

• There are times when an EAL will be a threshold point on a measurable continuous function, such as a primary system coolant leak that has exceeded technical specifications.

• At other times, the EAL and the IC will coincide, both identified by a discrete event that places the plant in a particular emergency class.

3.3 Recognition Categories ICs and EALs are grouped in one of several categories. This classification scheme incorporates symptom-based, event-based, and barrier-based ICs and EALs.

• R - Abnormal Rad Levels/Radiological Effluent

• C - Cold Shutdown / Refueling System Malfunction

• E - Independent Spent Fuel Storage Installation (ISFSI)

• F - Fission Product Barrier Degradation

• H - Hazards

• S - System Malfunction Some recognition categories are further divided into one or more subcategories depending on the types and number of plant conditions that dictate emergency classifications. An EAL may or may not exist for each subcategory at all four classification levels. Similarly, more than one EAL may exist for a subcategory in a given emergency classification when appropriate (i.e., no EAL at the General Emergency level but three EALs at the Unusual Event level).

3.4 Emergency Class Descriptions There are three considerations related to the emergency classes. These are:

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 9 of 19

• The potential impact on radiological safety, either as now known or as can be reasonably projected.

• How far the plant is beyond its predefined design, safety and operating envelopes.

• Whether or not conditions that threaten health are expected to be confined to within the site boundary.

The ICs deal explicitly with radiological safety affect by escalating from levels corresponding to releases within regulatory limits to releases beyond EPA Protective Action Guideline (PAG) plume exposure levels.

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

• Potential degradation of the level of safety of the plant is indicated primarily by exceeding plant technical specification Limiting Condition of Operation (LCO) allowable action statement time for achieving required mode change.

• Precursors of more serious events may be included because precursors represent a potential degradation in the level of safety of the plant.

• Minor releases of radioactive materials are included. In this emergency class, however, releases do not require monitoring or offsite response (e.g., dose consequences of less than 10 millirem).

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

SITE AREA EMERGENCY: 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 Guideline exposure levels beyond the site boundary.

• The discriminator (threshold) between Site Area Emergency and General Emergency is whether or not the EPA PAG plume exposure levels are expected to be exceeded outside the site boundary.

• This threshold, in addition to dynamic dose assessment considerations discussed in the EAL guidelines, clearly addresses NRC and offsite emergency response agency concerns as to timely declaration of a General Emergency.

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 10 of 19 GENERAL EMERGENCY: 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 be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

• The bottom line for the General Emergency is whether evacuation or sheltering of the general public is indicated based on EPA PAGs and, therefore, should be interpreted to include radionuclide release regardless of cause.

• To better assure timely notification, EALs in this category are primarily expressed in terms of plant function status, with secondary reliance on dose projection. In terms of fission product barriers, loss of two barriers with loss or potential loss of the third barrier constitutes a General Emergency.

3.5 Operating Mode Applicability The following definitions of operating modes are used in this document:

1 Power Operation Reactor mode switch in 'Run."

2 Startup Reactor mode switch in "Start to Hot Stby" or "Refuel" (with all reactor vessel head closure bolts fully tensioned).

3 Hot Shutdown Reactor mode switch in "Shutdown" with reactor coolant temperature GREATER THAN 212 0F.

4 Cold Shutdown Reactor mode switch in 'Shutdown" with reactor coolant temperature LESS THAN OR EQUAL TO 212 0F.

5 Refueling Reactor Mode Switch in "Shutdown" or "Refuel" with one or more reactor vessel head closure bolts less than fully tensioned.

In addition to these operating modes, NEI 99-01 [Ref. 1] defines the following additional mode:

D Defueled

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 11 of 19 All reactor fuel removed from Reactor Vessel (full core off load during refueling or extended outage)

The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action is initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.

Recognition categories are associated with the operating modes listed in the following matrix:

Recognition Category Mode R C F H S 1 - Power X X X X Operation 2 - Startup X X X X 3 - Hot X X X X Shutdown 4 - Cold X X X Shutdown 5 - Refueling X X . X D - Defueled X X X 3.6 Fission Product Barriers Many of the EALs derived from the NEI methodology are fission product barrier based. That is, the conditions that define the EALs are based upon loss of or potential loss to one or more of the three fission product barriers. "Loss" and "potential loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials and 'potential loss" means imminent loss of the barrier.

The primary fission product barriers are:

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 12 of 19

• Fuel Cladding (FC): Zirconium tubes which house the ceramic uranium oxide pellets along with the end plugs which are welded into each end of the fuel rods comprise the FC barrier.

• Reactor Coolant System (RCS): The reactor vessel shell, vessel head, vessel nozzles and penetrations and all primary systems directly connected to the reactor vessel up to the first containment isolation valve comprise the RCS barrier.

• Containment (CMT): The primary containment structure and all isolation valves required to maintain primary containment integrity under accident conditions comprise the CMT barrier.

3.7 Emergency Classification Based on Fission Product Barrier Degradation The following criteria are the bases for event classification related to fission product barrier loss or challenge:

• Notification of Unusual Event:

Any loss or any potential loss of Containment

• Alert:

Any loss or any potential loss of either Fuel Cladding or RCS

• Site Area Emergency:

Loss or potential loss of any two barriers

• General Emergency:

Loss of any two barriers and loss or potential loss of third barrier 3.8 EAL Relationship to EOPs Where possible, the EALs have been made consistent with and utilize the conditions defined in the DAEC Emergency Operating Procedures (EOPs). While the symptoms that drive operator actions specified in the EOPs are not indicative of all possible conditions which warrant emergency classification, they do define the symptoms, independent of initiating events, for which reactor plant safety and/or fission product barrier integrity are threatened. Where these symptoms are clearly representative of one of the NEI Initiating Conditions, they have been utilized as an EAL. This permits rapid classification of emergency situations based on plant conditions without the need for additional evaluation or event diagnosis. Although some of the EALs presented here are based on conditions defined in the EOPs, classification of

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 13 of 19 emergencies using these EALs is not dependent upon EOP entry or execution. The EALs can be utilized independently or in conjunction with the EOPs.

3.9 Symptom Based vs. Event Based Approach To the extent possible, the EALs are symptom based. That is, the action level is defined by values of key plant operating parameters that identify emergency or potential emergency conditions. This approach is appropriate because it allows the full scope of variations in the types of events to be classified as emergencies. But, a purely symptom based approach is not sufficient to address all events for which emergency classification is appropriate. Particular events to which no predetermined symptoms can be ascribed have also been utilized as EALs since they may be indicative of potentially more serious conditions not yet fully realized.

Category R - Abnormal Rad Levels/Radiological Effluent and Category F - Fission Product Barrier Degradation are primarily symptom-based. The symptoms are indicative of actual or potential degradation of either fission product barriers or personnel safety.

Other categories tend to be event-based. For example, System Malfunctions are abnormal and emergency events associated with vital plant system failures, while Hazards are those non-plant system related events that have affected or may affect plant safety.

3.10 Treatment of Emergency Class Upgrading The emergency class is based on the highest EAL reached. For example, two Alerts remain in the Alert category. Or, an Alert and a Site Area Emergency is a Site Area Emergency.

3.11 Treatment of Emergency Class Downgrading Another important aspect of usable EAL guidance is the consideration of what to do when the risk posed by an emergency is clearly decreasing. A combination approach involving recovery from General Emergencies and Site Area Emergencies and termination from Unusual Events and Alerts is used. Downgrading to lower emergency classes adds notifications but may has merit for higher levels of emergency classification.

3.12 Classifying Transient Events For some events, the condition may be corrected before a declaration has been made. For example, an emergency classification is warranted when automatic and manual actions taken within the control room do not result in a required reactor scram. However, it is likely that actions taken outside of the control room will be successful, probably before the Emergency Director classifies the event. The key consideration in this situation is to determine whether or not further plant damage occurred while the corrective actions were being taken. In some situations, this can be readily determined, in other situations, further analyses (e.g., coolant sampling, may be necessary).

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 14 of 19 In general, observe the following guidance: Classify the event as indicated and terminate the emergency once assessment shows that there were no consequences from the event and other termination criteria are met. For example, a momentary event, such as an ATWS or an earthquake, requires declaration even though the condition may have been resolved by the time the declaration is made.

• An ATWS represents a failure of a front line safety system (RPS) designed to protect the health and safety of the public.

• The affect of an earthquake on plant equipment and structures may not be readily apparent until investigations are conducted.

There may be cases in which a plant condition that exceeded an EAL threshold was not recognized at the time of occurrence, but is identified well after the condition has occurred (e.g.,

as a result of routine log or record review) and the condition no longer exists. In these cases, an emergency should not be declared. Reporting requirements of 10 CFR 50.72 are applicable and the guidance of NUREG-1 022, Rev. 1, Section 3 should be applied.

3.13 Imminent EAL Thresholds Although the majority of the EALs provide very specific thresholds, the Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded. While this is particularly prudent at the higher emergency classes (as the early classification may provide for more effective implementation of protective measures), it is nonetheless applicable to all emergency classes. Explicit EALs, specifying use of Emergency Director judgment, are given in the Hazards, ISFSI and Fission Product Barrier Degradation categories.

4. TECHNICAL BASES INFORMATION 4.1 Recognition Category Organization The technical bases of the EALs are provided under Recognition Categories R, C, E, F, H and S of this document. A table summarizing the Initiating Conditions introduces each category. The tables provide an overview of how the ICs are related under each emergency class. ICs within each category are listed according to classification (as applicable) in the following order:

Notification of Unusual Event, Alert, Site Area Emergency, and General Emergency.

The basis information for the fission barrier table indicators is organized similarly to the other basis information described above. For each barrier - fuel clad, RCS, and primary containment - basis

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 15 of 19 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 R, C, E, H, and S EALs described above. Indicators include Radiation/Core Damage, RPV Level, Leakage, Primary Containment Atmosphere, and Emergency Director Judgment.

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

4.2 Initiating Condition Structure ICs in Recognition Categories R, C, E, H and S are structured in the following manner:

• Recognition Category Title

• IC Identifier:

o First character identifies the category by letter (R, C, E, H and S) o Second character identifies the emergency classification level (U for Notification of Unusual Event, A for Alert, S for Site Area Emergency, and G for General Emergency) o Third character is the numerical sequence as given in Revision 4 of NEI 99-01

[Ref. 1] (e.g., SA2). Due to document revisions, certain NEI ICs have been deleted, leaving gaps in the numerical sequence.

• Emergency Class: Notification of Unusual Event, Alert, Site Area Emergency, or General Emergency

• IC Description

• Operating Mode Applicability: Refers to the operating mode during which the IC/EAL is applicable

• Emergency Action Level(s): EALs are the conditions applicable to the criteria of the IC and are used to determine the need to classify an event/condition. If more than one EAL is applicable to an IC, emergency classification is required when any EAL within the IC reaches the EAL threshold. To clarify this intent, ICs with multiple EALs include a parenthetical phrase in the EAL title line, indicating that each constitutes an emergency classification. For example, the phrase "(RA1.1 or RA1.2)" indicates that either EAL is a Notification of Unusual Event.

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 16 of 19

• Basis: Provides information that explains the IC and EAL(s). Plant source document references are provided as needed to substantiate site-specific information included in the EALs and bases.

4.3 EAL Identification The EAL identifier is the IC identifier followed by a period and sequence number (e.g., RU1.1, RU1.2, etc.). If only one EAL is assigned to an IC, the EAL is given the number one.

The primary purpose of the EAL identifier is to uniquely distinguish each classifiable condition.

Secondary purposes are to assist location of an EAL within the EAL classification scheme and to announce the emergency classification level.

5. DEFINITIONS In the ICs and EALs, selected words are in uppercase print. These words are defined terms.

Definitions are provided below.

AFFECTING SAFE SHUTDOWN: Event in progress has adversely affected functions that are necessary to bring the plant to and maintain it in the applicable HOT or COLD SHUTDOWN condition. Plant condition applicability is determined by Technical Specification LCOs in effect.

Example 1: Event causes damage that results in entry into an LCO that requires the plant to be placed in HOT SHUTDOWN. HOT SHUTDOWN is achievable, but COLD SHUTDOWN is not. This event is not"AFFECTING SAFE SHUTDOWN."

Example 2: Event causes damage that results in entry into an LCO that requires the plant to be placed in COLD SHUTDOWN. HOT SHUTDOWN is achievable, but COLD SHUTDOWN is not. This event is 'AFFECTING SAFE SHUTDOWN."

BOMB: refers to an explosive device suspected of having sufficient force to damage plant systems or structures.

CIVIL DISTURBANCE: is a group of or more persons violently protesting station operations or activities at the site.

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 2120F, 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.

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 17 of 19 CONFINEMENT BOUNDARY: is the barrier(s) between areas containing radioactive substances and the environment.

CONTAINMENT CLOSURE: (BWR) is considered to be Secondary Containment as required by Technical Specifications.

EMERGENCY DIRECTOR: Individual responsible for overall direction and control of the Emergency Response Organization (ERO). In the Control Room this would be the Operations Shift Manager (OSM) . In the Technical Support Center (TSC) this would be the Emergency Coordinator (EC). In the Emergency Operations Facility (EOF) this would be the Emergency Response and Recovery Director (ER&RD).

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

EXTORTION: is an attempt to cause an action at the station by threat of force.

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

HOSTAGE: is a person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE FORCE: one or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

HOT SHUTDOWN: As defined in Technical Specifications Table 1.1-1, the reactor mode switch is in the shutdown position and the reactor coolant temperature is greater than 21 21F and all reactor head closure bolts fully tensioned.

IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (IDLH): A condition that either poses an immediate threat to life and health or an immediate threat of severe exposure to contaminants which are likely to have adverse delayed effects on health.

INTRUSION / INTRUDER: is a person(s) present in a specified area without authorization.

Discovery of a BOMB in a specified area is indication of INTRUSION into that area by a HOSTILE FORCE.

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): The on site facility where the loaded Dry Shielded Canisters (DSCs) will be stored in Horizontal Storage Modules (HSMs).

The installation is intended for interim storage until the spent fuel is removed from the plant site.

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 18 of 19 LOWER FLAMMABILITY LIMIT (LFL): The minimum concentration of a combustible substance that is capable of propagating a flame through a homogenous mixture of the combustible and a gaseous oxidizer.

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

OPERATING MODES: The applicable operating modes for each IC/EAL is listed based on NEI 99-01 mode descriptions. DAEC EALs use the operating modes defined in Technical Specifications Table 1.1-1. These are:

REACTOR MODE AVERAGE REACTOR COOLANT MODE TITLE SWITCH POSITION TEMPERATURE (OF) 1 Power Operation Run N/A 2 Startup Refuel (a) or Startup/Hot N/A 2 HotShutdon ( Standby 3 Hot Shutdown (a) Shutdown >212 4 Cold Shutdown (a) Shutdown 52 12 5 Refueling (b) Shutdown or Refuel N/A (a) All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bolts less than fully tensioned.

PLANT PROTECTED AREA: is an area which normally encompasses all controlled areas within the security protected area fence. The ISFSI is located in its own protected area separate from the Plant.

SABOTAGE: is deliberate damage, mis-alignment, or mis-operation of plant equipment with the intent to render the equipment inoperable. Equipment found tampered with or damaged due to malicious mischief may NOT meet the definition of SABOTAGE until this determination is made by security supervision.

SAFE SHUTDOWN AREA: Any area containing equipment, systems, or components that are necessary to bring the plant to, and maintain it in a shutdown condition. In the EAL Bases Documents and Tables, Safe Shutdown Area is synonymous with Vital Area.

SIGNIFICANT TRANSIENT: is an UNPLANNED event involving one or more of the following:

(1) automatic turbine runback >25% thermal reactor power, (2) electrical load rejection >25% full electrical load, (3) Reactor Trip, (4) Safety Injection Activation, or (5) thermal power oscillations

>1 0%

EAL BASES DOCUMENT Rev. 0 SUPPORTING EAL BASES INFORMATION Page 19 of 19 STRIKE ACTION: is a work stoppage within the PROTECTED AREA by a body of workers to enforce compliance with demands. The STRIKE ACTION must threaten to interrupt NORMAL PLANT OPERATIONs.

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

VALID: an indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

VISIBLE DAMAGE: is damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g., paint chipping, scratches) should not be included.

VITAL AREA: is any area, normally within the PLANT PROTECTED AREA, which contains equipment, systems, components, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation. In the EAL Bases Documents and Tables, Safe Shutdown Area is synonymous with Vital Area. NOTE:

Vital areas for EAL purposes are not the same as vital areas for Security purposes.

EAL BASES DOCUMENT Rev. 0 REGULATORY CONTEXT Page 1 of 3 Effective Date:

Title 10, Code of Federal Regulations, Part 50 provides the regulations that govern emergency preparedness at nuclear power plants. Nuclear power reactor licensees are required to have NRC-approved "emergency response plans" for dealing with "radiological emergencies." The requirements call for both onsite and offsite emergency response plans, with the offsite plans being those approved by FEMA and used by the State and local authorities. This document deals with the utilities' approved onsite plans and procedures for response to radiological emergencies at nuclear power plants, and the links they provide to the offsite plans.

Section 50.47 of Title 10 of the Code of Federal Regulations (10 CFR 50.47),

entitled "Emergency Plans," states the requirement for such plans. Part (a)(1) of this regulation states that "no operating license will be issued unless a finding is made by NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency."

The major portion of 10 CFR 50.47 lists "standards" that emergency response plans must meet. The standards constitute a detailed list of items to be addressed in the plans. Of particular importance to this project is the fourth standard, which addresses "emergency classification" and "action levels." These terms, however, are not defined in the regulation.

10 CFR 50.54, "Conditions of licenses," emphasizes that power reactor licensees must "follow, and maintain in effect, emergency plans which meet the standards in Part 50.47(b) and the requirements in Appendix E to this part." The remainder of this part deals primarily with required implementation dates.

10 CFR 50.54(q) allows licensees to make changes to emergency plans without prior Commission approval only if: (a) the changes do not decrease the effectiveness of the plans and (b) the plans, as changed, continue to meet 10 CFR 50.47(b) standards and 10 CFR 50 Appendix E requirements. The licensee must keep a record of any such changes. Proposed changes that decrease the effectiveness of the approved emergency plans may not be implemented without application to and approval by the Commission.

10 CFR 50.72 deals with "Immediate notification requirements for operating nuclear power reactors." The "immediate" notification section actually includes three types of reports: (1) immediately after notification of State or local agencies

EAL BASES DOCUMENT Rev. 0 REGULATORY CONTEXT Page 2 of 3 (for emergency classification events); (2) one-hour reports; and, (3) four-hour reports.

Although 10 CFR 50.72 contains significant detail, it does not define either "Emergency Class" or "Emergency Action Level." But one-hour and four-hour reports are listed as "non-emergency events," namely, those which are "not reported as a declaration of an Emergency Class." Certain 10CFR50.72 events can also meet the Notification of Unusual Event emergency classification if they are precursors of more serious events. These situations also warrant anticipatory notification of state and local officials. (See Section 3.7, "Emergency Class Descriptions".)

By footnote, the reader is directed from 10 CFR 50.72 to 10 CFR 50 Appendix E, for information concerning "Emergency Classes."

10 CFR 50.73 describes the "Licensee event report system," which requires submittal of follow-up written reports within thirty days of required notification of NRC.

10 CFR 50 Appendix E, Section B, "Assessment Actions," mandates that emergency plans must contain "emergency action levels." EALs are to be described for: (1) determining the need for notification and participation of various agencies, and (2) determining when and what type of protective measures should be considered. Appendix E continues by stating that the EALs are to be based on: (1) in-plant conditions; (2) in-plant instrumentation; (3) onsite monitoring; and (4) offsite monitoring.

10 CFR 50 Appendix E, Section C, "Activation of Emergency Organization," also addresses "emergency classes" and "emergency action levels." This section states that EALs are to be based on: (1) onsite radiation monitoring information; (2) offsite radiation monitoring information; and, (3) readings from a number of plant sensors that indicate a potential emergency, such as containment pressure and the response of the Emergency Core Cooling System. This section also states that "emergency classes" shall include: (1) Notification of Unusual Events (NOUEs), (2) Alert, (3) Site Area Emergency, and (4) General Emergency.

These regulations are supplemented by various regulatory guidance documents.

A significant document that has dealt specifically with EALs is NUREG-0654/FEMA-REP-1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," October 1980.

Recognition Category E (Events Related to ISFSI) is applicable to licensees using their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR

EAL BASES DOCUMENT Rev. 0 REGULATORY CONTEXT Page 3 of 3 72.32. Recognition Category E is not applicable to stand alone ISFSls, Monitored Retrievable Storage Facilities (MRS), or ISFSIs that may process and/or repackage spent fuel. The emergency classifications for Recognition Category E are those provided by NUREG 0654/FEMA Rep.1 in accordance with 10 CFR 50.47. The classification of an ISFSI event under provisions of a 10 CFR 50.47 emergency plan should be consistent with the definitions of the emergency classes as used by that plan. A site-specific analysis would make this determination, but in most cases it is expected that classification of an NOUE would be appropriate. It is expected that the initiating conditions germane to a 10 CFR 72.32 emergency plan (described in NUREG-1 567) are subsumed within 10 CFR 50.47 emergency plan's classification scheme.

Usage Level F Information Use l Effective Date:

TECHNICAL REVIEW Prepared by: _ Date:

Reviewed by: Date:

Independent Reviewer Reviewed by: _ Date:

Operations Reviewer PROCEDURE APPROVAL I am responsible for the technical content of this procedure and for obtaining the necessary approval from the State and County Emergency Management officials prior to implementation.

Documentation of State and County Emergency Management approval is via NEP-Approved by: _ Date:

Manager, Emergency Planning

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION I Page 2 of 33 Table of Contents CUl RCS Leakage ................................................................. 3 CU2 Unplanned Loss of RCS Inventory with Irradiated Fuel in the RPV ...................... 4 CU3 Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes ........ 6 CU4 Unplanned Loss of Decay Heat Removal Capability with Irradiated Fuel in the RPV ................................................................. 7 CU5 Fuel Clad Degradation ................................................................. 9 CU6 Unplanned Loss of All Onsite or Offsite Communications Capabilities .. 10 CU7 Unplanned Loss of Required DC Power For Greater Than 15 Minutes .. 12 CU8 Inadvertent Criticality ................................................................ 14 CA1 Loss of RCS Inventory................................................................. 15 CA2 Loss of RPV Inventory with Irradiated Fuel in the RPV ...................................... 17 CA3 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses 19 CA4 Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV .....21 CS1 Loss of RPV Inventory Affecting Core Decay Heat Removal Capability ............. 24 CS2 Loss of RPV Inventory Affecting Core Decay Heat Removal Capability with Irradiated Fuel in the RPV ................ ................................................ 27 CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged with Irradiated Fuel in the RPV ........................................................ 30

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 3 of 33 CUI RCS Leakage EVENT TYPE: Coolant Leakage OPERATING MODE APPLICABILITY: Cold S/D EAL THRESHOLD VALUE:

CU1.1 Unidentified or pressure boundary leakage GREATER THAN 10 GPM.

OR CU1.2 Identified leakage GREATER THAN 25 GPM.

DAEC EAL INFORMATION:

This IC is included as a NOUE because it is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified and pressure boundary leakage was selected as it is sufficiently large to be observable via normally installed instrumentation or reduced inventory instrumentation such as level hose indication. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances). The EAL for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage. Prolonged loss of RCS Inventory may result in escalation to the Alert level via either IC CA1 or CA4.

The difference between CU1 and CU2 deals with the RCS conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown, the RCS will normally be intact and RCS inventory and level monitoring means such as makeup volume control tank levels are normally available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels CuM

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION I Page 4 of 33 CU2 Unplanned Loss of RCS Inventory with Irradiated Fuel in the RPV EVENT TYPE: RCS Level OPERATING MODE APPLICABILITY: Refueling EAL THRESHOLD VALUE:

CU2.1 Unplanned RCS level decrease LESS THAN the RPV flange for 15 minutes or longer.

OR CU2.2 RPV Level cannot be monitored AND Loss of RPV inventory as indicated by unexplained Drywell/Reactor Building Equipment or Floor Drain sump, or Torus, level increase.

DAEC EAL INFORMATION:

This IC is included as a NOUE because it may be a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that decrease RCS water level below the RPV flange are carefully planned and procedurally controlled. An UNPLANNED event that results in water level decreasing below the RPV flange warrants declaration of a NOUE due to the reduced RCS inventory that is available to keep the core covered. The allowance of 15 minutes was chosen because it is reasonable to assume that level can be restored within this time frame using one or more of the redundant means of refill that should be available. If level cannot be restored in this time frame then it may indicate a more serious condition exists. Continued loss of RCS Inventory will result in escalation to the Alert level via either IC CA2 or CA4.

The difference between CU1 and CU2 deals with the RCS conditions that exist between cold shutdown and refueling modes. In cold shutdown the RCS will normally be intact and standard RCS inventory and level monitoring means are available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.

CU2

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 5 of 33 In the refueling mode, normal means of core temperature indication and RCS level indication may not be available. Redundant means of RPV level indication will normally be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and Torus level changes. The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building. A rise in Torus water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Torus level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

Escalation to Alert would be via either CA2 or RCS heatup via CA4.

EAL 1 involves a decrease in RCS level below the top of the RPV flange that continues for 15 minutes due to an UNPLANNED event. This EAL is not applicable to decreases in flooded reactor cavity level (covered by RU2 ) until such time as the level decreases to the level of the vessel flange. For BWRs, if RPV level continues to decrease and reaches the Low-Low ECCS Actuation Setpoint then escalation to CA2 would be appropriate.

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels CU2

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 6 of 33 CU3 Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

CU3.1 Loss of all offsite power to Emergency Busses 1A3 and 1A4 is expected to last for greater than 15 minutes AND At least one Emergency Bus, 1A3 or 1A4, is powered by it's Standby Diesel Generator.

DAEC EAL INFORMATION:

Prolonged loss of AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete Loss of AC Power (e.g., Station Blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

REFERENCES:

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

2. UFSAR Section 8.2, Offsite Power System

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels CU3

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 7 of 33 CU4 Unplanned Loss of Decay Heat Removal Capability with Irradiated Fuel in the RPV EVENT TYPE: RCS Temperature OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

CU4.1 An unplanned event results in RCS temperature GREATER THAN 212 IF OR CU4.2 Loss of all RCS temperature and RPV level indication for GREATER THAN 15 minutes.

DAEC EAL INFORMATION:

This IC is included as a NOUE because it may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. In cold shutdown the ability to remove decay heat relies primarily on forced cooling flow. Operation of the systems that provide this forced cooling may be jeopardized due to the unlikely loss of electrical power or RCS inventory. Since the RCS usually remains intact in the cold shutdown mode a large inventory of water is available to keep the core covered. Incold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown conditions may be attained within hours of operating at power. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). In addition, the operators should be able to monitor RCS temperature and RPV level so that escalation to the alert level via CA4 or CA1 will occur if required.

CU4

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 8 of 33l During refueling the level in the RPV will normally be maintained above the RPV flange.

Refueling evolutions that decrease water level below the RPV flange are carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid increases in RCS/RPV temperatures depending on the time since shutdown. Escalation to the Alert level via CA4 is provided should an UNPLANNED event result in RCS temperature exceeding the Technical Specification cold shutdown temperature limit with CONTAINMENT CLOSURE not established.

Unlike the cold shutdown mode, normal means of core temperature indication and RCS level indication may not be available in the refueling mode. Redundant means of RPV level indication are therefore procedurally installed to assure that the ability to monitor level will not be interrupted. However, if all level and temperature indication were to be lost in either the cold shutdown or refueling modes, EAL 2 would result in declaration of a NOUE if either temperature or level indication cannot be restored within 15 minutes from the loss of both means of indication. Escalation to Alert would be via CA2 based on an inventory loss or CA4 based on exceeding its temperature criteria.

The Emergency Director must remain attentive to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded.

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels CU4

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION I Page 9 of 33 CU5 Fuel Clad Degradation EVENT TYPE: Fuel Clad Degradation OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

CU5.1 Reactor Coolant sample activity indicating fuel clad degradation GREATER THAN 2.0 ItCi/gm dose equivalent 1-131.

DAEC EAL INFORMATION:

This IC is included as a NOUE because it is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems.

This EAL addresses coolant samples exceeding coolant technical specifications for iodine spike.

The calculated radiation monitor setpoint indicative of fuel clad degradation greater than Technical Specification allowable limits is greater than the radiation monitor setpoints already in place via EAL RA2, which is applicable in all operating modes. Therefore, EAL RA2 will bound the expected NEI IC for EAL CU5 requiring a radiation monitor value indicative of fuel clad degradation when in the Cold S/D or Refueling operating modes.

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels

2. Engineering Calculation No. 04-004-A, Rev.1, 'Radiation Exposure Rates at Spent Fuel Pool ARM 9178 Following Gap Release with the Cavity Flooded and DEI = 2 micro-Cilgm".

CU5

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION I Page 10 of 33 CU6 Unplanned Loss of All Onsite or Offsite Communications Capabilities EVENT TYPE: Communication OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

CU6.1 Loss of ALL of the following onsite communication capabilities affecting the ability to perform routine operation:

• Plant Operations Radio System

• In-Plant Telephones

• Plant Paging System OR CU6.2 Loss of ALL of the following offsite communications capability:

• All telephone lines (commercial)

• Microwave Phone System

• FTS Phone System DAEC EAL INFORMATION:

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

The availability of one method of ordinary offsite communications is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (e.g., relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being utilized to make communications possible.

CU6

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION I Page 11 of 33

REFERENCES:

1. Emergency Plan, Section F, Emergency Communications

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels CU6

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 12 of 33 CU7 Unplanned Loss of Required DC Power For Greater Than 15 Minutes EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Cold SID, Refueling EAL THRESHOLD VALUE:

CU7.1 Unplanned Loss of Vital DC power to required DC busses based on bus voltage LESS THAN 105 VDC indicated.

AND Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.

DAEC EAL INFORMATION:

The purpose of this IC and its associated EALs is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss.

UNPLANNED is included in this IC and EAL to preclude the declaration of an emergency as a result of planned maintenance activities. Routinely plants will perform maintenance on a Train related basis during shutdown periods. It is intended that the loss of the operating (operable) train is to be considered. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will be per CA4 "Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV."

Bus voltage should be based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value should incorporate a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is usually near the minimum voltage selected when battery sizing is performed.

Typically the value for the entire battery set is approximately 105 VDC.

CU7

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Re 1 Page 13of 33l

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels CU7

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 14 of 33 CUB Inadvertent Criticality EVENT TYPE: Inadvertent Criticality OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

CU8.1 An unplanned extended positive period observed on nuclear instrumentation.

DAEC EAL INFORMATION:

This IC addresses criticality events that occur in Cold Shutdown or Refueling modes (NUREG 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States) such as fuel misloading events and inadvertent dilution events. This IC indicates a potential degradation of the level of safety of the plant, warranting a NOUE classification. This IC excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated) which are addressed in the companion IC SU8.

This condition can be identified using period monitors. The terms "extended" is used in order to allow exclusion of expected short term positive periods from planned fuel bundle or control rod movements during core alteration for BWRs. These short term positive periods are the result of the increase in neutron population due to subcritical multiplication.

Escalation would be by Emergency Director Judgment.

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels CU8

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 15 of 33 CAI Loss of RCS Inventory EVENT TYPE: RCS Level OPERATING MODE APPLICABILITY: Cold S/D EAL THRESHOLD VALUE:

CA1.1 Loss of RCS inventory as indicated by RPV level LESS THAN 119.5 inches.

OR CA1.2 Loss of RCS inventory as indicated by unexplained Drywell/Reactor Building Equipment or Floor Drain sump, or Torus, level increase and RCS level cannot be monitored for GREATER THAN 15 minutes DAEC EAL INFORMATION:

These example EALs serve as precursors to a loss of ability to adequately cool the fuel.

The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level decrease and potential core uncovery. This condition will result in a minimum classification of Alert. The inability to restore and maintain level after reaching 119.5 inches would therefore be indicative of a failure of the RCS barrier.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode.

Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). The above forms the basis for needing both a cold shutdown specific IC (CA1) and a refueling specific IC (CA2).

CA1

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 16 of 33 In the cold shutdown mode, normal RCS level and RPV level instrumentation systems will normally be available. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and Torus level changes. The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building. A rise in Torus water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Torus level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. The 15-minute duration for the loss of level indication was chosen because it is half of the CS1 Site Area Emergency EAL duration. The 15-minute duration allows CA1 to be an effective precursor to CS1. Significant fuel damage is not expected to occur until the core has been uncovered for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per the analysis referenced in the CS1 basis.

Therefore this EAL meets the definition for an Alert emergency.

The difference between CA1 and CA2 deals with the RCS conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and standard RCS inventory and level monitoring means are available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.

If RPV level continues to decrease then escalation to Site Area will be via CS1.

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels CA1

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 17 of 33 CA2 Loss of RPV Inventory with Irradiated Fuel in the RPV EVENT TYPE: RCS Level OPERATING MODE APPLICABILITY: Refueling EAL THRESHOLD VALUE:

CA2.1 Loss of RPV inventory as indicated by RPV level LESS THAN 119.5 inches.

OR CA2.2 Loss of RPV inventory as indicated by unexplained Drywell/Reactor Building Equipment or Floor Drain sump, or Torus, level increase and RPV level cannot be monitored for GREATER THAN 15 minutes DAEC EAL INFORMATION:

These example EALs serve as precursors to a loss of heat removal. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level decrease and potential core uncovery. This condition will result in a minimum classification of Alert. The inability to restore and maintain level after reaching 119.5 inches would therefore be indicative of a failure of l the RCS barrier.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode.

Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). The above forms the basis for needing both a cold shutdown specific IC (CA1) and a refueling specific IC (CA2).

In the refueling mode, normal means of RPV level indication may not be available.

Redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted.

CA2

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 18 of 33l However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and Torus level changes. The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building. A rise in Torus water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Torus level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. The 15-minute duration for the loss of level indication was chosen because it is half of the CS2 Site Area Emergency EAL duration. The 15-minute duration allows CA2 to be an effective precursor to CS2.

Significant fuel damage is not expected to occur until the core has been uncovered for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per the analysis referenced in the CS2 basis. Therefore this EAL meets the definition for an Alert.

The difference between CA1 and CA2 deals with the RCS conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and standard RCS inventory and level monitoring means are available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.

If RPV level continues to decrease then escalation to Site Area will be via CS1.

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels CA2

EAL BASES DOCUMENT  : EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 19 of 33 CA3 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Cold S/D, Refueling, Defueled EAL THRESHOLD VALUE:

CA3.1 Loss of all offsite power to Emergency Busses 1A3 and 1A4 AND Failure of A Diesel Generator (1G-31) and B Diesel Generator (1G-21) to supply power to emergency busses 1A3 and 1A4.

AND Failure to restore power to at least one emergency bus, 1A3 or 1A4, within 15 minutes from the time of loss of both offsite and onsite AC power.

DAEC EAL INFORMATION:

Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink. When in cold shutdown or refueling mode the event can be classified as an Alert, because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL. Escalating to Site Area Emergency IC SS1, if appropriate, is by Abnormal Rad Levels / Radiological Effluent, or Emergency Director Judgment ICs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to essential busses. Even though an essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel CA3

EAL BASES DOCUMENT - EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION I Page 20 of 33 makeup capability) are not operable on the energized bus then the bus should not be considered operable.

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for'Development of Emergency Action Levels CA3

- EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 21 of 33 CA4 Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV EVENT TYPE: RCS Temperature OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

CA4.1 With Secondary Containment and RCS integrity not established, an unplanned event results in RCS temperature GREATER THAN 212 OF OR CA4.2 With Secondary Containment established and either RCS integrity not established or RCS inventory reduced, an unplanned event results in RCS temperature GREATER THAN 212 IF for GREATER THAN 20 minutes.

(Note: Ifan RCS heat removal system is in operation within this time frame and RCS temperatureis being reduced then this EAL is not applicable.)

OR CA4.3 An unplanned event results in RCS temperature GREATER THAN 212 IF for GREATER THAN 60 minutes or results in an RCS pressure increase of GREATER THAN 10 psig. (Note: If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced then this EAL Is not applicable.)

DAEC EAL INFORMATION:

EAL 1 addresses complete loss of functions required for core cooling during refueling and cold shutdown modes when neither Secondary Containment nor RCS integrity are established. RCS integrity is in place when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). No delay time is allowed for EAL1 because the evaporated reactor coolant that may be released into the Containment during this heatup condition could also be directly released to the environment.

EAL 2 addresses the complete loss of functions required for core cooling for > 20 minutes during refueling and cold shutdown modes when Secondary Containment is CA4

EAL BASES DOCUMENT - EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 22 of 33l established but RCS integrity is not established or RCS inventory is reduced. As in EAL 1, RCS integrity should be assumed to be in place when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). The allowed 20 minute time frame was included to allow operator action to restore the heat removal function, if possible. The allowed time frame is consistent with the guidance provided by Generic Letter 88-17, "Loss of Decay Heat Removal" (discussed later in this basis) and is believed to be conservative given that a low pressure Containment barrier to fission product release is established. EAL 2 is not applicable if actions are successfulIn restoringan RCS heat removal system to operation and RCS temperature is being reduced within the 20 minute time frame.

EAL 3 addresses complete loss of functions required for core cooling for > 60 minutes during refueling and cold shutdown modes when RCS integrity is established. As in EAL 1 and 2, RCS integrity should be considered to be in place when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). The status of Secondary Containment in this EAL is immaterial given that the RCS is providing a high pressure barrier to fission product release to the environment. The 60 minute time frame should allow sufficient time to restore cooling without there being a substantial degradation in plant safety. The pressure increase covers situations where, due to high decay heat loads, the time provided to restore temperature control, should be less than 60 minutes. The RCS pressure setpoint chosen is 10 psig since this is the lowest value that can be read on installed Control Board instrumentation. EAL 3 is not applicable if actions are successfulIn restoring an RCS heat removal system to operation and RCS temperature Is being reduced within the 60 minute time frame assuming that the RCS pressure Increase has remained less than the site specificpressure value.

Escalation to Site Area would be via CS1 or CS2 should boiling result in significant RPV level loss leading to core uncovery.

A loss of Technical Specification components alone is not intended to constitute an Alert. The same is true of a momentary unplanned excursion above 200 OF when the heat removal function is available.

The Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded.

CA4

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION I Page 23 of 33

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels

2. NEP 2004-0034, EAL Submittal - Containment Pressure Indicator Justification CA4

.I EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 24of 33 CS1 Loss of RPV Inventory Affecting Core Decay Heat Removal Capability EVENT TYPE: RCS Level OPERATING MODE APPLICABILITY: Cold S/D EAL THRESHOLD VALUE:

CS1.1 With Secondary Containment not established:

a. RPV inventory as indicated by RPV level is LESS THAN 113.5 inches OR

b. RPV level cannot be monitored for GREATER THAN 30 minutes with a loss of RPV inventory as indicated by unexplained Drywell/Reactor Building Equipment or Floor Drain sump, or Torus, level increase OR CS1.2 With Secondary Containment established:

a. RPV inventory as indicated by RPV level is LESS THAN +15 inches OR

b. RPV level cannot be monitored for GREATER THAN 30 minutes with a loss of RPV inventory as indicated by either:

• Unexplained Drywell/Reactor Building Equipment or Floor Drain sump, or Torus, level increase

• Erratic Source Range Monitor Indication DAEC EAL INFORMATION:

Under the conditions specified by this IC, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RPV breach, pressure boundary leakage, or continued boiling in the RPV.

CS1

- EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION I Page 25 of 33 In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode.

Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> {site-specific} or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). The above forms the basis for needing both a cold shutdown specific IC (CS1) and a refueling specific IC (CS2).

In the cold shutdown mode, normal RCS level indication systems will normally be available. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and Torus level changes. The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building. A rise in Torus water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Torus level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

These example EALs are based on concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal, SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues, NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States, and, NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. A number of variables, (BWRs - e.g., such as initial vessel level, or shutdown heat removal system design) can have a significant impact on heat removal capability challenging the fuel clad barrier. Analysis in the above references indicates that core damage may occur within an hour following continued core uncovery therefore, conservatively, 30-minutes was chosen.

The 30-minute duration allowed when Secondary Containment is established allows sufficient time for actions to be performed to recover needed cooling equipment and is considered to be conservative given that level is being monitored via CS1 and CS2. For BWRs releases would be monitored and escalation would be via Category R ICs if required.

CS1

Thus, BWR declaration of a Site Area Emergency is warranted under the conditions specified by the IC. Escalation to a General Emergency is via CG1 or RG1.

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels CS1

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 27 of 33 CS2 Loss of RPV Inventory Affecting Core Decay Heat Removal Capability with Irradiated Fuel in the RPV EVENT TYPE: RCS Level OPERATING MODE APPLICABILITY: Refueling EAL THRESHOLD VALUE:

CS2.1 With SECONDARY CONTAINMENT NOT ESTABLISHED, EITHER of the following occurs:

(a) RPV inventory as indicated by RPV level is LESS THAN 113.5 inches (b) RPV level cannot be monitored with Indication of core uncovery as evidenced by one or more of the following:

• Containment High Range Rad Monitor reading GREATER THAN 10 Rem/hr.

• Erratic Source Range Monitor Indication OR CS2.2 With SECONDARY CONTAINMENT ESTABLISHED, EITHER of the following occurs:

(a) RPV inventory as indicated by RPV level is LESS THAN +15 inches (b) RPV level cannot be monitored with Indication of core uncovery as evidenced by one or more of the following:

• Containment High Range Rad Monitor reading GREATER THAN 10 Rem/hr.

• Erratic Source Range Monitor Indication CS2

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 28 of 33 DAEC EAL INFORMATION:

Under the conditions specified by this IC, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RPV breach or continued boiling in the RPV. Since BWRs have RCS penetrations below the setpoint, continued level decrease may be indicative of pressure boundary leakage.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode.

Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for typically 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or longer after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). The above forms the basis for needing both a cold shutdown specific IC (CS1) and a refueling specific IC (CS2).

These example EALs are based on concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal, SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues, NUREG-1449, Shutdown and Low-Power Operationat Commercial Nuclear PowerPlants in the United States, and, NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. A number of variables, (BWRs - e.g., such as initial vessel level, or shutdown heat removal system design) can have a significant impact on heat removal capability challenging the fuel clad barrier. Analysis in the above references indicates that core damage may occur within an hour following continued core uncovery therefore, conservatively, 30 minutes was chosen.

As water level in the RPV lowers, the dose rate above the core will increase. The dose rate due to this core shine will result in significantly increased Containment High Range Radiation Monitor readings. An unexplained reading of greater than 10 Rem/hr may be indicative of fuel damage. The basis for 10 Rem/hr is that it is sufficiently above the normal shutdown levels to avoid an unnecessary entry into the EAL. The 10 Rem/hr is also well below the containment radiation monitor reading of 2E+2 R/hr that would be indicative of 1% clad failure found in the following calculation:

CS2

-EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 29 of 33 Calculation of Drywell Radiation Monitor Reading Assuming 1% 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 = 2.9E+3Remlhr Drywell reading = 2.9E+3Remlhr x [1 % /20  %/J]= 1.45E+2 Rem/hr, round off as 2E+2 Rem/hr Post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

For EAL 2 in the refueling mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted.

For BWRs releases would be monitored and escalation would be via Category R ICs if required. Thus, BWR declaration of a Site Area Emergency is warranted under the conditions specified by the IC.

Escalation to a General Emergency is via CG1 or RGI.

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels CS2

EAL BASES DOCUMENT - EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 30 of 33 CGI Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged with Irradiated Fuel in the RPV EVENT TYPE: Inability to Reach or Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Cold S/D, Refueling EAL THRESHOLD VALUE:

CG1.1 (1) Loss of RPV inventory as indicated by unexplained Drywell/Reactor Building Equipment or Floor Drain sump, or Torus, level increase AND (2) RPV Level:

(a) LESS THAN +15 inches for GREATER THAN 30 minutes OR (b) Cannot be monitored with Indication of core uncovery for GREATER THAN 30 minutes as evidenced by one or more of the following:

• Containment High Range Rad Monitor reading GREATER THAN 10 Rem/hr.

• Erratic Source Range Monitor Indication AND (3) Indication of Secondary Containment challenged as indicated by one or more of the following:

• Drywell Hydrogen or Torus Hydrogen GREATER THAN 6%

AND Drywell Oxygen or Torus Oxygen GREATER THAN 5%

• Containment Pressure GREATER THAN 53 psig

• Secondary Containment not established

• Two or more Reactor Building areas exceed Max Safe Radiation Levels CG1

EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 31 of 33 DAEC EAL INFORMATION:

For EAL 1 in the cold shutdown mode, normal RCS level and RPV level instrumentation systems will normally be available. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and Torus level changes. Sump and Torus level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

For EAL 1 in the refueling mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and Torus level changes.

EAL 2 represents the inability to restore and maintain RPV level to above the top of active fuel. Fuel damage is probable if RPV level cannot be restored, as available decay heat will cause boiling, further reducing the RPV level.

These example EALs are based on concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal, SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues, NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States, and, NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.) A number of variables, (BWRs - e.g., such as initial vessel level, or shutdown heat removal system design) can have a significant impact on heat removal capability challenging the fuel clad barrier. Analysis in the above references indicates that core damage may occur within an hour following continued core uncovery therefore, conservatively, 30 minutes was chosen.

As water level in the RPV lowers, the dose rate above the core will increase. The dose rate due to this core shine will result in significantly increased Containment High Range Radiation Monitor readings. An unexplained reading of greater than 10 Rem/hr may be indicative of fuel damage. The basis for 10 Rem/hr is that it is sufficiently above the normal shutdown levels to avoid an unnecessary entry into the EAL. The 10 Rem/hr is also well below the containment radiation monitor reading of 2E+2 R/hr that would be indicative of 1% clad failure found in the following calculation:

CG1

EAL BASES DOCUMENT I EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION

. Page 32 of 33 Calculation of Drywell Radiation Monitor Reading Assuming 1% 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 = 2.9E+3Rem/hr Drywell reading = 2.9E+3Rem/hr x [1 % /20 %] = 1.45E+2 Rem/hr, round off as 2E+2 Rem/hr The GE is declared on the occurrence of the loss or imminent loss of function of all three barriers. Based on the above discussion, RCS barrier failure resulting in core uncovery for 30 minutes or more may cause fuel clad failure. With the Secondary Containment breached or challenged then the potential for unmonitored fission product release to the environment is high. This represents a direct path for radioactive inventory to be released to the environment. This is consistent with the definition of a GE.

In the context of EAL 3, Secondary Containment closure is the action taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Secondary Containment should not be confused with refueling containment integrity as defined in technical specifications. Site shutdown contingency plans typically provide for re-establishing Secondary Containment following a loss of heat removal or RCS inventory functions. If Secondary Containment is re-established prior to exceeding the temperature or level thresholds of the RCS Barrier and Fuel Clad Barrier EALs, escalation to GE would not occur.

For BWRs, the use of secondary containment radiation monitors should provide indication of increased release that may be indicative of a challenge to secondary containment. The site-specific radiation monitor values should be based on the EOP "maximum safe values" because these values are easily recognizable and have an emergency basis.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gasses in Secondary Containment. However, Secondary Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists.

CG1

N EAL BASES DOCUMENT EBD-C Rev. 0 COLD SHUTDOWN/REFUELING - SYSTEMS MALFUNCTION Page 33 of 33 l

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels CGI

EAL BASES DOCUMENT EBD-E Rev. 1 ISFSI ABNORMAL EVENTS CATEGORY PAGE 1 of 6 Usage Level INFORMATION I Effective Date:

TECHNICAL REVIEW Prepared and Verified by: Date:

Validated by: Date:

Emergency Planning Staff

[ PROCEDURE APPROVAL I am responsible for the technical content of this procedure and for obtaining the necessary approval from the State and County Emergency Management officials prior to implementation.

Documentation of State and County Emergency Management approval is via NEP-Approved by: Date:

Manager, Emergency Planning

EAL BASES DOCUMENT EBD-E Rev. 1 ISFSI ABNORMAL EVENTS CATEGORY PAGE 2 of6 Table of Contents EU1 Damage To A Loaded Cask Confinement Boundary ............................................ 3 EU2 Confirmed Security Event With Potential Loss Of Level Of Safety Of The ISFSI..5

-EAL BASES DOCUMENT EBD-E Rev. 1 ISFSI ABNORMAL EVENTS CATEGORY PAGE 3 of 6 EUI Damage To A Loaded Cask Confinement Boundary EVENT TYPE: Independent Spent Fuel Storage Installation (ISFSI)

OPERATING MODE APPLICABILITY: Not Applicable EAL THRESHOLD VALUE:

EU1.1 Any one of the following natural phenomena events with resultant visible damage to or loss of a loaded cask confinement boundary:

• Report by plant personnel of a tornado strike.

• Report by plant personnel of a seismic event.

OR EU1.2 The following accident condition with resultant visible damage to or loss of a loaded cask confinement boundary:

• A loaded transfer cask is dropped as a result of normal handling or transporting.

OR EU1.3 Any condition in the opinion of the Emergency Director that indicates loss of loaded fuel storage cask confinement boundary.

DAEC EAL INFORMATION:

The CONFINEMENT BOUNDARY is the barrier (Dry Shielded Canister (DSC)) that separates areas containing radioactive substances, spent nuclear fuel or high-level waste, and the environment.

During all packaging, transfer, and storage activities, the DSC is completely enclosed in one of two additional containers, the DSC transfer cask or the horizontal storage module, and is never exposed to the environment. Both of these devices provide physical missile protection and radiation shielding for the DSC.

Because the CONFINEMENT BOUNDARY is not directly accessible for visible inspection, the DAEC EAL definition of VISIBLE DAMAGE to the CONFINEMENT BOUNDARY is defined as:

damage to the DSC transfer cask or horizontal storage module that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of the dry shielded canister inside the transfer cask or horizontal storage module. Example damage includes: deformation due to heat, impact, or unplanned movement, EUI

-- EAL BASES DOCUMENT EBD-E Rev. 1 ISFSI ABNORMAL EVENTS CATEGORY PAGE 4 of 6 denting, penetration, rupture, cracking, or spalling of concrete to expose concrete reinforcing bar, or reduction in depth or configuration of radiation shielding materials. Surface blemishes (e.g.,

paint fading, paint chipping, concrete cracks or scratches) are not included in visible damage.

For the events of concern here, the key issue is whether the resultant damage to or loss of the loaded fuel cask CONFINEMENT BOUNDARY leads to the degradation of the fuel during transfer or storage, or poses an operational safety problem with respect to its removal from storage. The wind speed, earthquake intensity, height of loaded transfer cask drop as a result of normal handling or transporting etc., in and of themselves, are not the key issue.

EAL Threshold Values for earthquakes are addressed in accordance with AOP 901, "Earthquake."

EAL Threshold Values for Tornadoes are addressed in accordance with AOP 903, "Tornado."

For EAL 1 and EAL 2, the results of the ISFSI Safety Analysis Report (SAR) per NUREG-1536 or SAR referenced in the cask's Certificate of Compliance and the related NRC Safety Evaluation Report were used to develop the DAEC list of natural phenomena events and accident conditions. These EALs address responses to a dropped cask, a tipped over cask, or natural phenomena affecting a cask (e.g., seismic event, tornado, etc.) or a dropped cask.

(Reference Action Request OTH026062 for credible and non-credible event analysis.)

For EAL 3, any condition not explicitly detailed as an EAL threshold value, which, in the judgment of the Emergency Director, is a potential degradation in the level of safety of the ISFSI. Emergency Director judgment is to be based on known conditions and the expected response to mitigating activities within a short time period.

REFERENCES:

1. "Methodology for Development of Emergency Action Levels," NEI 99-01 Revision 4, January 2003 (DAEC EU1 is renamed from NEI E-HU1)

2. Abnormal Operating Procedure (AOP) 901, "Earthquake"

3. Abnormal Operating Procedure (AOP) 903, "Tornado"

4. NUREG-1 536, "Standard Review Plan for Dry Cask Storage Systems"

5. SAR referenced in the cask's Certificate of Compliance and the related NRC Safety Evaluation Report

6. Action Request OTH026062 EUI

EAL BASES DOCUMENT -  ; EBD-E Rev. 1 ISFSI ABNORMAL EVENTS CATEGORY PAGE 5 of 6 EU2 Confirmed Security.Event With Potential Loss Of Level Of Safety Of The ISFSI EVENT TYPE: Independent Spent Fuel Storage Installation (ISFSI)

OPERATING MODE APPLICABILITY: Not Applicable EAL THRESHOLD VALUE:

EU2.1 DAEC Security Supervision reports ANY of the following:

Suspected sabotage device affecting a horizontal storage module, dry shielded canister or transfer cask, or found inside ISFSI protected area.

• Confirmed tampering with a horizontal storage module, dry shielded canister or transfer cask.

• A hostage situation that disrupts normal ISFSI operations.

• Civil disturbance or strike that disrupts normal ISFSI operations.

• Internal disturbance that is not short lived or is not a harmless outburst involving one or more individuals within the ISFSI protected area.

• Intrusion into the ISFSI protected area by a hostile force.

• Any security event of increasing severity that persists for 30 minutes, or greater, which affects the ISFSI:

o Credible bomb threats o Hostage/Extortion o Suspicious fire or Explosion o Significant Security System Hardware Failure o Loss of Guard Post Contact DAEC EAL INFORMATION:

Security events which do not represent a potential degradation in the level of safety of the ISFSI are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72.

EU2

EAL BASES DOCUMENT EBD-E Rev. 1 ISFSI ABNORMAL EVENTS CATEGORY PAGE 6 of 6 Reference is made to DAEC Security Supervision because these individuals are the designated personnel qualified and trained to confirm that a security event is occurring or has occurred.

Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Security Plan. These IC's are based upon the Site Security Plan.

EAL 1 describes a suspected sabotage device affecting a Horizontal Storage Module, Dry Shielded Canister or Transfer Cask, or located inside the ISFSI Protected Area. This is considered a potential degradation -in the level of safety of the ISFSI.

EAL 2 is for confirmed tampering with a Horizontal Storage Module, Dry Shielded Canister or Transfer Cask located inside the Protected Area at the ISFSI as well as in transport to the ISFSI. This is considered a potential degradation in the level of safety of the ISFSI.

EAL 3 identifies a hostage situation that disrupts normal ISFSI operations. A hostage situation is considered to disrupt normal operations if it results in the inability to perform surveillance activities, or alters other activities at the ISFSI or during transport operations.

EAL 4 describes a civil disturbance or strike which are considered to be spontaneous activities that disrupt normal ISFSI operations. A civil disturbance or strike is considered to disrupt normal ISFSI operations ifit initially disrupts normal ingress or egress to the ISFSI Protected Area, or if it requires assistance from the Local Law Enforcement Agencies (LLEA) to control.

EAL 5 deals with suspicious internal disturbances that may have been planned by one or more individuals as a diversion to gain entry to the ISFSI Protected Area.

EAL 6 is an intrusion of a hostile force into the ISFSI Protected Area representing a potential substantial degradation of the level of safety of the ISFSI. A civil disturbance which penetrates the Protected Area can be considered a hostile force.

EAL 7 is for security events of increasing severity that persist for 30 minutes, or longer, affecting the ISFSI. A security event is considered to be "of increasing severity" ifthe event is NOT under control of the security force within 30 minutes.

REFERENCES:

1. Methodology for Development of Emergency Action Levels," NEI 99-01 Revision 4, January 2003 (DAEC EU2 is renamed from NEI E-HU2)

2. Abnormal Operating Procedure (AOP) 914, "Security Events" EU2

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 1 of 30 l ~Usage Level l

. ~Information Usel l Effective Date:

TECHNICAL REVIEW Prepared by: Date:

Reviewed by: Date:

Independent Reviewer Reviewed by: Date:

Operations Staff PROCEDURE APPROVAL I am responsible for the technical content of this procedure and for obtaining the necessary approval from the State and County Emergency Management officials prior to implementation.

Documentation of State and County Emergency Management approval is via NEP-Approved by: Date:

Manager, Emergency Planning

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 2 of 30 Table of Contents FUl Any Loss or Any Potential Loss of Primary Containment Barrier .......................... 3 FA1 Any Loss or Any Potential Loss of Either Fuel Clad Or RCS Barrier .................... 4 FS1 Loss Or Potential Loss of Any Two Barriers ....................................................... 5 FG1 Loss of Any Two Barriers AND Potential Loss of the Third Barrier ....................... 6 FISSION BARRIER: Fuel Clad ....................................................... 7 FISSION BARRIER: RCS ...................................................... 14 FISSION BARRIER: Primary Containment ...................................................... 22

-EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 3 of 30 FU1 Any Loss or Any Potential Loss of Primary Containment Barrier EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot Shutdown I EAL Threshold Values:

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

DAEC INFORMATION:

The entry conditions for this Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table. This logic is simplified from the generic NEI 99-01 logic based on the following considerations: I

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. Comprehensiveness - A comparison was made of the combinations of barrier losses and potential losses between the DAEC logic and the NEI 99-01 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 NEI 99-01 logic are significantly affected by the simplified logic when applied to I a BWR.

REFERENCES:

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

FUI

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 4 of 30 FAI Any Loss or Any Potential Loss of Either Fuel Clad Or RCS Barrier I EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot Shutdown I EAL Threshold Values:

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

DAEC INFORMATION:

The entry conditions for this Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table. 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. Comprehensiveness - A comparison was made of the combinations of barrier losses and potential losses between the DAEC logic and the NEI 99-01 logic. All six generic I 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 NEI 99-01 logic are significantly affected by the simplified logic when applied to a I BWR.

REFERENCES:

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

FA1

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 5 of 30 FS1 Loss Or Potential Loss of Any Two Barriers EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot Shutdown EAL Threshold Values:

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

DAEC INFORMATION:

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. Comprehensiveness - A comparison was made of the combinations of barrier losses and potential losses corresponding to Site Area Emergency between the DAEC logic and the NEI 99-01 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 NEI 99-01 logic are significantly affected by the simplified logic when applied to a BWR.

REFERENCES:

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

FS1

- - EAL BASES DOCUMENT EBD-F

. I Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 6 of 30 FG1 Loss of Any Two Barriers AND Potential Loss of the Third Barrier I EVENT TYPE: See Fission Barrier Table OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot Shutdown I EAL Threshold Values:

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

DAEC INFORMATION:

The entry conditions for this Initiating Condition are shown by the logic chart located to the right of the Fission Barrier Table. 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. Comprehensiveness - A comparison was made of the combinations of barrier losses and potential losses between the DAEC logic and the NEI 99-01 logic. All six generic I 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 NEI 99-01 logic are significantly affected by the simplified logic when applied to a I BWR.

REFERENCES:

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

FG1

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION l PAGE 7 of 30 FISSION BARRIER: Fuel Clad DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:

Clad Damage Determination LOSS: Fuel Damage assessment (PASAP 7.2) indicates at least 5% fuel clad damage.

POTENTIAL LOSS: None DAEC INFORMATION:

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 indicators described previously. This can be determined per FUEL DAMAGE ASSESMENT, PASAP 7.2.

REFERENCES:

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

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I I

Fuel Clad Barrier Radiation/Core Damage

. EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 8of30 FISSION BARRIER: Fuel Clad DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:

Drywell/Torus Radiation Monitoring LOSS: Drywell Area Hi Range Rad Monitor RIM-9184A or B reading GREATER THAN 7E+2 Rem/hr OR LOSS: Torus Area Hi Range Rad Monitor RIM-9185A or B reading GREATER THAN 3E+1 Rem/hr POTENTIAL LOSS: None DAEC INFORMATION:

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

There is no significant deviation from the generic "loss" indicator. Per NEI 99-01, the (site-specific) reading (Drywell/Torus Rad - above) is a value that 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 determinationof barrierloss be made whenever the indicatorthreshold 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 from the core. These calculations were based on "nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-Fuel Clad Barrier Radiation/Core Damage

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 9 of 30 conservative for determination of this EAL threshold. Two separate cases were evaluated.

In the first case, the released activity was assumed to be contained in the drywell atmosphere. This case is considered representative of conditions following a line break in which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus. This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results for each case were provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response characteristics of the detector systems. The figures show a drywell reading of about 2.9 x 103 Rem/hr or a torus reading of about 1.1 x 102 Rem/hr I associated with 20% gap release at two hours after shutdown. Scaling this down to 5%

gap release:

Calculation of Drywell and Torus Monitor Readings Assuming 5% Gap Release 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 103 Rem/hr Drywell reading = 2.9 x 10 3 Rem/hr x [5 % / 20 %] = 7.25 x 102 Rem/hr, round off as 7 E+2 Rem/hr 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 102 Rem/hr Torus reading = 1.1 x 1 o2 Rem/hr x [5 % / 20 %] = 2.75 x 101 Rem/hr, round off as 3 E+1 Rem/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

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I I

Fuel Clad Barrier Radiation/Core Damage

• EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 10 of 30 FISSION BARRIER: Fuel Clad DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:

Primary Coolant Activity Level LOSS: Coolant activity GREATER THAN 300 JACi/gm dose equivalent 1-131.

POTENTIAL LOSS: None DAEC INFORMATION:

There is no significant deviation from the generic indicator. Consistent with the generic methodology, DAEC uses a coolant activity value of 300 pICi/gm 1-131 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

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I Fuel Clad Barrier Radiation/Core Damage

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 11 of 30 FISSION BARRIER: Fuel Clad DAEC INDICATOR: RPV Level EAL THRESHOLD VALUE:

Reactor Vessel Water Level LOSS: RPV Level LESS THAN -25 Inches I POTENTIAL LOSS: RPV Level LESS THAN +15 Inches I DAEC INFORMATION:

The loss indicator is based on a value that corresponds to the minimum value to assure core cooling without further degradation of the fuel clad. DAEC uses the Minimum Steam Cooling RPV Water Level of -25 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 0F.

Consistent with the EOPs, an indicated RPV level below -25 inches is used.

The potential loss indicator corresponds to the water level at the top of the active fuel (TAF). Consistent with the EOPs, an indicated RPV level below +15 inches.

REFERENCES:

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

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

3. Emergency Operating Procedure (EOP) Basis, Curves and Limits, C5, Minimum Steam Cooling RPV Water Level

4. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I Fuel Clad Barrier RPV Level

EAL BASES DOCUMENT -  : EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 12 of 30 FISSION BARRIER: Fuel Clad DAEC INDICATOR: Emergency Director Judgment EAL THRESHOLD VALUE:

Emergency Director Judgment Any condition in the opinion 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.

Emergency Director considerations for determining whether any barrier "Loss" or "Potential Loss" include imminent barrier degradation, degraded barrier monitoring capability, and consideration of dominant accident sequences.

Any condition which in the judgement of the Emergency Director indicates a LOSS or POTENTIAL LOSS of the FUEL CLAD barrier such as, but not limited to:

• Degraded barrier monitoring capability from loss of/lack of reliable indicators.

• Consideration for instrumentation operability.

• Portable instrumentation readings.

• Offsite monitoring results.

• Complete loss of 125 VDC.

• Loss of decay heat removal.

• ATWS with failure of Standby Liquid Control.

• Prolonged station blackout.

• Loss of offsite power with early HPCI/RCIC failure Imminent means that no turnaround in safety system performance is expected and that General Emergency conditions can be expected to occur within two hours. Imminent fission barrier degradation must be considered by the Emergency Director to assure timely declaration of a General Emergency and to better assure that offsite protective actions can be effectively accomplished.

Fuel Clad Barrier ED Judgment

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 13 of 30 Degraded barriermonitoringcapability from loss of/lack of reliable indicators must also be considered by the Emergency Director when determining if a fission barrier loss or I potential loss has occurred.

This assessment should also include consideration for instrumentation operability and portable instrumentation readings.

Offsite monitoring results may be an indication of Fission Product Barrier degradation causing an unmonitored release.

Dominant accidentsequences 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 HPCI/RCIC failure. The Emergency Director should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification declaration.

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation

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

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels Fuel Clad Barrier ED Judgment I

EAL BASES DOCUMENT - EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 14 of 30 FISSION BARRIER: RCS DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:

Drywell Radiation Monitoring LOSS: Drywell Area Hi Range Rad Monitor RIM-91 84A or B reading GREATER THAN 5 Rem/hr after Reactor Shutdown POTENTIAL LOSS: None DAEC INFORMATION:

This loss indicator is based on conditions after reactor shutdown to assure that it is not misapplied, i.e., to exclude readings due to N-16 effects which are typically 5 to 8 Rem/hr at full power conditions.

The 5 Rem/hr value for this loss indicator corresponds to instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the drywell atmosphere.

The reading will be less than that specified for the loss indicator for Radiation/Core Damage that applies to the Fuel Clad barrier. Thus, this indicator would be indicative of a RCS leak only. If the radiation monitor reading increased to that value specified by the Radiation/Core indicator applying to the Fuel Clad barrier, fuel damage would also be indicated.

As documented by NG-88-0966, General Electric performed a study to predict dose rate readings from fuel damage calculations for emergency planning. The calculations were performed to obtain gamma ray dose rates at the locations of the containment atmosphere monitoring system radiation detectors in the drywell and torus locations for assumed releases of gap activity from the core. These calculations were based on "nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were evaluated. In the first case, the released activity was assumed to be contained in the drywell atmosphere. This case is considered representative of conditions following a line break in which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus. This could be applied for an event which results in vessel RCS Barrier Radiation/Core Damage

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 15 of 30 isolation and blowdown to the suppression chamber. The results for each case were provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response characteristics of the detector systems. The figures show a drywell reading of about 2.1 x 104 Rem/hr associated with a 100% gap release immediately after shutdown. Assuming 99.99% fuel clad integrity (0.01% gap release) and uniform dispersal of radionuclides into the drywell immediately after shutdown, a drywell monitor reading is calculated:

Calculation of Drywell Monitor Reading Assuming 0.01% Gap Release NG-88-0966 value for 100% Gap Release at 0.01 minutes = 2.1 x 104 Rem/hr (2.1 x 104 ) Rem/hr x [(1 x 1 -2 ) percent / 100 percent] = (2.1) x 104"4 Rem/hr = 2.1 x 10° Rem/hr = 2 Rem/hr To assure an indicator that is readily discernible on the drywell radiation monitor scale, DAEC uses a valid reading above 5 Rem/hr after reactor shutdown.

REFERENCES:

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

2. Technical Specification 3.4.5, Drywell Leak Detection Instrumentation

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I RCS Barrier Radiation/Core Damage

EAL BASES DOCUMENT - EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 16 of 30 FISSION BARRIER: RCS DAEC INDICATOR: RPV Level EAL THRESHOLD VALUE:

Reactor Vessel Water Level LOSS: RPV Level LESS THAN +15 Inches POTENTIAL LOSS: None DAEC INFORMATION:

There is no significant deviation from the generic indicator. This loss indicator corresponds to the water level at the top of the active fuel (TAF). In order to provide normal means to cool the fuel, water level must be maintained above the top of active fuel otherwise extraordinary means must be taken to assure that adequate core cooling exists. In certain failure event sequences reactor vessel water level may be procedurally lowered to the top of active fuel and the reactor coolant system depressurized to allow for steam cooling of the core. Even though fuel clad damage is not predicted under these conditions several safety system failures need to have occurred to reach the condition where steam cooling would be procedurally required. Therefore this is indicative of a loss of the reactor coolant system boundary. Water levels below this value indicate a challenge to core cooling which is a precursor to more serious events.

REFERENCES:

1. Emergency Operating Procedures (EOP) Basis, Breakpoints

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I RCS Barrier RPV Level

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 17 of 30 FISSION BARRIER: RCS DAEC INDICATOR: Leakage EAL THRESHOLD VALUE:

RCS Leak Rate LOSS: Unisolable Main Steamline Break as indicated by the failure of both MSIVs in any one line to close AND EITHER:

High MSL flow or high steam tunnel temperature annunciators

• Direct report of steam release POTENTIAL LOSS: RCS leakage GREATER THAN 50 GPM inside the drywell.

OR POTENTIAL LOSS: Unisolable primary system leakage outside drywell as indicated by area temperatures or ARMs exceeding the Max Normal Limits per EOP 3, Table 6.

DAEC INFORMATION:

There are no significant deviations from the generic potential loss indicators applying to RCS leakage and indications of unisolable primary system leakage.

If an SRV is stuck open or is cycling and no other emergency conditions exist, an emergency declaration may not be appropriate. RCS leakage inside the drywell excludes Safety-Relief Valve (SRV) discharge through the SRV discharge piping into the torus below the water line. However, if the fuel is damaged and the SRV is allowing fission products to escape into primary containment, a loss of RCS should be determined as having occurred. The EC/OSM should also consult SU5, RCS Leakage, to determine if RCS leakage exceeds the threshold required for declaration of an Unusual Event.

Unisolable primary system leakage is considered a Potential loss of RCS based on RCS leakage outside the drywell. Site-specific RCS leakage is determined from temperature or area radiation alarms (ARMs) exceeding the Max Normal limits listed in Table 6, EOP 3. Unisolable primary system leakage in the areas of the steam tunnel, main turbine generator, RCIC, HPCI, etc., indicates a direct path from the RCS to areas outside primary containment. It should be confirmed that the indicators are caused by RCS leakage. Area temperatures or area radiation alarms above Max Normal limits are the criteria for declaration of an Alert classification. An unisolable leak which is indicated RCS Barrier Leakage

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 18 of 30 by exceeding Max Safe limits escalates to a Site Area Emergency when combined with Primary Containment Barrier loss (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

An unisolable MSL break is a breach of the RCS barrier. Thus, this EAL is included for consistency with the Alert emergency classification.

I

REFERENCES:

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

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

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

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

5. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels RCS Barrier Leakage

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 19 of 30 FISSION BARRIER: RCS DAEC INDICATOR: Primary Containment Atmosphere EAL THRESHOLD VALUE:

Drywell Pressure LOSS: Drywell Pressure GREATER THAN 2 psig and not caused by a loss of DW Cooling POTENTIAL LOSS: None DAEC INFORMATION:

There is no significant deviation from the generic indicator. The 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 I Containment. During reactor operation, with drywell cooling in operation and the drywell inerted, the normal operating pressure in the drywell is between 0.5 and 1.0 psig. Analysis at the DAEC shows that a 50 gpm RCS leak would result in a 2 to 3 psig pressure rise over a six minute time period. Since a 2 psig rise would place DAEC above the ECCS initiation setpoint, (2 psig) it is necessary to select the DAEC ECCS initiation setpoint of 2 psig to indicate an actual loss of the RCS. Drywell cooling is not isolated at the 2 psig ECCS initiation setpoint, therefore 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

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

4. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels RCS Barrier Pri. Cont. Atmosphere

EAL BASES DOCUMENT - - EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 20 of 30 FISSION BARRIER: RCS DAEC INDICATOR: Emergency Director Judgment EAL THRESHOLD VALUE:

Any condition in the opinion 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. Emergency Director considerations for determining whether any barrier "Loss" or "Potential Loss" include imminent barrier degradation, degraded barrier monitoring capability, and consideration of dominant accident sequences.

Any condition which in the judgement of the Emergency Director indicates a LOSS or POTENTIAL LOSS of the RCS barrier such as, but not limited to:

• Degraded barrier monitoring capability from loss of/lack of reliable indicators.

• Consideration for instrumentation operability.

• Portable instrumentation readings.

• Offsite monitoring results.

• Complete loss of 125 VDC.

• Loss of decay heat removal.

• ATWS with failure of Standby Liquid Control.

• Prolonged station blackout.

• Loss of offsite power with early HPCI/RCIC failure Imminent means that no turnaround in safety system performance is expected and that General Emergency conditions can be expected to occur within two hours. Imminent fission barrier degradation must be considered by the Emergency Director to assure timely declaration of a General Emergency and to better assure that offsite protective actions can be effectively accomplished.

Degraded barrier monitoring capability from loss of/ack of reliable indicators must also be considered by the Emergency Director when determining if a fission barrier loss or potential loss has occurred.

RCS Barrier ED Judgment

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 21 of 30 This assessment should also include consideration for instrumentation operability and portable instrumentation readings.

Offsite monitoring results may be an indication of Fission Product Barrier degradation causing an unmonitored release.

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 HPCI/RCIC failure. The Emergency Director should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification For the RCS barrier, the Emergency Director should also consider safety-relief valves (SRVs) open or cycling. If an SRV is stuck open or is cycling and no other emergency conditions exist, an emergency declaration may not be appropriate. However, if the fuel is damaged and the SRV is allowing fission products to escape into primary containment, a loss of RCS should be determined as having occurred.

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation

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

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels RCS Barrier ED Judgment

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 22 of 30 FISSION BARRIER: Primary Containment DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:

Significant Radioactive Inventory in Containment LOSS: None POTENTIAL LOSS: Drywell Area Hi Range Rad Monitor RIM-9184A or B reading, GREATER THAN 3E+3 Rem/hr OR POTENTIAL LOSS: Torus Area Hi Range Rad Monitor RIM-9185A or B reading GREATER THAN 1E+2 Rem/hr DAEC INFORMATION:

There is no significant deviation from the generic indicators. The potential loss (site-specific) indicator 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 barriers with Potential Loss of the Primary Containment barrier, and would result in declaration of a General Emergency. The basis for the 20% fuel clad damage threshold is described under the 20% core damage assessment indicator. It is intended that determination of barrierpotential 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 from the core. These calculations were based on "nominal" estimates of fuel rod gap fission product inventory fractions, which are considered to be more appropriate for determining a minimum threshold reading than inventory assumptions found in the NRC Regulatory Guides. The Regulatory Guide inventory assumptions applicable to dose assessments are larger and therefore non-conservative for determination of this EAL threshold. Two separate cases were 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 Primary Containment Barrier Radiation/Core Damage

EAL BASES DOCUMENT -.. EBD-F I ......

Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 23 of 30 which activity is released directly into the drywell. In the second case, the released activity was assumed to be contained in the torus. This could be applied for an event which results in vessel isolation and blowdown to the suppression chamber. The results for each case were provided for each case in the form of gamma ray dose rate versus time profiles for assumed releases of 100% and 20% of the gap activity from the core. The dose rate calculations were carried out independent of any specific information on details of construction or response characteristics of the detector systems. The figures show a drywell reading of about 2.9 x 103 Rem/hr and a torus reading of about 1.1 x 102 Rem/hr I associated with 20% gap release at two hours after shutdown. These values are rounded to 3 E+3 Rem/hr and 1 E+2 Remlhr, respectively. The two hour point was picked I because it allows ample time for the Technical Support Center to be operational and core damage assessment to begin.

REFERENCES:

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

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I I

Primary Containment Barrier RadiationlCore Damage

EAL BASES DOCUMENT. EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 24 of 30 FISSION BARRIER: Primary Containment DAEC INDICATOR: Radiation/Core Damage EAL THRESHOLD VALUE:

Clad Damage Determination LOSS: None POTENTIAL LOSS: Fuel Damage assessment (PASAP 7.2) indicates at least 20%

fuel clad damage.

DAEC INFORMATION:

As a site-specific "potential loss" indicator, DAEC uses determination 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 used earlier with this Indicator. This can be determined using appropriate fuel damage assessment procedures. Regardless of whether primary containment integrity is challenged, it is possible for 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 allowable leakage rates. With or without primary containment challenge, however, a major release of radioactivity requiring off-site protective actions from core damage is not possible unless a major failure of the fuel clad barrier allows radioactive material to be 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 Duing Incident Response to Severe Nuclear Power Plant Accidents, October 1988

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels Primary Containment Barrier RadiationlCore Damage

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 25 of 30 FISSION BARRIER: Primary Containment DAEC INDICATOR: RPV Level EAL THRESHOLD VALUE:

Reactor Vessel Water Level LOSS: None POTENTIAL LOSS: Primary Containment flooding required.

DAEC INFORMATION:

The entry into the Primary Containment Flooding emergency procedure indicates reactor vessel water level can not be restored and that a core melt sequence is in progress. EOPs/SAGs direct the operators to enter Containment Flooding when Reactor Vessel Level cannot be restored to greater than a Site Specific value (generally 2/3 core height) or is unknown. The conditions in this potential loss EAL represent imminent core melt sequences which, if not corrected, could lead to vessel failure and increased potential for containment failure. In conjunction with and an escalation of the level EALs in the Fuel and RCS barrier columns, this EAL will result in the declaration of a General Emergency - loss of two barriers and the potential loss of a third. If the emergency operating procedures have been ineffective in restoring reactor vessel level above the RCS and Fuel Clad Barrier Threshold Values, there is not a "success" path and a core melt sequence is in progress. Entry into Containment flooding procedures is a logical escalation in response to the inability to maintain reactor vessel level. Severe accident analysis (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation with the reactor vessel in a significant fraction of the core damage scenarios, and the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow emergency operating procedures to arrest the core melt sequence. Whether or not the procedures will be effective should be apparent within the time provided. The Emergency Director should make the declaration as soon as it is determined that the procedures have been, or will be, ineffective. There is no "loss" EAL associated with this item.

REFERENCES:

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

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels Primary Containment Barrier Leakage

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 26 of 30 FISSION BARRIER: Primary Containment DAEC INDICATOR: Leakage EAL THRESHOLD VALUE:

Containment Isolation Valve Status After Containment Isolation Signal LOSS: Failure of both valves in any one line to close AND a downstream pathway to the environment exists.

OR LOSS: Unisolable primary system leakage outside the drywell as indicated by area temps or ARMs exceeding the Max Safe Limits per EOP 3, Table 6, when Containment Isolation is required.

OR LOSS: Primary containment venting per EOPs.

POTENTIAL LOSS: None DAEC INFORMATION:

The "loss" indicators used at DAEC directly correspond to the generic indicators. Venting of the primary containment can be performed in accordance with EOP 2 irrespective of the offsite radioactivity release rate that will occur and by defeating isolation interlocks 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.

This EAL is intended to cover the inability to isolate the containment when containment isolation is required. In addition, the presence of area radiation or temperature alarms above the Max Safe limits listed in Table 6, EOP 3 after a containment isolation, indicate an unisolable primary system leakage outside the drywell. The indicators should be confirmed to be caused by RCS leakage. Also, an intentional venting of primary containment for pressure control per EOPs to the secondary containment and/or the Primary Containment Barrier Leakage

EAL BASES DOCUMENT, EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 27 of 30 environment is considered a loss of containment. Containment venting for temperature or pressure when not in an accident situation should not be considered.

REFERENCES:

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

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

3. Emergency Operating Procedures (EOP) Bases, Breakpoints

4. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I Primary Containment Barrier Leakage

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 28 of 30 FISSION BARRIER: Primary Containment DAEC INDICATOR: Primary Containment Atmosphere EAL THRESHOLD VALUE:

Drywell Pressure/Atmosphere LOSS: Rapid unexplained decrease following initial increase in pressure.

OR LOSS: Drywell pressure response not consistent with LOCA conditions.

POTENTIAL LOSS: Torus Pressure reaches 53 PSIG and increasing.

OR POTENTIAL LOSS: Drywell or Torus H2 cannot be determined to be LESS THAN 6% and Drywell or torus 02 CANNOT be determined to be LESS THAN 5%.

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" indicator is torus pressure of 53 psig, which is the Primary Containment Pressure Limit (PCPL) used in the EOPs. The second "potential loss" indicator is based on determination of explosive mixture in accordance with the SAGs.

DAEC SAGs require control of drywell and torus atmosphere gas concentrations to less than 6% H2 and less than 5% 02 to assure that an explosive mixture does not exist. This "potential loss" indicator is written to be consistent with the SAGs.

REFERENCES:

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

2. Severe Accident Guideline - 3 (SAG-3), Hydrogen Control

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels Primary Containment Barrier Primary Containment Atmosphere

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 29 of 30 FISSION BARRIER: Primary Containment DAEC INDICATOR: Emergency Director Judgment EAL THRESHOLD VALUE:

Any condition in the opinion of the Emergency Director that indicates LOSS or POTENTIAL LOSS of the Containment Barrier.

DAEC INFORMATION:

There is no significant deviation from the generic indicator. Emergency Director considerations for determining whether any barrier "Loss" or "Potential Loss" include imminent barrier degradation, degraded barrier monitoring capability, and consideration of dominant accident sequences.

Any condition which in the judgement of the Emergency Director that indicates LOSS or POTENTIAL LOSS of the Primary Containment Barrier such as, but not limited to:

• Degraded barrier monitoring capability from loss of/lack of reliable indicators.

• Consideration for instrumentation operability.

• Portable instrumentation readings.

• Offsite monitoring results.

• Complete loss of 125 VDC.

• Loss of decay heat removal.

• ATWS with failure of Standby Liquid Control.

• Prolonged station blackout.

• Loss of offsite power with early HPCI/RCIC failure Imminent means that no turnaround in safety system performance is expected and that General Emergency conditions can be expected to occur within two hours. Imminent fission barrier degradation must be considered by the Emergency Director to assure timely declaration of a General Emergency and to better assure that offsite protective actions can be effectively accomplished.

Degraded barrier monitoring capability from loss of/lack of reliable indicators must also be considered by the Emergency Director when determining if a fission barrier loss or potential loss has occurred.

Primary Containment Barrier ED Judgment

EAL BASES DOCUMENT EBD-F Rev. 6 FISSION PRODUCT BARRIER DEGRADATION PAGE 30 of 30 This assessment should also include consideration for instrumentation operability and portable instrumentation readings.

Offsite monitoring results may be an indication of Fission Product Barrier degradation causing an unmonitored release.

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 HPCI/RCIC failure. The Emergency Director should also consult System Malfunction EALs, as appropriate, to assure timely emergency classification

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation

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

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels Primary Containment Barrier ED Judgment

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 1 of 33 I

Usage Level INFORMATION USE ll Effective Date:

TECHNICAL REVIEW Prepared and Verified by: _ Date:

Reviewed by: Date:

Emergency Planning Staff Reviewed by: _ Date:

Operations Reviewer PROCEDURE APPROVAL I am responsible for the technical content of this procedure and for obtaining the necessary approval from the State and County Emergency Management officials prior to implementation.

Documentation of State and County Emergency Management approval is via NEP-Approved by: Date:

Manager, Emergency Planning

- - EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 2 of 33 Table of Contents HUI Natural and Destructive Phenomena Affecting the Protected Area .................................. 3 HU2 Fire Within Protected Area Not Extinguished Within 15 Minutes of Detection .................. 7 HU3 Release of Toxic or Flammable Gases Deemed Detrimental to Normal Operation of the Plant ............................................................ 8 HU4 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant............................................................. 9 HU5 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of a NOUE ............................................................ 12 HAl Natural and Destructive Phenomena Affecting the Plant Vital Area ................................ 13 HA2 Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown ............................................................ 17 HA3 Release of Toxic or Flammable Gases Within or Contiguous to a Vital Area Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or Establish or Maintain Safe Shutdown . ............................................................ 20 HA4 Confirmed Security Event in a Plant Protected Area ....................................................... 22 HA5 Control Room Evacuation Has Been Initiated ............................................................ 24 HA6 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert ............................................................ 25 HSI Confirmed Security Event in a Plant Vital Area ............................................................ 26 HS2 Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established ............................................................ 28 HS3 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of Site Area Emergency ............................................................ 30 HG1 Security Event Resulting in Loss Of Physical Control of the Facility ............................... 31 HG2 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency ............................................................ 32

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 3 of 33 HUI Natural and Destructive Phenomena Affecting the Protected Area EVENT TYPE: Natural Disasters and Destructive Phenomena OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HU1.1 Earthquake detected per AOP 901, Earthquake I OR HU1.2 Report of a tornado touching down within the Plant Protected Area, or within the switchyard, with NO confirmed damage to a Safe ShutdownNital Area or Control Room indication of degraded performance of a System of Concern.

OR HU1.3 Report of winds greater than 95 mph within the Plant Protected Area, or within the switchyard, with NO confirmed damage to a Safe ShutdownNital Area or Control Room indication of degraded performance of a System of Concern.

OR HU1.4 Vehicle crash into plant structures or systems within the Plant Protected Area with NO confirmed damage to a Safe ShutdownNital Area or Control Room indication of degraded performance of a System of Concern.

OR HU1.5 Report of an unanticipated explosion within the Plant Protected Area resulting in visible damage to permanent structures or equipment.

OR HU1.6 Report of turbine failure resulting in casing penetration or damage to turbine or I generator seals.

OR HU1.7 River level above 757 feet. I OR HU1.8 Uncontrolled flooding in a Safe ShutdownNital Area that has the potential to affect safety related equipment needed for the current operating mode.

OR HU1.9 River level BELOW 725 feet 6 inches. I HUI

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 4 of 33 DAEC EAL INFORMATION:

The Plant Protected Area is the area within the security fence. This includes ISFSI and the Intake Structure. Although the switchyard is included in this EAL, it is not part of the Plant Protected Area.

DAEC EAL Threshold Value 1 addresses earthquakes that are detected in accordance with AOP 901. For DAEC, a minimum detectable earthquake that is indicated on panel 1C35 is an acceleration greater than i 0.01 Gravity.

DAEC EAL Threshold Value 2 addresses report of a tornado striking within the Plant Protected Area or within the plant switchyard.

DAEC EAL Threshold Value 3 is based on the assumption that high winds within the PROTECTED AREA may have potentially damaged plant structures containing functions or systems required for safe shutdown of the plant.

DAEC EAL Threshold Value 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. Escalation to Alert under HA1 would occur if damage was sufficient to affect the ability to achieve or maintain safe shutdown, e.g., damage made required equipment inoperable or structural damage was observed such as bent supports or pressure boundary leakage.

HU1

EAL BASES DOCUMENT . EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 5 of 33 Safe Shutdown/Vital Areas Category Area Switchyard, 1G31 DG and Day Tank Rooms, 1G21 DG and Day Tank Electrical Power Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Heat Sink Torus Room, Intake Structure, Pumphouse Containment Drywell, Torus Emergency NE, NW, SE Coiner Rooms, HPCI Room, RCIC Room, RHR Valve Room, Systems North CRD Area, South CRD Area, CSTs Other Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56 I Area, SBGT Room I

Systems of Concern

• Reactivity Control

• Containment (Drywellorus)

. RHR/Core SprayISRVs

• HPCIIRCIC

• RHRSWIRiver WaterIESW

• Onsite AC PowerIEDGs

• Offsite AC Power

. Instrument AC

• DC Power

. Remote Shutdown Capability DAEC EAL Threshold Value 5 addresses explosions within the Plant Protected Area. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that significantly imparts significant energy to near-by structures or equipment. Damage can be indicated by report to the control room, physical observation, or by Control Room/local control station instrumentation. Such items as scorching, cracks, dents, or discoloration of equipment or structures are addressed by this EAL. The Emergency Director needs to consider the security aspects of the explosion, if applicable.

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

I DAEC EAL Threshold Value 7 addresses the observed effects of flooding in accordance with AOP 902. Plant site finished grade is at elevation 757.0 ft. Personnel doors and railroad and truck openings at or near grade would require protection in the event of a flood above elevation 757.0 ft.

Therefore, EAL 6 uses a threshold of flood water levels above 757.0 ft.

DAEC EAL Threshold Value 8 addresses the effect of flooding caused by internal events such as component failures, equipment misalignment, or outage activity mishaps. The site-specific areas include those areas that contain systems required for safe shutdown of the plant, that are not HUI

EAL BASES DOCUMENT - EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 6 of 33 designed to be wetted or submerged. Escalation of the emergency classification is based on the damage caused or by access restrictions that prevent necessary plant operations or systems monitoring.

EAL Threshold Value 9 addresses the effects of low river water level. The intake structure for the safety-related water supply systems (river water, RHR service water, and emergency service water) is located on the west bank 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 stream bed flow and divert it to the intake structure. This makes the entire flow of the river available to the safety-related water supply systems. A minimum flow of 13 cubic feet per second (cfs) from a minimum 1000-year river flow of 60 cfs must be diverted. The top of the barrier wall is at elevation 725 ft. 6 in. River water level below this level represents a potential degradation in the level of safety of the plant and is addressed by EAL Threshold Value 9.

In this EAL, "Vital Area" is defined as plant structures or areas containing equipment necessary for a safe shutdown, i.e., synonymous with Safe Shutdown Area.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 901, Earthquake

2. Abnormal Operating Procedure (AOP) 902, Flood

3. Abnormal Operating Procedure (AOP) 903, Tomado

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

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

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

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

8. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels HUI

EAL BASES DOCUMENT - EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 7of33 HU2 Fire Within Protected Area Not Extinguished Within 15 Minutes of Detection EVENT TYPE: Fire OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HU2.1. Fire in buildings or areas contiguous to any Safe ShutdownNital Area not extinguished within 15 minutes of control room notification or verification of a control room alarm.

DAEC EAL INFORMATION:

The purpose of this EAL is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems. This includes such items as fires within the administration building, and security building yet excludes fires in the warehouse or construction support center, waste-basket fires, and other small fires of no safety consequence. As used here, Detection is visual observation and report by plant personnel or sensor alarm indication. The 15 minute time period begins with a credible notification that a FIRE is occurring, or notification of a VALID fire detection system alarm. Verification of a fire detection system alarm includes actions that can be taken within the control room or other nearby location to ensure that the alarm is not spurious. A verified alarm is assumed to be an indication of a FIRE unless it is disproved within the 15-minute period by personnel dispatched to the scene. In other words, a personnel report from the scene may be used to disprove a sensor alarm if received within 15 minutes of the alarm, but shall not be required to verify the alarm.

Per AOP 913, the location of a fire can be determined by observing 1C40B alarm messages, Zone Indicating Unit (ZIU) alarms, or fire annunciators on panels 1C40 and 1C40A. The location of a fire can also be determined by verbal report of the person discovering the fire. Verification of the alarm in this context means those actions taken to determine that the control room alarm is not spurious.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 913, Fire

2. Abnormal Operating Procedure (AOP) 914, Security

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels HU2

EAL BASES DOCUMENT' EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 8 of 33 HU3 Release of Toxic or Flammable Gases Deemed Detrimental to Normal Operation of the Plant EVENT TYPE: Other Hazards and Failures OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HU3.1 Report or detection of toxic or flammable gases that has or could enter the site area boundary in amounts that can affect normal plant operations.

OR HU3.2 Report by Local, County or State Officials for evacuation or sheltering of site personnel based on an offsite event.

DAEC EAL INFORMATION:

This IC is based on the existence of uncontrolled releases of toxic or flammable gas that may enter the site boundary and affect normal plant operations. It is intended that releases of toxic or flammable gases are of sufficient quantity, and the release point of such gases is such that normal plant operations would be affected. This would preclude small or incidental releases, or releases that do not impact structures needed for plant operation. The EALs are intended to not require significant assessment or quantification. The IC assumes an uncontrolled process that has the potential to affect plant operations, or personnel safety.

REFERENCES:

1. UFSAR Section 2.2, Nearby Industrial, Transportation, and Military Facilities

2. UFSAR Section 6.4, Habitability Systems

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels HU3

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 9 of 33 HU4 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HU4.1 Credible Security Threat OR HU4.2 DAEC Security Supervision reports any of the following:

• Suspected sabotage device discovered within plant Protected Area.

• Suspected sabotage device discovered outside the Protected Area, in the plant switchyard, or ISFSI.

• Confirmed tampering with safety related equipment.

• A hostage/extortion situation that disrupts normal plant or ISFSI operations.

• Civil disturbance or strike which disrupts normal plant or ISFSI operations.

• Internal disturbance that is not short lived or that is not a harmless outburst involving one or more individuals within the Protected Area or ISFSI.

• Malevolent use of a vehicle outside the Protected Area which disrupts normal plant operations.

DAEC EAL INFORMATION:

Security events are based upon the Site Security Plan.

Security events which do not represent at least a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. The term "suspected sabotage device" is used in place of "bomb device" for consistency with the DAEC Safeguards Contingency Plan.

Consultation with Security supervision is required to determine these Threshold Values.

EAL 1 ensures that appropriate notifications for the security threat are made in a timely manner. The emergency response to a Credible Security Threat is initiated through AOP 914, HU4

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 10 of 33 "Security Events" and EPIP 2.8, uSecurity Threat". Only the plant to which the specific threat is made need declare the Notification of Unusual Event.

EAL 2 describes a set of Security Events, as reported by DAEC Security Supervision, as follows:

Suspected sabotage device discovered within the Protected Area. It is a potential degradation of the level of safety of the plant and is an UNUSUAL EVENT.

Suspected sabotage device discovered outside the Protected Area, in the plant switchyard or ISFSI, representing a potential degradation of the level of safety of the plant.

Confirmed tampering with safety related equipment is adapted from the list of security plan contingencies.

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

A civil disturbance or strike if considered to be a spontaneous activity that disrupts normal plant or ISFSI operations. A civil disturbance or strike is considered to disrupt normal plant operations if it initially disrupts normal ingress or egress to the owner controlled or protected area, or if it requires assistance from the Local Law Enforcement Agencies (LLEA) to control.

Suspicious internal disturbances that may have been planned by unauthorized personnel as a diversion to gain entry to the site property is also added to this list of security events.

Malevolent use of a vehicle outside the Protected Area that disrupts normal plant operations, disrupts normal ingress/egress, or results in the inability to perform surveillance activities is also added to this list of security events.

Suspected sabotage devices discovered within the plant Vital Area would result in escalation via other Security EALs.

REFERENCES:

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

2. NMC Security Procedures

3. EPIP 2.8, "Security Threat" HU4

EALBASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 11 of 33 I

4. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels HU4

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 12 of 33 HU5 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of a NOUE EVENT TYPE: Emergency Director Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HU5.1 Other conditions exist which in the judgment of the Emergency Director indicate that 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.

DAEC EAL INFORMATION:

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

From a broad perspective, one area that may warrant Emergency Director judgment is related to likely or actual breakdown of site-specific event mitigating actions. Examples to consider include inadequate emergency response procedures, transient response either unexpected or not understood, failure or unavailability of emergency systems during an accident in excess of that assumed in accident analysis, or insufficient availability of equipment and/or support personnel.

REFERENCES:

1. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 4, January 2003 HU5

EAL BASES DOCUMENT Il EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 13 of 33 HAI Natural and Destructive Phenomena Affecting the Plant Vital Area EVENT TYPE: Natural Disasters and Destructive Phenomena OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HA1.1 Receipt of the Amber Operating Basis Earthquake Light and the wailing seismic alarm I on 1C35 (+/- 0.06 gravity).

OR HA1.2 Report of Tornado or high winds greater than 95MPH within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to a Safe ShutdownNital Area or Control Room indication of degraded performance of a System of Concern.

OR HA1.3 Vehicle crash within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE I to a Safe ShutdownNital Area or Control Room indication of degraded performance of a System of Concern.

OR HAA.4 Turbine failure-generated missiles result in any VISIBLE DAMAGE to or penetration of any of a Safe ShutdownNital Area.

OR HA1.5 River level above 767 feet. I OR HA1.6 Uncontrolled flooding in a Safe ShutdownNital Area that results in degraded safety system performance as indicated in the Control Room or that creates an industrial safety hazards (e.g., electric shock) that precludes access necessary to operate or monitor safety equipment.

OR HA1.7 River level below 724 feet 6 inches. I OR HA1.8 Report to control room of damage affecting a Safe Shutdown/Vital Areas. I HA1

DAEC EAL INFORMATION:

Forthe events of concern here, the key issue is not the wind speed, earthquake intensity, etc., but whether there is resultant damage to equipment orstructuresrequiredto achieve or maintain safe shutdown, regardlessof the cause. Determination of damage affecting the ability to achieve or maintain safe shutdown can be indicated by reports to the control room, physical observation or by Control Room/local control station instrumentation.

EAL Threshold Value 1 addresses OBE events that are detected in accordance with AOP 901.

For DAEC, the OBE is associated with a peak horizontal acceleration of +/- 0.06 Gravity.

DAEC EAL Threshold Value 2 addresses report of a tornado striking a plant vital area and addresses high wind speeds as measured by the 33-Foot or 156-Foot elevations on the Meteorological Tower. The design basis wind speed is 105 miles per hour. However, the meteorological instrumentation is only capable of measuring wind speeds up to 100 miles per hour. Thus the alert level for sustained high wind speed, 95 miles per hour, is selected to be on-scale for the meteorological instrumentation and to conservatively account for potential measurement errors.

DAEC EAL Threshold Value 3 addresses vehicle (automobile, aircraft, forklift, truck or train) confirmed crashes affecting plant vital areas.

DAEC EAL Threshold Value 4 addresses Turbine failure-generated missiles affecting safe shutdown areas. This threshold addresses the threat to safety related equipment from missiles generated by main turbine rotating component catastrophic failures.

Safe ShutdownNital Areas -

Category Area Switchyard, 1G31 DG and Day Tank Rooms, 1G21 DG and Day Tank Electrical Power Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Coolant Supply Torus Room, Intake Structure, Pumphouse Containment Drywell, Torus Emergency NE, NW, SE Comer Rooms, HPCI Room, RCIC Room, RHR Valve Room, Systems North CRD Area, South CRD Area, CSTs Other Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56 I

_ __ Area, SBGT Room HAI

EAL BASES DOCUMENT - - EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 15 of 33 Systems of Concern

• Reactivity Control

• Containment (DrywelVTorus)

• RHR/Core Spray/SRVs

• HPCVRCIC

• RHRSW/River Water/ESW

• Onsite AC Power/EDGs

• Offsite AC Power

• Instrument AC

• DC Power

. Remote Shutdown Capability DAEC EAL Threshold Value 5 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 from the "maximum probable flood" to which the site could be subjected. The design flood water is at elevation 767.0 ft. Major equipment penetrations in the exterior walls are located above elevation 767.0 ft. Openings below the flood level are either watertight or are provided with means to control the inflow of water in order to ensure that a safe shutdown can be achieved and maintained.

EAL Threshold Value 6 addresses the effect of internal flooding that has resulted in degraded performance of systems affected by the flooding, or has created industrial safety hazards (e.g.,

electrical shock) that preclude necessary access to operate or monitor safety equipment. The inability to operate or monitor safety equipment represents a potential for substantial degradation of the level of safety of the plant. This flooding may have been caused by internal events such as component failures, equipment misalignment, or outage activity mishaps. The site-specific areas includes those areas that contain systems required for safe shutdown of the plant, that are not designed to be wetted or submerged.

DAEC EAL Threshold Value 7 addresses the effects of low river water level. The intake structure for the safety-related water supply systems (river water, RHR service water, and emergency service water) is located on the west bank of the Cedar River. The overflow weir is at elevation 724 feet 6 inches. River level at or below this elevation will result in all river flow being diverted to the safety related water supply systems. The top of the intake structure around the pump wells is at elevation 724 feet. If the river water level dropped to this 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.

DAEC EAL Threshold Value 8 addresses a report to the control room of damage affecting Safe ShutdownNital areas. The reported damage can be from tornadoes, high winds, flooding, missiles, collisions, or any other cause. The missiles mentioned here can be from any cause, e.g., tornado-generated; turbine, pump or other rotating machinery catastrophic failure; or generated from an explosion.

HA1

EAL BASES DOCUMENT tU ' EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 16 of 33 In this EAL, "Vital Area" is defined as plant structures or areas containing equipment necessary for a safe shutdown, i.e., synonymous with Safe Shutdown Area.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 901, Earthquake

2. Abnormal Operating Procedure (AOP) 902, Flood

3. Abnormal Operating Procedure (AOP) 903, Tornado

4. Abnormal Operating Procedure (AOP) 913, Fire

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

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

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

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

9. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels HA1

EAL-BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 17 of 33 HA2 Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown EVENT TYPE: Fire OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HA2.1 Fire or explosion in any Safe Shutdown/Vital Area.

AND Affected system parameter indications show degraded performance or plant personnel report VISIBLE DAMAGE to permanent structures or equipment within the specified area.

DAEC EAL INFORMATION:

Of particular concern for this EAL are fires that may be detected in any Safe Shutdown/Vital Area. I 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 of the damage.

Per AOP 913, the location of a fire can be determined by observing IC40B alarm messages, Zone Indicating Unit (ZIU) alarms, or fire annunciators on panels 1C40 and 1C40A.

HA2

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 18 of 33 I

NOTE Scope of Systems and Equipment of concern was established by review of Appendix R Safe Shutdown credited systems. 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.

Systems of Concem

• Reactivity Control

• Containment (DrywellTorus)

• RHR/Core Spray/SRVs

• HPCI/RCIC

• RHRSW/River Water/ESW

• Onsite AC Power/EDGs

• Offsite AC Power

• Instrument AC

• DC Power

• Remote Shutdown Capability This EAL addresses a FIRE I EXPLOSION and not the degradation in performance of affected systems. System degradation is addressed in the System Malfunction EALs. The reference to damage of systems is used to identify the magnitude of the FIRE I EXPLOSION and to discriminate against minor FIREs I EXPLOSIONs. The reference to safety systems is included to discriminate against FIREs I EXPLOSIONs in areas having a low probability of affecting safe operation. The significance here is not that a safety system was degraded but the fact that the FIRE I EXPLOSION was large enough to cause damage to these systems. Thus, the designation of a single train was intentional and is appropriate when the FIRE I EXPLOSION is large enough to affect more than one component. Lagging fires, fires in waste containers or any miscellaneous fires that may be in the vicinity of safety systems, but do not cause damage to these systems, should NOT be considered for this EAL.

With regard to EXPLOSIONS, only those EXPLOSIONS of sufficient force to damage permanent structures or identified equipment required for safe operation, should be considered. As used here, an EXPLOSION is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to near-by structures and materials. The occurrence of the EXPLOSION with reports of evidence of damage (e.g.,

deformation, scorching) is sufficient for the declaration. The Emergency Director also needs to consider any security aspects of the EXPLOSIONS, if applicable.

I HA2

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 19 of 33

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

5. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I HA2

- EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 20 of 33 I

HA3 Release of Toxic or Flammable Gases Within or Contiguous to a Vital Area Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or Establish or Maintain Safe Shutdown.

EVENT TYPE: Other Hazards and Failures OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HA3.1 Report or detection of toxic gases within or contiguous to a Safe ShutdownNital Area in concentrations that may result in an atmosphere Immediately Dangerous to Life and Health (IDLH).

OR HA3.2 Report or detection of gases in concentration greater than the Lower Flammability Limit within or contiguous to a Safe ShutdownNital Area.

DAEC EAL INFORMATION:

This IC is based on gases that affect the safe operation of the plant. This IC applies to buildings and areas contiguous to Safe ShutdownNital areas or other significant buildings or areas (i.e.,

service water pump house). The intent of this IC is not to include buildings (e.g., warehouses) or other areas that are not contiguous or immediately adjacent to a Safe Shutdown/Vital Area. It is appropriate that increased monitoring be done to ascertain whether consequential damage has occurred. Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Levels / Radioactive Effluent, or Emergency Director Judgment ICs.

EAL #1 is met if measurement of toxic gas concentration results in an atmosphere that is IDLH within a VITAL AREA or any area or building contiguous to a Safe Shutdown/Vital Area. Exposure to an IDLH atmosphere will result in immediate harm to unprotected personnel, and would preclude access to any such affected areas.

EAL #2 is met when the flammable gas concentration in a Safe ShutdownNital Area or any building or area contiguous to a Safe ShutdownNital Area exceed the LOWER FLAMMABILITY LIMIT. Flammable gasses, such as hydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repair equipment/components (acetylene - used in welding). This EAL addresses concentrations at which gases can ignite/support combustion. An uncontrolled release of flammable gasses within a facility structure has the potential to affect safe operation of the plant by limiting either operator or equipment operations due to the potential for ignition HA3

EAL BASES DOCUMENT - EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 21 of 33 and resulting equipment damage/personnel injury. Once it has been determined that an uncontrolled release is occurring, then sampling must be done to determine if the concentration of the released gas is within this range.

PerAOPs 913 and 914, the following areas are identified as safe shutdown areas and are shown I on the EAL tables. This table is displayed as an aid to the Emergency Director in determining appropriate areas of concern.

-__ -Safe ShutdownNital Areas Category Area Switchyard, 1G31 DG and Day Tank Rooms, 1G21 DG and Day Tank Electrical Power Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room Coolant Supply Torus Room, Intake Structure, Pumphouse Containment Drywell, Torus Emergency NE, NW, SE Comer Rooms, HPCI Room, RCIC Room, RHR Valve Room, Systems North CRD Area, South CRD Area, CSTs Other Control Building, Remote Shutdown Panel 1C388 Area, Panel 1C55/56 Area, SBGT Room

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

5. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I HA3

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 22 of 33 HA4 Confirmed Security Event in a Plant Protected Area EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE HA4.1 Intrusion into the Plant Protected Area by a Hostile Force.

OR HA4.2 DAEC Security Supervision reports any of the following:

• Sabotage device discovered in the plant Protected Area.

• Standoff attack on the Plant Protected Area by a Hostile Force (i.e., sniper).

• Any of the following security events that persists for 30 minutes, or greater, affecting the Plant Protected Area:

o Credible bomb threats o Hostage/Extortion o Suspicious Fire or Explosion o Significant Security System Hardware Failure o Loss of Guard Post Contact DAEC EAL INFORMATION:

Security events are based upon the Site Security Plan.

This class of security events represents an escalated threat to plant safety above that contained in the NOUE. A confirmed INTRUSION report is satisfied if physical evidence indicates the presence of a HOSTILE FORCE within the PROTECTED AREA.

EAL 1 is an intrusion of a hostile force into the Protected Area representing a potential for a substantial degradation of the level of safety of the plant. A civil disturbance, which penetrates the Protected Area, can be considered a hostile force.

Reference is made to DAEC security supervision because these individuals are the designated personnel on-site qualified and trained to confirm that a security event is occurring or has HA4

EAL BASES DOCUMENT t EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 23 of 33 I

occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the plant Security Plan.

Security events represent an escalated threat to plant safety above that contained in the Unusual Event. 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. A security event is considered to be "of increasing severity" if events are NOT under control of the security force within 30 minutes. Intrusion into a vital area by a hostile force will escalate this event to a Site Area Emergency.

REFERENCES:

1. NMC Security Procedures I

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

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I HA4

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 24 of 33 I

HA5 Control Room Evacuation Has Been Initiated EVENT TYPE: Control Room Evacuation OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HA5.1 Entry into AOP 915 for control room evacuation. I DAEC EAL INFORMATION:

The applicable procedure for control room evacuation at DAEC is AOP 915.

Evacuation of the Control Room represents a potential for substantial degradation of the level of safety of the plant and therefore requires an ALERT declaration. Additional support, monitoring and direction is required and accomplished by activation of the Technical Support Center at the ALERT classification level. Inability to establish plant control from outside the Control Room will escalate this event to a Site Area Emergency. I

REFERENCES:

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

2. UFSAR Section 6.4, Habitability Systems

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I HA5

I EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 25 of 33 I

HA6 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert EVENT TYPE: Emergency Director Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HA6.1 Other conditions exist which in the judgment of the Emergency Director indicate that I events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

DAEC EAL INFORMATION:

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

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operations

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I HA6

EAL BASES DOCUMENT EBD-H

.. Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 26 of 33 HSI Confirmed Security Event in a Plant Vital Area EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HS1.1 Intrusion into a Safe ShutdownNital Area by a hostile force.

OR HS1.2 Security Supervision reports either of the following:

• A security event that results in the loss of control in a Safe ShutdownNital Area (other than the Control Room).

• A confirmed sabotage device discovered in a Safe ShutdownNital Area.

DAEC EAL INFORMATION:

Security events are based upon the Site Security Plan.

This class of security events represents an escalated threat to plant safety above that contained in the Alert IC in that a HOSTILE FORCE has progressed from the PROTECTED AREA to a Safe ShutdownNital Area.

Consideration should be given to the following types of events when evaluating an event against the criteria of the site specific Security Contingency Plan: SABOTAGE and HOSTAGE /

EXTORTION. The Safeguards Contingency Plan identifies numerous events/conditions that constitute a threat/compromise to a Station's security. Only those events that involve Actual or Likely Major failures of plant functions needed for protection of the public need to be considered.

The following events would not normally meet this requirement; (e.g., Failure by a Member of the Security Force to carry out an assigned/required duty, internal disturbances, loss/compromise of safeguards materials or strike actions).

Loss of Plant Control would escalate this event to a GENERAL EMERGENCY. Reference is made to Security Supervision because these individuals are the designated personnel on-site qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the plant Security Plan.

HS1

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 27 of 33 I

I I

REFERENCES:

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

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I HS1

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 28 of 33 HS2 Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established EVENT TYPE: Control Room Evacuation OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HS2.1 Control room evacuation has been initiated.

AND Control of the plant cannot be established per AOP 915 within 20 minutes.

DAEC EAL INFORMATION:

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. Control of the plant from outside the control room is assumed when the controls are transferred to remote shutdown panel 1C388 in accordance with AOP 915.

The Emergency Director is expected to make a reasonable, informedjudgment within the 20 minute time limit that control of the plant from the remote shutdown panel has been established.

The intent of the EAL is that control of important plant equipment and knowledge of important 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 by determining the duration of the core uncovery and the resulting peak cladding temperature (PCT). The PCT calculations were performed by incorporating the SAFE output into the Core Heatup Analysis code (CHASTE). The details of these calculations are provided in Section 4 of the final report for DAEC Appendix R analyses ("Safe Shutdown Appendix R Analyses for Duane Arnold Energy Center", MDE-44-036).

HS2

The required analyses include evaluation of the safe shutdown capability of the remote shutdown system for various control room fire events assuming: (1) no spurious operation of equipment, (2) spurious operation of a safety-relief valve (SRV) for 20 minutes, (3) spurious operation of a SRV for 10 minutes, and (4) spurious leakage from a one-inch line. The analyses show that the worst case spurious operation of SRV or isolation valves on a one-inch liquid line (high-low pressure interface) will not affect the safe shutdown ability of the remote shutdown system for DAEC in case of a fire requiring control room evacuation before the identified time limit for the necessary operator actions at the auxiliary shutdown panels. For the limiting cases of worst case spurious leakage from a one-inch line and spurious operation of a SRV, operator control within 20 minutes would not impact the integrity of the fuel clad, the reactor pressure vessel, and the primary containment.

REFERENCES:

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

2. General Electric Report MDE-44-0386, Safe Shutdown Appendix R Analysis for DAEC, March 1986

3. UFSAR Section 6.4, Habitability Systems

4. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels HS2

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 30 of 33 HS3 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of Site Area Emergency EVENT TYPE: Emergency Director Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HS3.1 Other conditions exist which in the Judgment of the Emergency Director indicate that I 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 Guideline exposure levels beyond the site boundary.

DAEC EAL INFORMATION:

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

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I HS3

EAL BASES DOCUMENT EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 31 of 33 HG1 Security Event Resulting in Loss Of Physical Control of the Facility EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HG1.1 A hostile force has taken control of plant equipment such that plant personnel are unable to operate equipment required to maintain safety functions as indicated by loss of physical control of either:

• A Safe ShutdownNital Area such that operation of equipment required for safe shutdown is lost OR

• Spent fuel pool cooling systems if imminent fuel damage is likely (e.g., freshly offloaded reactor core in the pool).

DAEC EAL INFORMATION:

This IC encompasses conditions under which a HOSTILE FORCE has taken physical control of a Safe ShutdownNital Area (containing vital equipment or controls of vital equipment) required to maintain safety functions and control of that equipment cannot be transferred to and operated from another location. Typically, these safety functions are reactivity control (ability to shut down the reactor and keep it shutdown) reactor water level (ability to cool the core), and decay heat removal (ability to maintain a heat sink) for a BWR.

If control of the plant equipment necessary to maintain safety functions can be transferred to another location, then the above initiating condition is not met.

This EAL should also address loss of physical control of spent fuel pool cooling systems if imminent fuel damage is likely (e.g., freshly off-loaded reactor core in pool). Loss of physical control of the control room or remote shutdown capability alone may not prevent the ability to maintain safety functions per se. Design of the remote shutdown capability and the location of the transfer switches should be taken into account.

REFERENCES:

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

2. UFSAR Section 6.4, Habitability Systems

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels HG1

EAL BASES DOCUMENT. - EBD-H Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 32 of 33 HG2 Other Conditions Existing Which in the Judgment of the Emergency I Director Warrant Declaration of General Emergency EVENT TYPE: Emergency Director Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

HG2.1 Other conditions exist which in the Judgment of the Emergency Director indicate that I 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 be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

DAEC EAL INFORMATION:

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

HG2

EAL BASES DOCUMENT. . EBD-H

. Rev. 9 HAZARDS & OTHER CONDITIONS AFFECTING PLANT SAFETY PAGE 33 of 33

REFERENCES:

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

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels HG2

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 1 of 26 Information Use I Effective Date:

TECHNICAL REVIEW Prepared by: Date:

Reviewed by: Date:

Independent Reviewer Reviewed by: Date:

Operations Reviewer PROCEDURE APPROVAL ]

I am responsible for the technical content of this procedure and for obtaining the necessary approval from the State and County Emergency Management officials prior to implementation.

Documentation of State and County Emergency Management approval is via NEP-Approved by: Date:

Manager, Emergency Planning

- EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 2 of 26 Table of Contents RU1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment That Exceeds Two Times the Offsite Dose Assessment Manual (ODAM) Limit and is Expected to Continue For 60 Minutes or Longer ................. 3 RU2 Unexpected Increase in Plant Radiation ............................................................... 8 RA1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200X the Offsite Dose Assessment Manual (ODAM) Limit and is Expected to Continue for 15 Minutes or Longer ................ 11 RA2 Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel ......... ........................ 15 RA3 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 .......... ........................ 18 RS1 Offsite 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 ........................................................ 20 RGI Offsite 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 Using Actual Meteorology ....... 23

EAL BASES DOCUMENT - EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 3 of 26 RUI Any Unplanned Release of Gaseous or Liquid Radioactivity to l the Environment That Exceeds Two Times the Offsite Dose Assessment Manual (ODAM) Limit and is Expected to Continue For 60 Minutes or Longer EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE: l RU1.1 Valid Reactor Building ventilation rad monitor (Kaman 314, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading that exceeds I E-3 pCi/cc and is expected to continue for 60 minutes or longer.

OR RU1.2 Valid Offgas Stack rad monitor (Kaman 9/10) reading that exceeds 2.0 E-1 l PCVcc and is expected to continue for 60 minutes or longer.

OR RU1.3 Valid LLRPSF rad monitor (Kaman 12) reading that exceeds 1.0 E-3 pCi/cc l and is expected to continue for 60 minutes or longer.

OR RU1.4 Valid GSW rad monitor (RIS-4767) reading that exceeds 3E+3 CPS and is l expected to continue for 60 minutes or longer.

OR RU1.5 Valid RHRSW & ESW rad monitor (RM-1997)reading that exceeds 8E+2 CPS l and is expected to continue for 60 minutes or longer.

OR RU1.6 Valid RHRSW & ESW Rupture Disc rad monitor (RM-4268) reading that exceeds 1E+3 CPS and is expected to continue for 60 minutes or longer.

OR RU1.7 Confirmed sample analyses for gaseous or liquid releases indicates l concentrations or release rates in excess of 2 times ODAM limit and is expected to continue for 60 minutes or longer.

RU1

-- EALBASES DOCUMENT - EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT I PAGE 4of 26 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.

UNPLANNED, as used in this context, includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g.,

minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. The Emergency Director should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has met or will likely exceed 60 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the Emergency Director should, in the absence of data to the contrary, assume that the release has exceeded 60 minutes.

The approach taken for calculation of gaseous radioactive effluent 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 EAL threshold values and in any case, are addressed by off-site dose assessment by MIDAS, which is the preferred method for determining this condition. The calculation methods for setpoint determination are from ODAM Section 3.4 and are based on Regulatory Guide 1.109 methodology. The table below lists the results of the gaseous effluent EAL calculations. The Kaman extended range capability is used because the General Electric Offgas Stack monitor has a limited range.

RU1

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT I PAGE 5of 26 GASEOUS EFFLUENT EALS Karna 9110 Turbine Bldg (Kaman 1/2) and Offgas Stack 9Reactor Bldg (Kaman 314, 5/6, 7/8)

Maximum flow (CFM) 10,000 72,000 ConcntrtionRelase Concentration Release Rate Release Limits ((Cpcc) Rate (gCi/cc) (lCi/sec)

Tech Spec 1.1E-1 5.2E+5 6.2E-4 2.1 E+4 Unusual Event (2 x TS) 2.OE-1 1.OE+6 1.2E-3 4.2E+4 Alert (60 x TS) J 6.OE+0 3.OE+7 3.7E-2 1.3E+6 LLRPSF Kaman 12 Maximum flow (CFM) 99,000 Release Limits Concentration (gC/cc) Release Rate (gCi/sec)

Tech Spec 5.9E-4 2.8E+4 Unusual Event (2 x TS) 1.OE-3 5.6E+4 Alert (200 x TS) 1.OE-1 5.6E+6 The off-gas stack is treated as an elevated release and the turbine building and reactor building vents are treated as mixed-mode releases. The ground level setpoints are taken from the default setpoint calculations from the quarterly surveillance tests performed by DAEC Chemistry technicians. Reactor Building, Turbine Building, LLRPSF (Low Level Radwaste Processing and Storage Facility) and Offgas Stack Noble Gas Monitor alarm setpoints are calculated based on achieving the Tech Spec/ODAM instantaneous release limit, assuming annual average meteorology as defined in the ODAM. The Tech Spec/ODAM 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 I E-3 pCi/cc. The corresponding offgas stack monitor value is 1.1 E-1 pCi/cc, rounded off to 1 E-1 pCVcc. The Tech Spec Limit currently for the LLRPSF building ventilation alarm setpoint is 5.9 E-04 pCi/cc. The DAEC EAL therefore addresses valid radiation levels exceeding 2 times the alarm setpoint for 60 minutes or greater. This corresponds to 1 E-3 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 liquid radwaste so that no release of radioactive liquid to the environment is allowed. The radwaste effluent 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. The other pathways to the environment (RHRSW -

RUI

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 6of 26 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 occurred in the service water systems at DAEC. These monitors are displayed on panels 1C02 and 1C1 0.

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 effluent discharge line (if such a release were to 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 ease of reading the monitor scales.

Monitor TS Limit Reading UE Level Alert Level GSW 1,555 CPS 1.5E+3 CPS 3E+3 CPS 3E+5 CPS RHRSW & ESW to cooling tower 413 CPS 4E+2 CPS 8E+2 CPS 8E+4 CPS RHRSW & ESW to Discharge Canal 507 CPS 5E+2 CPS I E+3 CPS I E+5 CPS DAEC does not have a telemnetered radiation monitoring system or an automatic real-timel dose assessment system. I RUI

- EAL BASES DOCUMENT - EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 7 of 26

REFERENCES:

1. Offsite Dose Assessment Manual

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

5. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 4, January 2003 RU1

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 8 of 26 .

RU2 Unexpected Increase in Plant Radiation EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

RU2.1 Unplanned valid Refuel Floor ARM reading increase with an uncontrolled loss of reactor cavity, fuel pool, or fuel transfer canal water level with all irradiated fuel assemblies remaining covered by water as indicated by any of the following:

• Report to control room

• Valid fuel pool level indication (LI-3413) LESS THAN 36 feet and lowering

• Valid WR GEMAC Floodup indication (LI-4541) coming on scale.

OR RU2.2 Any unplanned ARM reading offscale high or GREATER THAN 1000 times normal* reading.

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

DAEC EAL INFORMATION:

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

There are three methods to determine water level decreases of 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.

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 I C04 is placed in service by l&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 RU2

. . EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 9 of 26.

span of this flood up level indicator. A valid indication (e.g., not due to loss of compensating air signal or other instrument channel failure) of reactor 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 of inventory loss at DAEC include visual observation of decreasing water levels in reactor cavity or spent fuel storage pool, Reactor Building (RB) fuel storage pool radiation monitor or refueling area radiation monitor alarms, observation of a decreasing trend on the spent fuel pool water level indicator, and actuation of the spent fuel pool low water level alarm. To eliminate minor level perturbations from concern, DAEC uses L13413 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 1C02 and alarmed on Panel I C04B. The DAEC EAL has also been written to reflect the case where an ARM may go offscale high prior to reaching 1,000 times the normal reading.

NOTE: On Annunciator Panel 1C04B, 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.

RU2

EAL BASES DOCUMENT - EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 10 of 26l

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) Form TSC-40, ARM Locations

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

5. Surveillance Test Procedure (STP) 3.0.0.0-01 PA, Daily and Shift Instrument Checks

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

7. Core Alterations, RFP403, Procedure for Moving Core Components Between Reactor Core and Spent Fuel Pool, Within the Reactor Core, or Within the Spent Fuel Pool

8. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 4, January 2003.

RU2

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 11 of 26 RA1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200X the Offsite Dose Assessment Manual (ODAM) Limit and is Expected to Continue for 15 Minutes or Longer EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

RA1.1 Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading that exceeds 3 E-2 JCicc and is expected to continue for 15 minutes or longer.

OR RA1.2 Valid Offgas Stack rad monitor (Kaman 9/10) reading that exceeds 6 E+0 pCi/cc and is expected to continue for 15 minutes or longer.

OR RA1.3 Valid LLRPSF rad monitor (Kaman 12) reading that exceeds 1 E-1 itC icc and is expected to continue for 15 minutes or longer.

OR RA1.4 Valid GSW rad monitor (RIS-4767) reading that exceeds 3E+5 CPS and is expected to continue for 15 minutes or longer.

OR RA1.5 Valid RHRSW & ESW rad monitor (RM-1 997) reading that exceeds 8E+4 CPS and is expected to continue for 15 minutes or longer.

OR RA1.6 Valid RHRSW & ESW Rupture Disc rad monitor (RM-4268) reading that exceeds 1E+5 CPS and is expected to continue for 15 minutes or longer.

OR RA1.7 Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates with a release duration expected to continue for 15 minutes or longer in excess of 200 times ODAM limit.

RA1

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. In a case where data from Kaman readings is being used to determine whether an EAL threshold value has been exceeded, Valid means that flow through the associated Kaman Monitor has been verified and does exist as indicated in pCi/sec on SPRAD (Safety Parameter Display System (SPDS) screen).

UNPLANNED, as used in this context, includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g.,

minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has met or will likely exceed 15 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the Emergency Director should, in the absence of data to the contrary, assume that the release has exceeded 15 minutes.

GASEOUS EFFLUENT EALS OffgsKmanStck /10 Turbine Bldg (Kaman 112) and Offgas Stack Kaman 9110 Reactor Bldg (Kaman 314, 516, 718)

Maximum flow (CFM) 10,000 72,000 Concentration Release

.. Concentration Release Rate Release Limits ric cRate at~/c)(l~ec

____ __ _ __ __ _ __ __ _ __ __ _ (pCi/sec) (g~Ci/cc) (p~Cilsec)

Tech Spec 1.1E-1 5.2E+5 6.2E-4 2.1 E+4 Unusual Event (2 x TS) 2.0E-1 1 .OE+6 1.2E-3 4.2Ei+4 Alert (60 x TS) 6.OE+0 3.OE+7 3.7E-2 1.3E+6 LLRPSF Kaman 12 Maximum flow (CFM) 99,000 Release Limits Concentration (giCVcc) Release Rate (jiCi/sec)

Tech Spec 5.9E-4 2.8E+4 Unusual Event (2 x TS) 1.OE-3 5.6E+4 Alert (200 x TS) 1.OE-1 5.6E+6 The off-gas stack is treated as an elevated release and the turbine building and reactor building vents are treated as mixed-mode releases. The ground level setpoints are taken from the default setpoint calculations from the quarterly surveillance tests performed by DAEC Chemistry technicians. Reactor Building, Turbine Building, LLRPSF (Low Level Radwaste Processing and Storage Facility) and Offgas Stack Noble Gas Monitor alarm RAI

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 13 of 26.

setpoints are calculated based on achieving the Tech Spec instantaneous release limit assuming annual average meteorology as defined in the ODAM. The Tech Spec Limit currently corresponds to a reactor building or turbine building ventilation alarm setpoint of 6.2 E-4 j4Cicc. 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 RU1 through RA1 to RS1. The DAEC EAL therefore addresses valid l radiation levels exceeding 60 times the alarm setpoint for 15 minutes or longer. Rounded down, this corresponds to 3 E-2 ItCi/cc. The corresponding offgas stack monitor value is 6.6 pCi/cc, rounded down to 6 E+0 pCicc. The Tech Spec/ODAM Limit currently for the LLRPSF building ventilation alarm setpoint is 5.9 E-04 pCi/cc. The DAEC EAL therefore addresses valid radiation levels exceeding 200 times the alarm setpoint for 15 minutes or longer. This corresponds to 1 E-1 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 liquid radwaste so that no release of radioactive liquid to the environment is allowed. The radwaste effluent line which could be used as a batch release mechanism has a trip function that prevents exceeding the DAEC release limit, and therefore no EAL limits are provided. The other pathways to the environment (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 occurred in the service water systems at DAEC. These monitors are displayed on panels 1C02 and 1C1 0.

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 effluent discharge line (if such a release were to 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 ease of reading the monitor scales.

RA1

- -EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT I PAGE 14 of 26 Monitor - TS Limit Reading UE Level Alert Level GSW 1,555 CPS 1.5E+3 CPS 3E+3 CPS 3E+5 CPS RHRSW & ESWto cooling tower 413 CPS 4E+2 CPS 8E+2 CPS 8E+4 CPS RHRSW & ESWto Discharge Canal 507 CPS 5E+2 CPS IE+3 CPS IE+5 CPS DAEC does not have a telemetered radiation monitoring system or an automatic real-time dose assessment system.

REFERENCES:

1. Offsite Dose Assessment Manual Section 6.0, 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. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems

5. EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents

6. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 4, January 2003 RAI

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 15 of26l RA2 Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

RA2.1 Report of any of the following:

• Valid ARM Hi Rad alarm for the Refueling Floor North End (RM 9163),

Refueling Floor South End (RM 9164), New Fuel Storage (RM 9153), or Spent Fuel Storage Area (RM 9178).

• Valid Refueling Floor North End (RM-9163), Refueling Floor South End (RM-9164), or New Fuel Storage Area (RM-9153) ARM Reading GREATER THAN 10 mRem/hr

• Valid Spent Fuel Storage Area ARM (RM-9178) Reading GREATER THAN 100 mRem/hr OR RA2.2 Valid water level reading LESS THAN 450 inches as indicated on LI-4541 (floodup) for the Reactor Refueling Cavity that will result in Irradiated Fuel uncovering.

OR RA2.3 Valid Fuel Pool water level indication (LI-3413) LESS THAN 16 feet that will result in Irradiated Fuel uncovering.

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

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 1C04B, fuel pool ventilation exhaust monitors, and by Standby Gas Treatment (SBGT) System automatic RA2

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 16 of 26 start. Applicable area radiation monitors include RM-9163, RM-9164, RM-9153, and RM-9178. These monitors are located in the north end of the refuel floor, the south end of the refuel floor, the new fuel vault area, and near the spent fuel pool, respectively.

Per ARP 1C04B, the applicable area radiation monitor alarms actuate when radiation levels increase above 100 mRem/hr in the spent fuel pool area or above 10 mRemlhr in the other three areas of concern. 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 measured radiation levels in the spent fuel pool or north fuel area are used to address the generic concern consistent with DAEC design and procedures.

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 1C04 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 fuel pool level indicator Li 3413 is used to monitor refueling water level.

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

Therefore, postulated failures which drain the reactor cavity through the reactor vessel cannot uncover fuel in the spent fuel storage racks. However, valid indication of spent fuel pool level less than 16 feet would indicate that spent fuel in the storage racks may potentially become uncovered.

RFP403 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.

RA2

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 17 of 26 Following these actions, the refueling floor is to be evacuated of all personnel. The DAEC EAL is written to address the generic concern that a spent fuel assembly was not fully covered by water. This can either be by visual observation of an uncovered spent fuel assembly or by trending fuel pool level in the control room if a spent fuel assembly could not be placed in a safe storage location specified by RFP 403 as described above.

The level indication (LI3413) is also used for the Fuel Transfer Canal. Any loss of level resulting in the uncovering of irradiated fuel, regardless if in the Fuel Transfer Canal, Fuel Pool, or Reactor Refueling Cavity, will also cause the ARMs to alarm.

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) Form TSC-40 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. Core Alterations, RFP403, Procedure for Moving Core Components Between Reactor Core and Spent Fuel Pool, Within the Reactor Core, or Within the Spent Fuel Pool

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

9. Bechtel Drawing C-493, Reactor Building - Spent Fuel Liner Plan Elevations and Details, Sheet 1, Rev. 6

10. Holtec International Drawing No. 1045, Rack Construction - Spent Fuel Storage Racks, Rev. 3

11. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 4, January 2003 RA2

EAL BASES DOCUMENT - EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 18 of 26 RA3 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 EAL THRESHOLD VALUE:

RA3.1 Valid area radiation levels GREATER THAN 15 mRem/hr in any of the following areas:

• Control Room (RM 9162)

• Central Alarm Station (by survey)

• Secondary Alarm Station (by survey)

OR RA3.2 Valid area radiation monitor (RE-9168), reading GREATER THAN 500 mRem/hr affecting the Remote Shutdown Panel, 1C388.

DAEC EAL INFORMATION:

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

There are no significant deviations from the generic EALs. 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 Central Alarm Station (CAS) and Secondary Alarm Station (SAS) shall be included in this EAL as access to these areas must be maintained and the CAS & SAS are continuously occupied. The capability exists for plant shutdown from outside the main control room in the event that the control room becomes uninhabitable using remote shutdown panel 1C388. The RB 757 CRD North ARM, RE-9168 is in the vicinity of the Remote Shutdown Panel and is used to monitor radiation levels to determine habitability for that area.

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 RA3

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 19 of 26l momentary increases due to events such as resin transfers or controlled movement of radioactive sources result in emergency 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.

REFERENCES:

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

2. Abnormal Operating Procedure (AOP) 913, Fire

3. Abnormal Operating Procedure (AOP) 914, Security

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

5. Surveillance Test Procedure (STP) 3.0.0.0-01, Daily and Shift Instrument 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, Habitability Systems

9. Bechtel Calculation DA-4, Project Number 265-002, Control Room Habitability, 9/3/80

10. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 4, January 2003 RA3

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 20 of 26l RS1 Offsite 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 EAL THRESHOLD VALUE:

RS1.1 Dose assessment using actual meteorology indicates doses GREATER THAN 100 mRem TEDE or 500 mRem thyroid CDE at or beyond the site boundary. (Preferred method)

OR RS1.2 If Dose Assessment is unavailable, either of the following:

Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading GREATER THAN 6 E-2 lCi/cc and is expected to continue for 15 minutes or longer.

Valid Offgas Stack rad monitor (Kaman 9/10) reading GREATER THAN 4 E+1 pCi/cc and is expected to continue for 15 minutes or longer.

OR RS1.3 Field survey results indicate closed window dose rates exceeding 100 mRem/hr expected to continue for more than one hour at or beyond the site boundary; or analyses of field survey samples indicate thyroid CDE of 500 mRem for one hour of inhalation at or beyond the site boundary.

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. In a case where data from Kaman readings is being used to determine whether an EAL threshold value has been exceeded, Valid means that flow through the associated Kaman Monitor has been verified and does exist as indicated in piCVsec on SPRAD.

RS1

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT

. ~PAGE 21 of 26l The preferred method for declaration of RS1 is by means of Dose Assessment using the MIDAS computer model. However, if Kaman monitor readings are sustained for 15 minutes or longer and the required MIDAS dose assessments cannot be completed within this period, then the declaration can be made using Kaman readings PROVIDED the readings are not from an isolated flow path. If Kaman readings are not valid, field survey results may be utilized.

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

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 the CRD mix was used for releases via the turbine or reactor building vents. The source term for a release via the offgas stack is further impacted by the status of the standby gas treatment system. The status of that system was also taken into consideration.

Based on 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 (mph)

D 156feet 8-12 D 33 feet 8 - 12 E 156 feet 8 - 12 E 33 feet 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 1Cilcc 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.

RS1

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 22 of 26 Because the RS1 and RG1 KAMAN limits are to be based on a one-hour exposure, establishing concentration limits so these peak values match the NUMARC limits is acceptable.

Initiating Condition - lSite I AreaRS1I Emergency I General RGI Emergency Valid Turbine or Reactor Building ventilation rad monitor (KAMAN) reading, 0.06 pCi/cc 0.6 jCi/cc for 15 minutes or longer, above:

DAEC does not have a telemetered radiation monitoring system.

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 RU1.

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 of iodine'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 Threshold Value 1 is written in terms of TEDE and CDE Thyroid.

REFERENCES:

1. Offsite Dose Assessment Manual, Section 6.0, 6.1.2 and 7.1.2, Bases

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

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

4. Radiation Engineering Calculation No. 96-007-A, Determination of DAEC Radioactive Release Initiating Conditions for AS1 & AG1 Emergency Classifications, July 3, 1996

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

6. EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents

7. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 4, January 2003 RS1

EAL BASES DOCUMENT - EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 23of 26l RGI Offsite 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 Using Actual Meteorology EVENT TYPE: Offsite Rad Conditions OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

RG1.1 Dose assessment using actual meteorology indicates doses GREATER THAN 1000 mRem TEDE or 5000 mRem thyroid CDE at or beyond the site boundary. (Preferred method)

OR RG1.2 If Dose Assessment is unavailable, either of the following:

• Valid Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) or Turbine Building ventilation rad monitor (Kaman 1/2) reading GREATER THAN 6 E-1 pCi/cc and is expected to continue for 15 minutes or longer.

• Valid Offgas Stack rad monitor (Kaman 9/10) reading GREATER THAN 4 E+2 pCi/cc and is expected to continue for 15 minutes or longer.

OR RG1.3 Field survey results indicate closed window dose rates exceeding 1000 mRem/hr expected to continue for more than one hour at or beyond the site boundary; or analyses of field survey samples indicate thyroid CDE of 5000 mRem for one hour of inhalation at or beyond the site boundary.

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. In a case where data from Kaman readings is being used to determine whether an EAL threshold value has been exceeded, Valid means that flow RG1

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 24 of 26l through the associated Kaman Monitor has been verified and does exist as indicated in pCi/sec on SPRAD.

The preferred method for declaration of RG1 is by means of Dose Assessment using the MIDAS computer model. However, if Kaman monitor readings are sustained for 15 minutes or longer and the required MIDAS dose assessments cannot be completed within this period, then the declaration can be made using Kaman readings PROVIDED the readings are not from an isolated flow path. If Kaman readings are not valid, field survey results may be utilized.

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

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 the CRD mix was used for releases via the turbine or reactor building vents. The source term for a release via the offgas stack is further impacted by the status of the standby gas treatment system. The status of that system was also taken into consideration.

Based on 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 (mph)

D 156feet 8-12 D 33 feet 8 - 12 E 156 feet 8 - 12 E 33 feet 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 RGI

- EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 25 of 26 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 RS1 and RG1 Kaman limits are to be based on a one-hour exposure, establishing concentration limits so these peak values match the NUMARC limits is acceptable.

Condition Site Area Emergency General Emergency Initiating Cnion.I RSI RGI Valid Turbine or Reactor Building ventilation rad monitor (KAMAN) reading, 0.06 pCi/cc 0.6 pCi/cc for 15 minutes or longer, above:

Valid Offgas Stack ventilation rad monitor (Kaman) reading, for 15 minutes or 40 pCi/cc 400 pCi/cc longer, above: _

DAEC does not have a telemetered radiation monitoring system.

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 AU1.

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 of iodine'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 Threshold Value 3 is written in terms of TEDE and CDE Thyroid.

REFERENCES:

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

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

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

4. Radiation Engineering Calculation No. 96-007-A, Determination of DAEC Radioactive Release Initiating Conditions for AS1 & AG1 Emergency Classifications, July 3, 1996

5. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems RG1

EAL BASES DOCUMENT EBD-R Rev. 9 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT PAGE 26of 26l

6. EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents

7. NEI Methodology for Development of Emergency Action Levels NEI 99-01 Revision 4, January 2003 l RG1

EAL BASES DOCUMENT - EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 1 of 32 Usage Level Information Use l Effective Date:

TECHNICAL REVIEW Prepared by: _ Date:

Reviewed by: Date:

Independent Reviewer Reviewed by: _ Date:

Operations Reviewer PROCEDURE APPROVAL I am responsible for the technical content of this procedure and for obtaining the necessary approval from the State and County Emergency Management officials prior to implementation.

Documentation of State and County Emergency Management approval is via NEP-__

Approved by: Date:

Manager, Emergency Planning

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 2 of 32 Table of Contents SUI Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes.................................................................................................................4 SU2 Inability to Reach Required Shutdown Within Technical Specification Limits ........................................................... 5 SU3 Unplanned Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes ................................................ 6 SU4 Fuel Clad Degradation ............................................................ 8 SU5 RCS Leakage ........................................................... 10 SU6 Unplanned Loss of All Onsite or Offsite Communications Capabilities ............... 12 SU8 Inadvertent Criticality ........................................................... 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 ....... ............ 15 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 Unavailable ......................................... 17 SA5 AC Power Capability to Essential Busses Reduced to a Single Power Source for Greater Than 15 Minutes Such That Any Additional Single Failure Would Result in Station Blackout ........................................................... 199 SS1 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses................................................................................................................ 20 SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful ........... 22 SS3 Loss of All Vital DC Power ........................................................... 24 SS4 Complete Loss of Heat Removal Capability ....................................................... 26

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 3 of 32 SS6 Inability to Monitor a Significant Transient in Progress ....................................... 27l SG1 Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power to Essential Busses ............. .............................................. 29l 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 ......................................... 31

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY I Page 4 of 32 SUI Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D I EAL THRESHOLD VALUE:

SU1.1 Loss of all offsite power to Emergency Busses 1A3 and 1A4 is expected to last for greater than 15 minutes.

AND Emergency Busses 1A3 and 1A4 are powered by their respective Standby Diesel Generators. I DAEC EAL INFORMATION:

This event is a precursor of a more serious Station Blackout condition and is thus considered as a potential degradation of the level of safety of the plant. It is possible to be operating within Technical Specification LCO Action Statement time limits and make a declaration of an Unusual Event in accordance with this EAL.

The intent of this EAL is to declare an UNUSUAL EVENT when offsite power has been lost and both of the emergency diesel generators have successfully started and energized their respective ECCS bus.

REFERENCES:

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

2. UFSAR Section 8.2, Offsite Power System

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I Sul

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 5 of 32l SU2 Inability to Reach Required Shutdown Within Technical Specification Limits EVENT TYPE: Inability to Reach or Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

SU2.1 Plant is not brought to required operating mode within applicable Technical Specifications LCO Action Statement Time.

DAEC EAL INFORMATION:

Limiting Conditions for Operations (LCO) require the plant to be brought to a specific condition when an LCO has been entered. Depending on the circumstances this may or may not be an emergency or a precursor to a more serious event. In any case when a plant initiates a shutdown due to having entered an LCO action statement a one-hour report must be made under 10CFR50.72(b) non-emergency events. The plant is within its safety envelope when being shutdown within the allowable action statement time of a Technical Specification. An immediate classification of UNUSUAL EVENT should be made when the plant is NOT brought to the required mode within the allowable action statement time of any Technical Specification LCO. Declaration is based on the time at which the LCO Action Statement specified time period elapses and is NOT related to how long a condition may have existed.

REFERENCES:

1. DAEC Technical Specifications

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels SU2

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 6 of 32 SU3 Unplanned Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

SU3.1 Unplanned loss of most or all 1C03, 1C04 and 1C05 annunciators or indicators associated with Safety Systems for greater than 15 minutes.

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. 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 attendant and responsive to abnormal indications that relate to those systems and components that have lost annunciation.

This IC and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment.

Recognition of the availability of computer based indication equipment is considered (e.g., SPDS, plant computer, etc.).

Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions.

SU3

i EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 7 of 32 The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72. If the shutdown is not in compliance with the Technical Specification action, the NOUE is based on SU2 "Inability to Reach Required Shutdown Within Technical Specification Limits."

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

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

This NOUE will be escalated to an Alert if a transient is in progress during the loss of annunciation or indication.

Unplanned loss of critical safety function indicators (i.e., EOP/EAL parameters) for greater than 15 minutes may preclude operators from taking actions to mitigate a transient.

Annunciators on 1C03, 1C04, and 1C05 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1D1 3.

Indications of loss of annunciators associated with safety systems include:

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

• Loss of 'sealed in" annunciators at affected panels

• Failure of affected annunciator panels shiftily testing by plant operators

• Expected alarms are not received

• Computer point ID B350 indicates UNSS ANN DC LOSS 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

3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power

4. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels SU3

-EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY I Page 8 of 32 SU4 Fuel Clad Degradation EVENT TYPE: Fuel Clad Degradation OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

SU4.1 Pretreatment Offgas System (RM4104) Hi-Hi Radiation Alarm OR SU4.2 Reactor Coolant sample activity value GREATER THAN 2.0 PCilgm dose equivalent 1-131.

DAEC EAL INFORMATION:

This IC is included as a NOUE because it is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems.

EAL 1 - RM-4104 Hi-Hi Radiation Alarm has been chosen because it is operationally significant, is readily recognizable by the Control Room Operations Staff, and is set at a level corresponding to noble gas release rate, after 30-minute delay and decay of 1 Ci/sec. A Notification of Unusual Event is classified because the Offgas Pretreatment Hi-Hi radiation alarm is considered to be an indication of a potential degradation in the level of safety of the plant and a potential precursor of more serious problems.

EAL 2 - Coolant samples exceeding the short-term concentration permitted by Technical Specifications are representative of minor fuel cladding degradation. A Notification of Unusual Event is classified because Reactor Coolant Activity levels exceeding the maximum concentration in Technical Specification is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. Iodine equivalence is per Technical Specifications definition for Dose Equivalent 1-1 31.

Escalation of this IC to the Alert level is via the Fission Product Barrier Degradation Monitoring ICs.

SU4

EAL BASES DOCUMENT - EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 9 of 32 Though the referenced Technical Specification limits are mode dependent, it is appropriate that the EAL's be applicable in all modes, as they indicate a potential degradation in the level of safety of the plant.

The companion IC to SU4 for the Cold Shutdown/Refueling modes is CU5.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 672.2, Offgas Radiation/Reactor Coolant High Activity

2. Technical Specification 3.4.6, Coolant Chemistry

3. Annunciator Response Procedure (ARP) I C03A, Reactor and Containment Cooling and Isolation

4. PCP 8.6, Alarm Setpoints and Efficiency for OG Pretreatment

5. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels SU4

EAL BASES DOCUMENT - EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 10 of 32 SU5 RCS Leakage EVENT TYPE: Coolant Leakage OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot SID EAL THRESHOLD VALUE:

SU5.1 Unidentified or pressure boundary leakage GREATER THAN 10 gpm.

OR

.SU5.2 Identified leakage GREATER THAN 25 gpm.

DAEC EAL INFORMATION:

EAL Threshold Values 1 and 2 are precursors of more serious RCS barrier challenges and are thus considered as a potential degradation of the level of safety of the plant.

Thus, it is possible to be operating within Technical Specification LCO Action Statement time limits and make a declaration of an Unusual Event in accordance with these EALs.

Credit for the action statement time limit should only be given when leakage exceeds technical specification limits but has not yet exceeded the Unusual Event EAL thresholds described above.

The DAEC Tech Spec Section 3.4.4 coolant system leakage LCO limits are: (1) < 5 gpm unidentified leakage, (2) < 25 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) < 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.

DAEC EAL Threshold Value 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 room indications.

DAEC EAL Threshold Value 2 uses identified leakage set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.

SU5

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 11 of 32l

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) I C04C, Reactor Water Cleanup and Recirculation

6. UFSAR Section 5.2.5, Detection of Leakage through Reactor Coolant Pressure Boundary

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

8. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels SU5

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 12 of 32l SU6 Unplanned Loss of All Onsite or Offsite Communications Capabilities EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

SU6.1 Loss of ALL of the following onsite communication capabilities affecting the ability to perform routine operation:

• Plant Operations Radio System

• In-Plant Telephones

• Plant Paging System OR SU6.2 Loss of ALL of the following offsite communications capability:

• All telephone lines (commercial)

• Microwave Phone System

• FTS Phone System DAEC EAL INFORMATION:

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

The availability of one method of ordinary offsite communications is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (e.g., relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being utilized to make communications possible.

SU6

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY I Pg Page 13 of 32

REFERENCES:

1. Emergency Plan, Section F, Emergency Communications

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I SU6

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY I Page 14 of 32 SU8 Inadvertent Criticality EVENT TYPE: Inadvertent Criticality OPERATING MODE APPLICABILITY: Hot S/D EAL THRESHOLD VALUE:

SU8.1 An unplanned extended positive period observed on nuclear instrumentation.

DAEC EAL INFORMATION:

This IC addresses inadvertent criticality events. While the primary concern of this IC is criticality events that occur in Cold Shutdown or Refueling modes (NUREG 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States), the IC is applicable in other modes in which inadvertent criticalities are possible. This IC indicates a potential degradation of the level of safety of the plant, warranting a NOUE classification. This IC excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated). The Cold Shutdown/Refueling IC is CU8.

This condition can be identified using period monitors. The term "extended" is used in order to allow exclusion of expected short term positive periods from planned control rod movements for BWRs . These short term positive periods are the result of the increase in neutron population due to subcritical multiplication.

Escalation would be by the Fission Product Barrier Matrix, as appropriate to the operating mode at the time of the event, or by Emergency Director Judgment.

REFERENCES:

1. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels SU8

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 15 of 32 SA2 Failure of Reactor Protection System Instrumentation to I Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Power Operation, Startup EAL THRESHOLD VALUE:

SA2.1 Auto Scram failure AND ANY of the following operator actions to reduce power are successful in shutting down the reactor:

• Manual Scram Pushbuttons

• Mode Switch to Shutdown

• Alternate Rod Insertion (ARI)

DAEC EAL INFORMATION:

NOTE If the mode switch is in Startup and the rods are fully inserted (i.e., the reactor is shutdown) prior to the automatic signal failure, then declaration of an Alert would not be required.

In this case, the event would be reported under 10 CFR 50.72 (b) (2) as a four hour report.

This condition indicates 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 the plant safety has been compromised, and design limits of the fuel may have been SA2

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY I Page 16 of 32 exceeded. An Alert is indicated because conditions exist that lead to potential loss of fuel clad or RCS. Reactor protection system setpoint being exceeded, rather than limiting safety system setpoint being exceeded, is specified here because failure of the automatic protection system is the issue.

A manual scram is any set of actions by the reactor operator(s) at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical (e.g., reactor trip button, Alternate Rod Insertion). Failure of manual scram would escalate the event to a Site Area Emergency.

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. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I SA2

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY I Page 17 of 32 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 Unavailable EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot SID EAL THRESHOLD VALUE:

SA4.1 Unplanned loss of most or all I C03, 1C04 and 1C05 annunciators or indicators associated with Critical Safety Systems for greater than 15 minutes.

AND Either of the following conditions exist:

• A significant plant transient is in progress.

• 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 involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10% or greater.

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 attendant and responsive to abnormal indications that relate to those systems and components that have lost annunciation.

Therefore, the generic criterion related to specific opinion of the Operations Shift Manager SA4

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY I Page 18 of 32 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.

MOST - 75% of safety system annunciators or indicators are lost OR a significant risk that a degraded plant condition could go undetected exists. The use and definition of MOST is not intended to require a detailed count of lost annunciators or indicators but should be used as a guide to assess the ability to monitor the operation of the plant.

Unplanned loss of critical safety function indicators (i.e., EOP/EAL parameters) for greater than 15 minutes may preclude operators from taking actions to mitigate a transient.

Annunciators on 1C03,1 C04, and 1C05 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1DI 3.

Indications of loss of annunciators associated with safety systems include:

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

• Loss of 'sealed in" annunciators at affected panels

• Failure of affected annunciator panels shiftly testing by plant operators

• Expected alarms are not received

• Computer point ID B350 indicates UNSS ANN DC LOSS 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

3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power

4. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels SA4

- EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 19 of 32l SA5 AC Power Capability to Essential Busses Reduced to a Single Power Source for Greater Than 15 Minutes Such That Any Additional Single Failure Would Result in Station Blackout EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

SA5.1 AC power capability to 1A3 or 1A4 busses reduced to a single power source for greater than 15 minutes AND Any additional single failure will result in station blackout.

DAEC EAL INFORMATION:

The DAEC EAL is written to address the underlying concern, i.e., only one AC power source remains and if it is lost, a Station Blackout will occur. Under the conditions of concern, entry into AOP 301, Loss of Essential Electrical Power, would be made under Tab 1, Loss of One Essential 41 60V Bus, and/or under Tab 3, Loss of Offsite Power.

Indications/alarms related to degraded AC power are displayed on control room panel 1C08 and are listed in AOP 301 under "Probable Indications."

At DAEC, the Essential Busses of concern are 4160V Busses 1A3 and 1A4. Each of these busses feed their associated 480V and 120V AC busses through step down transformers. Onsite power sources at DAEC include the A and B Diesel Generators, I G-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

4. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels SA5

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 20 of 32 SSI Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

SS1.1 Loss of all offsite power to Emergency Busses 1A3 and 1A4 AND Failure of A Diesel Generator (1G-31) and B Diesel Generator (1G-21) to supply power to emergency busseslA3 and 1A4.

AND Failure to restore power to at least one emergency bus, 1A3 or 1A4, within 15 minutes from the time of loss of both offsite and onsite AC power.

DAEC EAL INFORMATION:

In accordance with the generic guidance, DAEC is using a threshold of 15 minutes for Station Blackout to exclude transient or momentary power losses.

Under the conditions of concern, entry into AOP 301.1, Station Blackout, would be made under Tab 1. Indications/alarms related to station blackout are displayed on control room panel 1C08 and are listed in the procedure under "Probable Indications."

Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to essential busses. Even though an essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not operable on the energized bus then the bus should not be considered operable. If this bus was the only energized bus, then a Site Area Emergency per SS1 should be declared.

SS1

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 21 of 32

REFERENCES:

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

2. Technical Specifications Section 3.8, Electrical Power Systems

3. UFSAR Chapter 8, Electric Power

4. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I SS1

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY I Page 22 of 32 SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful EVENT TYPE: RPS Failure OPERATING MODE APPLICABILITY: Power Operation, Startup EAL THRESHOLD VALUE:

SS2.1 Auto Scram failure AND NONE of the following operator actions to reduce power are successful in shutting down the reactor:

• Manual Scram Pushbuttons

• Mode Switch to Shutdown

• Alternate Rod Insertion (ARI)

DAEC EAL INFORMATION:

Automatic and manual scram are not considered successful if action away from the reactor control console was required to scram the reactor.

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

SS2

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 23 of 32 l

REFERENCES:

1. Integrated Plant Operating Instruction (IPOI) 5, Reactor Scram

2. ATWS Emergency Operating Procedure (EOP) - RPV Control

3. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels SS2

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY I Page 24 of 32 SS3 Loss of All Vital DC Power EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D EAL THRESHOLD VALUE:

SS3.1 Loss of Div 1 and Div 2 125V DC busses based on bus voltage LESS THAN 105 VDC Indicated for greater than 15 minutes.

DAEC EAL INFORMATION:

Under the conditions of concern, AOP 302.1, Loss 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 125V DC Systems ensure power is available for the reactor to be shutdown safely and maintained in a safe condition. The 125V System is divided into two independent divisions - Division I and Division II - with separate DC power supplies. These power supplies consist of two separate 125V batteries and chargers serving systems such as RCIC, RHR, EDGs, and HPCI. Complete loss of both 125V DC Divisions could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations.

Escalation to a General Emergency would occur by Abnormal Rad Levels/Radiological Effluent, Fission Product Barrier Degradation, or Emergency Director Judgment ICs.

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

SS3

- EAL BASES DOCUMENT - EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 25 of 32 l

REFERENCES:

I

1. Abnormal 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. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I SS3

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 26 of 32 SS4 Complete Loss of Heat Removal Capability EVENT TYPE: Inability to Reach or Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D. I EAL THRESHOLD VALUE:

SS4.1 EOP Graph 4 Heat Capacity Limit is exceeded. I DAEC EAL INFORMATION:

This EAL addresses complete loss of functions, including ultimate heat sink, required for hot shutdown with the reactor at pressure and temperature. Reactivity control is addressed in other EALs. The loss of heat removal function is indicated by the Heat Removal Capability Temperature Limit Curve being exceeded.

Under these conditions, there is an actual major failure of a system intended for protection of the public. Thus, declaration of a Site Area Emergency is warranted.

Escalation to General Emergency would be via Abnormal Rad Levels / Radiological Effluent, Emergency Director Judgment, or Fission Product Barrier Degradation ICs.

REFERENCES:

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

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

3. ATWS Emergency Operating Procedure (EOP) - RPV Control

4. Emergency Operating Procedure ALC - Alternate Level Control

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

6. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels I SS4

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 27 of 32 SS6 Inability to Monitor a Significant Transient in Progress I EVENT TYPE: Instrumentation/Communication OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D I EAL THRESHOLD VALUE:

SS6.1 Significant transient in progress and ALL of the following:

• Loss of most or all annunciators on Panels 1003, 1C04 and 1C05.

• Compensatory non-alarming indications are unavailable.

• Indicators needed to monitor criticality, or core heat removal, or Fission Product Barrier status are unavailable.

DAEC EAL INFORMATION:

"Planned" and "UNPLANNED" actions are not differentiated since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not a mitigating factor.

The DAEC EAL is written in terms of a significant transient in progress with loss of both safety system annunciators and loss of compensatory non-alarming instrumentation as well as loss of indications needed to monitor criticality, core heat removal, or fission product barrier status.

Significant 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 10%.

Compensatory non-alarming indications include the plant process computer, SPDS, plant recorders, or plant instrument displays in the control room. These indications are needed to monitor safety functions that are of concern in the generic EAL. I SS6

- EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 28 of 32 Control room panels 1C03, 1C04, and 1C05 contain the annunciators, and indicators, associated with safety systems at DAEC. Annunciators on 1C03, 1C04, and 1C05 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1D1 3.

MOST - 75% of safety system annunciators or indicators are lost OR a significant risk that a degraded plant condition could go undetected exists. The use and definition of MOST is not intended to required a detailed count of lost annunciators or indicators but should be used as a guide to assess the ability to monitor the operation of the plant.

Indications of loss of annunciators associated with safety systems include:

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

• Loss of "sealed in" annunciators at affected panels

• Failure of affected annunciator panels shiftly testing by plant operators

• Expected alarms are not received

• Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to I panels 1C03, IC04, and 1CO5)

REFERENCES:

1. Operating Instruction (01) No. 317.2, Annunciator System

2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power

3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power

4. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels SS6

EAL BASES DOCUMENT - EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 29 of 32 _

SGI Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power to Essential Busses I EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Power Operation, Startup, Hot S/D I EAL THRESHOLD VALUE:

SGI.1 Loss of all offsite power to Emergency Busses 1A3 and 1A4 AND Failure of A Diesel Generator (IG-31) and B Diesel Generator (1G-21) to supply power to emergency busses 1A3 and 1A4.

AND ANY ONE OF THE FOLLOWING:

• Restoration of power to either Bus 1A3 or 1A4 is not likely within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

• RPV level is indeterminate.

• RPV level is below +15 inches.

DAEC EAL INFORMATION:

Under prolonged Station Blackout (SBO) conditions, fission product barrier monitoring I capability may be degraded. Although it may be difficult to predict when power can be restored, it is necessary to give the Emergency Director a reasonable idea of how quickly I a General Emergency should be declared based on the following considerations:

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

SG1

EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY Page 30 of 32l The first part of this EAL corresponds to the threshold conditions for Initiating Condition SS1 - 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 cooling that continues to make RPV water level measurements unreliable, or (3) indications of inadequate core cooling. Each of these conditions is discussed below:

1. SBO Coping Capabilitv Exceeded DAEC has a SBO coping duration of four hours. The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the 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 public protective actions.

2. RPV Water Level Measurements Remaining Unreliable Flashing of the reference leg water will result in erroneously high RPV water level readings giving a false indication of actual water inventory and potentially indicating adequate core cooling when it may not exist. EOP Graph 1, RPV Saturation Temperature, defines the conditions under which RPV level instrument leg boiling may occur.

3. Indications of Inadequate Core Cooling DAEC uses the RPV level that is used for the Fuel Clad "potential loss" condition in the Fission Product Barrier Matrix. 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 uStation Blackout Evaluation Iowa Electric Light and Power Company Duane Arnold Energy Center," February 10, 1992

3. Emergency Operating Procedure (EOP)1 - RPV Control

4. Emergency Operating Procedure (EOP) ALC - Alternate Level Control

5. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels SGI

-EAL- BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY I Page 31 of 32 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: Power Operation, Startup EAL THRESHOLD VALUE:

SG2.1 Auto Scram failure AND NONE of the following operator actions to reduce power are successful in shutting down the reactor:

• Manual Scram Pushbuttons Mode Switch to Shutdown Alternate Rod Insertion (ARI)

AND Loss of adequate core cooling or decay heat removal capability as indicated by either:

• RPV level cannot be maintained GREATER THAN -25 inches.

• HCL Curve (EOP Graph 4) exceeded.

DAEC EAL INFORMATION:

This EAL addresses conditions where failure of an automatic scram has occurred and manual actions performed at the Reactor Control Console to reduce reactor power have been unsuccessful AND a subsequent loss of adequate core cooling or decay heat removal capability occurs. If either of these challenges exists during an ATWS, a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier matrix declaration to permit maximum offsite intervention time.

SG2

-EAL BASES DOCUMENT EBD-S Rev. 7 SYSTEM MALFUNCTION CATEGORY I Page 32 of 32 If injection with all available Preferred and Alternate ATWS Injection Systems fails to provide sufficient injection to restore and maintain level above -25 inches (Minimum Steam Cooling RPV Water Level), adequate core cooling is threatened and submergence of the core is attempted by flooding the primary containment. This is accomplished by transfer to and implementation of the DAEC Severe Accident Guidelines (SAGs).

The Heat Capacity Limit (EOP Graph 4) is defined to be the highest torus temperature at which initiation of RPV depressurization will not result in exceeding the Primary Containment Pressure Limit (the PCPL is 53 psig at the DAEC) before the rate of energy transfer from the RPV to the primary containment is within the capacity of the containment vent.

Control of torus temperature relative to the Heat Capacity Limit is directed in the Primary Containment Control Guideline, EOP 2. If the actions being taken in EOP 2 to preserve torus heat capacity are inadequate or not effective, RPV pressure must be reduced in order to remain below the Heat Capacity Limit. Therefore, actions in the RPV pressure control section of the ATWS EOP must accommodate these requirements. Failure to do so may lead to failure of the containment or loss of equipment necessary for the safe shutdown of the plant.

REFERENCES:

1. Emergency Operating Procedure ATWS EOP - RPV Control

2. NEI 99-01 Rev. 4, NEI Methodology for Development of Emergency Action Levels SG2

ATTACHMENT 2 CLEAN TECHNICAL BASIS DOCUMENT 183 pages follow

Standby Trans.

34.514.16KV 1X4 Start-up Trans.

161/4.16KV 1X3

/ Y" 320 1G21 1)

II1B21 I B91 lX31 141 DAEC 4160 VAC ESSENTIAL POWER DISTRIBUTION Information Only

I i [

I

-sol I

iERE:sE l I

I'sI e2

. ~ .- " . 7r n f.:<: ..-- .. .,. j.c'xv W. . . . .. . .. . -. - , .- .,. - ..:. . I .....

eto

• DAEC is submitting to the NRC, with your approval, a revision to the Emergency Action Levels (EALs) based upon implementation of NEI 99-01 Revision 4 EALs.

• This presentation is intended to convey the changes being made to our EALs and to obtain State and County approval of these new EALs for use at DAEC.

• The NRC has formally endorsed NEI 99-01 Revision 4 Emergency Action Levels (EALs) via RIS 2003-18, Supplement 1, dated 7/13/04.

• The NMC has committed to implement these EALs, as a fleet, as soon as reasonably possible. All NMC sites will be submitting these EALs to the NRC for approval.

• NRC approval could take anywhere from 6 to 18-months.

DAEC implementation of these EALs will occur within a few weeks of receiving NRC approval.

• *.- a- Vc -r r

.. - a. ' -,, . vor a r? Ag - -* . ., W..z..7CEn- :Y

• Please refer to the EAL Matrix.

I

y--q e..s ti. s -o

" coe - - n

• Any questions or comments?

• Thank you very much for your time!

Duane Arnold Energy Center N 3313 DAEC Road Commited to Nuclear Exte~nc Palo, L4 52324 Operatedby NuclearManagement Company LLC Memorandum DATE: 10110104

SUBJECT:

EAL Submittal: Containment Pressure Indicator Justification TO: File FROM: Don A. Johnson FILE: A221 / NEP 2004 - 0034 NEI 99-01, revision 4, states that the Containment Pressure Indicator value for EAL CA4.3 should be "The RCS pressure setpoint chosen should be lOpsig or the lowestpressure that the site can read on installedControlBoard instrumentationthat is equal to or greaterthan I Opsig.

For DAEC, the LR Reactor Pressure Indicators, PI 4590A and PI 4590B, are capable of reading 10 psig.

PI 4590 A/B MEERD MTEN ME.

LR Reactor Pressure 1I l0l 0-250 psig IGADUATIONS OR 10 si S lNOR GRADU-ATIONS' 5 psi This memo to file is intended to support the setpoint of 10 psig for EAL CA4.3.

Page 1 of I

Coannlt to Nucar Excls To: David L Miller, Iowa Homeland Security & Emergency Management Division Don Flater, Iowa Department of Public Health Scott Hansen, Benton County Emergency Management Agency Ned Wright Linn County Emergency Management Agency From: Don A. Johnson Date: 9/812004 Re: Revision to the EAL Basis Document and EAL Tables File: A-221b NEP-2004 0030 The Duane Amold Energy Center (DAEC) proposes modifying areas within the Emergency Action Level (EAL) Bases Document (EBD) and EAL Tables (EPIP Forms).

The revisions are in order to incorporate industry 'lessons leamed' given in revision 4 of NEI 99-01 and as endorsed by the NRC via RIS 2003-18. These changes are not philosophically different than those already in place. These changes merely serve to ensure DAEC Staff have dear, concise, and regulatory endorsed guidance when determining EAL's at DAEC. A Training Packet has been drafted detailing the existing EAL and the new proposed EAL DAEC respectfully asks for State & County approval of these EAL's prior to submitting these EAL's to the NRC as DAEC desires to ensure that the key stakeholders in the Iowa Emergency Preparedness Program have an understanding of these EAL's and approve their use at DAEC.

If you have any questions or desire additional information regarding this matter, please contact Paul Sullivan, Manager, Emergency Planning at (319) 851-7191 or Don A. Johnson at (319) 851-7872.

Sincerely, J Don A. JohnsonC (319)851-7872

Comtd toN arfe To: Paul R.Sullivan From: Ned Wright, inn County Emergency Management Agency Re: EAL Changes as Referenced in NEP 2004-0030 File: A-221a I concur with the changes to the EAL's inorder to incorporate NEI 99-01 Revision 4 EAL's. The Emergency Preparedness staff at DAEC has effectively answered any question I had regarding these changes.

114i'. / - C F S Name Agency It d 4 Date

Caulftdro NMC] aExcefe To: Paul R.Sullivan From: Scott Hansen, Benton County Emergency Management Agency Re: EAL Changes as Referenced in NEP 2004-0030 File: A-221a I concur with the changes to the EAL's in order to incorporate NEI 99-01 Revision 4 EAL's. The Emergency Preparedness staff at DAEC has effectively answered any question I had regarding these changes.

Name Rat___ s E Agency

/'q/l/II&

C4tnnftte fo NuLeAr Eca c~

To: Paul R. Sullivan From: Don Flater, Iowa Department of Public Health Re: EAL Changes as Referenced in NEP 2004-0030 File: A-221a I concur with the changes to the EAL's in order to incorporate NEI 99-01 Revision 4 EAL's. The Emergency Prepatedness staff at DAEC has effectively answered any question I had regarding these changes.

,U AAWA 1-2r /Wz vd, Name Agency

/'O/i/o /

Date

omfftd to NudoWr£ To: Paul R. Sulrivan.

From: David L MWer, Iowa Homeland Security &Emergency Management Division Re: EAL Changes as Referenced in NEP 2004-0030 File; A-221a I concurwith the changes to the EAL's Inorder to Incorporate NEI 99-01 RevisIon 4 EALs. The Emergency Preparedness staff at DAEC has effectively answered any question I had regarding these changes.

Agencyal,_z V Dat

W ENGINEERING CALCULATION COVER SHEET Duane Arnold Energy Center Engineering Calculation No. CAL-R04-002 Rev. 1 Radiation Exposure Rates at Spent Fuel Pool ARM RE9178 Following Gap Release with the Reactor Cavity Flooded and DEI = 2 micro-Ci/gm.

Radiological Engineering Calculation No. 04-004-A Rev. 1 August 27th, 2004 Reference Documents: AR DDC ECP/EMA Other NEI 99-01 REV. 4 Method of Verification: EJ" Design Review El Alternate Calculation n Qualification Testing 43 2 h, 1

" n A) Is JI 1 7z7-1-- e8I27/2004 _ _ _ _w__ _ 37_I

/_7___

Revision. Prepared/Date Vaified/Date Approved/ ate NG-007Z Rev 8 Rad. Engineering Calc. 04-004-A Page 1 of 10 Engineering Calc. CAL-R04-OOZ Revision 1

1. Table of Contents I. TABLE OF CONTENTS 2 II. PURPOSE: 3 III. OBJECTIVE: 3 IV. ASSUMPTIONS: 3 V. DESIGN INPUTS AND

REFERENCES:

3 VI. METHODOLOGY 4 VII. CALCULATION 5 A. Determine Iodine weighting factors. 5 B. Determine the Ratio to apply to the inventory to obtain 2 uCi/ml DEI. 6 C. Determine Isotopic Concentrations of Gap Release in the Water 7 D. Geometry 7 E. Microshield Case Results 8 Vill.

SUMMARY

AND CONCLUSIONS 8 IX. ATTACHMENT 1: MICROSHIELD CASE RESULTS: T = 8 HOURS 9 X. ATTACHMENT 2: MICROSHIELD CASE RESULTS: T =30 DAYS 10 Rad. Engineering Cabc. 04-004-A Page 2 of 10 Engineering Caic. CAL-R04-OOZ Revision 1

II. PURPOSE:

This calculation is being performed to investigate the basis for a new plant Emergency Action Level (EAL). NEI guidance document NEI 99-01 Rev 4. recommends that plant personnel have the ability to identify when reactor coolant Dose Equivalent Iodine (DEI) exceeds 2 tiCVgm. This identification should be from indications from installed plant radiation monitors. The EAL (CU5) is for plant modes 4 and 5. This calculation may be used to provide a basis for this indication from a plant Area Radiation Monitor (ARM) It would be applicable for plant mode 5 with the reactor cavity flooded up.

III. OBJECTIVE:

Calculate the expected gamma dose rates at Area Radiation Monitors (ARMs) on the refuel floor during refueling operations following a gap release accident where Dose Equivalent Iodine (DEL) reaches 2 uci/gram.

IV. ASSUMPTIONS:

The postulated source term is a gap release. (Equivalent to that seen during the postulated Control Rod Drop accident. (CRDA).

Additional details on assumptions of the CRDA are included in reference 4.

Immediate and complete mixing of radioactive material from the gap release is assumed throughout the vessel and reactor cavity Because the Tech Spec limit of 2 pCi/gm is based on a dose equivalent iodine concentration in the vessel, it is not necessary to determine the actual amount of material released into the water. If the relative amounts released into the water are known, these can be scaled appropriately to provide a water concentration for each isotope of interest.

Because a typical time to cold shutdown is 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> and recent refueling outages have lasted less than 30 days, the calculations were limited to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 30 days post shutdown. These two time periods adequately bound the conditions.

For reactor water, Icc = I gram = lml.

V. DESIGN INPUTS AND

REFERENCES:

1. Regulatory Guide 1.183, "Alternate Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Reactors", July, 2000.

2. General Electric Document 22A2703C, "Radiation Sources", Revision 2.

3. GE Project Task Report T0802 for the Duane Arnold Energy Center Asset Enhancement Program, "Radiation Sources and Fission Products", GE-NE-A22-00100-58-01, RO, Class m, dated May 2000.

4. CALROO-PUP-008, Rev.l, Non-LOCA Radiological Consequence Dose with Alternate Source Term.

5. Microsoft Excel, Version 9.0.4402 SR-I

6. Microshield Software Version 5.03-00278, Gamma Ray Point Kernel Shielding Code, Grove Engineering

7. Plant Drawings BECH-MOO5 Rad. Engineering Caic. 04-004-A Page 3 of 10 Engineering Caic. CAL-R04-OOZ Revision 1

8. Federal Guidance Report No. 11, Limiting Values ofRadionuclideIntake And Air Concentration andDose Conversion Factorsfor Inhalation,Submersion, A4nd Ingestion, US EPA, September 1988 VI. METHODOLOGY

1. The core fission product inventory and is taken from Reference 3, Appendix B, Table 3.

2. Per Reference 1, Table 3, the fraction of equilibrium core inventory assumed to be in the gap for the various radionuclides is as follows:

Fraction of Core Inventory in Gap 1-131 0.08 Kr-85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals (Cs, Rb) 0.12

3. To define the amount of radioactive material available for release to the reactor coolant, the above ratios are applied to the core fission product inventory.

4. Weighting factors for each of the iodines that are included in the DEI are determined.

5. Because the objective of the calculation is to determine a dose rate at 2 uci/gm DE1, a conversion factor can be applied to the quantity of rad. material available for release.

6. The geometry of the area radiation monitors on the refuel floor is determined from a walk-down and plant drawing BECH-MOO5.

7. With the geometry and the source concentrations known, Microshield cases are run.

8. A time frame between 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 30 days following reactor shutdown realistically bounds the potential for exceeding 210Ci/ml DEL.

Rad. Engineering Caic. 04-004-A Page 4 of 10 Engineering Catc. CAL-Ro4-OOZ Revision I

V1I. CALCULATION A. Determine Iodine weighting factors.

Exposure to dose factors are taken from FGR I1, (Reference 8). Weighting factors for Dose Equivalent Iodine are calculated with the equation:

DEI= Z Qm.x DCFJ,.

hEDCFe13 where:

DEI = Dose Equivalent Iodine 131 (Ci)

Qgil = Quantity of Iodine isotope "n*(Ci)

DCFn,, = Dose Conversion Factor for iodine isotope 'n' (Sv/Bq)

DCF,,31= Dose Conversion Factor for 1-131 (SvIBq).

An Excel Spreadsheet is used:

FGR Exposure-to-dose DEl Weighting Factor(Sv[Bq) Factor 1131 2.92E-07 1.OOE+00 1132 1.74E-09 5.96E-03 1133 4.86E-08 1.66E-01 1134 2.88E-10 9.86E-04 1135 8.46E-09 2.90E-02 Rad. Engineering Calc. 04-004-A Page 5 of 10 Engineering Calc. CAL-R04-OOZ Revision 1

B. Determine the Ratio to apply to the Inventory to obtain 2 ucilml DEI.

Because we are looking for a Dose Equivalent Iodine concentration of 2 tCtlml, the core inventory can be multiplied by a conversion factor to obtain the desired DEL.

Reg Guide 1.183 Gap Equivalence Factor to Release 8 Hours Weighting 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> convert to Fraction CIIMWth Available Factors 1131 DEI DEI= 2 1131 8.000E-02 2.691E+04 2.153E+03 1 .00E+00 2.15E+03 1.69E+00 1132 5.000E-02 3.721E+04 1.861E+03 5.96E-03 1.11E+01 8.68E-03 1133 5.000E-02 4.308E+04 2.154E+03 1.66E-01 3.59E+02 212.55E403: 2.81E-01 I134 5.000E-02 3.575E+02 1.788E+01 9.86E-04 1.76E-02 7.83E-04 1.38E-05 1135 5.000E-02 2.227E+04 1.114E+03 2.90E-02 3.23E+01 2.53E-02 2.55E+03 2.00E+00 uci/cc 8 Hours Ci/MWth Concentration 1131 2.691E+04 2.11 E+01 1132 3.721E+04 2.91 E+01 I133 4.308E+04 3.37E+01 I134 3.575E+02 2.80E-01 I135 2.227E+04 1.74E+01 Reg Guide 1.183 Gap Equivalence Factor to Release 30 Day Weighting 30 Days convert to Fraction CiUMWth Available Factors 1131 DEI DEi= 2 I131 8.000B-02 2.134E+03 1.707E+02 1.00E+00 1.71 E+02 2.OOE+00 I132 5.000E-02 6.763E+01 3.382E+00 5.96E-03 2.02E-02 1.58E-05 I133 5.OOOE-02 2.137E-06 1.069E-07 1.66E-01 1.78E-08 21171E+02 1.39E-1 I I134 5.000E-02 0.000E+00 O.OOOE+00 9.86E-04 O.OOE+00 1.17E-02 O.OOE+00 1135 5.000E-02 O.OOOE+00 O.OOOE+00 2.90E-02 O.OOE+00 O.OOE+00 1.71 E+02 2.00E+00 ucidcc 30 Day CUMWth Concentration I131 2.134E+03 2.50E+01 1132 6.763E+01 7.92E-01 I133 2.140E-06 2.51 E-08 I134 O.OOOE+00 O.OOE+00 I135 0.000E+00 O.OOE+OO Rad. Engineering Calc. 04-004-A Page 6 of 10 Engineering Calc. CAL-R04-OOZ Revision I

C. Determine Isotopic Concentrations of Gap Release in the Water As stated in section VI, only a fraction of the radioactive material in the fuel is available for release.

Applying the applicable fractions and the previously determined ratios to the gap source inventory, the source concentrations are determined:

Relative Isotopic Relative Core Inventories Rcg Guide composition for GAP Concentration Assuming 1.183 Gap 2uCifgm DEI (mCl/g)

(CUMWT) Release (Ci)

Release 8.0HR 30.0D Fraction E.OHR 30.0D 8.0HR 30.OD BR 82 2.1128+02 1.800E-04 0.05 1.06E-01 9.00E-06 BR 82 0.00829 1.05E-07 BR 83 3.414E+02 O.OOOE+00 0.05 1.71E+01 O.OOE00 BR 83 0.0134 0.000 BR 84 1.756E-01 O.OOOE+00 0.05 8.78E-03 O.OOE+00 BR 84 6.87E-06 0.000 0.00 CS132 1.011E+01 4.219E-01 0.12 1.21E+00 5.06E-02 CS132 9.50E-04 5.93E-04 CS134 1.065E+04 1.036E+04 0.12 1.28E+03 1.24E.03 CS134 1.00E+00 14.562 CS134M 3 .301E802 O.OOOE+O0 0.12 3.96E+01 O.OOE+00 CS134M 0.0310 0.000 CS135 2.919E-02 2.921E-02 0.12 3.SOE-03 3.51E-03 CS135 2.74E-06 4.11E-05 CS136 2.912E+03 6.062E+02 0.12 3.49E+02 7.27R801 CS136 0.274 0.852 CS137 5.233E803 5.223E+03 0.12 6.28E-02 6.27E#02 CS137 0.492 7.342 CS138 2.774E+00 O.OOOE+00 0.12 3.33E-O1 O.OOE+00 CS138 2.61E-04 0.000 9.690E+02 4 .436E-15 0.05 4.85E+01 2.22E-16 I130 0.0379 2.60E-18 I131 2.6912+04 2.134E*03 0.08 2.15E+03 1.71E+02 I131 1.685 2.000 I132 3 .721E+04 6.763E801 0.05 1.86E+03 3.38E+00 1132 1.457 0.0396 1133 4.308E+04 2 .137E-06 0.05 2. 15E.03 1.07E-07 1133 1.686 1.25E-09 1134 3.575t+02 O.OOOE+00 0.05 1.79E+01 0.00+.00 1134 0.0140 0.000 1135 2 .227E804 O.OOOE00 0.05 1.11E+03 O.OOE+00 1135 0.872 0.000 KR 83M 8.937E802 O.OOOE+00 0.05 4.47E+01 O.OOE.00 KR 83M 0.0350 0.000 KR 85 4.501E+02 4.478E+02 0.10 4.50E+01 4.48E+01 KR 85 0.0352 0.525 KR 87M 1.9681.03 O.OOOE00 O.OS 9.84E-01 O.OOE+00 KR 85M 0.0770 0.000 KR 87 1.645E+02 O.OOOE+00 O.OS 8.23C+00 O.OOE+00 KR 87 0 .00644 0.000 KR 88 2 .5441.03 O.OOOE+00 0.05 1.27E+02 0.003+00 KR 88 0.0996 0.000 0.00 RB 86 9.754E+01 3.240E.01 0.12 1.17E+01 3.89E+00 RB 86 0.00916 0.0455 RB 88 2.840E+03 O.OOOE+00 0.12 3.41E+02 O.OOE+00 RB 88 0.267 0.000 XE129M 4.136E-01 3.164E-02 0.05 2.07E-02 1.588E-03 KE123M 1.62E-05 1.85E-05 XE131M 3 .091E802 1.185E+02 0.05 1.55E+01 5.93E+00 XE131M 0.0121 0. 0694 XE133 5.265E+04 1.241E+03 0.05 2.63E+03 6.21E+01 XE133 2.061 0.727 XE133M 1.702E+03 2.103E-01 0.05 8.518E01 1.OSE-02 XE133M 0.067 1.23E-04 XE135 2 .574E+04 2.235E-19 0.05 1.298+03 1.12E-20 XE135 1.01 1.31E-22 D. Geometry The geometry of the reactor cavity and nearby area radiation monitors on the refuel is determined from a walk-down of the refueling area and from plant drawing BECH-MOO5. The Geometry is used in Microshield cases.

1. The Refuel Floor Spent Fuel Pool Area Radiation Monitor, RE9178 is approximately 10' from the edge of the Reactor Cavity. It is approximately 3' off the floor. It is described in the Microshield cases as Dose Point 1.

2. The Refuel Floor New Fuel Vault Area Radiation Monitor, RE9153 is approximately 18' from the edge of the Reactor Cavity. It is approximately 5' off the floor. It is described in the Microshield cases as Dose Point 2.

3. The radius of the cavity is approximately 16 '2' and is modeled at 22' deep.

Rad. Engineering Calc. 04-004-A Page 7 of 10 Engineering Calc. CAL-R04-OOZ Revision I

E. Microshield Case Results With source activities and geometry described, Microshield cases are run. Case output is attached as attachments I and 2. Case Results are summarized here:

File Name Description mR/hr cavity8.ms5 Fuel Pool ARM (RE 9178) dose rate w/ 236 Cavity flooded and DEI=2uci/cc.

(dose point 1) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after shutdown.

cavity8.ms5 New Fuel Vault ARM (RE 9153) dose rate w/ 142 Cavity flooded and DEI=2uci/cc.

(dose point 2) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after shutdown.

cavity30.rns5 Fuel Pool ARM (RT9178) dose rate wI 634 Cavity flooded and DEI-=2uci/cc.

(dose point 1) 30 Days after shutdown.

cavity3O.ms5 New Fuel Vault ARM (RE 9153) dose rate wI 383 Cavity flooded and DEI=2uci/cc.

(dose point 2) 30 Days after shutdown.

VIII.

SUMMARY

AND CONCLUSIONS When considering a RCS Dose Equivalent Iodine concentration of 2 pCi/gm of reactor coolant, the dose rates surrounding the flooded reactor cavity will increase over time. This is due to the fact that the iodine in the reactor coolant decays at a faster rate than other significant gamma emitters.

This calculation shows that a valid, sustained dose rate greater thanl40 mRlhr at plant Area Radiation Monitor RE9178 when the fuel pool is flooded up may be indicative of dose equivalent Iodine concentrations of at or near the Tech. Spec. limit of 2 gtCi/gram.

Rad. Engineering Caic. 04-004-A Page 8 of 10 Engineering Cabc. CAL-R04-OOZ Revision 1

IX. ATTACHMENT 1: Microshield Case Results: t = 8 Hours Attachment 1 to Engineering Calculation No. CAL-R04-002 CAVITY8.MS5 Attachment Rev. 1 Radiation Exposure Rates at Spent Fuel Pool ARM RE9178 Following Gap Release with the Reactor Cavity Flooded and DEI = 2 IiCi/gm.

Three pages + Cover Sheet 4

3 2 _ A__

-8/7 /2 04y--'--827/2004 /

Attachment I Prepared/Date Vqified/Date Revision.

Rad. Engineering Cabc. 04-004-A Attachment I Engineering Caic. CAL-R04-OOZ Revision 1

MicroShield v5.03a (5.03-00277)

Alliant-IES Industries

'age :1 File Ref:

DOS File: CAVITY8.MS5 Date:

Run Date: August 27, 2004 By:

Run Time: 12:41:52 PM Checked:

Duration : 00:00:06 Case

Title:

Cavity Dose Rate 8hr

Description:

Dose rates over the flooded cavity. DEI 2 ucilgm. t= 8hr Geometry: 8 - Cylinder Volume - End Shields Source Dimensions Height 670.56 cm 22 ft Radius 502.92 cm 16 ft 6.0 in Dose Points X Y z

1. 1 807.72 cm 762 cm 0cm 26 ft 6.0 in 25 ft 0.0 in

1. 2 1051.56 cm 822.96 cm 0cm 34 ft 6.0 in 27 ft 0.0 in 0.0 in Shields Shield Name Dimension Material Density Source 5.33e+08 cm 3 Water 1 Air Gap Air 0.00122 Wall Clad 609.6 cm Concrete 2.35 Immersion Air 0.00122 Source Input Grouping Method : Standard Indices Number of Groups: 25 Lower Energy Cutoff: 0.015 Photons < 0.015 : Excluded Library: Grove Nuclide curies becquerels tjCi/cm3 BqIcm3 Br-82 4.4171e+000 1.6343e+011 8.2900e-003 3.0673e+002 Br-83 7.1399e+000 2.6418e+011 1.3400e-002 4.9580e+002 Br-84 3.6605e-003 1.3544e+008 6.8700e-006 2.5419e-001 Cs-1 32 5.0618e-001 1.8729e+010 9.5000e-004 3.5150e+001 Cs-134 5.3283e+002 1.9715e+013 1.0000e+000 3.7000e+004 Cs-134m 1.6518e+001 6.1115e+01 1 3.1 OOOe-002 1.1470e+003 Cs-135 1.4599e-003 5.4018e+007 2.7400e-006 1.0138e-001 Cs-136 1.4599e+002 5.4018e+012 2.7400e-001 1.0138e+004 Cs-137 2.6215e+002 9.6996e+012 4.9200e-001 1.8204e+004 Cs-138 1.3907e-001 5.1455e+009 2.61 OOe-004 9.6570e+000 1-130 2.0194e+001 7.4718e+01 1 3.7900e-002 1.4023e+003 1-131 9.0048e+002 3.3318e+013 1.6900e+000 6.2530e+004 1-132 7.7633e+002 2.8724e+01 3 1.4570e+000 5.3909e+004 1-133 8.9834e+002 3.3239e+013 1.6860e+000 6.2382e+004 1-134 7.4596e+000 2.7600e+01 1 1.4000e-002 5.1 800e+002

Page * : 2 DOS File: CAVITY8.MS5 Run Date: August 27, 2004 Run Time: 12:41:52 PM Duration : 00:00:06 Nuclide curies becquerels uCilcm 3 Ba/cm 3 1-135 4.6462e+002 1.7191e+013 8.7200e-001 3.2264e+004 Kr-83m 1.8649e+001 6.9001 e+011 3.5000e-002 1.2950e+003 Kr-85 1.8755e+001 6.9395e+011 3.5200e-002 1.3024e+003 Kr-85m 4.1028e+001 1.5180e+012 7.7000e-002 2.8490e+003 Kr-87 3.4314e+000 1.2696e+011 6.4400e-003 2.3828e+002 Kr-88 5.3069e+001 1.9636e+012 9.9600e-002 3.6852e+003 Rb-86 4.8807e+000 1.8059e+011 '9.1600e-003 3.3892e+002 Rb-88 1.4226e+002 5.2638e+012 2.6700e-001 9.8790e+003 Xe-129m 8.6318e-003 3.1938e+008 1.6200e-005 5.9940e-001 Xe-131m 6.4472e+000 2.3855e+011 1.21 OOe-002 4.4770e+002 Xe-133 1.0982e+003 4.0632e+013 2.061 Oe+000 7.6257e+004 Xe-1 33m 3.5699e+001 1.3209e+012 6.7000e-002 2.4790e+003 Xe-135 5.3815e+002 1.9912e+013 1.0100e+000 3.7370e+004 Buildup The material reference is: Source Integration Parameters Radial 20 Circumferential 10 Y Direction (axial) 10 Results - Dose Point # I - (807.72,762,0) cm Energy Activity Fluence Rate Fluence Rate Exposure Rate Exposure Rate MeV photons/sec MeV/cm2 /sec MeV/cm2 /sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.03 2.449e+13 2.544e-03 2.851e-02 2.521e-05 2.825e-04 0.04 2.078e+1 1 8.487e-04 2.654e-02 3.754e-06 1.174e-04 0.06 6.732e+11 3.01 Oe-02 2.865e+00 5.978e-05 5.690e-03 0.08 1.612e+13 2.012e+00 2.479e+02 3.184e-03 3.923e-01 0.1 2.713e+10 6.614e-03 7.810e-01 1.012e-05 1.195e-03 0.15 2.116e+12 1.606e+00 1.265e+02 2.645e-03 2.082e-01 0.2 1.983e+13 3.349e+01 1.745e+03 5.911e-02 3.080e+00 0.3 7.259e+12 3.766e+01 1.065e+03 7.143e-02 2.021 e+00 0.4 2.931e+13 3.371e+02 6.232e+03 6.568e-01 1.214e+01 0.5 3.888e+1 3 8.213e+02 1.112e+04 1.612e+00 2.183e+01 0.6 6.536e+13 2.252e+03 2.408e+04 4.395e+00 4.699e+01 0.8 5.803e+1 3 4.262e+03 3.214e+04 8.106e+00 6.114e+01 1.0 2.267e+1 3 2.956e+03 1.743e+04 5.449e+00 3.213e+01 1.5 1.570e+13 5.583e+03 2.239e+04 9.394e+00 3.768e+01 2.0 5.223e+12 3.637e+03 1.166e+04 5.625e+00 1.803e+01 3.0 1.826e+1 1 3.067e+02 7.553e+02 4.161e-01 1.025e+00 4.0 9.694e+06 2.898e-02 6.114e-02 3.585e-05 7.564e-05 3.431e+01 6.507e+01 3.933e-02 5.0 7.547e+09 7.459e-02

Page ' : 3 DOS File: CAVITY8.MS5 Roun Date: August 27, 2004 Run Time: 12:41:52 PM Duration : 00:00:06 Enerqy Activity Fluence Rate Fluence Rate Exposure Rate Exposure Rate MeV photons/sec MeV/cm 2 /sec MeV/cm 2 /sec mRlhr mR/hr No Buildup With Buildup No Buildup With Buildup TOTALS: 3.061ee+14 2.026e+04 1.291e+05 3.583e+01 2.367e+02 Results - Dose Point # 2 - (1051.56,822.96,0) cm Enerny Activity Fluence Rate Fluence Rate Exposure Rate Exposure Rate MeV photons/sec MeV/cm 2 /sec MeV/cm 2/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.03 2.449e+1 3 8.880e-04 1.054e-02 8.801 e-06 1.044e-04 0.04 2.078e+1 1 3.688e-04 1.237e-02 1.631 e-06 5.473e-05 0.06 6.732e+1 1 1.424e-02 1.489e+00 2.827e-05 2.957e-03 0.08 1.612e+13 9.726e-01 1.337e+02 1.539e-03 2.116e-01 0.1 2.713e+10 3.248e-03 4.309e-01 4.969e-06 6.592e-04 0.15 2.116e+12 8.146e-01 7.201 e+01 1.342e-03 1.186e-01 0.2 1.983e+1 3 1.743e+01 1.01Oe+03 3.076e-02 1.782e+00 0.3 7.259e+1 2 2.032e+01 6.286e+02 3.855e-02 1.192e+00 0.4 2.931e+13 1.865e+02 3.715e+03 3.634e-01 7.239e+00 0.5 3.888e+1 3 4.626e+02 6.675e+03 9.081e-01 1.310e+01 0.6 6.536e+1 3 1.285e+03 1.451 e+04 2.509e+00 2.832e+01 0.8 5.803e+1 3 2.478e+03 1.944e+04 4.713e+00 3.698e+01 1.0 2.267e+1 3 1.739e+03 1.056e+04 3.205e+00 1.946e+01 1.5 1.570e+13 3.331 e+03 1.353e+04 5.604e+00 2.276e+01 2.0 5.223e+12 2.180e+03 7.003e+03 3.371 e+00 1.083e+01 3.0 1.826e+11 1.836e+02 4.483e+02 2.491e-01 6.082e-01 4.0 9.694e+06 1.725e-02 3.592e-02 2.135e-05 4.444e-05 5.0 7.547e+09 2.031 e+01 3.794e+01 2.329e-02 4.349e-02 TOTALS: 3.061e+14 1.191e+04 7.776e+04 2.102e+01 1.426e+02

X. ATTACHMENT 2: Microshield Case Results: t = 30 Days Attachment 2 to Engineering Calculation No. CAL-R04-002 CAVITY30.MS5 Attachment Rev. I Radiation Exposure Rates at Spent Fuel Pool ARM RE9178 Following Gap Release with the Reactor Cavity Flooded and DEI = 2 tiCi/gm.

Three pages + Cover Sheet 4 ___ 1 3 _ _ _ _ _ _ _ _ _ _ _ _ _ l _ _ _ _ _ _ _ _ _ _ _ l__

2 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

1 . -W- l _8/27/2004 _l Attachment 2 Prepared/Date Verified/Date Revision.

Rad. Engineering Caic. 04-004-A Attachment 2 Engineering Cabc. CAL-R04-OOZ Revision 1

MicroShield v5.03a (5.03-00277)

Alliant-IES Industries Page :1 File Ref:

DOS File: CAVITY30.MS5 Date:

Run Date: August 27, 2004 By:

Run Time: 12:40:49 PM Checked:

Duration : 00:00:05 Case

Title:

Cavity Dose Rate 30d

Description:

Dose rates over the flooded cavity. DEI = 2 ucilgm. t= 30d Geometry: 8 - Cylinder Volume - End Shields Source Dimensions Height 670.56 cm 22 ft Radius 502.92 cm 16 ft 6.0 in Dose Points X Y z

1. 1 807.72 cm 762 cm 0 cm 26 ft 6.0 in 25 ft 0.0 in

1. 2 1051.56 cm 822.96 cm 0cm 34 ft 6.0 in 27 ft 0.0 in 0.0 in Shields Shield Name Dimension Material Density Source 5.33e+08 cm3 Water 1 Air Gap Air 0.00122 Wall Clad 609.6 cm Concrete 2.35 Immersion Air 0.00122 Source Input Grouping Method : Standard Indices Number of Groups: 25 Lower Energy Cutoff: 0.015 Photons < 0.015: Excluded Library: Grove Nuclide curies becquerels uCi/cm 3 Bq/cm3 Br-82 5.6320e-005 2.0838e+006 1.0570e-007 3.9109e-003 Cs-1 32 3.1597e-001 1.1691e+010 5.9300e-004 2.1941e+001 Cs-1 34 7.7590e+003 2.8708e+014 1.4562e+001 5.3879e+005 Cs-135 2.1899e-002 8.1027e+008 4.11 OOe-005 1.5207e+000 Cs-136 4.5397e+002 1.6797e+013 8.5200e-001 3.1524e+004 Cs-1 37 3.9120e+003 1.4474e+014 7.3420e+000 2.7165e+005 1-131 1.0657e+003 3.9429e+013 2.0000e+000 7.4000e+004 1-132 2.11 00e+001 7.8070e+-0 1 3.9600e-002 1.4652e+003 1-133 6.6603e-007 2.4643e+004 1.2500e-009 4.6250e-005 Kr-85 2.7973e+002 1.0350e+013 5.2500e-001 1.9425e+004 Rb-86 2.4244e+001 8.9701e+011 4.5500e-002 1.6835e+003 Xe-129m 9.8573e-003 3.6472e+008 1.8500e-005 6.8450e-001 Xe-131m 3.6978e+001 1.3682e+012 6.9400e-002 2.5678e+003 Xe-1 33 3.8736e+002 1.4332e+013 7.2700e-001 2.6899e+004 Xe-1 33m 6.5538e-002 2.4249e+009 1.2300e-004 4.551 Oe+000

Page :2 DOS File: CAVITY30.MS5 Run Date: August 27, 2004 Run Time: 12:40:49 PM Duration : 00:00:05 Buildup The material reference is : Source Integration Parameters Radial 20 Circumferential 10 Y Direction (axial) 10 Results - Dose Point # I - (807.72,762,0) cm Energy Activity Fluence Rate Fluence Rate Exposure Rate Exposure Rate MeV photons/sec 2 MeV/cm /sec MeV/cm 2 /sec mR/hr mR/hr No Buildup With BuilduD No Buildup With Buildup 0.03 1.362e+1 3 1.415e-03 1.585e-02 1.402e-05 1.571 e-04 0.04 9.71 0e+1 1 3.966e-03 1.240e-01 1.754e-05 5.486e-04 0.06 2.093e+12 9.358e-02 8.908e+00 1.859e-04 1.769e-02 0.08 7.347e+12 9.168e-01 1.130e+02 1.451e-03 1.788e-01 0.1 6.866e+10 1.674e-02 1.976e+00 2.561e-05 3.024e-03 0.15 2.160e+12 1.639e+00 1.291e+02 2.700e-03 2.126e-01 0.2 2.494e+12 4.213e+00 2.195e+02 7.436e-03 3.874e-01 0.3 1.306e+1 3 6.777e+01 1.917e+03 1.286e-01 3.637e+00 0.4 3.202e+1 3 3.682e+02 6.808e+03 7.175e-01 1.326e+I01 0.5 4.728e+1 2 9.990e+01 1.353e+03 1.961e-01 2.655e+00 0.6 3.525e+14 1.214e+04 1.298e+05 2.370e+01 2.534e+02 0.8 2.885e+14 2.119e+04 1.598e+05 4.030e+01 3.040e+02 1.0 2.503e+13 3.263e+03 1.924e+04 6.014e+00 3.546e+01 1.5 8.848e+1 2 3.148e+03 1.262e+04 5.296e+00 2.124e+01 2.0 2.566e+10 1.787e+01 5.728e+01 2.763e-02 8.858e-02 TOTALS: 7.535e+14 4.030e+04 3.321e+05 7.640e+01 6.346e+02 Results - Dose Point # 2 - (1051.56,822.96,0) cm Energy Activity Fluence Rate Fluence Rate Exposure Rate Exposure Rate MeV photons/sec MeV/cm2/sec MeV/cm 2 /sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.03 1.362e+13 4.937e-04 5.860e-03 4.893e-06 5.807e-05 0.04 9.71 Oe+11 1.724e-03 5.783e-02 7.623e-06 2.557e-04 0.06 2.093e+12 4.426e-02 4.629e+00 8.792e-05 9.194e-03 0.08 7.347e+12 4.432e-01 6.093e+01 7.014e-04 9.641 e-02 0.1 6.866e+1 0 8.219e-03 1.090e+00 1.257e-05 1.668e-03 0.15 2.160e+12 8.316e-01 7.351e+01 1.369e-03 1.21 Oe-01 0.2 2.494e+12 2.192e+00 1.270e+02 3.870e-03 2.242e-01 0.3 1.306e+13 3.657e+01 1.131e+03 6.938e-02 2.146e+00 0.4 3.202e+13 2.037e+02 4.059e+03 3.970e-01 7.908e+00 0.5 4.728e+12 5.627e+01 8.119e+02 1.104e-01 1.594e+00 0.6 3.525e+14 6.932e+03 7.824e+04 1.353e+01 1.527e+02

X 3age ': 3 JOS File: CAVITY30.MS5 Run Date: August 27, 2004 Run Time: 12:40:49 PM Duration : 00:00:05 Energy Activity Fluence Rate Fluence Rate Exposure Rate Exposure Rate MeV photons/sec MeV/cm 2 /sec MeV/cm 2 /sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.8 2.885e+14 1.232e+04 9.666e+04 2.343e+01 1.838e+02 1.0 2.503e+13 1.919e+03 1.165e+04 3.537e+00 2.148e+01 1.5 8.848e+12 1.878e+03 7.625e+03 3.159e+00 1.283e+01 2.0 2.566e+10 1.071e+01 3.441 e+01 1.656e-02 5.321 e-02 TOTALS: 7.535e+14 2.336e+04 2.005e+05 4.426e+01 3.830e+02

PLANT CHEMISTRY PROCEDURES.3200 MANUAL PCP 8.6 ALARM SETPOINTS AND EFFICIENCY FOR Rev. 18 OG PRETREATMENT Page 1of 9 Usage Level Reference Use

[ EffectiveDate: AUG 2 6 2004l

• ........ ,..A.4 A

• Prepared by'. Y C/ 1 ° Date: ft C Validated by. _ _ _ _ _ _ _ _ _ Date: _ _-_8 .- mstyStf I am responsible for the technical content of this procedure.

Verified and Approved by Procedure Owner: Date: _ _ _ _

Chemistry Supervisor

PLANT CHEMISTRY PROCEDURES.3200 MANUAL PCP 8.6 ALARM SETPOINTS AND EFFICIENCY FOR Rev. 18 OG PRETREATMENT Page 2 of 9 1.0 PURPOSE To provide a procedure for the analysis of pretreat data and the determination of the alarm setpoint for the Offgas Pretreatment Radiation Monitoring System.

2.0 DISCUSSION The Offgas Pretreatment Radiation Monitoring System continuously monitors air from the steam jet air ejector at a point upstream of the holdup line. The sample stream is monitored with an ionization chamber that provides dose rate readout (mR/hr). To relate the dose rate readout to the release rate of the radioactive noble gases, a grab-sample is analyzed for 6 prominent, gamma emitting, fission gases. The activity of these 6 noble gases is used to determine the release rate for all 22 noble gas U-235 fission products. The alarm setpoint is determined in terms of the pretreatment dose rate (mRlhr). The setpoint equation is derived from the equation for the Steam Jet Air Ejector in the ODAM. The equation is as follows:

S = 1O6 RF PE: [(K-)(e-"')]

where:

S = Monitor Alarm Setpoint (mRlhr) 106 = The Tech Spec release rate RF = The fraction of the reading due to fission gases P = Sum of 22/sum of 6 ratio K; = Concentration of the fission gases (giCi/sec)

Xi = Radioactive decay constant of noble gas nuclide i (min ')

t = 30 min. decay time in delay line The Krypton isotopes may fall below the best fit line if burnout is large. That is, if the fuel contains more Pu-239 and decreased U-235 as the end of the fuel cycle approaches.

3.0 EQUIPMENT AND MATERIALS NEEDED

- 3.1.1 See PCP 2.6.

I PLANT CHEMISTRY PROCEDURES 3200 MANuAL PCP 8.6 ALARM SETPOINTS AND EFFICIENCY FOR Rev. 18 OG PRETREATMENT Page 3 of 9 4.0 PRECAUTIONS AND LIMITATIONS NOTE The E6 is an indication of fuel failure. If the value increases to 300 microcuries/second OR greater, the Chemistry Supervisor should be notified immediately so that Administrative Control Procedure (ACP) 102.15, "Fuel Reliability Program" may be implemented.

4.1.1 Take septas off of sample vials in the hood, let stand for a few hours before storing back in drawer for reuse. (If oil is present, rinse with isopropyl alcohol.)

5.0 PROCEDURE 5.1 SAMPLE COUNTING 5.1.1 Take a 14.8 mL sample of the offgas pretreatment in accordance with PCP 2.6.

5.1.2 Record the following information.

(1) Sample time and date (2) Monitor reading (mRlhr) Control Room Back Panels (3) Power level (MWT)

(4) Pretreatment (hold-up line) flow rate (cfm) FR-4132 Adsorber Outlet Flow. Control Room Back Panels (5) Total (both dean-ups) Reactor Clean-up flow (gpm).

(6) Previous quarterly setpoint (if performing setpoint determination).

5.1.3 Determine sample quantity (sec)

Calculate sample quantity using the following equation:

14.8 472(hold - uplineflow)

PLANT CHEMISTRY PROCEDURES 3200 MANUtAL.L .I. PCP 8.6 ALARM SETPOINTS AND EFFICIENCY FOR Rev. 18 OG PRETREATMENT Page 4 of 9 5.1.4 To determine correct time to count a pre-treat sample on the gamma spectroscopy I system, perform the following:

(1) Subtract five (5) minutes from Time of Sample. This is done to more accurately I determine activity at the fuel. This will be referred to as the "Decay Time".

(2) Prior to counting, allow sample to decay sixty (60) minutes, but not more than seventy I (70) minutes, from the Decay Time.

Example: Sampling was performed at 10:00. Minus 5 minutes give a decay time at 9:55.

By allowing the sample to decay a total of 60 to 70 minutes, the tech shall start the count at 11:00 +/- 5 minutes. (Note: Decay to 9:55.)

5.1.5 Analyze the sample in accordance with PCP 7.3, Attachment 1, Decay Sample to the Decay Time.

5.1.6 IF Xe-1 33 is not identified, report the MDA value, without the less than (<) symbol.

NOTE There are two ways of calculating data results. The first and preferred is to use the Pretreat Excel Spreadsheet using section 5.2. The second method is to use Form 328 by starting at section 5.3.

5.2 SPREADSHEET ANALYSIS 5.2.1 Open the PRETREAT file by double clicking on the Pretreat icon (if available).

Otherwise, the spreadsheet is located on the DEPT drive under CHEM, then FORMS.

5.2.2 Enable MACROS.

5.2.3 At the "QA INFO" screen, click Perform QA Check". After completion of the QA Check, a row of tabs appears at the bottom of the screen. These allow selection of the pretreat form and graph.

5.2.4 Select 'FORM" tab at bottom of screen.

NOTE The l 6 is an indication of fuel failure. If the value increases to 300 microcuries/second OR greater, the Chemistry Supervisor should be notified immediately so that Administrative Control Procedure (ACP) 102.15, "Fuel Reliability Program" may be implemented.

.- PLANT CHEMISTRY PROCEDURES 3200 MANUAL. PCP 8.6 ALARM SETPOINTS AND EFFICIENCY FOR Rev. 18 OG PRETREATMENT Page 5 of 9 5.2.5 Recorddata Items2-10andall pCVsecvalues. Items 11, 12and 18are automatically calculated by the program.

5.2.6 When performing the Quarterly Alarm Setpoint, obtain the value for Item 17 from the previous quarterly setpoint determination for Offgas Pretreatment.

5.2.7 The program will calculate all other necessary data and plot the results on the attached graph. ('GRAPH' tab at bottom of screen.)

5.2.8 Select "GRAPH' tab at bottom of screen to review plotted data.

5.2.9 Although the program does not draw the line connecting the data points, observe the data and follow the NOTE below.

5.2.10 If the best-fit line appears to be horizontal, return to the form and record '0" as Item 13, if not already there. If the data appears to be non-horizontal, contact the foreman or the Chemistry Supervisor and they will review the data.

5.2.11 Record appropriate P "sum of 22/sum of 6" from Attachment I as Item 14.

5.2.12 Based on the fractional change, record the setpoint value as the monitor HiHi Alarm, Item 23.

(1) If the fractional change is outside the band of +/-0.2, compare the calculated setpoint to 1E4 mRlhr. Use the most conservative (smaller) value as the new setpoint.

(2) If the fractional change iswithin the +/-0.2, use the previous setpoint value. Verify this is less than 1E4 mR/hr.

5.2.13 Print both the form and graph by returning to the form, then clicking on either of the boxes within the body of the form labeled 'FILE PRINT COMMAND" 5.2.14 If performing the monthly STP, record the 'Grand Total Activity" from the isotopic printout as the 'Total Gases Release Rate".

5.2.15 Enter NIDX into CDM.

5.2.16 Enter K Factor into CDM.

5.3 DATA ANALYSIS USING FORM 328 5.3.1 Record the data requested on Form 328, Items 1 through 11.

5.3.2 Record noble gas concentrations measured in the pretreat sample in Column B.

PLANT CHEMISTRY PROCEDURES 3200.MANUAL PCP 8.6 ALARM SETPOINTS AND EFFICIENCY FOR Rev. 18 OG PRETREATMENT Page 6of 9 5.3.3 Sum the measured concentrations of the fission gases in Column B AND record at Item 12 (Z 6).

NOTE The value of Item 12 from Attachment 1 (£ 6) is an indication of fuel failure. If the value jumps to 1000 microcuries/second OR greater, the Chemistry Supervisor should be notified immediately so that Administrative Control Procedure (ACP) 102.15, "Fuel Reliability Program" may be implemented.

5.3.4 Multiply Column B by Column C for each noble gas concentration. Record values in Column D.

5.3.5 Plot the six values in Column D on log-log graph paper (Form 328). Draw a straight horizontal line representing the best fit for the plotted values, particularly that of Xe-138. If the data appears to be non-horizontal, contact the foreman or the Chemistry Supervisor and they will review the data and determine whether a different line needs to be drawn.

5.3.6 Measure the rise (y2-yl) and the run (x2-xl). Calculate the slope of the line. (This is a geometric slope, measured with a ruler, not an arithmetic slope.) Normally, this is zero.

Slope = rise/run 5.3.7 Record the slope on Form 328 as Item 13.

5.3.8 On the pretreat analysis graph (Attachment 1), use the slope to find P "Sum of 22/Sum of 6" ratio. Record as Item 14.

5.3.9 Multiply Item 12 "Sum of 6" by Item 14 "P" and record as Item 15. This is the "Sum of 22".

5.3.10 Add the release rate values (pCi/sec) for Ar-41 and N-1 3 to Item 15 "Sum of 22".

Record as the total gases, Item 16.

5.3.11 Divide Item 16 "Total Gases" by Items 8 and 9. Record as the monitor K Factor, Item 16.b.

5.3.12 If performing monthly STP 3.7.6-01, record the "Grand Total Activity" from the isotopic printout as the "Total Gases Release Rate" in the STP.

5.3.13 Enter NIDX into CDM.

I

PLANT CHEMISTRY PROCEDURES 3200 MANUAL PCP 8.6 ALARM SETPOINTS AND EFFICIENCY FOR Rev. 18 OG PRETREATMENT Page 7 of 9 5.3.14 Enter K Factor into CDM.

5.4 CALCULATING ALARM SETPOINT Calculating the alarm setpoint is required only once a quarter. Use this method only if the spreadsheet is unavailable.

5.4.1 Use data assembled on Form 328, from Section 5.2.

5.4.2 Obtain the value for Item 17 from the previous quarterly setpoint determination for Offgas Pretreatment.

5.4.3 To Item 12, add the measured concentrations of Ar-41 and N-13. Record the total as Item 18.

5.4.4 Divide Item 12 by Item 18 AND multiply by the pretreat monitor reading, Item 8.

Record value as Item 19. This is the fraction of the reading due to fission gases.

5.4.5 Multiply each fission gas in Column B by the value in Column E. Record the results in Column F. This decays each isotope for the 30-minute hold-up time.

5.4.6 Sum the values in Column F AND record as Item 20.

5.4.7 Compute the monitor alarm setpoint, S, AND record as Item 21.

5.4.8 Compute the fractional change from the previous setpoint AND record as Item 22.

5.4.9 Based on the fractional change, record the setpoint value as the monitor Hi Hi Alarm, Item 23.

(1) IF the fractional change is outside the band of +/-0.2, compare the calculated setpoint to 1E4 mR/hr. Use the most conservative (smaller) value as the new setpoint.

(2) IF the fractional change is within the band of +0.2, use the previous setpoint value. Verify this value is less than 1E4 mR/hr.

5.4.10 Compute the Hi Alarm Setpoint AND record as Item 24.

5.4.11 Notify the Chemistry Foreman of the results.

'PLANT CHEMISTRY PROCEDURES 3200 MANUAL.. PCP 8.6 ALARM SETPOINTS AND EFFICIENCY FOR Rev. 18 OG PRETREATMENT Page 8 of 9 6.0 RECORDING AND REPORTING DATA NOTE Steps 6.1.1 and 6.1.2 are required when a cyclic STP or an efficiency is being performed.

6.1.1 Reanalyze the Isotopic data on the gamma spectrometer using 14.8 cc for the volume to obtain jICi/cc.

6.1.2 Divide monitor reading (mR/hr) by the total pCi/cc of the identified isotopes from step 6.1.1 to obtain the monitor efficiency.

7.0 REFERENCES

7.1.1 PCP 2.6, 'Offgas Pretreatment and Posttreatment Sampling" 7.1.2 EPRI"Failed Fuel Action Plan Guidelines", November 1987 8.0 ATTACHMENTS 8.1.1 Pretreatment Analysis Graph

PLANT CHEMISTRY PROCEDURES 3200 MANUAL PCP 8.6 ALARM SETPOINTS AND EFFICIENCY FOR Rev. 18 OG PRETREATMENT Page 9 of 9 ATTACHMENT I PRETREATMENT ANALYSIS GRAPH -

hWw~ of AII.% 11- X '*;;

. =

mm.- -_

a .,.

_ ___ ____i

'a j t.,~,* s S l n - .t.l . .-.-

V4- *ii~f i,,S*'l si 'iS -

'-1:- M.s-*

' *. *~~~iiii.4

. b:-:

AA !Bian- __ M 100 =.___ a I..W_ i _ .. .

-s- -....

-a -. ~r.. . .J.L - .:

- -.-..... == ==

_ ._. _=== .....m ............- f {

-. 1@

r

-t - .,.

=

a e.=~- W , .. =--:.;.-s: *_-

__D_=__

= . t-6t .'.--i

_Ir b

- . - .3

• _  : i- .o@ i *_, ,_._

- .:.:i.- i. 5 a _ X

-. _i.i.- i  :.: 2t-'-:w

________________=_-,,_________i________________

JDAEC EOP BASES DOCUMENT EOP BREAKPOINTS Rev. 6 Page 1of 15 lEffectiveDate: OCT 2 1 2003 TECHNICAL REVIEW Prepared by. C;4 Date: 16 la/1 Technical Date: loll1lO3 Review by: __ __ _to/__ __ __ __ __ _ Date:___ __ __ __ __ __ _

Operations Department Staff Technical =

Review by. ~..J Date: 10/Z /03 Operatl6ns Training Department Staff DOCUMENT APPROVAL I am responsible for the technical content of this procedure.

Approved by Procedure Owner: Date:inatoDt

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 2 of 15 A -

The purpose of the EOP breakpoint section is to illustrate the various action levels inside each EOP section which, through experience, are found to be significant enough to require additional familiarity. The action levels frequently correspond exactly to known plant limits and design characteristics.

The EOP breakpoints can be viewed as a simplistic one path flowchart of the significant actions in each EOP section, a short hand EOP. It is useful to understand, though not to memorize, the action levels. An understanding of when the next action point occurs for an item like torus temperature can free up operators and the EOP user to concentrate on other areas or to prioritize actions if resources are limited.

RO-CRS interaction can be enhanced if both parties are aware of the next most important criteria. An RO may assist the CRS by being cognizant of the next big picture step. EOP breakpoints can be used to assist in developing the RO-CRS interaction.

Focused communications will enhance use of the procedures by the operating crew.

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 3 of 15 COMMON POWER CONTROL BREAKPOINTS Reactor Item of Interest Significance Power 56% Minimum Flow at 100% Lowest power that can be reached if Recirc Rod Line Pumps runback to minimum 44% Natural Circ at 100% Lowest power that can be reached if Recirc Rod Line Pumps tripped 18% Bypass Valve Capacity Maximum power that can be handled through (20.6) the bypass valves. Other steam loads may handle an additional 5% power (Note 1) 15% APRM Scram Scram with the Reactor Mode Switch in the STARTUP position 10.2% Boron Injection Power SBLC injection required in ATWS EOP if power Level >10.2% for torus temp >110 'F (see BIIT

_ _Curve) 5% APRM Downscale Power below which level will not be lowered to choke power during ATWS Level Power Control

-20 Point of Adding Heat "Shutdown and no boron injected into the RPV" on IRM (POAH) means that the reactor is subcritical (power Range 8 decreasing), reactor power is below the POAH and no boron has been injected All Rods Minimum Subcritical Reactor is shutdown (no ATWS), exit ATWS Inserted to Banked Withdrawal EOP and enter EOP 1, enter IPOI 5 if exiting at least Position (MSBWP) EOP I for power control Position 00 Note 1: Plant analysis for implementing the Extended Power Uprate at the DAEC indicates that after EPU implementation, Bypass Valve Capacity will be around 20.6. Previous operating experience indicates that the actual value after power uprate may be around 18%.

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 4 of 15 ADDITIONAL ATWS POWER CONTROL BREAKPOINTS Value Item of Interest Significance 25% Pk-Pk Large Oscillation Boron injection required during ATWS.

Oscillations Threshold 47% SBLC Hot Shutdown Boron The least weight of soluble boron which, if Tank Level Weight injected into the RPV and mixed uniformly, will OR maintain the reactor shutdown under hot standby conditions. This weight is utilized to 2 barrels of assure the reactor will be shutdown Boric Acid irrespective of control rod position when RPV and water level is raised to uniformly mix the 2 barrels of injected boron.

Borax 16% SBLC Cold Shutdown Boron The least weight of soluble boron which, if Tank Level Weight injected into the RPV and mixed uniformly, will OR maintain the reactor shutdown under all conditions. This weight is utilized to assure 4 barrels of the reactor will remain shutdown irrespective Boric Acid of control rod position or RPV water and temperature.

4 barrels of Borax

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 5 of 15 RPV Pressure Item of Interest Significance (psig) 1240 Safety Relief Valve Pressure boundary protection l _ Setpoint 1140 ARI Trip Reactor scram and recirc pump trip, depressurizes the scram air header 1110, 1120, ADS SRV Setpoints SRVs open and cause fluctuations in level 1130,1140 . and power 1055 High RPV Pressure Control reactor pressure below the scram Scram setpoint EOP 1 Entry Condition 1035 Low-Low Set B Open signal to B LLS Logic 1030 Low-Low Set A Open signal to A LLS Logic

-948 Bypass Valves 100% Maximum pressure for bypass valves with open EHC Pressure Set at 940 psig 940 Bypass Valves 0% open Minimum pressure for bypass valves l_ (Controlled by EHC Pressure Set) 915 Low-Low Set B Close signal to B LLS Logic 910 Low-Low Set A Close signal to A LLS Logic 900 LPCI Loop Select LPCI Loop Select Logic Hold Point 630 Minimum Alternate RPV MARFP with I SRV open in ATWS EOP Flooding Pressure (MARFP)

-500 Condensate Pump Start of Condensate Pump Injection Shutoff Head 450 LPCI/CS Permissive LPCI & CS Injection Valves Open

-330 CS Pump Shutoff Head Start Core Spray Injection

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 6 of 15 RPV Pressure Item of Interest Significance (Psi9) 310 Minimum Alternate RPV MARFP with 2 SRVs open in ATWS EOP Flooding Pressure (MARFP)

-260 RHR Pump Shutoff Head Start LPCI Injection 200 Minimum Alternate RPV MARFP with 3 or more SRVs open in ATWS Flooding Pressure EOP (MARFP) 135 Shutdown Cooling Tech Spec Shutdown Cooling Interlock Setpoint (See Note 1) 150 Minimum Alternate RPV MARFP with 4 SRVs open in RPV/F & ATWS Flooding Pressure EOP (MARFP) 50 Minimum RPV Flooding Minimum dP to keep an SRV open Pressure (MRFP) __.

Note 1: This equates to a RPV pressure of about 90-100 psig.

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 7 of 15 RPV Level Item of Interest Significance (inches)

+211 High Level Trip Setpoint, . Loss of high pressure injection (FW, Main Turbine Trip HPCI, RCIC)

LLoss of 100% Heat Sink

+170 Low Water Level Scram,

• RPS defeats needed in ATWS PCIS Groups 2, 3,4 Isolations . Containment Isolation, l Shutdown Cooling Valves Close

+119.5 High Pressure Injection,

• HPCI/RCIC Auto Initiation PCIS Group 5 Isolation, . RWCU Isolation ARI *RC slto

• ARI Initiation & Recirc Pump ATWS Trip

+87 Two Feet Below During ATWS if power >5% or unknown, Feedwater Sparger lower level to +87 inches to reduce core inlet l _ subcooling

+64 ECCS Auto Start, . ADS Timers start PCIS Group 1 Isolation . CS/RHR Auto Initiation MSIVs close and result in loss of main condenser

+15 Top of Active Fuel (TAF)

• Loss of Adequate Core Cooling (ACC)

(Note 1) through core submergence

• Maximize injection with Alternate Injection

. Systems in ALC when level < +15" Note 1: +15 inches is used for TAF than 0 inches for the following reasons:

• To allow monitoring RPV level on the Wide Range instrumentation -

prevents risk of uncovering the core if using Fuel Zone instruments.

• Fuel Zone instruments use the same tap as jet pump instrumentation and any flow through the jet pumps including LPCI flow will cause the Fuel Zone instruments to read high.

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 8 of 15 RPVchese Item of Interest Significance

-25 Minimum Steam Cooling

• No guarantee that fuel cladding RPV Water Level temperature can be kept <1500 F (MSCRWL) . ED required in ALC before -25 inches if an injection source is available

. SAG Entry in ALC if cannot restore and maintain level above -25 inches with less than 2 ECCS Pumps available

. Lower end of level control band in ATWS level/power control

• Loss of ACC in ATWS Steam Cooling &

SAG Entry

-39 Elevation of top of Jet . RPV water level following DBA LOCA (P2/3 Core Height) a SAG Entry in ALC if cannot restore and maintain level above -39 inches with at least 2 ECCS Pumps available

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 9 of 15 Torus Level Item of Interest Significance (feet) 23.2 Bottom of Torus Vent Cannot vent torus (Note 1) 16 Highest Torus Level Torus level indication not reliable above the Indication upper tap at 16 feet 13.8 Maximum Pressure

• Bottom of ring header, loss of pressure (13.83) Suppression Pressure suppression capability above this level Primary Containment Watr Level n . Backpressure may cause damage to SRV Weater Leveltailpipe, supports or T-Quencher if SRVs SRV Tailpipe Level Limit are opened or lift (SRVTPLL) . Loss of Torus H2/02 Sampling and Torus CAM/PASS Sample Line CAM Radiation Monitor l _ Tap 13.5 Bottom of Torus-DW Negative containment pressure or torus-to-Vacuum Breakers drywell dP from drywell sprays may cause drywell or torus failure - if>13.5 ft terminate l _ injection into torus per EOP 2 10.4 EOP Value For Max LCO EOP 2 entry (10A3) 10.1 EOP Value For Min LCO EOP 2 entry (10.11) 7.1 Bottom of downcomer Loss of pressure suppression capability l openings below this level 5.8 HPCI Exhaust Containment pressurization if operate HPCI 4.5 SRV Quencher Top of SRV discharge devices, containment pressurization if operate SRVs while discharge is uncovered Note 1: Use 16 ft in EOPs due to limit on torus level indication

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 10 of 15 Temp Item of Interest Significance 340 SRV Qualification Emergency RPV Depressurization must be Temperature performed prior to exceeding SRV

._ qualification temperature 280 Drywell Design Initiate Drywell Sprays prior to 280 'F if (281) Temperature permitted by DWSIL Curve (EOP Graph 7),

ED required if cannot restore and maintain temperature <280 0F 150 Drywell Maximum EOP 2 Entry, Maximize drywell cooling Normal Operating Temperature 135 Drywell High Tech Spec LCO for High Drywell Temp ITemperature LCO

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 1 of 15 Temp Item of Interest Significance

-200 RHR NPSH Temperatures above this value may cause NPSH concerns for RHR pumps depending on flowrate and torus pressure

-190 Core Spray NPSH Temperatures above this value may cause NPSH concerns for Core Spray pumps l depending on flowrate and torus pressure 110 Initiate SBLC SBLC injection required in ATWS if power Tech Spec LCO >10.2%

Reactor scram required by Tech Specs 95 Tech Spec LCO EOP 2 entry, Maximize torus cooling with RHR Pumps not required for ACC

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 12 of 15 Pressure Item of Interest Significance (psig) 53 Primary Containment When PCPL is reached, containment venting (Torus) Pressure Limit (PCPL) is required.

30 CAD Isolation CAD Supply valves close if drywell pressure (Drywell) exceeds 30 psig

-21.9 Pressure Suppression Pressure Suppression Pressure exceeded for (Torus) normal torus level

>11 Drywell Sprays Drywell sprays may be initiated if drywell (Torus) parameters are within the Drywell Spray Initiation Limit and torus level is less than (11.15) 13.5 feet

<11 Torus Spray Initiation Start towus sprays prior to 11 psig, if possible.

(Torus) Pressure If pressure is exceeded before torus sprays are initiated - initiate them anyway (11.15) 7.2 Drywell Spray Curve Only for containment pressure less than 7.2 (Drywell) Break Point psig does the DWSIL change from 3500 F 2 Drywell High Pressure ECCS Initiation, Isolations and RPS defeats (Drywell) Scram Setpoint may be needed, EOP 1 and EOP 2 entry I Drywell N2 Makeup Drywell N2 makeup supply isolates if drywell (Drywell) Isolation pressure exceeds 1 psig

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 13 of 15 H2 Conc Item of Interest Significance 4.0 EOP 2 Entry High H2 alarm, enter or re-enter EOP 2

>0.4% H2 High Hydrogen in EOPs Initiate CAD to maintain 02 Concentration <4%

AND 24%02 0.4 Minimum Detectable Monitor 02 to determine if CAD operation is Concentration required

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 14 of 15 Cont. Item of Interest Significance Level (ft) 95 Drywell Vent Line Elevation of drywell vent, cannot flood above this point in SAGs Terminate injection from sources outside the Primary Containment per SAGs 82 Top of Active Fuel (TAF) TAF relative to primary containment water (81.8) level. Guarantees core submergence if level (81.8)_ this high in SAGs 74 Upper Drywell Spray Upper sparger no longer effective for spraying Sparger atmosphere, still may be used for injection 70 Bottom of Active Fuel BAF relative to primary containment water (69.8) (BAF) level 54 Upper Drywell H2/02 Isolate Upper CAMIPASS Sample Line, must CAM and PASS Sample use alternate sampling Line Tap 53 Bottom of RPV Continued injection may flood RPV through breach in bottom head 52 Lower Drywell Spray Lower sparger no longer effective for spraying Sparger atmosphere, still may be used for injection 43 Lower Drywell H2/02 Shift CAM/PASS Sample Line to upper drywell CAM and PASS Sample tap, isolate lower tap Line Tap 39 Bottom of MSL Drain Open Inboard MSL Drain MO-4423 prior to MOV reaching this level 28 4 feet of water on Drywell Sufficient water to cool core debris that has Floor breached the RPV, injection no longer needs to be maximized into the primary containment, coordinate any continued flooding with TSC AMT

DAEC EOP BASES DOCUMENT Rev. 6 EOP BREAKPOINTS Page 15 of 15 Levl Item of Interest Significance 25 Drywell Air Purge Line Cannot use air purge for H2 control if containment is flooded beyond this elevation 24 Drywell Floor Continued injection past this elevation will start to flood the drywell 23.2 Bottom of Torus Vent Cannot vent the torus Connection

ANNUNCIATOR RESPONSE PROCEDURE ARP IC03A REACTOR AND CONTAINMENT COOLING AND ISOLATION Revision 37 Usage Level Reference Use Effective Date: JUN 1 7 2004 I TECHNICAL REVIEW Prepared by: Date:

Validated by: A Date: S12o1 11U(perat I Verified by: Date:4wvY L/3104

.- 10Syste gixr Reviewed by: _ _ __ _ _ Date:

Operations Committee J PROCEDURE APPROVAL I am responsible for the technical content of this procedure.

Approved by Procedure Owner. GNUI/ Date:

Operations Approved by: __ Date:_ _ _6__

DAEC Plant Manager

REVISION CONTROL SHEET Page 1 of 1 ANNUNCIATOR RESPONSE PROCEDURES ARP IC03A REVISION NUMBER: 37 DATEPING TYRPING lI R f TYPING I COORD DATE REV.I COORD I DAT IREV._I COORD IDATE REV.

Grid Layout 7/28/90 0 * (C-1) 4/4/01 6 (C-2) 4/4/01 8 (A-1) 9/8/03 8 (C-3) 9/8103 7 (A-2) 414/01 7 (C4) 414/01 10 (A-3) .414101 7 (C-5) 4122/04 17 (A4) 4/4/01 10 (C-6) 4122/04 5 (A-5) 4/22/04 7 (C-7) 4/22/04 6 (A-6) 4/22/04 6 (C-8) 9/8/03 8 (A-7) 4/4101 4 (C-9) 4/4/01 7 (A-8) 4/4/01 4 (A-9) 4122/04 8 (D-1) 9/8/03 4 (D-2) 4/4/01 6 (B-1) 5/10/02 7 (D-3) 9/8/03 7 (B-2) 4/4/01 8 (D-4) 9/30/03 9 (B-3) 4/4101 8 (D-5) 4/22/04 7 (B4) 414/01 8 (D-6) 414101 3 (8-5) 414/01 (D-7) 4/4101 2 6

(B-6) 4/22/04 6 (D-8) 9/8103 7 (B-7) 4/22/04 7 (D-9) 1/14/02 7 (B-8) 9/8/03 6 (B-9) 4/4101 5 J _________________

_______________________ I A A-

ANNUNCIATOR RESPONSE PROCEDURES ALARM WINDOW ENGRAVINGS AND GRID LAYOUT IC03A I 2 3 4 5 6 7 8 9 A FUEL POOL EXHAUST POST TREAT PRETREAT OFFGAS VENT PIPE ADS ADS ADS A' CORE SPRAY 'A' CORE SPRAY RIS-4131AAB RM-4101AN3 RM-4104 RM-411OAB -A/B 2MINTIMER(S) RXLO-LO-LO LEVEL LO WATER LEVEL SYSTEM PUMP IP-211A HI-HI RAD HI-HI-HI RAD OR HI-HI RAD HI-HI RAD INITIATED AND CONFIRMED AUTO INITIATED TRIP OR RAD MONITOR INOP CORE SPRAY OR MOTOR OVERLOAD RHR PUMP RUNNING B FUEL POOL EXHAUST POST TREAT PRETREAT OFFGAS VENT PIPE ADS ADS ADS *A'CORE SPRAY *A OR B CORE SPRAY RR-4131 RR-4101 RM-4104 RM-4118AIB BOTH TIMERS CORE SPRAY OR CORE SPRAY OR DISCHARGE LINE 125 VDC HI RAD HI-HI RAD HI RAD HI RAD LOCKED OUT RHR PUMP RUNNING RHR PUMP RUNNING LO PRESSURE LOGIC POWER PERMISSIVE PERMISSIVE FAILURE C FUEL POOL EXHAUST POST TREAT PRETREAT OFFGAS VENT PIPE SRV/SV ADS/LLS SRV A CORE SPRAY *A' CORE SPRAY RIS4131AIB RM-4101A/B RM-4104 RM-411eANB TAILPIPE 125 VDC BELLOWS FAILURE SPARGER DISCHARGE LINE RAD MONITOR HI RAD RAD MONITOR RAD MONITOR HI PRESS OR CONTROL POWER LO AP HI PRESSURE DNSCLnNOP DNSCLItNOP ONSCLANOP HI TEMP FAILURE O POST TREAT POST TREAT PRETREAT OFFGAS VENT PIPE LLS ADS ADS LIQUID RBCCW OFFGAS SAMPLE RM4101A/B SAMPLE FLOW SAMPLE *A OR *B IN TEST STATUS BOTH TEST JACKS RAD MONITORS RM4820 HULO FLOW DOWNSCALE TROUBLE HU/LO FLOW ARMED INSTALLED DNSCLtlNOP HI RAD I~ ~ I.

(U1) 0~

to O5 -

10

ANNUNCIATOR PANEL: IC03A COORDINATES: A-:

REVISION: 8 DATE: 9/8/03 PAGE: 1 of 2 FUEL POOL EXHAUST RIS-4131A/B HI-HI RAD TITLE: FUEL POOL EXHAUST HIGH-HIGH RADIATION (RIS-4131A S) 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) / SETPOINT(S) 1.1 Fuel Pool Exhaust Relay D1 1-K2 via < 9 mR/hr (inc)

Radiation high-high RIS-4131A 1.2 Fuel Pool Exhaust Relay D11-K2 via < 9 mR/hr (Inc)

Radiation high-high RIS-4131B 2.0 AUTOMATIC ACTIONS 2.1 A[B] SBGT starts.

2.2 A Group 3 A[B] isolation occurs.

3.0 OPERATOR ACTIONS 3.1 Verify the AUTOMATIC ACTIONS have occurred. If the specified automatic response has failed to occur, manually initiate the appropriate actions. Verify completion of the Group 3 Isolation by one of the following means:

a. Verify that the individual valves, fans and dampers from Section 2.0 AUTOMATIC ACTIONS have shifted to isolation status, OR

b. On the PCIS Status Board, verify that the green ALL VALVES CLOSED lights are on to coincide with the amber ISOLATION SIGNAL lights that are on, allowing time for the valves to close, OR

c. Check the CIMS PRINTER PRT-1 which will print out Group 3 valves, fans and dampers which fail to shift to isolation status on an isolation signal.

3.2 If Refueling or Loading a Dry Spent Fuel Cask - stop Refueling or Loading Spent Fuel and evacuate the Refuel Floor.

3.3 At 1C02, confirm Fuel Pool Exhaust High Rad condition on REFUEL POOL EXHAUST VENT RAD MONITOR RR-4131A/B.

3.4 At 1C36, confirm Fuel Pool Exhaust High Rad on FUEL POOL EXHAUST RADIATION MONITOR RIS-4131ANB.

(Continued)

ANNUNCIATOR PANEL: IC03A COORDINATES: A-1 REVISION: 8 DATE: 9/8103 PAGE: 2 of 2 l

3.0 OPERATOR ACTIONS (Continued) 3.5 At 1C11, monitor the following:

a. RI-9153, NEW FUEL VAULT AREA MONITOR

b. RI-9163, NORTH REFUEL FLOOR MONITOR

c. RI-9164, SOUTH REFUEL FLOOR MONITOR

d. RI-9178, SPENT FUEL POOL AREA MONITOR 3.6 If an actual High Rad condition exists, send an Operator and a H.P. Technician I

to investigate the problem.

3.7 If Fuel Pool Exhaust is confirmed > 9 mR/hr, enter EOP 3 (Secondary Containment Control).

3.8 If due to a Fuel Pool low level condition, refer to ARP IC04B. A-4 FUEL POOL HVILO LEVEL.

3.9 If the FUEL POOL EXHAUST RADIATION MONITOR RIS-4131AIB Logic is in test status, reset RIS-4131ANB, Group 1II,SBGT and PCIS Logic when the test sianal has cleared.

4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checkedfrepaired and comply with the Technical Specification requirements for Primary Containment Isolation Instrumentation and Secondary Containment Isolation Instrumentation.

5.0 REFERENCES

5.1 APED-DI 1-002 <2A,7>

5.2 BECH-M141, BECH-M176 5.3 EOP3 5.4 01 879.1 5.5 ARP IC04B, A-4 5.6 DAEC Technical Specifications 5.7 MM 132, MM 141, MM 164, MM 140 5.8 DCP 1450 5.9 Tech. Spec. Amendment 193 5.10 CAL-E94-013 5.11 ECP 1646

ANNUNCIATOR PANEL: IC03A COORDINATES: A-2 REVISION: 7 DATE: 414101 PAGE: 1 of 2 TITLE: POST TREATMENT OFFGAS HIGH-HIGH-HIGH RADIATION OR INOPERATIVE (RM-4101A/B) 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) I SETPOINT(S) 1.1 Post Treatment Offgas Relay DII-K18 6xIO 5CPS (inc)

Radiation RM-4101A Hi-Hi-Hi 1.2 RM-4101A inoperative Relay D11-K15 Mode Switch in other than OPERATE, Module unplugged, High Voltage low.

1.3 Post Treatment Offgas Relay D11-K18 6 x 105 CPS (inc)

Radiation RM-4101B Hi-Hi-Hi 1.4 RM-4101 B inoperative Relay D 1-K15 Mode Switch in other than OPERATE, Module unplugged, High Voltage low.

2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 At ICIO, monitor RM-4101AIB to determine which Radiation Monitor is Hi-Hi-Hi or INOP and determine if the other Channel is in or near a tripped condition.

3.2 At 1C02, monitor OFFGAS POSTTREAT RAD MONITOR RR-4101 to confirm RM-4101AJB Hi-Hi-Hi or INOP condition.

3.3 On PPC, check Kaman System Monitor #10 to determine Offgas Stack Radiation trend.

(continued)

ANNUNCIATOR PANEL: 1C03A COORDINATES: A-2 REVISION: 7 DATE: 414101 PAGE: 2 of 2 3.0 OPERATOR ACTIONS (continued) 3.4 Ifan actual Hi-Hi-Hi Offgas Radiation Condition exists, perform the applicable section of AOP 672.2 (Offqas Radiation/Reactor Coolant-High Activity).

a. Notify the Radiation Protection Department to perform a "dose assessment,"

if applicable.

b. Verify the Offgas charcoal adsorbers in service per 01 672 (Off-gas and Recombiner System).

3.5 Refer to EPIP, for EAL Classification, if applicable.

3.6 Ifthe RM-4101A/B is in test status, reset the RM-4101AIB when the test signal has cleared.

4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the ODAM requirements for Radiation Monitoring.

I 4.2 Reset the affected RM-4101 A/B per O0879.1 (Process Radiation Monitoring System) after repairs have been completed.

5.0 REFERENCES

5.1 APED-D1 1-002 <2,4A>

5.2 BECH-M014 5.3 AOP 672.2 5.4 01 879.1, 01 672 5.5 ARP 1C07B, D-9 5.6 EPIP 1.1 5.7 DAEC Technical Specifications 5.8 MM 132, MM 131, MM 139 5.9 QA Audit Report 1-87-10 5.10 DCP 1451 5.11 Tech. Spec. Amendment 193 5.12 EMA A44812

ANNUNCIATOR PANEL: IC03A COORDINATES: A-3 REVISION: 7 DATE: 414101 PAGE: 1 of 2 PRETREAT RM-4104 HI-HI RAD TITLE: PRETREATMENT OFFGAS SYSTEM (RM-4104) HI-HI RADIATION 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I SETPOINT(S)

• 1.1 Pretreatment Offgas Relay D1I-K6 via Variable System activity RM-4104 high-high NOTE Setpoints for Pretreat Radiation Monitor:

a) High Trip - Set at a level corresponding to noble gas release rate, after 30 minute delay and decay of 0.16 Ci/sec.

b) HIGH-HIGH Trip - Set at a level corresponding to noble gas release rate, after 30 minute delay and decay of 1 Ci/sec.

2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 At 1C06, monitor Offgas System flow on FR- 374 for abnormalities.

3.2 At I C02, monitor the following recorders for indication of Off-Gas Pretreat high activity:

a. RR-4104, Off-Gas Pretreatment Rad Monitor

b. RR-4448, Main Steam Line Radiation Monitor 3.3 At 1C10, monitor RM-4104, Off-Gas Pretreat Radiation Monitor for indication of high activity.

3.4 At 1C36, monitor Main Steam Line Radiation on Main Steam Line Radiation Monitors, RM4448A/B/C/D.

(continued)

ANNUNCIATOR PANEL: 1C03A COORDINATES: A-3 REVISION: 7 DATE: 414101 PAGE: ' 2 of 2 3.0 OPERATOR ACTIONS (Continued) 3.5 On PPC, check Kaman System Monitor #10 to determine Offgas Stack Rad trend.

3.6 If Offgas Activity and Main Steam Line Activity are trending upward at a fast rate, perform AOP 672.2 (Offcias Radiation/Reactor Coolant High Activity).

3.7 Have Chemistry verify that the Offgas equivalent noble gas activity at the pretreat rad monitor is

• 1.0 Ci/sec. If equivalent noble gas activity is > 1.0 Ci/sec, perform the Tech Spec required actions for Main Condenser Offgas activity.

3.8 Refer to EPIP Section 1.1 for EAL Classification, if applicable.

3.9 If the OFFGAS PRETREAT RAD MONITOR RM-4104 is in test status, reset RM-4104 trip Logic when the test signal has cleared.

4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the ODAM requirements for Radiation Monitoring.

.1 4.2 Reset RM-4104 per 01 879.1 (Process Radiation Monitoring System) when repairs have been made.

5.0 REFERENCES

5.1 APED-Di 1-002, APED-DI 1-004 5.2 BECH-M141 5.3 01672,01879.1 5.4 AOP 672.2 5.5 EPIP 1.1 5.6 DAEC Technical Specifications 5.7 MM132,MM131,MM139 5.8 DCP 1451 5.9 Tech. Spec. Amendment 193 5.10 EMA A44812

ANNUNCIATOR PANEL: IC03A COORDINATES: A-4 REVISION: 10 DATE: -- 414101 PAGE: I of 3 OFFGAS VENT PIPE RM-411 6AB HI-HI RAD TITLE: OFFGAS VENT PIPE HIGH-HIGH RADIATION (RM-4116AIB) 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I SETPOINT(S) 1.1 Offgas Vent Pipe Relay A71 B-K75 via 6.0 X 105 CPS Radiation High-High RM-4116A 1.2 Offgas Vent Pipe Relay A71B-K76 via 6.0 X 105CPS Radiation High-High RM-41 16B 2.0 AUTOMATIC ACTIONS 2.1 Group 3 A Isolation valves close via OFFGAS STACK VENT PIPE A RAD MONITOR RM-4116A.

2.2 Group 3 B Isolation valves close via OFFGAS STACK VENT PIPE B RAD MONITOR RM-4116B.

2.3 The SBGT starts.

3.0 OPERATOR ACTIONS 3.1 Verify completion of the Group 3 Isolation by one of the following means:

a. Verify that the individual valves, fans and dampers from Section 2.0 AUTOMATIC ACTIONS have shifted to isolation status, OR

b. On the PCIS status board, verify that all green ALL VALVES CLOSED lights are on to coincide with the amber ISOLATION SIGNAL lights that are on, allowing time for the valves to close, OR

c. Check the CIMS PRINTER PRT-1 which will print out Group 3 valves, fans and dampers which fail to shift to isolation status on an isolation signal.

3.2 At 1C02, confirm Offgas High High Radiation condition on OFFGAS STACK VENT PIPE RAD MONITOR RR-4116.

3.3 At I C1 0, confirm Offgas High High Radiation condition on OFFGAS STACK VENT PIPE RAD MONITOR RM-4116A[B].

3.4 Check Offgas System operation on 1C34.

(Continued)

ANNUNCIATOR PAN!EL: IC03A COORDINATES: A-4 REVISION: 10 DATE: 414101 PAGE: 2 of 3 3.0 OPERATOR ACTIONS (continued) 3.5 Check Kaman System Monitor #10 to determine Offgas Stack Rad trend.

3.6 At 1C36, check A[B, C, D] MAIN STEAM LINE RADIATION MONITOR RM-4448A, 4448C, 4448B and RM-4448D for signs of fuel failure.

3.7 Notify the Radiation Protection Department to perform a "dose assessment."

3.8 Refer to EPIP for EAL classification, if applicable.

3.9 If the Off-Gas Vent Pipe RM-4116A/B Logic is in test status, reset Off Gas Vent Pipe RM41 16A/B Logic when the test signal has cleared.

3.10 RM-4116A and RM-4116B are required to be operable during venting or purging any time when primary containment integrity is required. (Technical Specifications for Primary Containment Isolation Instrumentation) 3.11 Ifeither RM-41 16A or RM-41 16B becomes inoperable during conditions which require them to be operable, restore the inoperable channel within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> close the primary vent and purge valves or establish administrative control of those valves with continuous monitoring of alternate instrumentation.

3.12 Ifboth RM-4116A and RM-4116B become inoperable during conditions which require them to be operable, restore isolation capability within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If isolation capability is not restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, then close the primary vent and purge valves or establish administrative control of those valves with continuous monitoring of alternate instrumentation within the next hour.

4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the Technical Specification requirements for Primary Containment Isolation I

Instrumentation.

4.2 Administrative control requires (1) an Operator is stationed at the valve controls, and (2) that Operator is instructed to terminate venting or purging when procedures direct valve closure.

ANNUNCIATOR PANEL: IC03A COORDINATES: A4 REVISION: 10 DATE: 414101 PAGE: 3 of 3

5.0 REFERENCES

5.1 APED-D11-002 <2,5>

5.2 BECH-M141 5.3 01879.1 5.4 ODAM 617 5.5 EPIP 1.1 5.6 DAEC Technical Specifications 5.7 DCR 1270 5.8 MM 132, MM 164 5.9 QA Audit Report 1-87-10 5.10 DCP 1451 5.11 Tech. Spec. Amendment 184, 193, 209 5.12 Rad Engineering Calculation No. 95-OIOC 5.13 DDC 3200

ANNUNCIATOR PANEL: 1C03A COORDINATES: A-5 REVISION: 7 DATE: 4122/04 PAGE: 1 I of 2 ADS "ANB" 2 MIN TIMER(S)

INITIATED TITLE: AUTOMATIC DEPRESSURIZATION SYSTEM TIMERS INITIATED 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) / SETPOINT(S) 1.1 ADS Channel A Timer Relay B21 C-K5A for Energized I KS4400A initiated on 120 Sec Timer 1C03 KS-4400A 1.2 ADS Channel B Timer KS4400B initiated on Relay B21 C-K5B for 120 Sec Timer Energized I IC03 KS-4400B 2.0 AUTOMATIC ACTIONS 2.1 ADS timer indicating light(s) at 1C03 energize and timer(s) KS-4400A and/or KS-4400B start timing out. If after 120 seconds, timer(s) have not been RESET or OVERRIDDEN, ADS Relief Valves PSV-4400, 4402,4405 and/or PSV-4406 open to blowdown the Reactor Pressure Vessel.

NOTE ADS valves PSV-4400, PSV-4402, PSV-4405, and PSV-4406 open 120 seconds after the ADS "A/1B 2 MIN TIMER(S) INITIATED (1C03A, A-5) annunciator activates if Reactor water level remains less than +64 inches and either an RHR Pump is operating with a discharge pressure of 125 psig or a Core Spray Pump isoperating with a discharge pressure of 145 psig and confirmatory low Reactor level of +170 inches is present.

3.0 OPERATOR ACTIONS 3.1 Determine if the ADS timers should be reset, overridden, or allowed to time out per the EOPs.

3.2 Ifthe ADS Initiation Logic Channel A or B is in test status, reset the ADS Initiation Logic when the test signal has cleared.

ANNUNCIATOR PANEL: 1C03A COORDINATES: A-5 REVISION: 7 DATE: 4122/04 PAGE: 2 of 2 4.0 SUPPLEMENTAL ACTIONS 4.1 If the alarm is a result of failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the Technical Specification requirements for Emergency Core Cooling System (ECCS) Instrumentation

5.0 REFERENCES

5.1 APED-B21-018 5.2 IPOI 4, IPOI 5 5.3 EOP 1 5.4 DAEC Technical Specifications 5.5 MM 132 5.6 DCP 1410, DCP 1451, DCP 1475 5.7 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: IC03A COORDINATES: Ax REVISION: 6 DATE: 4122104 PAGE: 1 of 3 ADS RX LO-LO-LO LEVEL AND CORE SPRAY OR RHR PUMP RUNNING TITLE: AUTOMATIC DEPRESSURIZATION SYSTEM RELAYS ENERGIZED 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) I SETPOINT(S) 1.1 ADS Initiation Logic Relay B21 C-K7A Energized I Channel A energized due to:

a. Core Spray Pump Relay B21C-K9A via 145 psig (inc)

I P-211 A running E21A-K23A and OR PS-2107A

b. RHR Pump 1P-229A Relay B21C-K9A via 125 psig (inc) or C running EI1A-KI01A and AND PS-2023A or PS-2024A

c. Reactor Low-Low-Low LIS4533 +63.8 inch (dec) water level AND

d. A LOGIC ADS TIMER HS-4461 Not depressed RESET pushbutton on or Panel 1C03 Overridden (Continued)

ANNUNCIATOR PANEL: IC03A COORDINATES: A-6 REVISION: 6 DATE: 4122104 PAGE: 2 of 3 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) I SETPOINT(S)

.(Continued) 1.2 ADS Initiation Logic Relay B21 C-K7B Energized I Channel B energized due to:

a. Core Spray Pump Relay B21 C-K9B via 145 psig (inc) 1P-211 B running E21A-K23B and OR PS-2127A

b. RHR Pump 1P-229B Relay B21 C-K9B via 125 psig (inc) or D running ElA-K10B and AND PS-1 917A or PS-I 925A

c. Reactor Low-Low-Low LIS-4534 +63.8 inch(dec) water level AND

d. B LOGIC ADS TIMER HS-4462 Not depressed RESET pushbutton on or Panel 1C03 Overridden 2.0 AUTOMATIC ACTIONS 2.1 ADS 120 second timers KS-4400A and/or KS4400B start timing out and after 120 sec, if not reset or overridden, ADS Relief Valves SV-44001440214405/4406 open to blowdown the Reactor Pressure Vessel.

NOTE ADS valves PSV-4400, PSV-4402, PSV-4405, and PSV4406 open 120 seconds after the ADS "ANB" 2 MIN TIMER(S) INITIATED (1C03A, A-5) annunciator activates if Reactor water level remains less than +64 inches and either an RHR Pump is operating with a discharge pressure of 125 psig or a Core Spray Pump is operating with a discharge pressure of 145 psig and confirmatory low Reactor level of +170 inches is present.

ANNUNCIATOR PANEL: IC03A COORDINATES: A-6 REVISION: 6 DATE: 4122/04 PAGE: 3 of 3 3.0 OPERATOR ACTIONS 3.1 Determine if ADS timers should be reset, overridden, or allowed to time out per the EOPs.

3.2 If the ADS Initiation Logic is in test status, reset ADS Initiation Logic when the test signal has cleared.

4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the Technical Specification requirements for Emergency Core Cooling System (ECCS) Instrumentation.

5.0 REFERENCES

5.1 APED-B21-018, APED-El 1-007, APED-E21-006 5.2 IPOI 4, IPOI 5 5.3 EOP 1, EOP 2 5.4 DAEC Technical Specifications 5.5 MM 132 5.6 DCP 1410, DCP 1451 5.7 CAL-E92-006 5.8 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: 1C03A COORDINATES: A-7 REVISION: 4 DATE: 414/01 PAGE: I of 2 ADS LO WATER LEVEL CONFIRMED TITLE: AUTOMATIC DEPRESSURIZATION SYSTEM LOW WATER LEVEL CONFIRMED 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I SETPOINT(S) 1.1 ADS Reactor Low Water LIS-4561 +170 inches (dec)

Level Channel A tripped 1.2 ADS Reactor Low Water LIS-4562 +170 inches (dec)

Level Channel B tripped 2.0 AUTOMATIC ACTIONS 2.1 If simultaneous with Reactor Low-Low-Low water level signal, the 120 Second ADS Timer starts timing out.

2.2 If the ADS permissive signals of 2.1 above do not exist, no AUTOMATIC ACTIONS occur.

3.0 OPERATOR ACTIONS 3.1 Verify the AUTOMATIC ACTIONS have occurred. If the specified automatic response has failed to occur, manually initiate the appropriate actions.

3.2 If Reactor water level is below 170", enter EOP I (RPV Control).

3.3 If the ADS Initiation Logic is in test status, reset ADS Initiation Logic when the test signal has cleared.

ANNUNCIATOR PANEL: IC03A COORDINATES: A-7 REVISION: 4 DATE: 414101 PAGE: 2 of 2 4.0 SUPPLEMENTAL ACTIONS 4.1 If a Plant emergency condition does not exist as determined in Step 3.2 above, perform Reactor Recovery as appropriate per IPOI 4 (Shutdown) or IPOI 5 (Reactor Scram).

4.2 If the cause of the alarm is due to a failed sensor/logic channel, initiate a Work Request Card to check/repair the faulty sensor/logic channel and comply with the Technical Specification requirements for Emergency Core Cooling System (ECCS) Instrumentation.

5.0 REFERENCES

5.1 APED-B21-018 5.2 IPOI 4, IPOI 5 5.3 EOP 1 5.4 DAEC Technical Specifications 5.5 MM 132 5.6 DCP 1451 5.7 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: IC03A COORDINATES: AZ REVISION: 4 DATE: 414101 PAGE: I of 2 "A" CORE SPRAY SYSTEM AUTO INITIATED TITLE: CORE SPRAY SYSTEM "A" ACTUATED 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) / SETPOINT(S) 1.1 Core Spray System A Relay E21-K12A via +63.8" (dec)

Logic Reactor Water LIS-4531, 4533,4532,4534 Level Low initiated arranged in one-out-of-two-twice logic 1.2 Core Spray System A Relay E21-K12A via 2.0 psig (inc)

Logic Drywell High PS-4310B, 43128, 4311B, Pressure initiated 4313B arranged in one-out-of-two-twice logic 2.0 AUTOMATIC ACTIONS 2.1 A CORE SPRAY PUMP 1P-211A starts and the Core Spray System A lines up for injection.

3.0 OPERATOR ACTIONS 3.1 Verify the AUTOMATIC ACTIONS have occurred. If the specified automatic response has failed to occur, manually initiate the appropriate actions.

3.2 If Drywell pressure is 2.0 psig or greater, enter EOP 1 (RPV Control) and EOP 2 (Primary Containment Control).

3.3 If Reactor water level is less than 170 inches, enter EOP 1 (RPV Control).

3.4 If the Core Spray System A Initiation Logic is in test status, reset the Core Spray System A initiation Logic when the test signal has cleared.

ANNUNCIATOR PANEL: IC03A COORDINATES: A-8 REVISION: 4 DATE: 414101 PAGE: 2 of 2 4.0 SUPPLEMENTAL ACTIONS 4.1 When the Core Spray System A is no longer required for operation as determined in Step 3.2 above, return the Core Spray System A to standby readiness per 01 151 (Core Spray System).

4.2 If the cause of the alarm was due to a failed sensor/trip channel, initiate a Work Request Card to have that sensor/trip channel check/repaired and comply with the Technical Specification requirements for Emergency Core Cooling System (ECCS) Instrumentation.

5.0 REFERENCES

5.1 APED-E21-006 <2,3>

5.2 BECH-E121 <3>

5.3 EOP 1, EOP 2 5.4 01151 5.5 DAEC Technical Specifications 5.6 MM 132 5.7 DCP 1451 5.8 CAL-E92-006 5.9 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: IC03A COORDINATES: A-9 REVISION: 8 DATE: 4122/04 PAGE: 1 of 3

"'A" CORE SPRAY PUMP I P-211 A TRIP OR MOTOR OVERLOAD TITLE: "A" CORE SPRAY PUMP 1P-21 IA TRIP OR MOTOR OVERLOAD CONDITION 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) / SETPOINT(S) 1.1 Core Spray Pump Relay 174 via Ranges from 120 amps 1P-21 1A inverse time delay overcurrent Relay 304-151 for > 230 seconds to 170 amps for > 63 I

seconds 1.2 Core Spray Pump Bkr. Aux. Contacts 1200 Amps (inc)

I P-211A Motor 152-304/b via Instantaneous Relay 304-150 overcurrent 1.3 Core Spray Pump Bkr. Aux Contacts Ground fault detected I P-21 IA Motor ground 152-304/b via fault Relay 304-15OG 1.4 Bus 1A3 Undervoltage Relay 194-31 via 20% Bus Voltage (dec)

Relay 127-3 1.5 Breaker 1A304 racked-out Aux Contact Not in Operating I 304-151POS Position 1.6 a. CS Suction Valves Relay K-26A via Energized MO-2100 or MO-2147 LS-3 (MO-2100) or start to cdose LS-3 (MO-2147)

AND

b. HS-2103A out of HS-2103A contacts Normal position BYPASS position 1.7 a. Attempt to start Core HS-2003 START and NORMAL Spray Pump 1P-211A AFTER START AND AND

b. Breaker 1A304 fail Breaker. Aux. Contacts Closed to close 152-304/b 1.8 a. Automatic initiation Relay 74-K0304 5.25 seconds after signal initiation and power AND available

b. Breaker 1A304 open Breaker Aux. Contacts Closed 152-304/b

ANNUNCIATOR PANEL: IC03A COORDINATES: A-9 REVISION: 8 DATE: 4122/04 PAGE: 2 of 3 2.0 AUTOMATIC ACTIONS 2.1 If due to CAUSE 1.1, A Core Spray Pump I P-21 1A will trip after time delay.

Time delay ranges from 230 seconds at 120 amps to 63 seconds at 170 amps.

2.2 If due to CAUSES 1.2 through 1.6 or 1.8, A CORE SPRAY PUMP 1P-21 IA trips.

2.3 If due to CAUSE 1.2 or 1.3, Pump 1P-21 IA Lockout Relay on Bkr 1A304 also trips.

2.4 If due to CAUSE 1.7, Pump IP-211A does not start.

3.0 OPERATOR ACTIONS 3.1 At 1C03, confirm Pump IP-211A trip by monitoring A CORE SPRAY PUMP DISCH PRESSURE INDICATOR P1-2106 and checking 1P-21 IA amps.

3.2 If 1P-211A is needed to supply adequate core cooling and has tripped, attempt to restart I P-21 IA. If unable to restart, refer to EOP 1 to maintain RPV level.

,. =11na,.

Do not reset Pump I P-211AIAockout relay without permission from the CRS.

3.3 Determine the cause of the Pump 1P-21 1A trip by sending an Operator to Bus 1A3 to check Bkr 1A304 targets displayed and/or the position of Bkr 1A304.

3.4 If due to an overload non-trip condition only, at 1C03, perform the following:

a. Check amps on A CORE SPRAY PUMP IP-211A to ensure that alarm is not malfunctioning.

b. Confirm amps > 120 on Ammeter labeled A CORE SPRAY PUMP I P-21 1A

c. Reduce Pump 1P-211A amps to < 120 by reducing pump flow and check that the alarm resets.

d. If flow cannot be reduced or the alarm does not clear and 1P-21 1A is not required to ensure adequate core cooling, then remove CS Pump 1 P-21 IA from service.

ANNUNCIATOR PANEL: JC03A COORDINATES: A-9 REVISION: 8 DATE: 4/22104 PAGE: 3 of 3 4.0 SUPPLEMENTAL ACTIONS 4.1 If Pump 1P-211A was not tripped and rated flow could not be attained at < 120 amps, send an Operator to 1P-211A with a portable vibration monitor to check the following:

a. Abnormal noise or vibration (i.e., 2 3 mils or 1.0 in/sec velocity).

b. Check for abnormal leakage, motor overheating and/or valve misalignment.

4.2 Comply with the Technical Specification requirements for ECCS - Operating or ECCS - Shutdown.

4.3 If Pump 1P-21 IA tripped due to any electrical fault, initiate a Work Request Card to have the pump checked, meggered, and repaired as necessary.

4.4 Once the cause of the Pump I P-21 1A trip has been determined and corrected, return CS Pump 1P-211A to service per 01 151 (Core Spray System).

5.0 REFERENCES

5.1 APED-E31-006 <2>

5.2 BECH-E104 <25,26>

5.3 BECH-E121 <3,3A,41,41A>

5.4 BECH-E511 <3,12N>

5.5 BECH-M121 5.6 EOP 1 5.7 01151 5.8 DAEC Technical Specifications 5.9 DCP 1355, DCP 1451 5.10 MM 132 5.11 AR 971153.01

ANNUNCIATOR PANEL: IC03A COORDINATES: B-1 REVISION: 7 DATE: 5110102 PAGE: I of 2 FUEL POOL EXHAUST RR-4131 Hi RAD TITLE: FUEL POOL EXHAUST HIGH RADIATION (RR-4131) 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I SETPOINT(S) 1.1 Fuel Pool Exhaust RR-4131A via Red Pen 4 mRlhr (inc)

Radiation High 1.2 Fuel Pool Exhaust RR-4131B via Black Pen 4 mR/hr (ind)

Radiation High 2.0 AUTOMATIC ACTIONS 2.1 None; however, if radiation continues to increase, a SBGT System initiation and a Group IlIl isolation occur.

3.0 OPERATOR ACTIONS 3.1 Perform the following:

a. At 1C02, confirm Fuel Pool Exhaust High Rad condition on RR-4131.

b. At 1C36, confirm Fuel Pool Exhaust High Rad on FUEL POOL EXHAUST RADIATION MONITOR RIS-4131ANB.

c. At I C11, monitor the following:

1) RI-9153, NEW FUELVAULTAREAMONITOR

2) RI-9163, NORTH REFUEL FLOOR MONITOR

3) RI-9164, SOUTH REFUEL FLOOR MONITOR

4) RI-9178, SPENT FUEL POOL AREA MONITOR

d. Check Kaman System for Reactor Building Radiation Release trends.

3.2 If actual High Rad condition exists, send an Operator and a H.P. Technician to investigate the problem.

3.3 If due to a low Fuel Pool level, refer to ARP 1C04B, A4 (FUEL POOL HIILO LEVEL).

3.4 If due to Dry Fuel Storage activities including the discharge of exhaust from the blowdown and vacuum drying of the DSC into the fuel pool, evaluate the situation and take appropriate remedial action.

3.5 If the Fuel Pool Exhaust RR4131 logic is in test status, reset Fuel Pool Exhaust RR-4131 Logic when the test signal has cleared.

ANNUNCIATOR PANEL: IC03A COORDINATES: B-1 REVISION: 7 DATE: 5/10/02 PAGE: 2 of 2 4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the Technical Specification requirements for Primary Containment Isolation Instrumentation and Secondary Containment Isolation Instrumentation.

5.0 REFERENCES

5.1 APED-D11-002 <2A,7>

5.2 BECH-M141, BECH-M176 5.3 ARP 1C04B, A-4 5.4 DAEC Technical Specifications 5.5 MM 132, MM 140 5.6 DCP 1450 5.7 Tech. Spec. Amendment 193 5.8 ECP 1646

ANNUNCIATOR PANIFL: _IC03A COORDINATES: B-2 REVISION: 8 DATE: 4141/01 PAGE: 1 of 2 POST TREAT RR-4101 HI-HI RAD TITLE:. POSTTREATMENTOFFGAS HIGH-HIGH RADIATION (RR-4101) 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) / SETPOINT(S) 1.1 Post Treatment Offgas RR-4101 (Red Pen) 5 CPS (inc) 1 X lo0 Radiation high (Red Pen) 1.2 Post Treatment Offgas RR-4101 (Black Pen) I X 105CPS (inc)

Radiation high (Black Pen) 2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 At 1C34, monitor Offgas System parameters and verify Charcoal Adsorbers in service per O 672 (Offaas and Recombiner System).

3.2 At I C02, confirm Post Treatment Offgas Radiation > 1 X 105 CPS on OFFGAS POSTTREAT RAD RECORDER RR-4101.

3.3 At 1CI0, confirm Post Treatment Offgas Radiation conditions > 1 x 105 CPS on OFFGAS POSTTREAT RAD MONITORs RM-4101AIB.

3.4 Check Kaman System Monitor #10 to determine Offgas Stack Radiation trends on PPC and RR-4176 at 1CCO.

3.5 Notify the Radiation Protection Department to perform a "dose assessment," if applicable.

3.6 Refer to EPIP, Section 1.1 for EAL Classification, if applicable.

3.7 If RM-4101AIB is In TEST status, reset RM-4101A/B Logic when the test signal has cleared.

ANNUNCIATOR PANEL: JC03A COORDINATES: B-2 REVISION: 8 DATE: 414/01 PAGE: 2 of 2 4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the ODAM requirements for Radiation Monitoring.

4.2 Reset the affected RM-4101 ANB per 01 879.1 (Process Radiation Monitoring System) after repairs have been completed.

5.0 REFERENCES

5.1 APED-D1 1-002 <2,4A>

5.2 BECH-M014 5.3 01672,01879.1 5.4 EPIP 1.1 5.5 DAEC Technical Specifications 5.6 MM 132, MM 131, MM 139 5.7 QA Audit Report 1-87-10 5.8 DCP 1451 5.9 Tech, Spec. Amendment 184, 193 5.10 EMA A44812

ANNUNCIATOR PANEL: 1C03A COORDINATES: B-3 REVISION: 8 DATE: 414101 PAGE: -I of 2 PRETREAT RM-4104 HI RAD TITLE: PRETREATMENT (RM-4104) RADIATION HIGH 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) / SETPOINT(S) 1.1 Pretreatment Off Gas Relay D11-K10 Variable System Activity high via RM-4104 NOTE Setpoints for Pretreat Radiation Monitor.

a) High Trip - Set at a level corresponding to noble gas release rate, after 30 minute delay and decay of 0.16 Ci/sec.

b) HIGH-HIGH Trip - Set at a level corresponding to noble gas release rate, after 30 minute delay and decay of 1 Cilsec.

2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 At 1 C02, monitor the following recorders for indication of Offgas Pretreatment high activity:

AK RR-4104, OFFGAS PRETREAT RADIATION MONITOR B. RR-4448, MAIN STEAM LINE RADIATION MONITOR 3.2 At 1C10, monitor RM-4104, OFFGAS PRETREATMENT RADIATION MONITOR for indication of high activity.

3.3 At 1C36, monitor Main Steam Line radiation on MAIN STEAM LINE RADIATION MONITORS, RM-4448AIB/C1D.

3.4 On PPC, check Kaman System Monitor #10 for Offgas Stack Radiation trend.

3.5 If Offgas Activity and Main Steam Line Activity are trending upward at a fast rate perform, AOP 672.2 (Offqas RadiationlReactor Coolant High Activity).

3.6 If due to actual increase in offgas activity, reference Tech. Spec. Surveillance Requirement for Main Condenser Offgas which may require performance of an STP if activity has increased by > 50%.

continued

ANNUNCIATOR PAN EL: IC03A COORDINATES: B-3 REVISION: 8 DATE: -141/401 PAGE: 2 of 2 3.0 OPERATOR ACTIONS (continued) 3.7 Have Chemistry verify that the Offgas equivalent noble gas activity at the Pretreatrment Radiation Monitor is

• 1.0 Ci/sec. If equivalent noble gas activity is > 1.0 Cisec, perform the Technical Specification required actions for Main Condenser Offgas activity.

3.8 Refer to EPIP for EAL classification, if applicable.

3.9 If the Off-Gas Pretreatment RM-4104 is in test status, reset RM-4104 trip Logic when the test signal has cleared.

4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the ODAM requirements for Radiation Monitoring.

I 4.2 Reset RM-4104 per 01 879.1 (Process Radiation Monitorinq System) when repairs have been made.

5.0 REFERENCES

5.1 APED-DI 1-002, APED-DI 1-004 5.2 BECH-M141 5.3 01672,01879.1 5.4 EPIP 1.1 5.5 AOP 672.2 5.6 DAEC Technical Specifications 5.7 MM 132, MM 131, MM 139 5.8 DCP 1451 5.9 Tech. Spec. Amendment 193

ANNUNCIATOR PANI EL: 1C03A COORDINATES: B-4 REVISION: 8 DATE: 414101 PAGE: I of 2 OFFGAS VENT PIPE RM-4116AIB Hi RAD TITLE: OFFGAS VENT PIPE HIGH RADIATION (RM-4116A/B) 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) / SETPOINT(S) 1.1 Offgas Vent Pipe Relay D11 -K1 3 via 1.0 X 10 4 CPS (inc)

Radiation high RM-4116A 1.2 Offgas Vent Pipe Relay D11-K13 via 1.0 X 104 CPS (inc)

Radiation high RM-4116B 2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 Perform the following:

a. At 1C02, confirm Offgas High Radiation condition on OFFGAS STACK VENT PIPE RAD MONITOR RR-4116.

b. At 1C1 0, confirm Offgas High Radiation condition on OFFGAS STACK VENT PIPE A[B] RAD MONITOR RM-4116A[B].

c. Check Kaman System Monitor 10 for Offgas Stack Radiation trend.

d. At 1C34, monitor Offgas System parameters.

e. If Offgas activity is trending upward at a fast rate, perform the appropriate section of AOP 672.2 (Offoas Radiation/Reactor Coolant High Activity).

3.2 If a slow increase in Offgas activity is noted, send an Operator to place Standby Offgas After Filter I F-214ANB in service per 01 672 (Offcias and Recombiner System).

3.3 If the OFFGAS STACK VENT PIPE ANB] RAD MONITOR RM-4116AfB logic is in test status, reset OFFGAS STACK VENT PIPE A[B] RAD MONITOR RM-4116A/B Logic when the test signal has cleared.

3.4 RM-4116A and RM-4116B are required to be operable during venting or purging any time when primary containment Integrity is required. (Technical Specifications for Primary Containment Isolation Instrumentation).

(Continued)

ANNUNCIATOR PANEL: IC03A COORDINATES: B-4 REVISION: 8 DATE: 414101 PAGE: 2 of 2 3.0 OPERATOR ACTIONS (Continued) 3.5 If both RM-4116A and RM-4116B become inoperable during conditions which require them to be operable, restore isolation capability within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If isolation capability is not restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, then close the primary vent and purge valves or establish administrative control of those valves with continuous monitoring of alternate instrumentation within the next hour.

4.0 SUPPLEMENTAL ACTIONS 4.1 Administrative control requires (1)an Operator is stationed at the valve controls and (2) that Operator is instructed to terminate venting or purging when procedures direct valve closure.

4.2 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the Technical Specification requirements for Primary Containment Isolation I

Instrumentation.

4.3 Reset the affected RM-4116 A/B per 01 879.1 (Process Radiation Monitoring System) when repairs have been completed.

5.0 REFERENCES

5.1 APED-D1I-002 <2,5>

5.2 BECH-M141 5.3 01 672, 01 879.1 5.4 AOP 672.2 5.5 DAEC Technical Specifications 5.6 MM 132, MM 131, MM 139 5.7 DCP 1451 5.8 Tech. Spec. Amendment 184, 193, 209 5.9 Rad Engineering Calculation No. 95-OIOC

ANNUNCIATOR PAN EL: IC03A COORDINATES: B-5 REVISION: 5 DATE: 4/4/01 PAGE: A. of 1 ADS BOTH TIMERS LOCKED OUT TITLE: AUTOMATIC DEPRESSURIZATION SYSTEM TIMER LOCKED OUT 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) / SETPOINT(S) 1.1 A[B] LOGIC ADS HS-4461 and HS-4462 Both switches in RESET TIMER RESET switches or OVERRD position on 1C03 NOTE The timer reset pushbuttons will also reclose the ADS valves if ADS has already initiated.

2.0 AUTOMATIC ACTIONS 2.1 ADS Channel A[B] is disabled when A[B] LOGIC ADS TIMER RESET switch I HS-4461[4462] is in the RESET (depressed) or OVERRD (depressed and turned).

3.0 OPERATOR ACTIONS 3.1 Observe that A[B] LOGIC ADS TIMER RESET switches HS-4461 and HS-4462 are either depressed or Inthe OVERRD position.

4.0 SUPPLEMENTAL ACTIONS 4.1 Comply with Technical Specification requirements for Emergency Core Cooling System (ECCS) Instrumentation.

4.2 Return AB] LOGIC ADS TIMER RESET Switches HS4461 and 4462 to normal as soon as permitted by plant conditions.

5.0 REFERENCES

5.1 APED B21-018 <1,2,3>

5.2 EOP1 5.3 DAEC Technical Specifications 5.4 DCP 1268/1 5.5 MM 132 5.6 DCP 1410, DCP 1451 5.7 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: 1C03A COORDINATES: Be REVISION: 6 DATE: 4122104 PAGE: 1 of 2

[ ADS CORE SPRAY OR RHR PUMP RUNNING lEPERMISSIVE TITLE: AUTO DEPRESSURIZATION SYSTEM CORE SPRAY OR RESIDUAL HEAT REMOVAL PUMP RUNNING 1.0 PROBABLE CAUSE(S) INITIATING DEVICE(S) / SETPOINT(S) 1.1 a. ADS Core Spray Relay B21C-K9A via E21A- 145 psig (inc) I Pump 1P-211A K23A and PS-2107A running Channel A OR

b. ADS RHR Pump Relay B21C-K9A via EllA- 125 psig (inc) I 1P-229A or C KI01A and PS-2023A or running Channel A 2024A 1.2 a. ADS Core Spray Relay B21C-K9B via E21A- 145 psig (inc) I Pump I P-211 B K23B and PS-2127A running Channel B OR

b. ADS RHR Pump Relay B21C-K9B via El1A- 125 psig (inc) I I P-229B or D KI01B and PS-1917A or running Channel B PS-1925A 2.0 AUTOMATIC ACTIONS 2.1 None; however, ADS Permissive for Core Spray or RHR pump running is made up.

3.0 OPERATOR ACTIONS 3.1 If a Core Spray or RHR pump is running due to an auto initiation signal, enter EOP 1 (RPV CONTROL).

3.2 If the ADS Core Spray or RHR Pump Running Logic Channel A and/or B is in the test status, reset ADS Core Spray or RHR Running Channel A and/or B Logic when the test signal has cleared.

ANNUNCIATOR PANEL: IC03A COORDINATES: B-6 REVISION: 6 DATE: 4122/04 PAGE: 2 of 2 4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the ADS Logic alarm was due to detector/trip channel logic failure, initiate a Work Request Card to have that detector checked/repaired as necessary and comply with the Technical Specification requirements for Emergency Core Cooling System (ECCS) Instrumentation.

5.0 REFERENCES

5.1 APED-El 1-007; APED-B21-018 5.2 IPOI 4, IPOI 5 5.3 EOP1 5.4 DAEC Technical Specifications 5.5 MM 132 5.6 DCP 1451 5.7 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: IC03A COORDINATES: B-7 REVISION: 7 DATE: 4122104 PAGE: 1 of 2 ADS1 CORE SPRAY OR RHR PUMP RUNNING PERMISSIVE TITLE: AUTO DEPRESSURIZATION SYSTEM CORE SPRAY OR RESIDUAL HEAT REMOVAL PUMP RUNNING 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I SETPOINT(S) 1.1 a. ADS Core Spray Relay B21C-KIOA via 145 psig (inc) I Pump 1P-211A E21A-K025A and running Channel A PS-2107B OR

b. ADS RHR Pump Relay B21C-KIOA via 125 psig (inc) I 1P-229A or C E1IA-K1OIA and running Channel A PS-2023B or 2024B 1.2 a. ADS Core Spray Relay B21 C-K1 OB via 145 psig (inc) I Pump 1P-211 B E21A-K025B and running Channel B PS-2127B OR

b. ADS RHR Pump Relay B21C-K1OB via 125 psig (inc) I IP-229B or D E11A-K1O1B and running Channel B PS-1917B or PS-1 925B 2.0 AUTOMATIC ACTIONS 2.1 None; however, ADS Permissive for Core Spray or RHR pump running is made up.

3.0 OPERATOR ACTIONS 3.1 If a Core Spray or RHR pump Is running due to an auto initiation signal, enter EOP 1 (RPV CONTROL).

3.2 If the ADS Core Spray or RHR Pump Running Logic Channel A and/or B is in test status, reset ADS Core Spray or RHR Running Channel A and/or B Logic when the test signal has cleared.

ANNUNCIATOR PANEL: IC03A COORDINATES: B-7 REVISION: 7 DATE: 4122104 PAGE: 2 of 2 4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the ADS Logic alarm was due to detector/trip channel logic failure, initiate a Work Request Card to have that detector checked/repaired as necessary and comply with the Technical Specifications requirements for Emergency Core Cooling System (ECCS) Instrumentation.

5.0 REFERENCES

5.1 APED-El1-007; APED-B21-018<1>

5.2 IPOI 4. IPOI 5 5.3 EOPI1 5.4 DAEC Technical Specifications 5.5 MM 132 5.6 DCP 1451 5.7 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: IC03A COORDINATES: B-8 REVISION: 6 DATE: 918103 PAGE: 1 of 2 "A" CORE SPRAY DISCHARGE LINE LO PRESSURE TITLE: CORE SPRAY SYSTEM A LOOP LOW DISCHARGE PRESSURE 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) / SETPOINT(S) 1.1 Core Spray System A PS-2116B 47.5 psig (dec)

Loop Discharge pressure low 2.0 AUTOMATIC ACTIONS 2.1 None; however, a computer message is received.

3.0 OPERATOR ACTIONS 3.1 At 1C03, verify System A low pressure condition on A Core Spray Pump discharge pressure indicator P1-2106.

3.2 Send an Operator to perform the following:

a. Verify RHRICore Spray Fill Pump IP-70 running.

b. Fill and vent Core Spray System per 01 151 (Core Spray System).

c. Verify RHR/Core Spray Fill Pump 1P-70 to A Core Spray Pump lineup per 01 149 (Residual Heat Removal System).

3.3 Verify proper Core Spray System Panel lineup per 01 151.

3.4 Ifthe Core Spray System A low pressure condition cannot be corrected immediately, notify the CRS and comply with the Technical Specification I requirements for ECCS - Operating and ECCS - Shutdown.

ANNUNCIATOR PANEL: IC03A COORDINATES: B-8 REVISION: 6 DATE: 918/03 PAGE: 2 of 2 4.0 SUPPLEMENTAL ACTIONS 4.1 If necessary, initiate a Work Request Card to have RHR/Core Spray Fill Pump 1P-70 tested/repaired as necessary.

4.2 Comply with the Technical Specification requirements for ECCS - Operating and ECCS - Shutdown.

5.0 REFERENCES

5.1 APED-E21-006 <2>

5.2 BECH-M119, BECH-M121, BECH-M400 5.3 DAEC Technical Specifications 5.4 01149,01151 5.5 MM 132 5.6 DCP 1451, DCP 1451/2 5.7 DDC 1885 I

ANNUNCIATOR PANEL: IC03A COORDINATES: B-9 REVISION: 5 DATE: 414101 PAGE: I of 2 A" OR "B" CORE SPRAY 125 VDC LOGIC POWER FAILURE TITLE: CORE SPRAY SYSTEM I AND (OR) 2 LOGIC POWER FAILURE t >^

- -

• we-  : gf -. .< . r ' o ~,>¢-

-:- ~ . . -

- _ CAU )N~;3.

The following Auto Initiations are inoperative:

a)2 psig Hi D/W Pressure - Core Spray, HPCI, Diesel Generator, D/W Cooling fans to slow b) Lo-Lo-Lo RPV Level - Core Spray, RHR, Diesel Generator, RBCCW isolation 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) / SETPOINT(S) 1.1 Loss of 125 VDC Relay Relay E21A-K1A Deenergized Logic Power to Core Spray System 1 1.2 Loss of 125 VDC Relay Relay E21A-K1 B Deenergized Logic Power to Core Spray System 2 2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 Send an Operator to 125 VDC. Panel 1D11 to close to verify closed Ckt 15 for Core Spray System 1 Logic Power and to Panel I D21 to close or verify closed Ckt 15 for Core Spray System 2.

3.2 If the above alarm does not clear.

a. At 1C43, check condition of Fuse E21A-F1A at Fuse Position F-1 and Fuse E21A-F2A at Fuse Position F-2 for Relay Logic A on Terminal Block AA for Core Spray System 1.

b. At 1C44, check condition of Fuse E21A-F1 B at Fuse Position F-1 and Fuse E21A-F2B at Fuse Position F-2 for Relay Logic B on Terminal Block AA for Core Spray System 2.

(continued)

ANNUNCIATOR PANEL: IC03A COORDINATES: B-9 REVISION: 5 DATE: 414101 PAGE: 2 of 2 I 3.0 OPERATOR ACTIONS 3.3 If Logic power cannot be restored, comply with the Technical Specification requirements for Primary Containment Isolation Instrumentation, Emergency Core Cooling System (ECCS) Instrumentation, ECCS - Operating, ECCS -

Shutdown, AC Sources - Operating, and AC Sources - Shutdown.

4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed logic power monitor, initiate a Work Request Card to have that logic power monitor checked/repaired.

5.0 REFERENCES

5.1 APED-E21-005<1>; APED-E21-006 <2>

5.2 BECH-E840 5.3 DAEC Technical Specifications 5.4 MM 132 5.5 DCP 1451 5.6 Tech. Spec. Amendment 193

ANNUNCIATOR PAN EL: IC03A COORDINATES: C-1 REVISION: 6 DATE: 414101 PAGE: 1 of 2 FUEL POOL EXHAUST RIS-4131AJB RAD MONITOR DNSCLJINOP TITLE: FUEL POOL EXHAUST RAD MONITOR DOWNSCALE/INOPERATIVE (RIS-41 31 AIB) 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I SETPOINT(S) 1.1 Fuel Pool Exhaust Relay D11-K1 via .01 mRlhr (dec) downscale RIS-4131A 1.2 Fuel Pool Exhaust Relay D1I-K1 via Mode Switch in other inoperative RIS-4131A than OPERATE, module unplugged, High Voltage supply low.

1.3 Fuel Pool Exhaust Relay D11-KI via .01 mR/hr (dec) downscale RIS-4131 B 1.4 Fuel Pool Exhaust Relay D1 1-K1 via Mode Switch in other inoperative RIS-4131 B than OPERATE, module unplugged, High Voltage supply low.

2.0 AUTOMATIC ACTIONS 2.1 If MODE SWITCH is in other than OPERATE:

a. A[B] Standby Gas Treatment starts.

b. A Group 3 A[B] Isolation occurs.

ANNUNCIATOR PANEL: . 1C03A COORDINATES: C-1 REVISION: 6 DATE: 414101 PAGE: .. 2 of 2 3.0 OPERATOR ACTIONS 3.1 Verify completion of the Group 3 Isolation by one of the following means:

a. Verify that the individual valves, fans, and dampers from Section 2.0 AUTOMATIC ACTIONS have shifted to isolation status, OR

b. On the PCIS status board, verify that the green ALL VALVES CLOSED lights are on to coincide with the amber ISOLATION SIGNAL lights that are on, allowing time for the valves to close, OR

c. Check the CIMS PRINTER PRT-1 which will print out Group 3 valves, fans, and dampers which fail to shift to isolation status on an isolation signal.

3.2 At 1C36, monitor FUEL POOL EXHAUST RADIATION MONITORS RIS-4131A/B to determine if downscale, inoperative or mode switch out of operate.

3.3 Check Computer Point B558.

3.4 If due to a RIS-4131A/B downscale condition, notify H.P. Technician to investigate problem.

3.5 If due to Mode Switch out of OPERATE, reset Group 3, SBGT, and PCIS Logic when Mode Switch Is returned to OPERATE.

4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the Technical Specification requirements for Primary Containment Isolation I

Instrumentation and Secondary Containment Isolation Instrumentation.

4.2 Reset the RIS-4131 trip logic per 01 879.1 (Process Radiation Monitoring System) when the repairs have been completed.

5.0 REFERENCES

5.1 APED-DI 1-002 <2A,7>; APED-D1 1-028 <1>

5.2 BECH-M141, BECH-M176 5.3 01 879.1 5.4 DAEC Technical Specifications 5.5 MM 132, MM 164 5.6 DCP 1451 5.7 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: IC03A COORDINATES: C-2 REVISION: 8 DATE: 414/01 PAGE: I of 2 POST TREAT RM-4101AIB Hi RAD TITLE: POST TREATMENT OFFGAS HIGH RADIATION 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) / SETPOINT(S) 1.1 Post Treatment Relay D1 -K17 4 X 104 CPS (inc)

Offgas Radiation high (RM-41 01 A) 1.2 Post Treatment Relay D1I-K17 4 X 104 CPS (inc)

Offgas Radiation high (RM-41 01 B)

NOTE The Upscale high radiation setpoint corresponds to the bypass limit and provides signals to the Offgas System to close Bypass Valve CV-4134B and open Treatment Valve CV-4134A if MODE SELECT Handswitch HS-4103 on 1C34 is in AUTO.

2.0 AUTOMATIC ACTIONS 2.1 If MODE SELECT Handswitch HS-4103 is in the AUTO Position, Bypass Valve CV-41 34B doses and Inlet Valve CV-41 34A opens to place Charcoal Adsorbers in service.

3.0 OPERATOR ACTIONS 3.1 Verify Charcoal Adsorbers in service in series per 01 672 (Offqas and Recombiner System) at Offgas Panel 1C34 and monitor Offgas System parameters.

3.2 At 1C02, monitor A. RR-41 01, OFFGAS POST TREAT RAD MONITOR B. RR-4116, OFFGAS VENT PIPE RAD MONITOR 3.3 At 1C10, monitor A. RM-4101A/B, OFFGAS POST TREAT RAD MONITORS B. RM-4116ANB, OFFGAS VENT PIPE RAD MONITORS

ANNUNCIATOR PANEL: IC03A COORDINATES: C-2 REVISION: 8 DATE: 414101 PAGE: 2 of 2 3.0 OPERATOR ACTIONS (Continued) 3.4 Monitor Kaman #10 on PPC and RR4176 at 1C02 OFFGAS STACK RAD RECORDER.

3.5 If time permits, place standby OFFGAS POST FILTER I F-214A/B in service per 01 672, Section 6.9.

3.6 If Offgas Post Treat Offgas Stack activities continue to trend upward at a fast rate, perform appropriate section of AOP 672.2 (Offgas Radiation/Reactor Coolant High ActivitV).

3.7 Refer to EPIP for EAL classification if applicable.

3.8 If RM-4101A/B is in TEST status, reset OFFGAS POST TREAT RAD MONITORS RM-41OIAIB when test signal has cleared.

4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the ODAM requirements for Radiation Monitoring.

I 4.2 Reset the affected RM4101 A/B per 01 879.1 (Process Radiation Monitoring System) after repairs have been completed.

5.0 REFERENCES

5.1 APED-DI 1-002 <2,4A>

5.2 BECH-M014 5.3 01672,01879.1 5.4 AOP 672.2 5.5 EPIP 1.1 5.6 ODAM 6/7 5.7 MM 132, MM 162 5.8 DCP 1451 5.9 Tech. Spec. Amendment 184, 193

. ANNUNCIATOR PANEL: IC03A COORDINATES: C-3 REVISION: 7 DATE: 918103 PAGE: 1 of 2 PRETREAT RM-4104 RAD MONITOR DNSCLIINOP TITLE: PRETREATMENT OFF-GAS DOWNSCALE OR INOPERATIVE (RM-4104) 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) / SETPOINT(S) 1.1 Pretreatment Off-Gas Relay D1 1-K7 via 0.5 mR/hr (dec) monitor downscale RM-4104 1.2 Pretreatment Off-Gas Relay D11 -K8 via Mode Switch in other monitor inoperative RM-4104 than OPERATE, Module unplugged, High voltage supply low.

2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS NOTE Expected alarm at low RX power level.

3.1 At 1CCO, monitor Indicating light to determine if OFFGAS PRETREAT RAD MONITOR RM-4104 is downscale or inoperative.

3.2 If the indicating lights are not lighted, perform the following:

a. Send an Operator to 1Y21 to verify ON/RESET CKT 14.

b. In ICIO, check the condition of Fuse D11A-F8 on Terminal Block AA Fuse Position F-3.

NOTE Low or erratic Offgas flow may indicate premature recombination of 02 and H2 in the Offgas System. Refer to AOP 672.3 (Off-gas Premature Recombination Events).

3.3 If DOWNSCALE, notify Chemistry/Health Physics to investigate problem.

3.4 If inoperative due to an Instrument failure, initiate a Work Request Card to have failed RM-4104 tested/repaired, notify the CRS and comply with ODAM requirements for Radiation Monitoring.

I (Continued)

ANNUNCIATOR PANEL: 1C03A COORDINATES: C-3 REVISION: 7 DATE: 9/8103 PAGE: 2 of 2 3.0 OPERATOR ACTIONS (Continued) 3.5 If due to a sample problem, perform ARP I C03A, D-3 (PRETREAT SAMPLE FLOW TROUBLE).

3.6 If the Pretreatment Offgas Monitor RM-4104 is in test status, reset RM-4104 trip Logic when the test signal has cleared.

4.0 SUPPLEMENTAL ACTIONS 4.1 Reset RM-4104 per 01 879.1 (Process Radiation Monitoring System) when repairs have been completed.

5.0 REFERENCES

5.1 APED-D1 1-002 <2,4>; APED-Hi 1-061 5.2 BECH-M141 5.3 01879.1 5.4 AOP 672.3 5.5 ARP 1C03A, D-3 5.6 MM 132, MM 139 5.7 DCP 1451 5.8 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: IC03A COORDINATES: C4 REVISION: 10 DATE: 414101 PAGE: I of 2 OFFGAS VENT PIPE RM-4116AIB RAD MONITOR DNSCLIINOP TITLE: OFF-GAS VENT PIPE RAD MONITOR (RM-4116A/RM-4116B)

DOWNSCALE/INOPERATIVE 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) / SETPOINT(S) 1.1 Radiation Monitor Relay D11-K12 1 CPS (dec) 4116A downscale 1.2 RM-4116A inoperative Relay D11-K11 Mode Switch not in OPERATE, Module unplugged, Low high voltage.

1.3 RM-4116B downscale Relay D11-K12 I CPS (dec) 1.4 RM-4116B inoperative Relay D11-K11 Mode Switch not in OPERATE, Module unplugged, Low high voltage.

2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 At IC10, monitor OFFGAS STACK VENT PIPE A[B] RAD MONITOR RM-4116A/B to determine if DOWNSCALE or INOPERATIVE.

3.2 If DOWNSCALE due to loss of power, send an Operator to the following Panels to verify the breakers closed/reset:

Monitor Panel Breaker RM4116A 1D50 03 RM4116B 1D60 03 3.3 RM-4116A and RM4116B are required to be operable during venting or purging any time when primary containment integrity is required. (Technical Specifications for Primary Containment Isolation Instrumentation) 3.4 If either RM-4116A or RM4116B becomes inoperable during conditions which require them to be operable, restore the inoperable channel within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> close the primary vent and purge valves, or establish administrative control of those valves with continuous monitoring of alternate instrumentation.

(Continued)

ANNUNCIATOR PANEL: IC03A COORDINATES: C-4 REVISION: I10 DATE: 414101 PAGE: i 2 of 2 3.0 OPERATOR ACTIONS (Continued) 3.5 Ifboth RM-4116A and RM-4116B become inoperable during conditions which require them to be operable, restore isolation capability within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Ifisolation capability is not restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, close the primary vent and purge valves or establish administrative control of those valves with continuous monitoring of alternate instrumentation within the next hour.

4.0 SUPPLEMENTAL ACTIONS 4.1 Administrative control requires (1) an operator is stationed at the valve controls, and (2) that operator is instructed to terminate venting or purging when procedures direct valve closure.

4.2 Initiate a Work Request Card to have inoperable radiation monitor(s) I repaired/tested and comply with the Technical Specification requirements for Primary Containment Isolation Instrumentation.

4.3 Reset the affected RM4116ANB per 01 879.1 (Process Radiation Monitoring System) when repairs have been completed.

5.0 REFERENCES

5.1 APED-DI1-002 <2,5>

5.2 BECH-M141 5.3 01 879.1 5.4 ODAM 6/7 5.5 DAEC Technical Specifications 5.6 MM 132, MM 139 5.7 DCP 1451 5.8 Tech. Spec. Amendment 184, 193, 209

ANNUNCIATOR PANEL: IC03A COORDINATES: C-5 REVISION:. 17 DATE: 4122104 PAGE: 1 of 3 SRV/SV TAILPIPE HI PRESS OR HI TEMP TITLE: SRV/SV TAILPIPE HIGH PRESSURE OR HIGH TEMPERATURE 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I SETPOINT(S) 1.1 PSV-4400 through TR4400A[B, C, DI 2500F (inc) 4407 Tailpipe temperature high 1.2 PSV-4400A/B/C Relay 95-K4400 - 25 psig (inc) through 4407A/B1C 95-K4407 via Tailpipe high pressure PS-4400AIB/C through 4407AIB1C (2 out of 3 PS's on each relief) 2.0 AUTOMATIC ACTIONS NOTE Refer to EAL SU5 (RCS Leakage) for a SRV that is open and cannot be shut.

2.1 If due to CAUSE 1.2, the following actions occur

a. At I C21, the affected white TAILPIPE PRESSURE NORMAL light turns off.

b. At 1C03, an amber light turns on for SAFETY/RELIEF VALVE OPENING indication.

c. The affected PSV-4400 - 4407 OPEN computer message is received.

d. A signal to SPDS is transmitted.

e. The Low-Low-Set logic receives one of its required arming signals.

Low-Low-Set will actuate if a high RPV pressure scram signal is received.

ANNUNCIATOR PANEL: IC03A COORDINATES: C-5 REVISION: 17 DATE: 4122/04 PAGE: 2 of 3 3.0 OPERATOR ACTIONS 3.1 Enter AOP 683, Abnormal Safety Relief Valve Operation, if not previously entered.

ANNUNCIATOR PANEL: IC03A COORDINATES: C-5 REVISION: 17 DATE: 4122/04 PAGE: 3 of 3 4.0 SUPPLEMENTAL ACTIONS 4.1 If necessary, refer toOl 183.1 for instructions to deenergize ADS/LLS Valve Control Circuits.

4.2 Comply with the Technical Specification requirements for Safety Relief Valves (SRVs) and Safety Valves (SVs).

5.0 REFERENCES

5.1 APED-B21-018 <1,2,3,3A>

5.2 APED-B21-114 5.3 BECH-E121, <2B,2C,2D>

5.4 BECH-M114, BECH-M141 5.5 01149, 01183.1, 01 264 5.6 DAEC Technical Specifications 5.7 EOP 1, EOP 2 5.8 IPOI 4, IPOI 5 5.9 CMAR 083062 5.10 MM 149, MM 132, MM 139 5.11 DCP 1410, DCP 1451 5.12 Tech. Spec. Amendment 193 5.13 AR 96-1078 5.14 AR 18756 5.15 AR 23595

ANNUNCIATOR PANEL: IC03A COORDINATES: C-6 REVISION: 5 DATE: 4122/04

• PAGE: I of 4

'ADSILLS 125 VDC CONTROL POWER

- FAILURE TITLE: AUTOMATIC DEPRESSURIZATION SYSTEM/LOW LOW SET CONTROL POWER FAILURE 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) / SETPOINT(S) 1.1 Loss of 125 VDC power to Relay B21C-KIA Deenergized ADS logic A 1.2 Loss of normal 125 VDC Relay B21 C-K1 B Deenergized I power to ADS logic B 1.3 Loss of normal 125 VDC Relay B21 C-K1 IA Deenergized power to PSV-4400 1.4 Loss of normal 125 VDC Relay B21 C-K1 1B Deenergized power to PSV-4401 and LLS ChannelA 1.5 Loss of normal 125 VDC Relay B21 C-KI IC Deenergized power to PSV-4402 1.6 Loss of normal 125 VDC Relay B21 C-K1 ID Deenergized power to PSV-4405 1.7 Loss of normal 125 VDC Relay B21 C-K1 I E Deenergized power to PSV-4406 1.8 Loss of normal 125 VDC Relay B21 C-KI IF Deenergized power to PSV-4407 and LLS Channel B 1.9 Loss of 125 VDC Relay B21 C-K22A Deenergized power to Low Low Set ChannelA 1.10 Loss of 125 VDC Relay B21C-K22B Deenergized power to Low Low Set Channel B 2.0 AUTOMATIC ACTIONS 2.1 If due to CAUSE 1.2 to 1.8, a transfer to the alternate 125 VDC power source occurs.

ANNUNCIATOR PANEL: IC03A COORDINATES: C-6 REVISION: 5 DATE: 4/22104 PAGE: 2 of 4 3.0 OPERATOR ACTIONS 3.1 At 1C03, monitor SRV CLOSED (green) indicating lights are ON to confirm power Is available for SRV operation.

3.2 If indication Is lost for any SRV, check both normal and alternate power sources/fuses (Attachment 1) for the affected valve.

3.3 At 1C45, visually check the status of initiating devices (HFA relays) for CAUSE 1.1 through 1.8.

a. Relays are normally energized.

b. Ifa relay is deenergized, check normal power source fuses (Attachment 1).

c. If multiple relays are deenergized, send an Operator to check the condition of circuit breaker for circuit 14 on 1D13 or 1D23 (Attachment 1).

3.4 If the indicated HFA relays are all energized, monitor Low-Low Set indicating lights on 1C45.

a. If DS12A is OFF, check fuse F20A (Attachment 1).

b. If DS12B is OFF, check fuse F20B (Attachment 1).

3.5 If no visual indications of failure are observed, check fuses F21A and F21B (Attachment 1).

3.6 IfADS/Low Low Set Logic circuit(s) is/are determined to be inoperative, notify the CRS and comply with the Technical Specification requirements for Emergency Core Cooling System (ECCS) Instrumentation and Low-Low Set (LLS) Instrumentation.

3.7 If the ADS/Low Low Set Control Power Logic is in test status, reset ADS/Low Low Set Control Power Logic when the test signal has cleared.

4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the Technical Specification requirements for Emergency Core Cooling System (ECCS) Instrumentation and Low-Low Set (LLS) Instrumentation.

ANNUNCIATOR PANEL: IC03A COORDINATES: C-6 REVISION: 5 DATE: 4122104 PAGE: 3 of 4

5.0 REFERENCES

5.1 APED-B21-018 <1,2,3>; APED-121-112 <1,2,3>

5.2 DAEC Technical Specifications 5.3 DCR 1183 5.4 MM 132 5.5 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: IC03A COORDINATES: C-6 REVISION: 5 DATE: 4122/04 PAGE: 4 of 4 ATTACHMENT 1 ADS/LLS POWER DISTRIBUTION NORMAL POWER SOURCES - LOSS CAUSES ALARM Terminal Circuit ID Power Monitor Fuses Block Power Source ADS LogicA B21C-K1A F1, F2 CC 1D13 CKT 14 ADS Valve PSV-4400 B21C-K11A F3A, F4A CC 1D13 CKT 14 LLS Channel A, PSV4401 B21C-K11B F3B, F4B CC 1D13 CKT 14 ADS Valve PSV-4402 B21C-K11C F3C, F4C CC 1D13 CKT 14.

ADS Valve PSV-4405 B21C-K11D F3D, F4D CC 1D13 CKT 14 ADS Valve PSV-4406 B21C-K11E F3E, F4E CC 1D13 CKT 14 LLS ChannelA B21C-K22A F20A, F21A HH ID13 CKT 14 ADS Logic B B21C-K1B F5B, F6B CC 1D23 CKT 14 LLS Channel B, PSV-4407 B21C-K11F F7F, F8F CC 1D23 CKT 14 LLS Channel B B21C-K22B F20B, F21B KK 1D23 CKT 14 ALTERNATE POWER SOURCES - PROVIDE POWER ON LOSS OF NORMAL SOURCE Terminal Circuit ID Fuses Block Power Source ADS Valve PSV-4400 F7A, F8A CC 1D23 CKT 14 LLS Channel A, PSV-4401 F7B, F8B CC 1D23 CKT 14 ADS Valve PSV-4402 .F7C, F8C CC 1D23 CKT 14 ADS Valve PSV-4405 F7D, F8D CC 1D23 CKT 14 ADS Valve PSV-4406 F7E, F8E CC 1D23 CKT 14 ADS Logic B F5A, F6A CC ID13 CKT 14 LLS Channel B, PSV-4407 F3F, F4F CC 1D13 CKT 14

ANNUNCIATOR PAN EL: IC03A COORDINATES: C-7 REVISION: 6 DATE: 4122104 PAGE: 1 of 2 SRV BELLOWS FAILURE TITLE: SRVISV TAILPIPE BELLOWS FAILURE 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I SETPOINT(S) 1.1 PSV4400 [01, 02, 05, Relay B21C-MSPX via 25 psig (inc) I 06, 07] bellows PS-4400[01, 02, 05, 06, 07]

pressure high 2.0 AUTOMATIC ACTIONS 2.1 None; however, a white BELLOWS INTEGRITY light turns off at 1C21 for the affected bellows.

3.0 OPERATOR ACTIONS NOTE Bellows failure does not necessarily mean the valve is leaking; however, it does make the safety (mechanical) actuation portion of the relief valve inoperative.

3.1 At 1C03, check amber lights above ADS Safety/Relief Valve Control Switches.

3.2 At I C21, monitor Tail Pipe temperatures at TR-4400A/B/CID and observe white BELLOWS INTEGRITY light.

3.3 If due to a bellows leakage problem or any Safety/Relief Valve indicates open, notify the CRS and comply with the Technical Specification requirements for Safety Relief Valves (SRVs) and Safety Valves (SVs).

4.0 SUPPLEMENTAL ACTIONS 4.1 Initiate a Work Request Card to have leaking bellows tested/replaced as necessary.

4.2 If necessary, refer to 01-183.1 for instructions to deenergize ADSILLS Valve Control Circuits.

ANNUNCIATOR PANEL: IC03A COORDINATES: C-7 REVISION: 6 DATE: 4122104 PAGE: 2 of 2

5.0 REFERENCES

5.1 APED-B21-018 <3>

5.2 BECH-E121, <2A>

5.3 BECH-M141 5.4 DAEC Technical Specifications 5.5 01183.1 5.6 EOP1 5.7 MM 132, MM 139, MM 149 5.8 DCP 1410, DCP 1451 5.9 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: 1C03A COORDINATES: C-8 REVISION: 8 DATE: 9/8103 PAGE: I of 2 "A"CORE SPRAY SPARGER LO AP TITLE: CORE SPRAY SYSTEM I HEADER TO TOP OF CORE PLATE LOW DIFFERENTIAL PRESSURE NOTE This is a normal alarm while shutdown, while shutting down (i.e., high core flow with relatively low load line).

1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) / SETPOINT(S) 1.1 Core Spray System I PDIS-21 19 2.46 psid (dec)

Header to top of Core plate differential pressure low 2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS NOTE This alarm may be an indication of a Core Spray line break inside the Reactor vessel.

3.1 Send an Operator to I C55 to verify proper lineup on PDIS-2119 and that PDIS-2119 reading exceeds the alarm setpoint.

3.2 Evaluate Reactor loadline and core flow conditions. A low loadline and high core flow condition has been observed to be a cause for this alarm.

a. If this condition exists, reduce core flow and pull rods to raise loadline to clear alarm condition, if the Reactor Engineer and CRS agree.

I 3.3 If it is determined that the A CORE SPRAY SPARGER LO AP alarm is due to a line break in the A CORE SPRAY SPARGER, notify the CRS and comply with Technical Specification requirements for ECCS - Operating and ECCS -

Shutdown.

ANNUNCIATOR PAN EL: I C03A COORDINATES: C-8 REVISION: 8 DATE: 918103 PAGE: 2 of 2 4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/Core Spray sparger, initiate a Work Request Card to have that sensor/core spray sparger checked/repaired as necessary.

5.0 REFERENCES

5.1 APED-E21-006 <2>

5.2 BECH-Ml 15, BECH-Mi 21 5.3 DAEC Technical Specifications 5.4 MM 132 5.5 DCP 1451 5.6 Tech. Spec. Amendment 193 5.7 AR 18756

ANNUNCIATOR PANEL: IC03A COORDINATES: C-9 REVISION: 7 DATE: 414101 PAGE: I of 2 "A" CORE SPRAY DISCHARGE LINE HI PRESSURE TITLE: "A"CORE SPRAY SYSTEM A HIGH DISCHARGE PRESSURE NOTE This is a normal alarm when "A" Core Spray Pump is running.

1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) / SETPOINT(S) 1.1 Core Spray System A PS-2116A 100 psig (inc) discharge pressure high 2.0 AUTOMATIC ACTIONS 2.1 None; however, a computer message is received.

3.0 OPERATOR ACTIONS 3.1 At 1C03, verify system high pressure condition on (A CORE SPRAY) PUMP DISCHARGE PRESSURE indicator P1-2106.

3.2 If high, send an Operator to vent Core Spray System "A" per 01 151 (Core Spray System).

a. After venting, monitor system pressure closely.

NOTE Leaving MO-2115 in a closed position for other than cycling MO-2117 will require a Core Spray LCO. The opening function of the valve is not safety related.

3.3 If pressure starts increasing, perform the following:

a. Close MO-2115.

b. Open MO-2117.

c. Close MO-2117.

d. Open MO-2115.

ANNUNCIATOR PANEL: IC03A COORDINATES: C-9 REVISION: 7 DATE: 414101 PAGE: - 2 of 2 4.0 SUPPLEMENTAL ACTIONS 4.1 None I

5.0 REFERENCES

5.1 APED-E21-006 <2>

5.2 BECH-M121, BECH-M400 5.3 01151 5.4 MM 132 5.5 DCP 1451 5.6 Tech. Spec. Amendment 193 5.7 AR 960729.01

ANNUNCIATOR PANEL: 1C03A COORDINATES: D-1 REVISION: 4 DATE: 918103 PAGE: 1 of 2 POST TREAT OFFGAS SAMPLE HI/LO FLOW TITLE: POST TREATMENT OFFGAS SAMPLE HIGH/LOW FLOW 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) / SETPOINT(S) 1.1 Post Treatment Offgas PS4101, (1C134) 10" Hg (inc)

Sample flow high 1.2 Post Treatment Offgas PS-4101, (1C134) 1" Hg (dec)

Sample flow low 2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 Notify Chemistry/Health Physics to investigate Post Treatment Offgas Sample flow problem and to adjust flow as necessary.

3.2 If due to a loss of power to Sample Pumps I P-246A/B, send an Operator to Panel IY1 I to check the status of circuit 17.

3.3 If OFFGAS POSTTREAT A[B] RAD MONITOR RM41 01A/B are Downscale due to the Offgas Sample Hi/Lo Flow Condition, notify the CRS and comply with the I ODAM requirements for Radiation Monitoring.

4.0 SUPPLEMENTAL ACTIONS 4.1 Initiate a Work Request Card to have failed sample pump 1P-246A/B or cogged filter repaired/replaced as necessary.

ANNUNCIATOR PANEL: 1C03A COORDINATES: D-1 REVISION: 4 DATE: 9/8/03 PAGE: 2 of 2

5.0 REFERENCES

5.1 APED-DI 1-002 <2,4A>

5.2 BECH-M141 5.3 DAEC Technical Specifications 5.4 MM132,MM139 5.5 DCP 1451

ANNUNCIATOR PANEL: IC03A COORDINATES: D-2 REVISION: 6 DATE: 414101 PAGE: I of 2 POST TREAT RM-410IA/B DOWNSCALE TITLE: POST TREATMENT OFFGAS DOWNSCALE (RM-4101A/B) 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) / SETPOINT(S) 1.1 Post Treatment Offgas Relay D11-K16 via I CPS (dec)

Radiation Monitor RMA41 01A RM-4101A downscale 1.2 Post Treatment Offgas Relay D11K6 via 1 CPS (dec)

Radiation Monitor RM4101B RM-4101B downscale 2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 At 1 C02, monitor OFFGAS POSTTREAT RAD MONITOR RR-41 01 to confirm RM-41I1A/B DOWNSCALE condition.

3.2 At 1C10, monitor RM-41OIAB indication and lights to help determine which radiation monitor is downscale and to determine if the other channel Is in or near a tripped condition. Attempt to reset tripped radiation monitor if unit indicates

> 1 cps.

3.3 If the RM 4101AB is in test status, reset RM-4101A/B Logic when the test signal has cleared.

4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/channel checked/repaired and comply with the ODAM requirements for Radiation Monitoring.

I 4.2 Reset the affected RM-4101AB per 01 879.1 (Process Radiation MonitorinQ System) after repairs have been completed.

ANNUNCIATOR PANEL: IC03A COORDINATES: D-2 REVISION: 6 DATE: 4/4101 PAGE: 2 of 2

5.0 REFERENCES

5.1 APED-D11-002 <2,4A>

5.2 BECH-M141 5.3 AOP 691 5.4 01 879.1 5.5 ARP 1C07B, D-9 5.6 DAEC Technical Specifications 5.7 MM 131, MM 132, MM 136, MM 139 5.8 DCP 1451 5.9 Tech. Spec. Amendment 193 5.10 EMA A44812

ANNUNCIATOR PANEL: IC03A COORDINATES: 0-3 REVISION: 7 DATE: 918103 PAGE: 1 of 2 PRETREAT SAMPLE FLOW TROUBLE TITLE: PRETREATMENT OFFGAS SAMPLE FLOW TROUBLE 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) / SETPOINT(S) 1.1 Pretreatment Off-Gas PDIS-4120 1" Hg (dec)

Sample Flow low 1.2 Pretreatment Off-Gas PDIS-4120 10" Hg (inc)

Sample Flow high 2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 At I CIO, monitor Pretreatment Offgas radiation on OFFGAS PRETREAT RAD MONITOR RM-41 04.

3.2 Notify Chemical Lab of Pretreatment Offgas Sample Flow abnormal flow condition.

3.3 If due to a loss of power problem, send an Operator to 1L1-20 to perform the following:

a. Verify status of Ckt 15 for Offgas Sample Heater.

b. Verify status of Ckt 23 for Sample Pump I P-222.

ANNUNCIATOR PANEL: IC03A COORDINATES: D-3 REVISION: 7 DATE: 918103 PAGE: 2 of 2 4.0 SUPPLEMENTAL ACTIONS 4.1 Initiate a Work Request Card to have clogged filter/failed Pump I P-222 replaced/repaired as necessary.

4.2 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the ODAM requirements for Radiation Monitoring.

5.0 REFERENCES

5.1 APED-D11-002 <2,4>

5.2 MM 132, MM 139 5.3 DCP 1451 5.4 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: IC03A COORDINATES: D-4 REVISION: 9 DATE: 9/30103 PAGE: 1 of 2 OFFGAS VENT PIPE SAMPLE HI/LO FLOW TITLE: OFFGAS VENT PIPE SAMPLE HIGH/LOW FLOW 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) I SETPOINT(S) 1.1 Offgas Vent Pipe PS-4129 13" Hg (inc) I Sample Flow low 1.2 Offgas Vent Pipe PS-4129 1" Hg (dec) I Sample Flow high 2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 Notify Chemist and/or Operator to check the Offgas Vent Pipe Sample Station:

a. If flow is low, perform the following:

1) Start Standby Pump IP-247ANB.

2) Check for evidence of water accumulation.

3) Check to see Iffilter is clogged or wet.

4) If during cold weather, check heat tracing 1Li 40 Ckt 11 and for a possible frozen return line.

5) If due to a loss of power, check 1L140 Ckt 01.

b. If flow is high, perform the following:

1) Check if both sample pumps are running - only one should be in service.

2) Check filter to see if intact.

3) Have Chemist adjust to proper sample flow rate.

(Continued)

ANNUNCIATOR PANEL: IC03A COORDINATES: D-4 REVISION: 9 DATE: 9130103 PAGE: 2 of 2 3.0 OPERATOR ACTIONS (Continued) 3.2 RM-4116A and RM4116B are required to be operable during venting or purging any time when primary containment integrity is required. (Technical Specification requirements for Primary Containment Isolation Instrumentation) 3.3 If either RM-4116A or RM-4116B becomes inoperable during conditions which require them to be operable, restore the inoperable channel within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> close the primary vent and purge valves or establish administrative control of those valves with continuous monitoring of alternate instrumentation.

3.4 If both RM41 16A and RM-41168 becomes inoperable during conditions which require them to be operable, restore isolation capability within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If isolation capability is not restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, then close the primary vent and purge valves or establish administrative control of those valves with continuous monitoring of alternate instrumentation within the next hour.

4.0 SUPPLEMENTAL ACTIONS 4.1 Administrative control requires (1)an Operator is stationed at the valve controls, and (2) that Operator is instructed to terminate venting or purging when procedures direct valve closure.

4.2 If the filter appears clogged, notify Plant Chemist to have changed.

4.3 Initiate a Work Request Card to have failed sample pump 1P-247AIB replaced/repaired as necessary.

4.4 Initiate a Work Request Card to have inoperable radiation monitor(s) repaired/tested as necessary.

5.0 REFERENCES

5.1 APED-DI 1-002 <2,5>

5.2 BECH-E029 <1,2>

5.3 ODAM 617 5.4 MM 132, MM 141 5.5 DCP 1451 5.6 Tech. Spec. Amendment 184,193, 209

ANNUNCIATOR PANEL: IC03A COORDINATES: D-5 REVISION: 7 DATE: 4122/04 PAGE: 1 of 3 ALLS "A" OR "B" ARMED TITLE: LOW-LOW SET (LOGIC) CHANNELS A OR B ARMED 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I SETPOINT(S) 1.1 Low-Low Set Channel A Relay C71A-K27A via Energized 1055 psig armed due to RPS Relays C71A-K5A or (inc)

Reactor High Pressure C71A-K5C Logic Al or A2 tripped AND

a. Low-Low Set Channel Relay B21 C-K23B Energized I B armed (see Sec. 1.2)

OR

b. PSV-4400 tail pipe Relay 95-04KR401 via Energized 25 psig (inc) pressure high PS-4400A, 4400B or OR 4400C arranged in a I 2 out of 3 logic configuration

c. PSV-4401 tail pipe Relay 95-2KR201A via Energized 25 psig (Inc) pressure high PS-4401A, 4401B or OR 4401 C arranged in a 2 out of 3 logic configuration

d. PSV-4402 tail pipe Relay 95-2K201 B via Energized 25 psig (inc) pressure high PS-4402A, 4402B, or 4402C arranged in a 2 out of 3 logic configuration (Continued)

ANNUNCIATOR PANEL: IC03A COORDINATES: D-5 REVISION: 7 DATE: 4122/04 PAGE: 2 of 3 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) / SETPOINT(S)

(Continued) 1.2 Low-Low Set Channel B Relay C71A-K27B via Energized 1055 psig armed due to RPS Reactor Relays C71A-K5B or (inc)

High Pressure Logic B1 or C71A-K5D I B2 tripped AND

a. Low-Low Set Channel A Relay B21 C-K23A Energized I armed (Sec. 1.1 above)

OR

b. PSV-4405 tail pipe Relay 95-1KR101A via Energized 25 psig (inc) pressure high PS-4405A, 4405B or OR 4405C arranged in a 2 out of 3 logic configuration

c. PSV-4406 tail pipe Relay 95-3KR301A via Energized 25 psig (inc) pressure high PS4406A, 4406B or OR 4406C arranged in a 2 out of 3 logic configuration

d. PSV-4407 tail pipe Relay 95-1 KR1 01 B via 25 psig (inc) pressure high PS-4407A, 4407B or PS-4407C arranged in a 2 out of 3 logic configuration 2.0 AUTOMATIC ACTIONS 2.1 When any Safety Relief Valve lifts, the LLS Valves PSV-4401 and PSV-4407 will open and reset at 900 psig and 905 psig respectively to reduce Reactor pressure to prevent opening of other relief/safety valves.

2.2 Until LLS A and B LOGIC LO-LO SET ARMING RESET pushbutton PB-4401 and PB-4407 are both depressed, PSV-4401 will continue to reopen at 1030 psig and then reset at 910 psig, and PSV-4407 will continue to reopen at 1035 psig and then reset at 915 psig.

ANNUNCIATOR PANEL: 1C03A COORDINATES: D-5 REVISION: L7 DATE: 4122/04 PAGE: 3 of 3 3.0 OPERATOR ACTIONS 3.1 Confirm the LLS Channel A/B armed condition at 1C45 by observing:

a. Light DS10A Is out (LLS Ch. A armed).

b. Light DS10B Is out (LLS Ch. B armed).

3.2 Verify the AUTOMATIC ACTIONS have occurred. If the specified automatic response has failed to occur, manually initiate the appropriate actions.

3.3 If Reactor pressure of 1055 psig or greater is confirmed, enter EOP 1 (RPV Control).

3.4 If the LLS Logic Channel A/B is in test status, reset the LLS Channel A1B Logic when the test signal has cleared.

4.0 SUPPLEMENTAL ACTIONS 4.1 If a Plant emergency condition does not exist as determined in Step 3.3 above, perform Reactor Recovery as appropriate per IPOI 4 (Shutdown) or IPOI 5 (Reactor Scram).

4.2 If the cause of the alarm is due to a failed sensor/trip channel, initiate a Work Request Card to have that sensor/trip channel checked/repaired and comply with Technical Specification requirements for Reactor Protection System (RPS)

Instrumentation and Low-Low Set (LLS) Instrumentation.

5.0 REFERENCES

5.1 APED-B21-018; APED-B21-114; APED-C71-004 5.2 BECH-E121 <2B,2C,2D>

5.3 IPOI 4, IPOI 5 5.4 EOP I 5.5 DAEC Technical Specifications 5.6 MM 132 5.7 DCP 1451 5.8 Tech. Spec. Amendment 193 5.9 EMA A58851 I

ANNUNCIATOR PANEL: IC03A COORDINATES: D-6 REVISION: 3 DATE: 414101 PAGE: 1 of 1 ADS INTEST STATUS TITLE: AUTOMATIC DEPRESSURIZATION SYSTEM IN TEST STATUS 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) / SETPOINT(S) 1.1 Test Switch installed in Test Switch assembly Inserted B21C-J1A receptacle contacts 1.2 Test Switch installed in Test Switch assembly Inserted B21C-JIB receptacle contacts 1.3 Test Switch installed in Test Switch assembly Inserted B21 C-J2A receptacle contacts 1.4 Test Switch installed in Test Switch assembly Inserted B21 C-J2B receptacle contacts 2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 If no test is in progress, send an Operator to 1 C45 to check Switch Receptacle J1A, JIB, J2A, and J2B.

3.2 Comply with Technical Specification requirements for Emergency Core Cooling System (ECCS) Instrumentation.

3.3 If an STP is in progress, ensure that all Test Switches are removed at the completion of testing.

4.0 SUPPLEMENTAL ACTIONS 4.1 None I

5.0 REFERENCES

5.1 APED-B21-018 <3>

5.2 DCP 1451 5.3 Tech. Spec. Amendment 193

ANNUNCIATOR PANEL: IC03A COORDINATES: D-7 REVISION: 2 DATE: 414101 PAGE: 1 of 2 ADS BOTH TEST JACKS INSTALLED TITLE: AUTOMATIC DEPRESSURIZATION SYSTEM TEST PROCEDURE FAULTY 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I. SETPOINT(S) 1.1 a. Test Switch installed Test Switch assembly Installed in Receptacle contacts B21C-J1A AND

b. Test Switch installed* Test Switch assembly Installed in Receptacle contacts B21 C-J2A 1.2 a. Test Switch installed Test Switch assembly Installed in Receptacle contacts B21C-JIB AND

b. Test Switch installed Test Switch assembly Installed in Receptacle contacts B21 C-J2B 2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 If no test is in progress, send an Operator to 1C45 to check that the Test Switches have been removed from Receptacle B21C-J1A, J2A, J1B, and J2B.

3.2 Comply with Technical Specification requirements for Emergency Core Cooling System (ECCS) Instrumentation.

3.3 If an STP is in progress, ensure that only one Test Switch is being used as per the applicable STP.

ANNUNCIATOR PANEL: IC03A COORDINATES: D-7 REVISION: 2 DATE: 414101 PAGE: 2 of 2 4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a damaged test receptacle/trip channel, initiate a Work Request Card to have the damaged test receptacle/trip channel checked/repaired as necessary and comply with the Technical Specification requirements for Emergency Core Cooling System (ECCS) Instrumentation.

5.0 REFERENCES

5.1 APED-B21-018 <3>

5.2 DAEC Technical Specifications 5.3 01 183;2 5.4 MM 132 5.5 DCP 1451

ANNUNCIATOR PANEL: IC03A COORDINATES: D-8 REVISION: 7 DATE: 918/03 PAGE: 1 of 2 LIQUID RAD MONITORS DNSCUINOP TITLE: LIQUID RADIATION MONITORS DOWNSCALE OR INOPERATIVE NOTE This is a normal alarm during surveillances affecting the liquid Rad Monitors.

1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I SETPOINT(S) 1.1 Radwaste Effluent Relay D 1I-K3 via 1 CPS (dec) downscale RM-3972 1.2 Radwaste Effluent Relay D 1I-KI via Mode Switch in other inoperative RM-3972 than OPERATE, Module unplugged, High Voltage supply low.

1.3 GSW Effluent Relay D11 -K1 2 via 1 CPS (dec) downscale RIS-4767 1.4 GSW Effluent Relay D1 -K 1 via Mode Switch in other inoperative RIS-4767 than OPERATE, Module unplugged, High Voltage supply low.

1.5 RBCCW downscale Relay D11-K16 via 1 CPS (dec)

RIS-4820 1.6 RBCCW Relay DII -K15 via RIS-4820 Mode Switch in other Inoperative than OPERATE, Module unplugged, High Voltage supply low.

1.7 RHRSW/ESW Relay D 1I-K8 via 1 CPS (dec) downscale RM-1997 1.8 RHRSW ESW Relay DI14-K7 via Mode Switch in other inoperative RM-1997 than OPERATE, Module unplugged, High Voltage supply low.

1.9 RHRSW Emergency Relay D 1I-K12 via 1 CPS (dec)

Service Water (Rupture RM-4268 Disc) downscale 1.10 RHRSW Emergency Relay D11-KI1 via Mode Switch in other Service Water (Rupture RM-4268 than OPERATE, Disc) Inoperative Module unplugged, High Voftage supply low.

ANNUNCIATOR PANEL: IC03A COORDINATES: D-8 REVISION: , 7 DATE: 9/8/03 PAGE: 2 of 2 2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 At 1C1 0, monitor the following to determine if downscale or inop condition exists:

a. RM-3972 Radwaste Effluent to Discharge Canal Rad Monitor

b. RIS-4767 General Service Water Rad Monitor C. RIS-4820 RBCCW Rad Monitor

d. RM-1997 RHRSW and ESW Discharge to Cooling Tower Rad Monitor

e. RM-4268 RHRSW and ESW Discharge to Canal Rad Monitor 3.2 If downscale/inoperative due to loss of power, send an Operator to verify the status of the following breakers:

a. At 1D60, Ckt 03 for RIS-4767 and 4820.

b. At I D50, Ckt 03 for RM-3972, 1997 and 4268.

3.3 If above Radiation Monitor(s) islare inoperative, notify CRS and comply with the I Monitoring requirements of ODAM 6/7.

4.0 SUPPLEMENTAL ACTIONS 4.1 If the cause of the alarm was due to a failed sensor/logic channel, initiate a Work Request Card to have that sensor/logic channel checked/repaired and comply with the Limiting Conditions for Operation of ODAM 6/7.

4.2 Reset the above Radiation Monitor per 01 879.1 (Process Radiation Monitoring System) when repairs have been completed.

5.0 REFERENCES

5.1 APED-Di 1-002 <2A,6>

5.2 01879.1 5.3 ODAM 6/7 5.4 STP NS790704, NS790304, NS790305, NS790303, NS790306, NS790301, NS790307 5.5 DCP 612, DCP 1451 5.6 MM 132, MM 139 5.7 Tech. Spec. Amendment 184

ANNUNCIATOR PANEL: IC03A COORDINATES: D-9 REVISION: 7 DATE: 01/14102 PAGE: 1 of 3 RBCCW RM-4820 HI RAD TITLE: REACTOR BUILDING CLOSED COOLING WATER HIGH RADIATION 1.0 PROBABLE CAUSE(S) / INITIATING DEVICE(S) / SETPOINT(S) 1.1 RBCCW Radiation High Relay D11-K17 via Variable RIS-4820 2.0 AUTOMATIC ACTIONS 2.1 None 3.0 OPERATOR ACTIONS 3.1 At lClO, confirm RBCCW high radiation at RBCCW RAD MONITOR RIS-4820.

3.2 If due to a loss of 24 VDC power, send an Operator to I D60 to verify status of Ckt 03.

High radiation Indicates a probable leak into the RBCCW System from one of the components being cooled. All leakage from the RBCCW System, as well as surge tank overflow, is to be considered POTENTIALLY CONTAMINATED.

3.3 Send an Operator to monitor RBCCW Surge Tank 1T-78 level gauge LG-4807.

3.4 If Surge Tank 1T-78 level is above the indicating range of LG-4807, check the overflow line to the open Radwaste funnel for possible surge tank overflow.

3.5 Drain Surge Tank IT-78 to establish an accurate level indication.

3.6 Check the Demin Makeup valve.

3.7 Notify Radwaste of excessive waste water.

(Continued)

ANNUNCIATOR PANEL: IC03A COORDINATES: D-9 REVISION: 7 DATE: 01114102 PAGE: 2 of 3 3.0 OPERATOR ACTIONS (Continued) 3.8 Isolate the following components unless required for plant operations:

a. RWCU Non-Regenerative Heat Exchangers 1E-215A and B (V-1 2-1, V-12-51).

b. RB Equipment Drain Sump Heat Exchanger 1E-33 (V-12-25, V-12-26).

c. Reactor Building Sample Coolers (V-12-96, V-12-97).

d. Turbine Building Sample Coolers (V-12-61, V-12-60).

e. Fuel Pool Cooling Heat Exchangers 1E-21 1A and B (V-1 2-34, V-1 2-33, V-1 2-28, V-12-30).

f. Hydrogen Water Chemistry Finishing Sample Cooler (V--12-119, V-12-118).

g. CAV Sample Cooler (V-1 2-1 00, V-12-101).

h. CRD Pump Cooler for the idle CRD Pump. Ifdeemed necessary, swap CRD Pumps to allow isolation of the CRD Pump Cooler for the running CRD pump

[IE-216A, V-12-44, V-12-461, [IE-216B, V-12-43, V-12-45].

i. RWCU Pump Cooler for the idle RCWU Pump. Ifdeemed necessary, swap RWCU Pumps to allow isolation of the RWCU Pump Cooler for the running RWCU Pump [IE-217A, V-12-53, V-12-57], [1E-217B, V-12-54, V-12-58].

3.9 Ifthe leak was isolated, unisolate one component at a time, while checking for leaks.

3.10 When leaking components(s) is(are) identified, reisolate it (them) and restore the other components to normal operation.

Restore RBCCW flow to the Reactor Recirculation Pump Heat Exchangers within 10 minutes or Ifpump seal cavities reach 2500 F at TR-4600 on I C21.

3.11 Ifthe leak was not isolated, isolate the Drywell RBCCW header, serving the Reactor Recirculation Pump Heat Exchangers and the Drywell Equipment Drain Sump Cooler, by closing MO-4841A and B with HS-4841 on 1C06.

(Continued)

ANNUNCIATOR PANEL: IC03A COORDINATES: D-9 REVISION: 7 DATE: 01/14/02 PAGE: 3 of 3 3.0 OPERATOR ACTIONS (Continued) 3.12 Fmaeurel 4

If continued cooling of the Reactor Recirc Pumps is not possible, scram the Reactor per IPOI 5 (Reactor Scram) and semuay scRa theReacirc Pumps within 5 minutes or if seal ocavity temperatures reach 250'F on TR4600 on I C21.

3.13 If the leak is in the Drywell RBCCW header, accept the leak, if possible, until the Reactor can be shut down and depressurized.

3.14 If the leak in the Drywell RBCCW header can be accepted, unisolate the Drywell RBCCW header by opening MO-4841A and B closely monitoring RBCCW rad level.

3.15 If necessary, shutdown the Reactor per IPOI 4 (Reactor Shutdown) to repair leaking component(s).

4.0 SUPPLEMENTAL ACTIONS 4.1 Initiate a Work Request Card to have leaking component tested/repaired as necessary.

5.0 REFERENCES

5.1 APED-D11-002 <2D,6>

5.2 BECH-M112 5.3 01414 5.4 IPOI 4, IPOI 5 5.5 MM131,MM132 5.6 AR 971239 5.7 AR 23595

1 A q - l I__..__-__.._.I_.

e l r P

__=-,__:_D'

_._.=--._._

l- @ -

~"l:[1-g i V, i

II '

srn

=._

. rTl--rrTI-T--Erw _. l ,, j 4,, j

_,=._s _._.

_ b'__ _. = ...... ....

_J rTr l rs rfrrJz _.S

_.... _._._.. w

. __..____'rZrarl-, 4 I F 1_1 __

-70 Y.T-rl'-, vr*rrnel

_@_*_W_

_.. . j,, ,, ,) .- ~ b 4-=X2 l

_- 0 S K__

_ =w w__ b

= . GA

.,_ ' I ' I l 1-<_

__r.o_*__

rs r<<r7rnl

- ~ = -.- j

_ __.__._..  : __arsi --

__._ __ - _w__

_ d I _- > _ _ _*__

t_.__._w_

-I _.-1s . rnl _ __ _ i __w_

_ _ r rb l _ _ =S _

l_

r_ __ _____

._ . ........................ _ ... b*s__*

CD=_ _,,, j, ,, s1w

__ _ _ -[=,.n 1=_ _

,_._.. rmlral_:

,*.,, . r .. l ,-.M. - .- -- -

___z.__ I f - - - f I IT'IIWI __N,--._ I ;= C:zr-rfl=

[ . ,.,, jr I o-- l bn

- t

-s---< ___e___ I

___..____ I s--w __Z._.____.. I Ctlx-r)-rfT-tlw) ._& ._ .... _ . I I __ I w _we__ I I l l-0-wo l l- l K-- l I _t e___ I a- l z-@ l zl-aL,22 I

@ -¢-l-- l

. e I l - l 1:-^s I . [3 1 e _ I L= -_=

r__ T-f _ I

.. . ,I I- r__

.- =_=.=.=

L_"Ltl nZ1=Irfl L_ I I .  ;

-C--'-- ' I- - I- I-=

2 _*3 1s Lt L {tS @ ri^ I_ @ Ltl- l . . .,, ,

I .______.____ I __.___*.__ I _____ .............................................................. __ I ___

I_ X___ _- -r . rl =_ [_ _ _

l _ __ l * *----'-- l

• _- _ l *______

I I I r t_ I --

I I I'- --

-, 03 -, 0 I _ . _

b ,-1 -, rTq-rT-IT-rT-rT-9

=_

I __S __ _.

I _=__ ._=_. __--'-,,-=

I _____

Modes:- CO Mt _O _ iDEU lI--]_ M._.4110 Modes: E0Z Uo m 4.5.DE

^ __ l.__,^

• l

_ , j .

--- - -- I I- -- -a II - * - -- . .I

_ ,--Z_ _,j---

- Cn-i r W ... __.... __

_. cm X _ _

__] _. ._b _ -

--- -,-11

_.,~rm~h, w__

rIT -.. -- _

. =.,::r.a -r1TrrTTmE~ -a. irar

._E.-a._...b

-. 1CaLm

-. cmTTrr!ja:

-. CKI

_a. 'T1'rffUU -. 0-- --

a' t , r. a I _ __ _ __

I

-U 2 _-, . ., I , D- . .. 1. .=.

- rrr

.__~

- Tr1w1 -- U _______~

- cmIr1Trnj rzrribj I I I ________

r, r.=m

- or~r~r1w) _U

-

• rlw - Cll=-W-J _

- . r. _r.

=r.=.==-

.. _ _3

- 0rlwl I I . _

I ,_.. _ _] .

.=_

_. I :I_ s______

=Zr-, - - , ..

___e_

. m_ _.

-Se ,,,_.,

r-v- _. 3*. 3, t7-, ,,,,- -. d T1T1 _ _ .____

-t I- *I .

ZC____.__

- cuzrrnmai _-. r . ms e_______

s

=.__ ___

_ i __

- clxlTwf Z7fr~wJ r= _-_*_.

r_ -___ -

I __ - II. - __ - . -.

"Z .P_ - _w - - -

..=}

.-G

_......... G .........

0.,,, ,,._e-r.ir.a:s............... 1..... - t F  : _,,

E _a r .. _ I

_u cm  : Q__ __~~_- rz _ _ _I _j_ -I ,=_,

__~ .. _

Tb.b FV.1 FISSION PRODUCT BARRIER MATRIX r 1

~

i

_-b-@

fts 6.,a-

.cS

..imrl -Z __ - .. . - 4, - ---- -

- rmrkimw

a. _c-a ,Z=:

I V-.__ Ell-PIE.

-. ni

-m I u_ I

.i P -[ 7 1

_ - -__ .. _-.. LtL-z l__ l I - - ---- -- -

_ _o~- I9=

lW___

I;;

I .. ____ I-=-=,1 I_

1:-

I W__

l_

-I

, _ 1 I

I I-

-4111PIT

- U - U -a - - a . - - l l l__ -l D1EI1 c~n:EI

1. 1- l_

T4I-wJ ..  :-. n..

'-U l -__

F=

1=- 1

. -- I l l __.__

l l

_._._I

_~~~ 1_

.__ W__ -

_- I F,-_r =-- =,Z, l--.

I_wS_

l l

I Modes: ...-=,2-.3 fa-, ..L'.

£.. Eil L.

r~l- 0-] M4n 1.2.3 l -) ) i '4...

Id¢.. __ Modes: Ct,  ; E-E -.- , ,- I od"-12,3 l - -I - --

RCS Specific Activity 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Specific Activity LCO 3.4.6 The specific activity of the reactor coolant shall be limited to DOSE EQUIVALENT 1-131 specific activity < 0.2 gCi /gm.

I APPLICABILITY: MODE 1, MODES 2 and 3 with any main steam line not isolated.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor coolant -------------NOTE-----------

specific activity LCO 3.0.4 is not applicable.

> 0.2 pCi/gm and

• 2.0 pCi/gm DOSE EQUIVALENT 1-131. A.1 Determine DOSE Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> EQUIVALENT I-131.

AD A.2 Restore DOSE 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> EQUIVALENT 1-131 to within limits.

B. Required Action and B.1 Determine DOSE Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion EQUIVALENT I-131.

Time of Condition A not met. AND OR B.2.1 Isolate all main 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> steam lines.

Reactor Coolant specific activity > 2.0 OR pCi/gm DOSE EQUIVALENT I-131.

(continued)

DAEC 3.4-13 Amendment 240

Nuclear Management Company Page 1 of 2 State Change History AR Pre- AR AR Screening Screen Submit to Screening Screening Que Done Initiate Screening Team Que Update 811912003 Complete & Close 101112003 8/1512003 8)15/2003 12:31:56 PM 6:17:12 AM by ZIMMERMAN, 6:09:45 PM by HARRISON, by MCVICKER, by SCHANBACHER, THOMAS A Owner 10:05:19 PM Owner EILEEN A Owner EDWARD E Owner DAEC WLUAM A SULLIVAN, (None)

(None) CAP Admin PAUL R Section 1 Activity Request Id: CAP028632 Activity Type: CAP Submit Date: 8115/2003 6:09:45 PM One iUne

Description:

EAL Threshold values not consistent with Initiating Condition - AU 1 Detailed

Description:

8/15/2003 6:09:45 PM - ZIMMERMAN, THOMAS A:

During EAL I Bases Document revisions it was found that EAL AU1 Threshold Values for Offgas Stack Kaman 9/10 and LLRPSF Kaman 12 were referenced at the ODAM release limit instead of referencing 2 times that value, as stated In the Initiating Condition. The Gaseous Effluent EALs table in the DAEC EAL Information section of the Bases has both values Indicated correctly.

Offgas Stack Kaman 9/10 EAL Threshold value currently: >1 .OE-1 mCI/cc, should be: >2.OE-1 mCi/cc LLRPSF Kaman 12 EAL Threshold value currently: >5.0E-4mClcc, should be: >1 .OE-3mCVcc Radiation Protection was contacted to verify that the Tech. Spec. (ODAM) values referenced in the table are correct. (see attached e-mail) Using the ODAM value as an EAL Threshold instead of 2 times the ODAM value is conservative. EALs for this Initiating Condition would have been entered in a timely manner (sooner than required).

Initiator: ZIMMERMAN, THOMAS A Initiator Department: BEP Emergency Planning DA Daternme of Discovery: 8/1512003 2:42:28 PM Date/Time of Occurrence: 8/1512003 2:42:28 PM Identified By: Site-identified System: (None)

Equipment# (1st): (None) Equipment Name (1st): (None)

Equipment # (2nd): (None) Equipment Name (2nd): (None)

Equipment# (3rd): (None) Equipment Name (3rd): (None)

Site/Unit: Duane Arnold Why did this occur?: 8/1512003 6:09:45 PM - ZIMMERMAN, THOMAS A:

Unknown.

Immediate Action Taken: 8/1512003 6:09:45 PM - ZIMMERMAN, THOMAS A:

Radiation Protection was contacted to verify that the Tech. Spec. (ODAM) values referenced in the table are correct (see attached e-mail)

EAL Threshold Values were verified to be appropriate for the remaining items in AU1 and also for all items in AA1.

Recommendations: 8/15/2003 6:09:45 PM -ZIMMERMAN, THOMAS A:

Initiate and track actions to raise the EAL Threshold value for Offgas Stack (Kaman) rad monitor reading to >2.OE-1mCicc and LLRPSF (Kaman) rad monitor reading to >1.0E-3mCi/cc In Initiating Condition AU1 for EAL Table, EAL-01 and in EAL Bases Document, EBD-A.

8/1812003 1:33:46 PM - MCVICKER, WILLIAM A:

AR Admin CAP Recommendations Significance Level: C Level of Effort: CA Assignment: review issues and Initiate any actions Responsible Work Group: Rad Protection Additional Reviews: N Mode Change Restraints: N Potential MRFF: N INPO OE: N Good Catch: N Hot Buttons: N SRO Review Required?: N Section 2 Operability Status: NA Compensatory Actions: N Basis for Operability: 8/15/2003 10:05:19 PM - HARRISON, EDWARD E:

Plant equipment operability not effected. As the initiator states the Unusal event would be entered conservatively-early under the current EAL document and the ALERT values are correct.

http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Template=printitem&recordid=47 1909&tableid=I 000 10/19/2004

Nuclear Management Company Page 2 of 2 Unplanned TSAC Entry: N External Notification: N !

Section 3 Screened?: Y Significance Level: C INPO OE Reqd?: N Potential MRFF?: N QAINuclear Oversight?: N Licensing Review?: N Good Catch/Well Doc'd?: NA Section 4 Inappropriate Action:

Process: EP - Emergency Preparedness Activity: AN - Analysis & Calcds Human Error Type: UNK- Unknown Human Perf Fail Mode: K6 - Inadeq. Standards Knowledge Equip Failure Mode: (None) Process Fail Mode: UNK - Unknown OrglMgt Failure Mode: UNK - Unknown Group Causing Prob: (None)

Hot Buttons: (None)

Section 5 CAP Admin: SULLIVAN, PAUL R CAP Owner. (None)

Project: Corrective Action Process (CAP) State: Done Activennactive: Inactive Submitter. ZIMMERMAN, THOMAS A Owner: (None) Last Modified Date: 10/812003 2:44:38 PM Last Modifier JOHNSON, DON A Last State Change Date: 1011/2003 6:17:12 AM Last State Changer SCHANBACHER, EILEEN A Close Date: 10/11M003 6:17:12 AM NUTRK ID:

1. of Children: 0

References:

Update:

Prescreen Comments:

Import Memo Field:

OPR Completed?: N OLD ACTION NUM:

sub tsid: 0 originaLprojectjid: 61 originalIssueId: 0228632 Site: Dtjane Arnold Cartridge and Frame:

Attachments and Parent/Child Links Offgas Stack / LLRPSF KAMAN setpoints (2094 bytes) by ZIMMERMAN, THOMAS A (8/1512003 5:45:22 PM)

Principal to CA029218: verne that the Tech. Spec.(ODAM) values referenced in the EAL table are correct by MCVICKER, WILUAM A (8/1912003 12:34:30 PM)

PriniDal to CA029219: EAL Threshold values not consistent wth Initiating Condition - AUI by MCVICKER, WILUAM A (8/19/2003 12:38:32 PM)

Change History 10/112003 6:17:12 AM by SCHANBACHER, EILEEN A Last State Changer Changed From HARRISON, EDWARD ETo SCHANBACHER, EILEEN A Close Date Changed From Unassigned To 10/1/2003 6:17:12 AM 10/8/2003 2:44:38 PM by JOHNSON, DON A Process Changed From (None) To EP - Emergency Preparedness Activity Changed From (None) To AN -Analysis & Caic's Human Error Type Changed From (None) To UNK - Unknown Human Perf Fall Mode Changed From (None) To K6 - Inadeq. Standards Knowledge Process Fail Mode Changed From (None) To UNK - Unknown Org/Mgt Failure Mode Changed From (None) To UNK - Unknown Last Modified Date Changed From 1011/2003 6:17:12 AM To 101812003 2:44:38 PM Last Modifier Changed From SCHANBACHER, EILEEN A To JOHNSON, DON A http://enws02ltmtrack/tmtrack.dll?IssuePage&Template=printitem&recordid=471909&tableid=1000 10/19/2004

From: Funk, Steve Sent: Friday, August 15, 2003 14:11 To: Zimmerman, Tom Cc: Wheeler, Dennis; Giorgio, Herb; Keith, Wendell; Sullivan, Paul

Subject:

Stack and LLRPSF KAMAN Setpoints

Tom, This note is in response to your request for information on EAL AU 1. This information should help to settle a discrepancy between the EAL Table EAL-01 and its corresponding basis document.

Section 5 of NEI 97-03, (Methodology for Development of Emeraencv Action Levels), provides guidance for the development of EALs. Page 5-A-3 of that document states that the AUI EAL is based on an effluent radiation monitor readout that is twice the ODCM (ODAM) set point It later alludes that the ODAM limit of interest is the limit that corresponds to the 10 CFR 20 limit which is Dose RATE related. (ODAM 6.2.2)

We demonstrate compliance with this limit by never releasing effluents at a rate that would result in an offsite dose RATE of 500mRem per YEAR. (-0.114 mRem per hour). The KAMAN Effluent monitors on all of the gaseous release points have High-High setpoints that correspond to a release rate of 500mRemlyear. The setpoint can vary based on the mix of isotopes and efficiency of the detector, but, in general there are only minimal fluctuations in the values over time.

The High-High Setpoints for the KAMANs are:

• K2- Turbine Building: 6.2E-4 uCVcc

• K4, K6, K8 - Reactor Building: 6.2E-4 uCVcc

• K10 Offgas Stack: 1.1E-1 uCi/cc

• K12 LLRPSF Building: 5.9E-4 uCi/cc sf

Nuclear Management Company Page 1 of 2 State Change History Assign Work Assign Work Conduct Review &

811912003 8/19/2003 Work Work Approval Initiate 12:34:28 PM Reassign 3:38:03 PM Assign Complete 9122/2003 811912003 by MCVICKER. Owner by WINDSCHILL, Owner by WHEELER, 3:43:08 PM 6:05:17 AM vMLLIAM A JOHN DENNIS by FUNK, Owner WINDSCHILL, WHEELER, Owner FUNK. STEPHAN M WHEELER.

JOHN DENNIS STEPHAN M DENNIS Quality Check Complete and Done Approved 9/22/2003 Close by WHEELER, 6:56:37 AM 9122/2003 Owner DAEC by WHEELER, 6:56:50 AM DENNIS DENNIS Owner (None)

CAP Admin Section 1 Activity Request Id: CA02921 8 Activity Type: Corrective Action Submit Date: 8/19/2003 12:34:28 PM Site/Unit: Duane Arnold One Line

Description:

verify that the Tech. Spec.(ODAM) values referenced in the EAL table are correct Activity Requested: Verify that the Tech. Spec. (ODAM) values referenced in the EAL table are correct. Furnish results to Emergency Planning.

Steve, this was sent to rad pro manager, is this an ODAM issue? If not please advise who is best qualified to perform the required corrective actions.

Denny CATPR: N Mode Change Restraint (None)

Initiator. ZIMMERMAN, THOMAS A Initiator Department: BEP Emergency Planning DA Responsible Group Code: DCHM Chemistry/Environmental DA Responsible Department: Plant Activity Supervisor: WHEELER, DENNIS Activity Performer: FUNK, STEPHAN M Section 2 Priority: 4 Due Date: 912512003 Management Exception From PI?: N QAINuclear Oversight?: N Licensing Review?: N NRC Commitment?: N NRC Commitment Date: Significance Level: C Section 3 Activity Completed: 9/22/2003 6:05:17 AM - FUNK, STEPHAN M:

EAL Form EAL-01, Rev. 4 lists offsite rad conditions In AUl and AU2 consistent with the current ODAMITS limits for the Reactor Building, Offgas Stack, Turbine Building and LLRPSF Building. The values listed are also consistent with what is described in the EAL Basis Document.

Hot Buttons: (None)

Section 4 QA Supervisor (None) Licensing Supervisor (None)

Section 5 Project: Other Corrective Action State: Done Activelinactive: Inactive Owner (None)

Submitter MCVICKER, WILLIAM A Assigned Date: 8/1912003 Last Modified Date: 9/22/2003 6:56:50 AM Last Modifier: WHEELER, DENNIS Last State Change Date: 9/22/2003 6:56:50 AM Last State Changer. WHEELER, DENNIS Close Date: 9/22/2003 6:56:50 AM NUTRK ID:

Child Number: 0

References:

Update:

Import Memo Field:

CAP Admin: DAEC CAP Admin Site: Duane Arnold OLD ACTION NUM:

Cartridge and Frame:

Attachments and ParentlChild Links Subtask from CAP028632: EAL Threshold values not consistent with Initiating Condition - AUI by MCVICKER, WILLIAM A (8/19/2003 12:34:30 PM) http://enwsO2/tmtrack/tmtrack.dU?IssuePage&RecordId=472596&Tableld= 1OOO&Template-printitem 10/19/2004

Nuclear Management Company Page 2 of 2 Change History 9/2212003 6:56:37 AM by WHEELER, DENNIS Last Modifier Changed From FUNK, STEPHAN M To WHEELER, DENNIS Last State Change Date Changed From 912212003 6:05:17 AM To 912212003 6:56:37 AM Lost State Changer Changed From FUNK, STEPHAN M To WHEELER DENNIS CAP Admin Changed From (None) To DAEC CAP Admin 912212003 6:56:50 AM by WHEELER, DENNIS State Changed From Quality Check To Done Via Transition: Complete and Close Activelinactive Changed From Active To Inactive Owner Changed From DAEC CAP Admin To (None)

Last Modified Date Changed From 912212003 6:56:37 AM To 9)2212003 6:56:50 AM Last State Change Date Changed From 9/22/2003 6:56:37 AM To 9122/2003 6:56:50 AM Close Date Changed From Unassigned To 9/2212003 6:56:50 AM http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Recordld=472596&TableId=I 000&Template=printitem IO/19/2004

Page 1 of2 Nuclear Management Company Company Page I of 2 State Change History x Assign Work Conduct Work Review &

Approval Quality Initiate 8/1912003 Assign 8/19/2003 3:02:04 Work Complete Approved Check 12:38:06 PM PM 8/29/2003 by MCVICKER. by JOHNSON, by ZIMMERMAN, 10:00:05AM 8/29/2003 Owner Owner by JOHNSON. 10:57,44 AM WILLIAM A SULLIVAN, DON A ZIMMERMAN, THOMAS A Owner DON A SULLIVAN, Owner DAEC I PAUL R THOMAS A CAP Admin PAUL R I

- iI Complete and Done Close 829/2003 by JOHNSON, 10:57:58 AM DON A Owner (None)

Section I Activity Request Id: CA029219 Activity Type: Corrective Action Submit Date: 8119/2003 12:38:06 PM SitelUnit: Duane Arnold One Line

Description:

EAL Threshold values not consistent with Initiating Condition - AU1 Activity Requested: Make changes, as necessary, to the EAL Table and the EAL Bases Document after ODAM Threshold values are determined correct or not. (Work with PCR 28641)

CATPR: N Mode Change Restraint: (None)

Initiator: ZIMMERMAN, THOMAS A Initiator Department: BEP Emergency Planning DA Responsible Group Code: BEP Emergency Planning DA Responsible Department: Business Activity Supervisor: SULLIVAN, PAUL R Activity Performer: ZIMMERMAN, THOMAS A Section 2 Priority: 3 Due Date: 812912003 Management Exception From PI?: N QACNuclear Oversight?: N Licensing Review?: N NRC Commitment?: N NRC Commitment Date: Significance Level: C Section 3 Activity Completed: 8/29/2003 10:00:05 AM - ZIMMERMAN, THOMAS A:

Reviewed the EAL Bases Document, EBD-A and NEI 99-01, Rev. 4. After review and discussion with Rad. Protection, it has been verified that the correct value for the Unusual Event entry is 2 times the ODAM limit, as stated in the Bases Document, Gaseous Effluent EALs table. The EAL AUI Threshold Values for Offgas Stack Kaman 9/10 and LLRPSF Kaman 12 have been made 2 times the ODAM limit.

For EAL Bases Document, EBD-A, the PWR 21931, has been completed. Appropriate training was performed per TMAR TM2003-1074.

Revision 6, of EBD-A is approved and being distributed. Effective date of change is 8/29103.

For EAL Table, EAL-01, the PWR 21932, has been completed. Appropriate training was performed per TMAR TM2003-1074.

Revision 4, of EAL-01 Is approved and being distributed. Effective date of change is 8129103.

This activity has been completed and is ready for closure.

Hot Buttons: (None)

Section 4 QA Supervisor: (None) Licensing Supervisor: (None)

Section 5 Project Other Corrective Action State: Done Active/lInactive: Inactive Owner: (None)

Submitter: MCVICKER, WILLIAM A Assigned Date: 8/19/2003 t Last Modified Date: 8/29/2003 10:57:58 AM Last Modifier, JOHNSON, DON A Last State Change Date: 8/29/2003 10:57:58 AM Last State Changer.JOHNSON, DON A Close Date: 8/2912003 10:57:58 AM NUTRK ID:

Child Number 0

References:

Update:

Import Memo Field:

CAP Admin: DAEC CAP Admin Site: Duane Arnold http://enwsO2/tmtrack/tmtrack.dll?IssuePage&Recordld=472598&TableId=I OOO&Template=printitem 10/19/2004

Nuclear Management Company Page 2 of 2 OLDACTIONNUM:

Cartridge and Frame:

Attachments and Parent/Child Links Subtask from CAP028632: EAL Threshold values not consistent with Initiatina Condition - AU1 by MCVICKER, WILLIAM A (8/1912003 12:38:32 PM)

Change History 812912003 10:57:44 AM by JOHNSON, DON A Last Modifier Changed From ZIMMERMAN, THOMAS A To JOHNSON, DON A Last State Change Date Changed From 8129/2003 10:00:05AM To 812912003 10:57:44 AM Last State Changer Changed From ZIMMERMAN, THOMAS A To JOHNSON, DON A CAP Admin Changed From (None) To DAEC CAP Admin 812912003 10:57:58 AM by JOHNSON, DON A State Changed From Ouality Check To Done Via Transition: Complete and Close Activellnactive Changed From Active To Inactive Owner Changed From DAEC CAP Admin To (None)

Last Modified Date Changed From 812912003 10:57:44 AM To 82912003 10:57:58 AM Last State Change Date Changed From 812912003 10:57:44 AM To 812912003 10:57:58 AM Close Date Changed From Unassigned To 8/29/2003 10:57:58 AM http://enwsO2/tmtrack/tmtrack.dll?lssuePage&Recordld=472598&Tableld=I O00&Template=printitem 10/19/2004

I ATTACHMENT 3 JUSTIFICATION MATRIX 36 pages follow

Last SERAppro ved EAL's NEI99-1 Rev. sEAL's 4 Proposed'EAL's DEfference or Explanation Devilation,-

AU1 AU1 RUI DIFFERENCE

• Do not have a perimeter radiation monitoring system Any Unplanned Release of Gaseous or Liquid Any UNPLANNED Release of Gaseous or Liquid Any Unplanned Release of Gaseous or Liquid

• Lsting specific values with instrument iDs and names Radioactivity to the Environment That Exceeds Radioacthivty to the Environment that Exceeds Radioactivity to the Environment That Exceeds Two Times the Radiological Technical Two Times the Radiological Effluent Technical Two Times the Offhite Dose Assessment Manual

• and is expected to continue for 60-minutes or longer Specifications For 60 Minutes or Longer Specifications for 60 Minutes or Longer. (ODAM) Umit and Is Expected to Continue For 60 meets Intent of for 60-minutes or longer and Is Op. Modes: ALL Minutes or Longer consistent with NEI 99-1 rev 4 bases Infbrmation.

Operating Mode Applicability All

• Consistent with current NEI EAL and last SER approved Operating Modes: ALL EAL Valid Reactor Building or Turbine Building ventilation VALID reading on any effluent monitor that exceeds

• See the EAL Bases Document for more information.

(Kaman) rad monitor reading ABOVE 1 E-3 pCtolcfor two times the alarm selpoInt established by a current RUI11 Valid Reactor Building ventilation rad monitor

• DAEC does not have automatic dose assessment more than 60 minutes. radioadvity discharge permit for 60 minutes or longer. (Kaman 3/4.56,7/8) or Turbine Building OR ventilation red monitor (Kaman 1/2) reading capablity.

that exceeds I E-3 pCVcc and Is expected to

• Changing A to R for IC designator to signify Rad Valid Offgas Stack (Kaman) red monitor reading VALID reading on one or more of the foliowing continue for 60 minutes or longer. Table.

ABOVE 6 E-1 isCVcc for more than 60 minutes. radiation monitors that exceeds the reading shown Otr OR 60 minutes or ionger (site-specific lst) OR

• DAEC Corredve Adon CAP 28632 reverifed the OR RU .2 Vanid Offgas Stack red monitor (Kaman 9/10) selpoints as attached below. The Initial approved SER Vaid LLRPSF (Kaman) red monitor reading ABOVE 9 reading that exceeds 2.0 E-1 ICi/cc and is had the Incorrect red values for the Offgas Stack and E-4 ItCic for more than 60 minubts Confirmed sample anaiytes for gaseous or liquid expected to continue for 60 minutes or longer. LLRPSF red monitors. CAP 28632 was initiated to releases Indicates concentrations or release rates. vereisy this the table and seilyoin were corrct and to OR with a release duration of 60 minutes or longer, In OR revise thb EAL accordingly Valid GSW rad monitorreading ABOVE 3E+3 CPSfor excess of two times (site specki technical RUI.3 Vald LLRPSF red monitor (Kaman 12) Added sequence Identifiers for EALs to more than 60 mrJutem spedlications). reading that exceeds 1.0 E-3 CICVoe and is help delineate the EALs and aid In OR OR expected to continue for 60 minutes or longerh communicating these EALs to the State &

Vand RHRSW & ESW rad monitor reading ABOVE VALID reading on perimeter radiafon montoring OR Counties. (NOTE: THIS COMMENT IS 8E.2CPSformorethan60minutes. system greater than 0.10 mRlhrABOVE normal RUIA Valid GSWrad monitor (RIS-4767) reading APPLICABLE THROUGHOUT THIS OR background sustahied for 60 minutes or longer [for that exceeds 3E+3 CPS and Is expected to MATRIX) sites having telemetered perimeter monitors], continue for 60 minutes or longer.

Valid RHRSW & ESW Discharge Canal rad monior OR OR reading ABOVE 1E4-3CPS for more than 60 minutes. ____________________

VALID indication on automatic real-time dose RUt15 Valid RHRSW & ESW red monitor (RM-1997) GASEOUS EFFLUENTEALS OR assessment capability greater than (site specific reading that exceeds 8E+2 CPS and Is oftlesebcl Ksr""r0 5515 , aidii K ln.2Jd Conlirmed sample analyses for gaseous or liquid value) for 60 minutes or longer [for sites having such expected to continue for 60 minutes or longer. 10LW ._i 7 releases indicates concentrations Inexcess of 2 times capability]. OR ( r l Conceetatlai Releas *RX ODAM bmiritsfor greater than 60 minutes O eeoeU. Pcicc Rate n b ReoRe RUI .6 Valid RHRSW & ESW Rupture Disc red eleastt sO (.CVci ((iteo OR monitor (RM-4268) reading that exceeds ESt- (5 I 5I e1 &4e . 2 I E.4 Dose assessment determines houriy dose outside the IE+3 CPS and Is expected to continue for 60 Al.t tEa60 t/Z TIS 2 t l OLt7 3 7L-2 11 2Lt6 site boundary ABOVE 0.1 mrem TEDE. minutes or longer. LLRPSFKmmn12 ORAa,,,~,nii 111( FA. 0900 ORlhas ku L n-s Loft Conson hWT 141ebo RUI .7 Confirmed sample analyses for gaseous or 1-11 SWC 5.9-t 2 e+4 liquid releases Indicates concentrations or Umnmwl Eren(l xS aZ I OL43 O 0S 5 release rates In excess of 2 times ODAM Al." (2D1aX I OL-I 5 01.6 limits and Is expected to continue for 60 minutes or longer.

/

Difference or Ex- nto Last SER Approved EAL's NEI 990 Rev. 4 EAL's Proposed EAL's Deviation AU2 AU2 RU2

• Change In format Need to ensure that the Decision Makers understand that there are 2 separate ICs for this Unexpected Increase In Plant Radiation Unexpected Increase In Plant Radiation. Unexpected Increase In Plant Radiation EALk DAEC does not number the ICs unless absolutely Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL necessary for human factors considerations.

• Consistent with current NEI EAL Uncontrolled loss of reactor cavity or fuel pool water VAUD (sitespecfic) Indication of uncontrolled water RU2.1 Unplanned valid Refuel Floor ARM reading

• toss of lever Is equivalent to tevel decrease.

level with an spent uel assembies remaining water level decrease Inthe reactor refueling cavity, spent Increase with an uncontrolled loss of reactor covered as Indicated by ANY of the following: fuel pool, or fuel transfer canal with al irradiated fuel cavity, fuel pool, or fuel transfer canal water

• valld dnrct area radation montor' Is equivalent to assemblies remaining covered by water, level with all Irradiated fuel assemblies 'Reh Floor ARM.

• Report to controi mom AND remaining covered by water as indicated by

• Added 'any to the ARM threshold to ensure an ARMs

• Valid fuel pool level Indication (Ll-3413) any of the followfhg are considered.

BELOW 36 feet and lowering Unplanned VALID (site-specific) Direct Area Radiation

• Report to control room . Changing A to R for IC designator to signify Rad Monitor reading Increases*Reottcoto om*Cagn'AtofrI sntrtosiyRd

• Valid WR GEMAC Floodup Indcathn (- o r OR

• Valid fuel pool level indication (L.- DIFFERENCE Table.

4541) coming on scale. OR3413) LESS THAN 36 feet and Unplanned VAUD Direct Area Radiation Monitor lowering OR readings Increases by a factor of 1000 over normar

• Valid WR GEMAC Floodup Indication Unexpected ARM reading offscale high or ABOVE (LI4541) coming on scale.

1000 times normal reading. OR

'Normal levels can be considered as the highest reading In the past twenty-four hours excluding the RU2.2 Any unplanned ARM reading offscale high or current peak value. GREATER THAN 1000 times normar reading.

'Normal levels can be considered as the highest reading hIthe past twentytourhours excluding the current peak value.

LastSERApprovedEALs NEt994tRev 4EAL's [ Proposed EAL's Dlotio, Ex.tton AA1 AAI RAI DIFFERENCE

• Do not have a perimeter radiation montoring system.

Any Unplanned Release of Gaseous or Liquid Any UNPLANNED Release of Gaseous or Liquld Any Unplanned Release of Gaseous or Liquid

• and Is expected to continue for 15 minutes or longer Radloactiviy to the Environment That Exceeds Radloactivty to the Environment that Exceeds 200 Radloactivty to the Environment that Exceeds meets Intent of 'for 15 -mnutes or longer' and Is 200 Times the Radiological Technical Times the Radiological Effluent Technical 200X the Offsaite Dose Assessment Manual consistent with NEI 99-01 rev 4 bases Information.

Specifications For 15 Minutes or Longer Specifications for 15 Minutes or Longer. (ODAM) Limit and Is Expected to Continue for tS Minutes or Longer

• Consistent with current NEI EAL and last SER approved EAL Operating Modes: ALL Op. Modes: ALL Operating Modes: ALL

• See the EAL Bases Document for more Information.

• DAEC does not have automatic dose assessment RA .1 Valid Reactor Building ventilation red monRitor Valid Reactor Bulding or Turbine Building ventilation VALID reading on any effluent monitor that exceeds capability.

(Kaman 3t4. 5M.718) or Turbine Bulding (Kaman) red monitor reading ABOVE 3 E-2 pCiko for 200 times the alarm set;oint established by a current ventilation red monitor (Kaman 1/2) reading

• Changing 'A to 'R' for IC designator to signify Rad more than 15 minutes. radioactivty discharge permit for 15 minutes or longer. that exceeds 3 E-2 pCVce and is expected to Table.

OR OR continue for 15 minutes or longer.

• DAEC Corrective Action CAP 28632 reverified the Valid Offgas Stack (Kaman) red monitor reading VALID reading on one or more of the following OR setpoints as attached below. The Initial approved SER ABOVE 2 E+1 pCI/c formore than 15 minutes. radiation monitors that exceeds the reading shown for had the incorrect red values for the Offgas Stack and RA1.2 Valid Offgas Stack red monitor (Kaman 9/1 0) LLRPSF red montors. CAP 28632 was Initiated to 15 minutes or longer (site-specific list)

OR reading that exceeds 8 E+0 pClcc and Is verify that the table and setpoints were correct and to OR expected to continue for 15 minutes or longer. revise this EAL accordingly.

Valid LLRPSF (Kaman) red monitor reading ABOVE 9 E-2 pCVcc for more than 15 minutes. Confirmed sample analyses for gaseous of liquid OR releases Indicates concentrations or release rates.

OR with a release duration of 15 minutes or longer, In RA1.3 Valid LLRPSF red monitor (Kaman 12) excess of 200 times (site speilfle technical reading that exceeds 1 E-1 pCice and Is CASEOUSE£FLULENT Vaiid GSW red monitor reading ABOVE 3E+5 CPS for expected to continue for 15 minutes or longer.

specifications). 0119Kwn 31ck M .,..,erdgtI.nw 112..*d more than 15 minutes.

OR OR Mn,0o.C 0:00o 72OM OR VALID reading on perimeter radiation monitoring RA1A Valid GSW red monitor (RIS-4767) reading that Valid RHRSW & ESW red monitor reading ABOVE exceeds 3E+S CPS and Isexpected to 8E+4 CPS for more than 15 minutes. system greater than 10.0 mRfl ABOVE normal background sustained for 15 minutes or longer [for continue for 5 minutes orlonger.

I 2LMk003 241:J-4 OR sites having telemetered perimeter monitors). OR ~jorj*1 7L-2 I 3L.6I 3ZU*Z Valid RHRSW & ESW Discharge Canal red monitor OR RA1.5 Valid RHRSW& ESWrad monkor (RM-1997)

LLRPSFK~ammn 12 reading ABOVE 1E+5 CPS for more than 15 minuties. reading that exceeds 8E+4 CPS and Is VALID Indication on automatic real-time dose OR assessment capability greater than (site specific expected to continue for 15 minutes o onger. AI..t x 75 1200 Confirmed sample analyses for gaseous or liquid value) for 15 minutes or longer [for sites having such OR releases Indicates concentrations In excess of 200 times capability]. RA .6 Valid RHRSW & ESW Rupture Disc rad ODAM limits for greater than 15 minutes. monitor (RM-4268) reading that exceeds 1E+5 OR CPS and is expected to continue for 15 minutes or longer.

Valid field survey reading outside the site boundary 1O mRlhr or '50 mRlhr CDE Thyroid. OR OR RA1.7 Confirmed sample analyses for gaseous or liquid releases indicates concentrations or Dose assessment determines hourly dose outside the release rates with a release duration expected site boundary ABOVE 10 mrern TEDE. to continue for 15 minutes or longer in excess of 200 times ODAM limit.

Diference, orExlntn Last SER ApprovedEAL's NEI 99-01 Rw. 4 EAL's ProposedEAL's D.aanttlon AA2 AA2 RA2

• Usting specific values with Instrument IDs and names Major Damage to Irradbated Fuel or Loss of Water Damage to Irradiated Fuel or Loss of Water Level Damage to Irradiated Fuel or Loss of Water Level

• Requiring valid water level reading does not change Level that Has or Will Result In the Uncovering of that Has or Will Result In the Uncovering of that Has or Will Result In the Uncovering of Intent and Is consistent with NEI 99d01 rev 4 Irradiated Fuel OutsIde the Reactor Vessel Irradiated Fuel OutsIde the Reactor Vessel. Irradiated Fuel Outside the Reactor Vessel

• Consistent with current NEI EAL.

Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL

• The level Indication (113413) Is also used for the Fuel Transfer Canal.

Report of ANY of the following: A VALID (site-specilic) alarm or reading on one or RA2.1 Report of any of the following:

• Changing A to 'R for IC designator to signify Rad

• Vald ARM HI RAD aiamm for the more of the following radiation monitors: (site-specdfc

• Valid ARM Hi Rad alarm for the Rehleling Table.

Refueling Floor North End, Reftuelng m Floor North End (RM 9163). Refueling Floor South End, New Fuel Storage Area, Floor South End (RM 9164), New Fuel or Spent Fuel Storage Area Rehel FloorArea Radiation Monitor Storage (RM 9153), or Spent Fuel Storage

• Valid RehUeling Floor North End, Refueling Area (RM 9178).

Floor South End, or New Fuel Storage

• Vald Refueling Floor North End (RM-Area ARM ReadFg ABOVE 10 mR/hr Fuel Handling Building Ventilation Monitor 9163), Refueling Floor South End (RM DIFFERENCE

• ARM Vaiid SpentFueiStorage Area Vaid pentFue StoageAreaARM91 64),ARM g9153) or New Fuel Storag Reading GREATERAreaTHAN (RM- 10 Reading ABOVE 100 mRlhr Refueling Bridge Area Radiation Monitor mRem/hr OR OR

• Valid Spent Fuel Storage Area (RM-91 78)

Report or visual observation of rradiated Fuel Water level less than (site-specific) feet for the reactor ARM Reading GREATER THAN 100 unovered refueling cavity, spent fel pool and uel transfer canal mRemrrr OR that will result in Irradiated fuel uncovering. OR Water level reading BELOW 450' as Indicated on RA22 Valid water level reading LESS THAN 450 L14541 (floodup) for the Reactor Refellng Cavity that Inches as indicated on L-4541 (floodup) for the will result In Irradiated Fuel uncovering Reactor Refueling Cavity that witl result In OR Irradiated Fuel uncoverng.

Vaitd Fuel Pool water level Indication (Ll-3413) BELOW OR 16 feet. RA2.3 Valid Fuel Pool water level Indication (Li-3413)

LESS THAN 16 feet that will result In irradiated Fuel uncovering.

NEI 99.01 Rev. 4 EAL's Proposed EAL's Deaonor*

Differenc* Explanation Last SER Approved EAL's AA3 AA3 RA3

• Added Instrument IDs and names.

Release of Radioactive Material or Increases In Release of Radioactive Material or Increases In Release of Radioactive Material or Increases In

• Consistent with last SER Approved EAL and current NEI Radiation Levels Within the Facility That Impedes RadiatIon Levels Within the Facillty That Impedes Radiation Levels Within the Facility That Impedes EAL Operation of Systems Required to MaIntaIn Safe Operation of Systems RequIred to MaIntain Safe Operation of Systems RequIred to MaintaIn Safe Operations or to Establish or to Maintain Cold Operations or to Establish or Maintain Cold Operations or to Establish or to MaIntaIn Cold

• Although the Rad Waste Control Room Is continuously Shutdown Shutdown Shutdown staffed, It Is not required to maintain safe plant operation or perform a safe plant shutdown.

• Changing A to R for IC designator to signify Rad Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL Table.

• Added the Central Alarm and Secondary Alamm Stations Vaild area radiation monitor(RE9162) reading greater VALID (site-specilic) radiation monitor readings RA3.1. Valid area radiation levels GREATER THAN DIFFERENCE to RA3.1 as these stations are necessary to ensure than 15 mR/hr Inthe Control Room. GREATER THAN 15 mR/hr Inareas requirIng 15 mRemlhr Inany of the following areas: consistent access to the plant and are normally OR occupancy tohit continuous(Site-specific) maintain plant safety

• Contmi Room (RM 9162) occupied functions:

Valid area radiation monitor (RE91 68) reading greater

• Central Alarm Station (by survey) than 500 mRlhrr adthe Remote Shutddown Panel. 1C88. OR CetaAarSaio(bsuvy VALID (site-specidic) radiation monitor readings

• Secondary Alarm Station (by survey)

GREATER ThAN 'site spedfifc vaiues Inareas OR requiring Infrequent access to maintain plant safety functions. (Slle speciti~c) list RA32. Valid area radiatbon mionilor (RE-9168). reading GREATER THAN 500 mRemlhr affecting the Remote Shutdown Panel. 1C388.

Last SER Approved EAL's NEt 9S-Ot Rev. 4 EAL's - Proposed EALt's Dference or D Cnsstntwih as SER ppratovenA n urn E ASI ASI RS1 *Consistent vwih last SER Approved EAL and current NEI Offslte Dose Resulting from an Actual or Imminent EAL Site Boundary Dose Resulting from an Actual or Offslte Dose Resulting from an Actual or Imminent Imminent Release of Gaseous Radloacttrty Release of Gaseous Radioactivity Exceeds 100 Release of Gaseous RadIoactIvity Exceeds 100

• Do not have a perimeter red monItoring system.

Exceeds 100 mrem TEDE or 500 mrem CDE mR TEDE or 500 mR Thyroid CDE for the Actual or mRem TEDE or 500 mRem CDE ThyroId for the Projected Duration of the Release. Actual or Projected Duration of the Release

• GREATER THAN ....and Is expected to continue for XX Thyroid for the Actual or Projected Duration of the minutes or longer Is equivalent to 'that exceeds or Is Release expected to continue for XX minutes or longes.

Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL

• Changing A to 'R for IC designator to signify Red Table.

Valid Reactor Building or Turbine Building ventlation VALID reading on one or more of the following RS1.1 Dose assessment using actual meteorology

• Changed order of EALs to put preferred method first (Kaman) rad monitor reading ABOVE 6 E-2 pCVxc for radiaffon monitors that exceeds or Isexpected to Indicates doses GREATER THAN 100 mRem more than 15 minutes. (Dose assessment not

• Added at or beyond the site boundary to RS1.3 for the exceed the reading shown for 15 minutes or longer TEDE or 500 mRemn thyroid CDE at or beyond dose rate threshold to ensure that EAL Dedsion-Makers available) (site-spectifc list) the sie boundary. (Preferred method) understand the boundaries of the thresholds.

OR OR OR Vaiid Offgas Stack (Kaman) red monitor reading Dose assessment using actual meteorology indicates RS12 If DoseAssessment I unavaiable anyofthe ABOVE 4 E+t pCVcc for morm than 15 minutes. (Dose doses greater than 100 mR TEDE or 500 mR thyroid fotlowing:

assessment not available) CDE at or beyond the sie boundary.

• Valid Reactor Building ventilation red DIFFERENCE OR OR monitor (Kaman 3/4. 56, 718) or Vaild field survey reading outside the site boundary A VALID reading sustained for 15 minutes or longer Turbine Building ventilation rad monitor

>100 mRthror >500 mRlhr CDE Thyroid. on perimeter radiation monitoring system greater than (Kaman 112)reading GREATER THAN 100 mR/hr. Ifor sites having telemetered perimeter 6 E-2 ptCVcc and Is expected to continue OR for 15 minutes or longer.

monitors]

Dose assessment determines Integrated accident dose projection outside the site boundary ABOVE 100 OR

• Vald Offgas Stack rad monitor (Kaman mrem TEDE or ABOVE 500 mremn CDE Thyroid. 9110) reading GREATER THAN 4 E+l Field survey results Indicate dosed window dose rates exceeding 100 mR/hr expected to continue for more pCVcc and Isexpected to continue for 15 than one hour; or analyses of field survey samples minutes or longer.

Indicate thyroid CDE of 500 mR for one hour of OR Inhalation, at or beyond the sHe boundary. RSI .3 Field survey results Indicate dosed window dose rates exceeding 100 mRem/hr expected to continue for more than one hour at or beyond the sie boundary; or analyses of field survey samples indicate thyroid CDE of 500 mRem for one hour of Inhalation at or beyond the sie boundary.

Last SER Approved EAL's 1

NEI 99-01 Rev. 4 EAL's 1 Proposed EAL's Dlfference or D~evition Eplanatlori AGl AGI RGI *Consistent with last SER Approved EAL and current NEI Offslte Dose Resulting from an Actual or Imminent EAL Site Boundary Dose Resulitng from an Actual or' Offslte Dose Resulting from en Actual or Imminent ImmInent Release of Gaseous Radloactivty Release of Gaseous Radloactivty Exceeds 1000 Release of Gaseous Radloact"t that Exceeds

• Do not have a perimeter radiation monitoring system Exceeds 1000 mrem TEDE or 5000 mrem COE mR TEDE or 5000 MR Tyroid CDE for the Actual 1000 mRem TEDE or 5000 mRem CDE ThyroId for the Actual or Projected Duration of the Release

• Added equipment IDs and names Thyroid for the Actual or Projected Duration of the or Projected DuratIon of the Release Using Actual Release Meteorology. UsIng Actual Meteorology * 'and Is expected to continue ..... meets Intent of 'for xx-minutes or longer' and Isconsistent with NEIl9M-1 rev 4 bases Information.

Op. Modes: ALL. Operating Modes: ALL Operating Modes: ALL

• Reformatted for clarity and to ensure preferred method Is listed first.

Valid Reactor Buildling or Turbine Building ventIlatIon VALID reading on one or more of the foliowing RG1.1 Dose assessment using actual meteorology

• Changing IA to 'R for ICdesignator to signify Rad (Kaman) rad monitor reading ABOVE 6 E-1 pCI/cc for radiation monitors that exceeds or expected to exceed Indicates doses GREATER THAN 1000 Table.

more than 15 minutes. (Dose assessment not the reading shown for 15 minutes or longer~(sute- mRern TEDE or 5000 mRern thyroid CDE at available) specific list) or beyond the site boundary. (Preferred

• Changed order of EAts to put preferred method first.

method)

• Added 'at or beyond the site boundary' to RGt .3for the OR OR OR dose rate threshold to ensure that EAL Decision-Makers Valid Offgas Stack (Kaman) rad monitor raig Dose assessment using actual Meteorology Indicates; understand the boundaries of the thresholds.

ABOVE 4 E+2 pCI/cc for more than 15 ml=te.(Dose doses greater than 1000 mR TEDE or 5000 MR RG1.2 iffDose AssesmentIs unaailable, ethrofvth assessment not available) Ihyrid CDE at or beyond the site boundary. foilowing:

DIFFERENCE OR OR

• Valid Reactor Building ventilation rad A VALID reading sustained for 15 minutes or longer monitor (Kaman 3/4, 5/8.7/8) or Valid fleld survey reading outside the site boundary Turbine Building ventilation rad

'1,000 mR/hr or >5.O00mRflrr CDE Thyroid. on perimeter radiation monitoring system greater than 1000 mR/hr. [for sites having telemetered perimeter monitor (Kaman 112) reading OR monitors] GREATER THAN 6 E-I pCVcc and Is expected to continue fortS5 minutes Dose assessment determines Integrated accident OR or longer.

dose prolectlon outside the site boundary ABOVE 1,000 mrnrm TEDE or ABOVE 5.000 mrenm CDE Field survey results indicate closed window dose rates

• Vaid Offgas Stack rad monitor Thyroid. exceeding 1000 mnR/hr expected to continue for more (Kaman 9/10) reading GREATER than one hour~ or analyses of field survey samples THAN 4 E+2 pCI/cc and Isexpected to Indicate thyroid CDE of 500 MR for one hour of Inhalation, at or beyond site boundary. continue for 15 minutes or longer.

OR RGI .3 Field survey results Indicate closed window dose rates exceeding 1000 mRermlhr expected to continue for more than one hour at or beyond the site boundary; or analyses of field survey samples Indicate thyroid CDE of 5000 mnRemn for one hour of Inhalation at or beyond the site boundary.

FsSpo 1Difference or I Last SER Approved EAL's NEI 99.01 Rev. 4 EAL's Proposed EAL's Druto Explanation H1u HU1 HUI DIFFERENCE

• Added reference to the Safe ShutdownfVltal Areas Table Natural and Destructive Phenomena Affecting the Natural and DestructIve Phenomena Affecting the and Systems of Concern Table Inthe applicable ICs to Natural and Destructive Phenomena Affecting the ensure consistent understanding of what area and/or Protected Area PROTECTED AREA. Protected Area systems we are concerned about. These tables are also included on the EAL Boards used by the ERO.

Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL

• Consistent with last SER Approved EAL and current NEI EAL Earthquake detected per AOP 901, Earthquake. (SIte-Specfldc) method indicates felt earthquake. HU1.1 Earthquake detected per AOP 901, Earthquake

• repor Is equivalent to report by plant personner OR OR OR

• Added swltchyard' to IC's with Plant Protected Area.

Report of tornado touching down within plant protected Report by plant personnel of tornado or high winds HU1.2 Report of a tornado touching down within the

• Added wllh NO conlirmed damage... to several ICs to area or within swItchyard. greater than (site-specific) mph striking within Plant Protected Area, or withi the switchyard, clarify that this Is a NOUE and to be consistent with NEI PROTECTED AREA boundary. with NO confirmed damage to a Safe OR Shutdowrsltal Area or Control Room indication bases.

OR W.. 5hd-,, A-s I Assessment by the control moon that an event has of degraded performance of a System of occurred. Vehicle crash Into plant structures or systems within Concern. IGI UlJiL)- y I W-3dy~ k 1W L, lUv.o~ I~

PROTECTED AREA boundary. O.drIraiPr. Mr Ho BallWy Rw¢ a w HocRae. co Sraing OR OR OR Hoar icTagsR... WA, S8t,*.PFib ww Vehide crash Into plant structures or systems within HU1.3 Report of winds greater than 95 mph within 0.r protected area boundary. Report by plant personnel of an unanticipated the Plant Protected Area. or within the switchyard. with NO confirmed damage to a UM I~L C c.. oo.m 'C Ro .. CIC Hoc,,. RHH ya Rns H om EXPLOSION within PROTECTED AREA boundary OR resulting InVISIBLE DAMAGE to permanent structure Safe Shutdown/ital Area or Control Room s .P, =CR=.. R Report of an unanticipated explosion within the or equipment. indication of degraded perfornance of a protected area boundary resullAng In visible damage to System of Concern.

OR structures or equipmient. OR Report of turbine failure resulting Incasing penetration OR or damage to turbine or generator seals. HU1.4 Vehicle crash Into plant strudures or systems Turbine failure resulting In casing penetration or within the Plant Protected Area with NO eeoncrnaVJ- Syse" OR confirmed damage to a Safe ShutdowNltal damage to turbine or generator seals.

Uncontrolled flooding In(site-specific) areas of the Area or Control Room ndication of degraded

• ROG&tA C Pe4. W OR plant that has the potential to affect safety related performance of a System of Concern.

River level ABOVE 757 feeL equipment needed for the current operating mode. OR

• OnhobAC POWE0 OR OR HU1.5 Report of an unanticipated explosion within the

• kuhrr.wtAC Any area water level ABOVE Max Normal Operating (Site-Specilc) occurrences affecting the PROTECTED Plant Protected Area resulting Invisible

• OCPMWe damage to permanent structures or equipment.

• R wNIde 5,c~ceoCace v vy LIMrL AREA.

OR OR River level BELOW 725 feet 6 inches. HU1.6 Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

OR HUI.7 River level ABOVE 757 feet.

OR HUI.8 Uncontrolled flooding In a Safe ShutdownNtal Area that has the potential to affect safety related equipment needed for the current operating mode.

OR HUI.9 River level BELOW 725 feet 6 inches.

Last SER Approved EIL's NE 99-01 Rev. 4 EAL's Proposed EAL's Deviftron Eorplanation HU2 HU2 HU2

• Consistent with current NEI EAL Fire Within Safe Shutdown Areas Not Extinguished FIRE Within PROTECTED AREA Boundary Not Fire Within Protected Area Boundary Not Within 15 Minutes of Detection Extinguished Within 15 Minutes of Detection. Extinguished Within 15 Minutes of Detection

• Areas of concern are the Safe ShutdowrVital Areas.

Chose to define a single source of affected areas for EAts as a human factors improvement opportunity. By Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL using the Safe Shutdown/tal Areas table as a consistent reference for applicable EALs. EAL decision-nakers do not need to waste time referring to other Fire In buildings orareas contguousto any ofthe FIRE in bulidings or areas contiguous to any ofthe HU2.1 Firein butidngsorareas contguous to any NONE tables or lists of impacted areas. This Is a conservative following areas not extinguished within 15 minutes of following (site-specfic) areas not extinguished within Safe ShutdownVtal Area not extinguished decision that has been made to aid in timely EAL control room noti1cation or verification of a control room 15 minutes of control room notification or verification within 15 minutes of contro room notfication or decision making.

alarm: of a control room alarm: (Site-spedcf) list verification of a control room aiarm.

• Reactor, turbine, control, admintsecurity

• Intake structure

• PFump house HU3 HU3 HU3

• Consistent with last SER Approved EAL and current NEI Reiease of Toxic or Fiammable Gases Deemed Reliase of Toxic or Flammable Gases Doemed Release of Toxic or Flammabie Gases Deemed EAL Detrimental to Safe Operation of the Plant Detrimental to Normal Operation of the Plant. Detrimental to Normai Operation of the Plant.

Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL Toxic or flammable gas release affecting normal Report or detection of toxic or flammable gases that HU3M1Report or detection of toxic or flammable NONE operation, has or could enter the site area boundary Inamounts gases that has or could enter the site area OR that can affect NORMAL PLANT OPERATIONS, boundary in amounts that can affect normal Report by iocal, county or State ofidail for potential OR piant operations.

evacuation of site personnel based on offsite event. Report by Local, County or State Offidals hr OR evacuation or sheltering of site personnel based on an HU3.2 Report by Local, County or State Officials for offsite event, evacuation or sheltering of site personnel based on an offsite event.

Difference or £pssir Last SER Approved EAL's NEI 99-01 Rev. 4 EAL's Proposed EAL's DxorplanaVo HU4 HU4 HU4

• Consistent wth last SER Approved EAL and current NEI Confirmed Security Event Which Indicates a Confirmed Security Event Which Indicates a Confirmed Security Event Which Indicates a EAL Potential Degradation In the Level of Saehty of the Potential Degradation in the Level of Safety of the Potential Degradation in the Level of Safety ofthe

• Ust of events Is derived from DAEC Security Plan.

Plant Plant. Plant Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL Suspected sabotage device discovered within plant Security events as determined from (slte-spectfic) HU4.1 Credible Security Threat protected area and outside plant vital area. Safeguards Contingency Plan and reported by the OR (site-specilic) security shift supervision OR OR HU42 DAEC Securty Supervsion reports any of the Suspected sabotage device discovered inplant tfotcng swtchyard. A credible site specific security threat notification

• Suspected sabotage device discovered within plant Protected Area.

• Suspected sabotage device discovered outside the Protected NONE Area, in the plant swftchyard, or ISFSI.

• Confirmed tampering with safety related equipment.

• A hostagelextortion situation that disrupts normal plant or ISFSI operations.

• CMi disturbance or strike which disrupts normal plant or ISFSI operations.

• Internal disturbance that Is not short lived or that Is not a harmless outburst Inving one or more individuals within the Protected Area or ISFSI.

• Malevolent use of a vehicle outside the Protected Area which disrupts normal plant operations.

HU5 HUS Hus

• Consistent with last SER Approved EAL and current NEI Other Conditions Existing Which In the Judgment of Other Conditions Existing Which In the Judgment Other Conditions Existing Which In the Judgment EAL.

the ECIOSM Warrant Declaration of an Unusual of the Emergency DIrector Warrant Declaration of of the Emergency Director Warrant Declaration of Event a NOUE. aNOUE.

Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL NONE Other condtions exist which Inthe judgment of the Other conditions exist which In the Judgment of the HU5.1 Other conditions exist which In the judgment EC/OSM Indicate potential degradation of the level of Emergency Director Indicate that events are in of the Emergency Director Indicate that events safety of the plant. process or have occurred which indicate a potential are In process or have occurred which degradation of the level of safety of the plant. No Indicate a potential degradation of the level of releases of radioactive material requiring offsite safety of the plant. No releases of radioactive response or monitoring are expected unless further material requiring offste response or degradation of safety systems occurs, monitoring are expected unless further degradation of safety systems occurs.

Last SER ApprovedEAL's mEI 99-01 Rev. 4 EAL's Proposed EAL's Devlatlon Erplanation HAI HAI HA1 DIFFERENCE

• Consistent with current NEI EAL.

Natural and Dstructive Phenomena Affecting the Natural and Destructive Phenomena Affecting the Natural and DestructIve Phenomena Affecting the

• Added reference to the Safe Shutdownsn Areas Table Plant Vital Area Plant VITAL AREA. Plant Vital Area and Systems of Concern Table Inthe applicable ICs to Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL ensure consistent understanding of what area and/or systems we are concemed about. These tables are also Induded on the EAL Boards used by the ERO.

(Site-Specitic) method Indicates Seismic Event Earthquake peak horizontal acceleration ABOVE t 0.06 greater than Operating Basis Earthquake (OBE). HAl11 Receipt of the Amber Operating Basis Gravity. OR Earthquake Light and the walling seismic alarm Tomado or high winds greater than (site-specific) mph on 1C35 (t 0.06 gravity).

OR within PROTECTED AREA boundary and resulting In OR Report of tomado strtking plant vital area. VISIBLE DAMAGE to any of the following plant structures I equipment or Control Room Indication of HA12 Report ofTornado or high winds greater than OR 95MPH within PROTECTED AREA boundary degraded performance of those systems.

Report to cnV room of damage affecting safe Reactor Building and resulting In VISIBLE DAMAGE to a Safe ShutdownMrtal Area or Control Room Indication shutdown areas. Intake Building Ultimate Heat Sink of degraded performance of a System of S.f. sh~fldoo.,VI A-o OR Concen.

. Reftuelng Water Storage Tank v o.Y K;21W Vehicle crash affecting plant vital areas. Diesel Generator Building OR deod I Powrs~ f I G3I Wx dwU &* ia'Q-O0s.ISLi.d ir E~dI~P~rIRooms saw Rooms. Essentd Sw~rhgr Rooms. Cat* Spreood Turbine Building HA1.3 Vehicle crash within PROTECTED AREA Ross OR Condensate Storage Tank boundary and resufling In VISIBLE DAMAGE to Sustained wind speed ABOVE 95 MPH. . Control Room a Safe ShutdownWVtal Area or Control Room cr- ,p* W ROOM Waoke 59Ohft PWAOrros Other (Site-Spectic) Structures O Indication of degraded performance of a OR OR L L. Nw.COO.,.Saly CO A.

• V~ WC -. RT, - HR . o.

System of Concern.

Missiles affecting safe shutdown areas. Vehicle crash withIn PROTECTED AREA boundary om o* S."dng. R-OWlSrSMowS P84 IC388Al*., F.81 I C5SV AR "~,1TR-m and resulting In VISIBLE DAMAGE to any of the OR OR following plant stbuches or equipment therein or HA1.4 Turbine failture-generated missiles result In RIver lv ABOVE 767 feet. control Indication of degraded performance of those any VISIBLE DAMAGE to or penetration of systems: any of a Safe ShutdownNital Area.

OR Reactor Building Water level ABOVE Max Safe Operating Limit In2 or Intake Building OR more areas AND Reactor shutdown Is required. Ultimate Heat Sink 5y.twou of Cescom HAI.5 River le ABOVE 767 feet.

Refueling Water Storage Tank OR Diesel Generator Building OR *1 'ReaMCo" R Rlver le BELOW 724 feet 6 Inches. Turbine Building

• onumateverWAS HA1.6 Uncontrolled flooding In a Safe Condensate Storage Tank Shutdown~ital Area that results In degraded Control Room safety system performance as Indicated In the

• OnsfeA.CPowerIEOGe Other (Site-Specitfic) Structures.

• 0O05,ACPoowe Control Room or that creates an industrial

• ku5,,sslAC OR Turbine fallure-generated missiles result In any safety hazards (e.g.. electric shock) that

• DCoe precludes access necessary to operate or

• Rea. lM.f Caoeftr VISIBLE DAMAGE to or penetration of any of the following plant areas: (site-specific) list. monitor safety equipment OR OR Uncontrolled flooding In (site-specific) areas of the plant that results In degraded safety system HA1.7 River level BELOW 724 feet 6 inches.

performance as Indicated In the control room or that OR creates Industrial safety hazards (e.g., electric shock) that precludes access necessary to operate or monltor HA1.8 Report to control room of damage affecting a safety equipment Safe ShutdwnVnNital Area.

OR (Site-Specific) occurrences within PROTECTED AREA boundary and resulting InVISIBLE DAMAGE to plant structures containing equipment necessary for safe shutdown, or has caused damage as evidenced by control room Indication of degraded performance of

_ __ __ _ __ ___ __ _ _ _ _ _ _ _ ______ ____ I W 5_ i __ _ _ __ _ _ __ _ _ __ _ _ __ _ _ __ L_ _ _ _ __ _ _ _ __ _ _ __ _ _ __ _ _ __ _ _ __ _ _

Deviationor Difference Epaa o LastSER Approved'EAL's NEI 99.01 Rev. 4 EAL's Proposed EAL's Erpbnatlon HA2 HA2 HA2

• Consistent with current NEI EAL Fire Affecting the Operability of Plant Safety FIRE or EXPLOSION Affecting the OperabilIty of Fire or Explosion Affecting the Operability of Plant

• Aas of concern are the Safe Shutdownital Areas.

Systems Required to Establish or MaIntain Safe Plant Safety Systems Required to Establish or Safety Systems Required to Establish or MaIntaIn Chose to define a single source of affected areas for Shutdown Maintain Safe Shutdown. Safe Shutdown EALs as a hudnan factors Improvement opportunity. By Operating Modes: ALL using the 'Safe ShutdownNftal Areas table as a Op. ode: AL Opratig Mdes:ALLconsistent reference for applicable EALs, EAL decision-Op. Modes: ALL Operating Modes: ALL makers do not need to waste time referring to other Fire or explosion In any of the lollowing areas: FIRE or EXPLOSION In any of the following (site- HA2,1 Fire or exposion Inany Safe ShutdowntVital tables or lists of Impacted areas. This Is a conservative specific) areas: (Site-specifc) list Area decision that has been made to aid Intimely EAL

• htor, tne AND AND NONE decision making.

• I Affected system parameter Indications show degraded Affeded system parameter Indications show Pump house p performance or plant personnel report ViSIBLE degraded perfomance or plant personnel AND DAMAGE to permanent stnJctures or equipment report VISIBLE DAMAGE to permanent within the specified area. structures or equipoment within the specified Affected system parameter indlcations show degraded area.

performance or plant personnel report visble damage to permanent structures or equipment within the specidted area.

HAA3 HA3 HA3

• Consistent with last SER Approved EA- and current NEI Release of Toxic or Flammable Gases Within a Release of Toxic or Flammable Gases Within or Release of Toxic or Flammable Gases Within or EAL Facility Structure Which Jeopardizes Operation of Contiguous to a ViTAL AREA Which Jeopardizes Contiguous to a Vital Area Which Jeopardizes

• Change Informat no change In Intent.

Systems Required to Maintain Safe Operations or to Operation of Systems Required to Maintain Safe Operation of Systems Required to Maintain Safe Establish or Maintain Cold Shutdown Operations or Establish or Maintain Safe Operations or Establish or Maintain Safe

• Added reference to Safe ShutdowntVltal Area to ensure Shutdown. Shutdown, consistent understanding of areas of concern.

Op. Modes:ALL Operating Modes: ALL Operating Modes: ALL Toxic or flammable gas making safe shutdown areas Report or detedion of toxic gases withI or contiguous HA3.1 Report or detection of toxic gases within or DIFFERENCE unnaial rIacsrl.to a VITAL AREA In concentrations that may result In contiguous to a Safe ShutdownVital Area In unhhabitabe or inaccessble an atmosphere IMMEDIATELY DANGEROUS TO concentrations that may resuit In an LIFE AND HEALTH (IDLH). atmosphere Immediately Dangerous to Life OR and Health (DUH).

Report or detection of gases In concentration greater OR than the LOWER FLAMMABILITY LIMIT within or HA3.2 Report or detection of gases In concentration contiguous to a ViTAL AREA. greater than the Lower Flammability Umit within or contiguous to a Safe ShutdowntVital Area.___________________________

Last SER Approved E4L' NE199-1 Rev.4EAL'sl Prpoosed EAL's Dmerence orl HA4 HA4 HA4

• Consistent with current NEI EAL Security Event In a Plant Protected Area Confirmed Security Event In a Plant PROTECTED Confirmed Security Event In a Plant Protected Area

• GREATER THAN Is equivalent to > and minimizes AREA. the use of symbols.

Op. Modes: ALLO Operating Modes: ALL

• Llst of events Is derived from DAEC Security Plan.

Operating Modes ALL HA4.1 Intrusion Into the Plant Protected Area by a hostile force.

IntrsloIhto pbant pnotected area by a hostile force INTRUSION Into the plant PROTECTED AREA by a OR IOtee aHOSTILE FORCE. HA4.2 DAEC Security Supervision reports any of the OR OR following:

Sabotage device discovered Inthe plant protected area, Other security events as determined from(srte-

• Sabotage device discovered in the plant specific) Safeguards Contingency Plan and reported Protected Area.

by the (site-specific) security shift supervisIon

• Standoff attack on the Plant Protected Area DIFFERENCE by a Hostile Force 0.e. sniper).

• Any of the following security events that persists for 30 minutes. or greater.

affecting the Plant Protected Area:

o Credible bomb threats o Hostage/Extortlon o Suspicious Fire or Explosion o Significant Security System Harrware Failure o Loss of Guard Post Contact HAS HAS HA5

• Consistent with last SER Approved EAL and current NEI Control Room Evacuation Has Been Initiated Control Room Evacuation Has Been Initiated. Control Room Evacuation Has Been InitIated EAL Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL NONE Control room evacuation initiated per AOP 915, Entry Into (site-speciflc) procedure for control room HAS51 Entry into AOP 915 brcontrol room Shutdown Outside Control Room, evacuation, evacuation.

HAS HAS HAS

• Consistent with last SER Approved EAL and current NEI Other Conditions Existing Which In the Judgment of Other Conditlons Existing Which In the Judgment Other Conditions Existing Which In the Judgment EAL the ECtOSM Warrant Declaration of an Alert of the Emergency DIrector Warrant Declaration of of the Emergency Director Warrant Declaration of an Alert. an Alert.

Op Modes: ALL Operating Modes: ALL Operating Modes: ALL Other conditions exist whk h in the ?dgment of the Other conditions exist which In the judgment of the HAS1 Other conditions exist which Inthe judgment NONE Other condictinexit pian tem be degraded Emergency Director Indicate that events are In of the Emergency Director Indicate that events ECIOSM indicate that plant systems mayb derdd process or have occurred which Involve actual or are In process or have occurred wlhich Involve and that increased monitoring of plant functons likely potential substantial degradation of the level of actual or ikety potential substantial warranted safety of the plant Any releases are expected to be degradation of the level of safety of the plant lImited to small fractions of the EPA Protective Action Any releases are expected to be Imited to Guideline exposure levels, small fractions of the EPA Protective Action Guideline exposure levels.

Difference orEplnto Last SER ApprovedEAL's NEI 99-01 Rev. 4 EALa Proposed EAL's Deviation Exptanaton HSI HS1 HSI

• Consistent with current NEI EAL Security Event In a Plant Vital Area Confirmed Security Event In a Plant VITAL AREA. Confirmed Security Event In a Plant Vital Area

• Change In format, no change InIntent Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL Intrusion Into plant vital area by a hostile force. INTRUSION Into the plant VITAL AREA by a HS1.1 Intrusion Into a Safe ShutdownNital Area by a OR HOSTILE FORCE. hostile force.

Sabotage device discovered Inthe plant vital area. OR OR DIFFERENCE Other security events as determrined from (sie- HS1 .2 Security Supervision reports either of the specific) Safeguards Contingency Plan and reported following:

by the (site-specific) security shift supervision

• A security event that resuits In the loss of control In a Safe ShutdownNtal Area (other than the Control Room).

• A confirmed sabotage device discovered In a Safe ShutdownNltal Area.

HS2 HS2 HS2

• Consistent with iast SER Approved EAL and cunent NEI Control Room Evacuation Has oeen InItiated and Control Room Evacuation Has Been Initiated and Control Room Evacuation Has Been InitIated and EAL Plant Control Cannot Be Established Plant Control Cannot Be Established. Plant Control Cannot Be Established Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL NONE Control room has been evacuated AND control of plant Control room evacuation has been initiated. HS2.1 Control Room evacuation has been Initiated.

from Remote Shutdown Panel 1C388 NOT established AND AND within 20 minutes.

Control of the plant cannot be established per (site- Control of the plant cannot be established per specific) procedure withIn (site-specific) minutes. AOP 915 within 20 minutes.

HS3 HS3 HS3

• Consistent with current NEI EAL Other CondItions Existing Which In the Judgment of Other Conditions Existing Which In the Judgment Other Conditions Existing Which In the Judgment the ECIOSM Warrant Declaration of a Site Area of the Emergency Director Warrant Declaration of of the Emergency Director Warrant DeclaratIon of Emergency SHe Area Emergency. Site Area Emergency.

Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL Other conditions exist which Inthe judgment of the Other conditions exist which In the judgment of the HS3.1 Other conditions exist which Inthe judgment NONE ECIOSM Indicate actual or ikely major failures of plant Emergency Director Indicate that events are In of the Emergency Director indicate that events functions needed for protection of the public, process or have occurred which Involve actual or are In process or have occurred which Involve likely major failures of plant functions needed for actual or likely major failures of plant functions protection of the public. Any releases are not needed for protection of the publc. Any expected to result In exposure evels which exceed releases are not expected to result In EPA Protective Action Guideline exposure levels exposure levels which exceed EPA Protective beyond the site boundary. Action Guideline exposure levels beyond the site boundary.

Last SER Approved EAL.' NEl9901Rev.4 LA'.- Difference or rlnto Proposed EAL's Deviation Explnaton HG1 HG1 HG1

• ConsistentwithaeunentNEIEALi Security Event Resulting In Loss of Ability to Reach Security Event ResultIng In Loss Of PhysIcal Secu Event CotyResulting In Loss Of PhysicalDn Infona not Intent of NEI and Mslntsin Cold Shutdown Control of the Facility. Control of the Facility. 0Dfeec nfra.ntItn fNI Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL Loss of physical control of the Control Room. A HOSTILE FORCE has taken control of plant HG1 .1 A hostile force has taken control of plant OR equipment such that plant personnel are unable to equipment such that plant personnel are operate equipment required to maintain safety unable to operate equipment required to Loss of physical control of remote shutdown capabiity. functions, maintain safety functions as Indicated by loss DIFFERENCE of physical control of either

• A Safe Shutdown/Vtal Area such that operation of equipment required for safe shutdown is lost OR

• Spent fuel pool cooling systems If Imminent fuel damage Is Ikely (e.g.,

freshly offloaded reactor core Inthe pool).

HG2 HG2 H02

• Consistent with ast SER Approved EAL end curent NEI Other Conditions Existing Which In the Judgment of Other ConditIons ExIsting Which In the Judgment Other Conditions ExIsting Which In the Judgment EA-the ECIOSM Warrant Declaration of a General of the Emergency Director Warrant Declaration of of the Emergency DIrector Warrant Declaration of Emergency General Emergency. General Emergency.

Op. Modes: ALL Operating Modes: ALL Operating Modes: ALL Other conditions exist which In the judgment of the Other conditions exist which In the judgment of the HG2.1 Other conditions exist which In the judgment ECIOSM indicate EITHEER Emergency Director Indicate that events are In of the Emergency Director Indicate that events

• A al __ process or have occurred which involve actual or are In process or have occurred which involve NONE d

ACual or immmii p btantialcrl Imminent substantial core degradation or meiffng wIth actual or Imminent substantial core degradation with potential for ss of potential for loss of containment integrity. Releases degradation or melting with potential for loss oniainment can be reasonably expected to exceed EPA of containment Integrity. Releases can be

• Potential for uncontrolled radbinudide Protective Action Guideline exposure levels offsite for reasonably expected to exceed EPA releases which can reasonably be more than the Immediate site area. Protective Action Guideline exposure levels expected to exceed EPA PAG plume offsate for more than the Immediate sils area.

exposure leveis outside the site boundary.

LastSERApprovedEAL's NEI 99401 Rev. 4 EAL's ProposedEAL's Differenetorl SUl SUl SUl

• Consistent with current NEI EAL Loss of All Oftite Powerto Essential Busses for Loss of All OffsKle Power to Essential Busses for Loss of All Offaste Powerto Essential Busses for

• DAECstatestheergencybusses (ste-speclflc Greater Than 15 Minutes Greater Than 15 Minutes. Greater Than 15Minutes Infomtion) poee by the applicable trnfrmers to Op Modes: ALL Operating Modes: Power Operation, Startup, Hot Operating Modes: Power Operation, Startup, Hot minimize any potential confusion whereby the Standby, Hot Shutdown SID transformer would have power, but the essential bus would not. The Intent ofthe EALIs aloss ofoffshle LsofOftePwratigMeThn1Mnue. Loss of power to (sitespeif)tra nsformer fo IFEEC power to the essential busses and each essentala bus Is Loss of Oftste Power iastinti More Than 15 Minutes. greater than 15 minutes. SU1.1 Loss of a te perto Emergency Busses DIFFERENCE geting power from applicable Standby Diesel AND 1A3 and 1A4 Is expeed to last for greater than Generator.

At least (sfte-specific) emergency generators are 15 minutes supplying power to emergency busses. AND Emergency Busses 1A3 and 1A4 are powered by their respective Standby Diesel Generators.

SU2 SU2 SU2

• Consistent with last SER Approved EAL and current NEI Inability to Reach Required Shutdown Within Inability to Reach Required Shutdown Within Inability to Reach Required Shutdown Within EAL Technical Specification Limits Technical SpecificatIon Limits. Technical Specification Limits Op. Modes: Run, Startup, Hot SID Operating Modes: Power Operation, Startup, Hot Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SID NONE Plant NOT brough to requbred mode within applicable LCO Action Statement Time Lmlta. Plant Isnot broughtto required operating mode within SU Pode wathsntbplcablhTtoechni oeatin (sLee-spectlc) Technical Specifications LCO Action mode within aCOlcaboe T aehnical Sttment Tipmeiain.IOAtonSaeen ie SU SU3 5133

• Consistent with current NEI EAL Unplanned Loss ofAII Safety System UNPLANNED Loss of Most orAll Safelty System Unplanned Loss of Most or All Safety System Annunicatlon or Indication In the Control Room Annunciation or Indication In The Control Room AnnuncIatIon or Indication In the Control Room for Greater Than 15 MInutes for Greater Than 15 Minutes for Greater Than 15 Minutes Op. Modes: Run, Startup, Hot SID Operating Modes: Power Operation, Startup, Hot Operating Modes: Power Operation, Startup, Hot Standby. Hot Shutdown S/D NONE Unplanned loss of most annunciators on panels 1C03, 513.1 Unplannedlossofmostorat C03, C04and 1C04 and 1COS lasting more than 15 minutes AND UNPLANNED loss of most or all (site-specific) I C05 annunciators or Indicators associated compensatory non-alarming Indications are evallable. annunciators or Indicators associated wih safety with Safety Systems for greater than 15 systems for greater than 15 minutes. minutes.

I

Difforence orExanlo Lsst SERApprovd EAL's NEt99.01 Rv. AEAL's Proposed EAL!s Deviation Explanation SU4 SU4 SU4

• Changed format, no change InIntent Fuel Clad Degradation Fuel Clad Degradation. Fuel Clad Degradation

• RM-4104 HI-Hi Radiation Alarm has been chosen Op. Modes: Run, Startup, Hot SID Operating Modes: Power Operation, Startup, Hot Operating Modes: Power Operation, Startup, Hot because it Is operationally significant Is readily Standby, Hot Shutdown S/D recognizable by t.e Control Room Operations Staff, and isset at a level corresponding to noble gas release rate, after 30-minute delay and decay of I Cl~sec. A VAl i T 1 r31 (Site-specific) racdiatio monitor readings Indicating aHINotification of Unusual Event Is classified because the 4E+3 mRdad degradatton greatertan SU4.1 Pretreatment 2f dose (RMi414 DIFFERENCE 0as g Pretreatment HIl-HIradiation alarm Is consIdered Specification allowable lmitsn to be an indication of a potential degradation Inthe level OR OR OR of safety of the plant and apotential precursor ofmore Coolant activity ABOVE 1.2 pClhnI DOSE (SRte-pecific) coolant sample activity value iniaig SU4.2 Reactor Coolant sample activiy valje seflous problems, EQUIVALENTo geatr1henTeh3ialGREATER THAN 2.0 piCtgm dose equivalent

• 2.0 pCI~ff dose equivalent 1-1 31 Isthe maximum fuelcladderaation greoatler mthas Tcna 1-131. concentfration In DAECs Technical Specifications.

SUS SUS SUS

• Consistent with last SER Approved EAL and current NEI RCS Leakage RCS Leakage. RCS Leakage EAL with the exception of the 'mnahisteanmline break Op. Modes: Run, Startup, Hot SID Operating Modes: Power Operation, Startup, Hot Operating Modes: Power OperatIon, Startup, Hot and remved ase LOSS froTh e Fissdon Br rie Tabl Stanby.HotShutownBID(RCS BARRIER - LEAKAGE). Decided to go wIth NEI Unidentified of pressure boundary leakage grater 99-01 rev 4 format by removing this IC from SUS and than 10 GPM. Unidentified or pressure boundary leakage greater SUS1 Unidentitied or pressure boundary leakage NONE Is more appropriate and It gives an input intothe logic OR than 10 gpm. GREATER THAN 10 gpm. for determining the appropriate classification level for Identified leakage greater than 25 GPM. OR OR Fission Barrier losses.

OR Identified leakage greater than 25 gpm. SU5S2 Identified leakage GREATER THAN 25 gpm.

Main steam iine break as determined from annunciators or plant personnel report.

D~ffearencip or'Epanto LastSERApprovedEAL's NEI 99-01 Rev. 4 EAL's ProposedEAL's Devlaton Exp/nation su6 SUe SUe

• Consistent with current NEI EAL.

Unplanned Loss of All OnsHts or OffisKe UNPLANNED Loss of All Onshte or Offlbte Unplanned Loss of Al Onslte or Offslti Communications Capabilities CommunicatIons CapabilIties. Communicatlons CapabilItIes Op. Modes: ALL Operating Modes: Power Operation, Startup, Hot Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SID Loss of all (sIte-specific list) onshe communications SU6.1 Loss of ALL of the following onsste Loss of ALL onsite telephone and radio capablity affecting the ability to pWrform routine communication capabilities affecting the ability communication methods (PABXydirect-ring, UHF, and operations. to perform routine operation:

radiological sey o syst) OR

• Plant Operations Radio System Loss of ALL elecronic LossofALcoR unication methods v*h eectrniccommnictionmetods ith Loss of all (ste-specifc list) offsle communications capability.

• In-Plant Telephones NONE government agencies (PABX. direct-ring, ENS.

• Plant Paging System microwave and police radio). OR SU6.2. Loss of ALL of the following offsilte communications capabilIty:

• All telephone lnes (commercial)

• Microwave Phone System

• FTS Phone System SUS SUe e Consistent with current NEI EAL Inadvertent Criticality Inadvertent Criticality

• Used BWR specific EAL NEW Operating Modes: Hot Standby, Hot Shutdown Operating Modes: Hot SID DIFFERENCE An UNPLANNED extended positive period observed SU8.1 An unplanned extended positve period on nuclear hnstrumentabion. observed on nuclear Instrumentation.

Difference or Epaa o Last SER Approved EAL's NEI 99-01 Rev. 4 EAL's Proposed EAL's Exphnat/on SA2 SA2 SA2

• DAEC chose to define available methods of manually Failure of Reactor Protection System Failure of Reactor Protection System Failure of Reactor Protection System scramting the reactor from within the Control Room InstrumentatIon to Complete or Initiate an Instrumentation to Complete or Initlate an Instrumentation to Complete or Initbb an (ICOS 'reactor control console-) to darify acceptable Automatic Reactor Scram Once a Reactor Automatic Reactor Scram Once a Reactor Automatic Reactor Scram Once a Reactor actions for the Alert level dedaration or the Ste Area Protection System Setpoint Has Boen Exceeded Protection System SetpoInt Has Been Exceeded Protection System Setpoint Has Been Exceeded Emergency dedaratlon.

and Manual Scram Was Successful and Manual Scram Was Successful, and Manual Scram Was Successful

• Consistent with current NEI EAL Op. Modes: Run. Startup Operating Modes: Power Operation, Startup. Hot Operating Modes: Power Operation, Startup

• Wording format changed to be consistent with escalation Standby EALs (SS2 SG2).

SA2.1 Auto Scram failure DIFFERENCE * 'Auto Scram failure bsequNalent to 'Indication(s) exist Fiueoauoaisca.Indication(s) exist that Indicate that reactor protection that reactor protection system setpoint was exceeded Fallure of sutomatic sca system selpoint was sytmston xeddadatmtcsrmAND eeeded a utoaticrscramdo and N and n automatic scram did not occur' u~mbcsrmnidn~er e and Isage less wordy to did not occur, and a successful manual scram of the following operator actions to better facilitate timely EAL deciaration and improved occurred, reduce power ar5i sc s i In shutting down human factors on the EAL Board.

the reactor

• Manual Scram Pushbuttons

• Mode Switch to Shutdown

• Altemate Rod Insertion (ARI)

SA4 SA4 SA4

• Consistent wth last SER Approved EAL and curent NEI Unplanned Loss of Most orAin Safety System UNPLANNED Loss of Most orAil Safety System Unplanned Loss of Most or All Safety System EAL Annunication or Indication In Control Room With Annunciation or Indication In Control Room With Annunciation or Indication In Control Room With

• Added 'conditions exist to second IC. No change In Either (1)t SlgnificantTransient In Progress. or Either (1)a SIGNIFICANTTRANSIENTIn Progress, Either (1)a Significant Translent In Progress, or AddedI (2) Compensatory Non.Alarming Indicators are or (2) Compensatory Non-Alarming Indicators are (2) Compensatory Non-Alarming Indicators Unavailable Unavailable. Unavailable Op. Modes: Run, Startup. Hot SID Operating Modes: Power Operation, Startup, Hot Operating Modes: Power Operation. Startup, Hot Standby, Hot Shutdown SID Unplanned loss of most annuniators on pane 1C03 UNPLANNED loss of most or al (site-specific) SA4.1 Unplannedlossofmostorail 1C03. 1C04 and DIFFERENCE 1UC04lanned1C05of most a hats on pminutes and annuncators or indicators associated with safety I COSannunciators or Indicators associated EiTi aER systems for greater than 15 minutes. with Safety Systems for greater than 15

• Signlticant transIent Inprogress. AND minutes

• Loss of compensatory non-alarming Either of the following: (sorb)

IndIcations, a. A SIGNIFICANT TRANSIENT is in pngess. Either of the following conditions exist OR

• A significant plant transient is In progress.

b. Compensatory non-alarming indications are unavailable. I Compensatory non-alabming Incilcabonsare unavailable.

Last SER Approved EAL's NEI 99-01 Rev. 4 EAL'S Proposed EAL's DEffennt orE DeviationEpanio n SAS SA5 SAS

• Consistent with current NEI EAL AC Power Capability to Essntial Buses Rduced AC power eapability to essentIal busses reduced AC Power Capability to Essential Busses Reduced

• DAEC defied essentia busses as 1A3 and IA4.

to a Single Power Source for Greater Than 15 to a single power source for greater than 15 to a Single Power Source for Greater Than 15 Minutes Such That Any Additional Single Failure minutes such that any additional singie failure Minutes Such That Any Additlonal Single Failure Would Result In Station Blackout would result In station blackout. Would Result In Station Blackout Op. Modes: Run, Startup, Hot SID Operating Modes: Power Operation, Startup, Hot Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown StD NONE AC power capability to site-specifi essential busses SA5.1 ACpowercapabieityto1A3or1A4busses Only one AC power source remains available to reduced to a single power source for greater than 15 reduced toea single power source for greater supply Bus IA3 or Bus 1A4 AND If ItIslost a Station minutes than 15 minutes Blackout will occur.

AND AND Any additional single failure wiii result in station Any additional single failure will result In blackout station blackout LS` o S a SS1

• Consistent with last SER Approved EAL and current NEI Loss of AN OPtwe Powr and Loss oe AsOnshe AC Loss of All Ofldb Pow r and Loss of All Onsita Loss of An Ost Power and Loss of Al Onsite AC EAL Pow toEss ntir Busss AC Power to Essential Busses. Powerto Essential Busses

• Definedemergencybusses IA3 end 1A4.

Op. Modes: Run, Startup. Not SID Operating Modes: Power Operation. Startup, Hot Operating Modes: Power Operation. Startup, Hot Standby. Not Shutdown SID

• DAEC states the emergency busses (site-specific Information) powered by the applicable transformners to minimize any potential confusion whereby the Loss ofVoltae on Buses 1A3 and 1 lsng mre Lo" of power to (site-specific) transformers. SS11 Loss of ailoffste powerto EmergencyBusses transformer would have power, but the essential bus toss of Volutaes onNBses1A3and1Andasingmo would not. The Intentof the EA. Isa loss ofofl'ite DIFFERENCE power to the essential busses, both Diesel Generators Failure of (site-specitfic) emergency generators to AND fail, and one essential bus cannot be restored.

supply power to emergency busses. Failure ofA Diesel Generator (1G-31) and B AND Diesel Generator emergency busses(1G-21) 1A3 andto 1A4.

supply power to Failure to restore power to at least one emergency bus within (site-specdic) minutes from the tme of loss AND of both ofste and onsie AC power. Faiure to restore power to at least one emergencybus, 1A3orMA4 within15minutes ornm the time of loss of both ofsfte and onsite AC power.

Last SER Approved EAL's NEI 9 Rev. 4 EAL's I01 Proposed EAL's Drne Deviationor Explanatbn SS2 582 S52

• DAEC chose to define available methods of manually Failure of Reactor Protection System Failure of Reactor Protection System Failure of Reactor Protection System scramming the ractor from wlthin the Control Room Instrumentation to Complete or InitIate on Instrumentation to Complete or Initiate an Instrumentation to Complete or Initiate an (1COS 'reactor control console') to clarify acceptable Automatic Reactor Scram Once a Reactor Automatic Reactor Scram Once a Reactor Automatic Reactor Scram Once a Reactor aetions for n the Acy level declaration or the Site Area Protection System Setpolnt Has Been Exceeded Protection System Setpolnt Has Been Exceeded Protection System Setpolnt Has Been Exceeded Emergency ecarat and Manual Scram Was NOT Successful and Manual Scram Was NOT SuccessfuL and Manual Scram Was NOT Successful

• Consistent with current NEI EAL Op. Modes: Run, Startup Operating Modes: Power Operation, Startup Operating Modes: Power OperatIon, Startup

• Wording & format consistent with SA2 and SG2.

• Auto Scram failure Isequivalent to lIndicatlon(s) exist Failure of automatic and manual scram Indication(s) exist that automatic and manual scram 552.1 Auto Scram failure DIFFERENCE that reactor protection system selpointwas exceeded were not successfl and automatic scram did not occur' and Is less wordy AND AND to better facilitate timely EAL declaration and Improved Power remains ABOVE 5% NMAQof the following operator actions to human factors on the EAL Board.

reduce power are sucess In shutting down the reactor Bonon Injection requred.

• Manual Scram Pushibuttons

• Mode Switch to Shutdown

• Altemate Rod Insertion (ARI)

SS3 S53 SS3

• Provided more detaillfor IC no change In intent Lose of All Vital DC Power Loss of All Vital DC Power. Loss of Al Vital DC Power

• Div I and Div2 125v DC busses are considered vItal at Op. Modes: Run, Startup, Hot SID Operating Modes: Power Operation. Startup, Hot Operating Modes: Power Operation, Startup, Hot DAEC.

Standby. Hot Shutdown S/D

• Under the conditions of concern, AOP 302.1, Loss of 125 VDC Power, would be entered Lunder Tab 3.

Complete Loss of 125 VDC. Consequentiy, the DAEC Unpn osoDh an151e Loss of Al Vital DC Power based on (site-specific) 553.1 Loss of Dlv land Dlv2 125VDC busses addeseslossofboth sons of the 125VDDC More Tha 15v Minutes.

Unlastieg 2 V use bus voltage Indications for greater than 15 minutes. based on bus voltage LESS THAN 105 VDC system consistent with AOP. At DAE-C. the 125V DC Latig oe ha 1 inte.Indicated for greater than 15 minutes. DIFFERENCE Systems ensure power Isavailable for the reactor to be shutdown safely and maintained in a safe condition. The 125V System Is divided Into two Independent divisions -

Divison I and Division I1- with separate DC power suppies. These power supplies consist of two separate 125V batteries and chargers serving systems such as RCIC. RHR. EDGs, and HPCI Complete loss of both 125V DC Divslons could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations.

• 105 VDC ls the setcolnt for a loss of 125 VDC busses.

SS4 SS4 SS4

• Consistent with current NEI EAL Complete Loss of Function Needed to Achieve or Complete Loss of Heat Removal Capability. Complete Loss of Heat Removal Capability

• EOP Graph 4 IsDAECs Heal Capacity Temperature Maintain Hot Shutdown Operating Modes: Power Operation, Startup, Hot Operating Modes: Power Operation, Startup, Hot

• mriat Curve.

Op. Modes: Run, Startup, Hot S/D Standby, Hot Shutdown SD1 DIFFERENCE EOP Graph 4 Heat Capacity Limit Is exceeded Heat Capacity Temperature Umit Curve exceeded 5S4.1 EOP Graph 4 Heat Capacty Umit Is OR (BWR). exceeded Reactor CANNOT be brought subcritleal.

LastSERApproved Ai's NEI 99-f Rev. 4EAL's Pro AL'sDeviation psed Difference or Explaneffon 556 SS5 556

• Consistent with last SER Approved EAL and current NEI Inability to Monitor a Significant Transient In Inability to Monitor a SIGNIFICANT TRANSIENT In Inability to Monitor a SIgnificant Transient In EAL Progress Progress. Progress

• Format Is different than NEI.

Op. Modes: Run, Startup, Hot SID Operating Modes: Power Operation. Startup, Hot Operating Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SID Signicant transient in progress and BOTH of the Loss of most or all (sie-specific) annunciators SS5.1 Significant transient in progress and ALL of folloicng associated with safety systems. the following:

• Loss of annundators on panels 1C03. AND

• Loss of most or all annunciators on DIFFERENCE 1CC04 and 1COS Compensatory non-alarming Indications are Panels I C03. 1C04 and I COS.

AND unavailable.

• Compensatory non-alanming

• Loss of compensatory non-alaning AND Indications are unavailable.

Indleatons a Itdo ations needed to monitor (ste safety

• ispedtq needed to monitor crticality.

Indicators cunctions an unavaelableb or core heat removal, or Fission Produft Baraier status are AND unavailable.

SIGNIFICANT TRANSIENT In progress.

SG1 SG1 SG1

• Consistent wth current NEI EAL Prolonged Loss of An Offste Power and Prolonged Prolonged Loss of All Offte Power and Prolonged Loss of All Offite Power and Loss of All Onsie AC Power Prolonged Loss of All Onrit AC Power to Prolonged Loss of All Onstle AC Pow r to

• Format h diferent than NEI.

Op. Modes: Run, Startup, Hot SID Essential Busses. Essential Busses * 'restoration of power to either bus Is etuvialent to Operating Modes: Power Operation, Startup, Hot Operating Modes: Power Operation, Startup, Hot restoration of at least one..' since DAEC has only two Standby, Hot Shutdown SID emergency busses.

DAEC stales the emergency busses (site-specific Loss of Voltage on Buses 1A3 and 1A4 and ANY of the Loss of power to (site-specific) transformers. SG11 Loss of ai offsIte power to Emergency Busses information) powered by the applicable transformers to follow4ng:ADA3adA4 mrininimizeanty potential confujsion whereby the Restotlon of powertoerherBus1A3or1A4h. a transformer would have power, but the essential bus NOT likely within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Failure of (site-specific) emergency diesel generators AND pould no The essent ia the EAL Is a loss e oeffsate

• RFVlevel hdetermhab to supply power to emergency busses. Failure ofA Diesel Generator (1G-31) and B fai, and one essential bussannbothb Dresel Generatohn AND Diesel Generator (IG-21) to supply power to hrs or an Indication of potential Fission Product Banrier

• RPV Leve BELOW +15 hncihes. emqec busses 1A3 and 1A4,.r ra niaino oenflFsinPoutBne Either of the following: (a or b) AND DIFFERENCE degradation exists.

• DAECs site specific Indication of continuing degradation ANY ONE OF THE FOLLOWING: of core cooling based on Fission Product Banier

a. Restoration of at least one emergency bus within

• Restoration of power to either Bus monitoring Is:

(sie-spectfic) hours Is hO2 Ilkely RA3or IA4 Is not likely wItoin 4 1. RPV level indelerminate - FIashkg ofthe reference OR hours, leg water wllf resut hI errneously high RPV water b (SileSpecdic) Indfeation ofcontinuing degradatfon RPV level IsIndeterminale. lovetarengs gMng afalse Indication of actual of (Se-Speiooc)lindicationoFksont g degrado water knentory and potentiallyIndkcatirg adequate ofcore cooling based on issi B

• RPV Level Is LESS THAN +15 core coolng when Itrmaynotexist EOP Graph 1, mo g Indches. RPVSafturstri Tenmperature, defines the condiflons under whIkh RPV bvelosurymentleg bolilng may occur.

2. RPV Level Is LESS THAN +15 Inches - DAEC uses the RPV level that Is used for the Fuel Clad

'pofenfial /oss condtiOn In the Fission Poduct BanflerMatrx. hlIs sRPVlevelbelow 15iches

Lost SER Approved EAL's NEI 99-01 Rv. 4 EAL's Proposed EAL's Deviation Explanallon 5G2 302 SG2

• Consistent with last SER Approved EAL and current Failure of the Reactor Protection System to Failure of the Reactor ProtectIon System to FaIlure of the Reactor Protection System to NEI EAL Complete an Automatic Scram and Manual Scram Complete an Automatic Scram and Manual Scram Complete an Automatic Scram and Manual Scram

• DAEC chose to define avallable methods of manually was NOT successful and There Is IndicatIon of an was NOT Successful and There Is Indication of an was NOT successful and There is Indication of an scramming the reactor from within the Control Room Extreme Challenge to the AbilIty to Cool the Core Extreme Challenge to the Ability to Cool the Core. Extreme Challenge to the Ability to Cool the Core (1C05 'reador control console) to clarify acceptable Op. Modes: Run, Startup Operating Modes: Power Operation, Startup Operating Modes: Power OperatIon, Startup actions for the Alert level declaration or the Site Area Emergency declaration.

• -25 Inches. DAECs Minimum Steam Cooling RPV Entry Into ATWS EOP- RPV Control Is requIred Indicallons exist that automatic and manual scram SG2.1 Auto Scram failure Water Level, and the HCL Curve (EOP Graph 4). gives were not successful. an Indication of challenged core cooling andtor decay AND AND AND heat removal capability.

RPV levei cannot be maintained ABOVE 430 Inches. NO of the following operator actions to Either of the following: (a or b) reduce power are succes In shutting down

• Wording and format consistent with SA2 & 552.

OR

a. Indication(s) exists that the core cooling Is the reador

• Auto Scram fallure is equivalent to 'Indication(s) exist EOP Graph 4 Heat Capacity Limit Is exceeded DIFFERENCE extremely challenged.

• Manual Scram Pushbuttons that reactor protection system setpolnt was exceeded and automatic scram did not occur' and Is less wordy OR

• Mode Switch to Shutdown to better facilitate timely EAL declaration and Improved

b. Indicatlon(s) exists that heat removal Is extremely human factors on the EAL Board.

• Alternate Rod Insertion (ARI) chalienged.

AND Loss of adequate core cooling or decay heat removal capablity as indicated by either

• RPV level cannot be maintained GREATER THAN -25 Inches.

• HCL Curve (EOP Graph 4) exceeded.

Last SER Approved`EAL's NEI 99-01 Rev. 4 EAL's - Proposed EAL's Ol1fferenc.

Deationor Explanation.

FUEL CLAD BARRIER FUEL CLAD BARRIER FUEL CLAD BARRIER

• Consistent with last SER Approved EAL and current NEI RADIATIONICORE DAMAGE RADIATIONICORE DAMAGE EAL L Fuel damage assessment (PASAP 7.2) L Drywall Radiation monitor reading GREATER L Fuel damage assessment (PASAP 7.2) Indicates

• Sffe-spedffc for OAEC Inciudes the fuel damage determines at least 5% fuel dad damage OR Fuel THAN (slte-spectfic) R/hr at least 5% fuel dad damage assessment and ToUs Area High Range Monitor damage Is Indicated by any of the fdowing: OR OR Indication as per cuiginal SER.

L (Site specific) as applicable L Drywall Area HI Range Red Monitor, RIM-9184A

• NOTE: Fission Barrier Table logic Isas follows for L Vald drywel red mnonitor reading ABOVE 7E'2 OR or B reading GREATER THAN 7E+2 Rem/hr a FIssIon

! Barrier EALs. ThIs logic diagram Is on Riahdrywl e oio edn BV E2O DAEC'a EAL Boards.

OR L Coolant Activity GREATER THAN (site-spiedfic) OR G w Value L Torus Area Hi Range Rad Monitor, RIM-9t85A or L Valid tonrs red monitor reading ABOVE 3E+1 R/hr B reading GREATER THAN 3E+1 Rem/hr _____

OR OR II_

L Coolant activity ABOVE 300tiCVgm DOSE L Coolant actiMty GREATER THAN 300pCVgrr, i 1 JL_

EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 DIFFERENCE _

M- H.F W .

FUEL CLAD BARRIER FUEL CLAD BARRIER

• Stating RPV Lever so Operators know what level to RPV LEVEL FUEL CLAD BARRIER RPV LEVEL refer to.

L RPV Level BELOW -30 Inches L RPV Level LESS THAN -25 Inches

• Consistent with last SER Approved EAL and current NEI P RPV Level BELOW 15 inches L Level LESS THAN (site-specic value) P RPV Level LESS THAN +15 inches DIFFERENCE EAL P Level LESS THiAN (site-spcific value) * -25 Inches Is DAECs Minimum Steam Cooling RPV

• L HvWater Level.

1- Ice Is DAEC's TAP.

Last SER ApprovedEAL's NEI 99-01 Rev. 4 EAL's Proposed EAL's DIfference or Explnaton Deiation FUEL CLAD BARRIER FUEL CLAD BARRIER FUEL CLAD BARRIER

• Consistent with last SER Approved EAL and current NEI EMERGENCY DIRECTOR JUDGMENT EMERGENCY DIRECTOR JUDGMENT EMERGENCY DIRECTOR JUDGMENT EAL Any conditIon which Inthe ECIOSMs judgment Any condition Inthe opinion of the Emergency Any condition Inthe opinion of the Emergency Indicates loss or potential loss of the fuel dad barrter Director that Indicates Loss or Potential Loss of the Director that Indicates Loss or Potential Loss of the NONE due to: Fuel Clad Banler Fuel Clad Banrer.

• Imminent banfler degradation

• Degraded fission banrer monitoring capability RCS BARRIER RCS BARRIER

• Consistent with last SER Approved.

RADIATIONICORE DAMAGE RCS BARRIER RADIATIONICORE DAMAGE

• Llsled specifIc equipment ID and name.

L Valid drywel rad monitor reading ABOVE 5 Rhr L Drywell Area HI Range Rad Monitor, RIM-9184A after reactor shutdown or B reading GREATER THAN 5 Rem/hr after DIFFERENCE

• after reactor shutdown: This loss Indicator Is based on L DryweaRadiation monitor reading GREATER reactor shutdown condItIons after reactor shutdown to assure that ItIs not THAN (site-speclfc) R/hr misapplied. L/.. to exclude readings due to N-15 effects which are typically 5 to 8 Rlhr at full power conditions.

Approved vla last SER.

RCS BARRIER RCS BARRIER

• Slating RPV Lever so Operators know what level to RPV LEVEL RCS BARRIER RPV LEVEL refer to.

L RPV Level BELOW 15Inches L RPVLevel LESSTHAN l15hnches DIFFERENCE

• Consistentwth last SER Approved EALand current NEI L Level LESS THAN (site-speclic value) EAL

• M 5 Inches Is DAEC's TAF.

RCS BARRIER RCS BARRIER RCS BARRIER

• Added the specific reference for Max Normal, no LEAKAGE LEAKAGE change InIntent.

P RCS Leakage Is ABOVE 50 GPM L (Site-specific) Indication of an unisolable Main L Unisolable Main Steamline Break as Indicated by

• Max Normal Limits per EOP 3 are the values used to OR Steambine Break the failure of both MSIVs In any one lne to dose detemine a Potentlal Loss d an RCS Banler. Max OR AD EIHER:Safe Uimits per EOP 3 are the values used to RCS leakage GREATER THAN S gpm Inside the dP

• High MSL flow or high steam tunnel determine a Loss of a Primary Containment Banler.

dryweaas Indicated by area temps or ARMs temperature annunciators

• Consistent with current NEI EAL OR

• Direct report of steam release DIFFERENCE

• Added site-spechtic indication of an unisolable main P UnIsolable primary system leakage outside P RCS Leakage Is GREATER THAN 50 GPM Inside steamrline leak.

drywefl as Indicated by area temperature or area the drywell radiation alarm OR P Unisolable primary system leakage outside the drywell as Indicated by area temps or ARMs exceeding the Max Normal Limits per EOP 3, Table 8.

Last SER Approved EALa NEI 99.01 Rev. 4 EAL's Proposed EAL's Dferce orathn RCS BARRIER RCS BARRIER RCS BARRIER

• Consistent with last SER Approved EAL and current NEI PRIMARY CONTAINMENT ATMOSPHERE PRIMARY CONTAINMENT ATMOSPHERE EA-oDAEC uses a GE Mark I Containment During reactor L Drywel pressure ABOVE 2 psig and not caused L Pressure GREATER THAN (site-specfic) PSIG L Drywel pressure GREATER THAN 2 psig and not operation, with drywel cooing Inoperation and the drywell by a loss of DW Cooling caused by a loss of DW Cooling inerted, the normal operating pressure in the drywell is between 0.5 and 1.0 psig. Analysis at the DAEC shows DIFFERENCE 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 hiffatIon setpoint (2 pslg) it Is necessary to select the DAEC ECCS Initiation setpoint of 2 psig to Indicate an actual loss of the RCS.

Dryweil cmoo"g Is not Isolated at the 2 psig ECCS IriatIon setpoint. therefore further pressure rise would be indicative of a RCS leak. Approved via last SER.

RCS BARRIER RCS BARRIER RCS BARRIER

• Consistent with last SER Approved EAL and current NEI EMERGENCY DIRECTOR JUDGMENT EMERGENCY DIRECTOR JUDGMENT EMERGENCY DIRECTOR JUDGMENT EAL Any condition which In the EC/OSM's Judgment Any condition In the opinion of the Emergency Any condition In the opinion of the Emergency Indicates loss or potential loss or the RCS barrier due Director that Indicates Loss or Potential Loss of the Dirctor that Indicates Loss or Potential Loss of the NONE to: RCS Barrier RCS Barrier

• Imminent banier degradation

• Degraded flsson banier monitoring capablity PRIMARY CONTAINMENT BARRIER PRIMARY CONTAINMENT BARRIER PRIMARY CONTAINMENT BARRIER

• ConsIstent with last SER Approved EAL and current NEI RADIATION/CORE DAMAGE RADIATION/CORE DAMAGE EAL P VaffddryweffradrnonltorreadhgABOVE3E+3 P Drywef Radiation rnonllorreading GREATER P DrywelIArea Hi Range Rad Monitor RIM-9184A Rthr THAN (site-spectic) R/hr or B reading GREATER THAN 3E+3 Rem/hr OR OR OR NONE P ValId torus rad monitor reading ABOVE 1E+2 R/hr P (Site specitic) as applicable P Torus Area Hi Range Rad Montor RIM-91 85A or OR B reading GREATER THAN i E+2 Rem/hr P Core damage assessment determines at least OR 20% fuel dad damage P Fuel damage assessment (PASAP 7.2) Indicates at least 20% fuel dad damage PRIMARY CONTAINMENT BARRIER PRIMARY CONTAINMENT BARRIER PRIMARY CONTAINMENT BARRIER

• Consistent with current NEI EAL RPV LEVEL RPV LEVEL NONE P RPV Level BELOW -40 Inches P Primary containment flooding required P Primary containment flooding required PRIMARY CONTAINMENT BARRIER PRIMARY CONTAINMENT BARRIER PRIMARY CONTAINMENT BARRIER

• Consistent with last SER Approvd EAL and current NEI LEAKAGE LEAKAGE EAL L Failure of both Isolation valves and a downstream L Failure of both valves In any one ine to dose L Failure of both valves In any one line to close and dMax Noemal LimPts per EOP 3 are the values used to pathway to the environment exIsts AND downstream pathway to the environment a downstream pathway to the environment exists detersine a Potential Loss of an RCS Barrier. Max Safe OR exst Limits per EOP 3 are the values used to determinhe a OR ests OR Loss of a Primary Containment Barrier.

L Unisolable primary system leakage outside the OR L Unisolable primary system leakage outside the DIFFERENCE

• Adding 'when Containment Isolation Is required Is In drywet as Indicated by area temps or ARMs L Unisolable primary system leakage outside drywet as Indicated by area temps or ARMs accordance with the basis for thIs EAL and Is added to R drywell as Indicated by area temperature or area exceeding the Max Safe Limits per EOP 3.Table the IC to ensure consistent understanding of when this O radiation alarm a, when Containment Isolation Is required. IC Is applIcable.

L Primary containment venting perfonred per EOPs OR OR L Intentional venting per EOPs L Primary containment venting per EOPs

Difference or Explanation Last SER Approved E4Lt NEI 99-01 Rev. 4 EAL's Proposed EAU's Dewal-lon PRIMARY CONTAINMENT BARRIER PRIMARY CONTAINMENT BARRIER PRIMARY CONTAINMENT BARRIER

• Consistent with last SER Approved EAL and current NEI PRIMARY CONTAINMENT ATMOSPHERE PRIMARY CONTAINMENT ATMOSPHERE EAL.

L Rapid unexplained decrease following Initial L Rapid unexplained decrease following Initial L Rapid unexplained decrease following initiai

• Added in pressure to ensure clarity.

Increase in pressure Increase Increase In pressure

• DAEC chose to define explosive mixture.

OR OR OR L Drywell pressure response not consistent with L Drywell pressure response not consistent with L Drywall pressure response not consistent with LOCA conditions LOCA conditions LOCA conditions DIFFERENCE P Tomus pressure reaches 53 psig P (Site-specific) PSIG and Increasing P Torus pressure reaches 53 psig and Increasing OR OR OR P Drywall or torus H2 CANNOT be determined to be P Explosive mixture exists P Drywall or Torus H2 cannot be determined to be BELOW 6% AND Drywell or torus O0CANNOT be LESS THAN 6% and Drywall or Torus Oz cannot determined to be BELOW 5% be determined to be LESS THAN 5%

PRIMARY CONTAINMENT BARRIER PRIMARY CONTAINMENT BARRIER PRIMARY CONTAINMENT BARRIER

• Consistent wih last SER Approved EAL and current NEI EMERGENCY DIRECTOR JUDGMENT EMERGENCY DIRECTOR JUDGMENT EMERGENCY DIRECTOR JUDGMENT EAL Any condition which In the EClOSM's Judgment Any condition Inthe opinion of the Emergency Any condition Inthe opinion of the Emergency Indicates loss or potential loss of the primary Director that Indicates Loss or Potential Loss of the Director that indicates Loss or Potential Loss of the NONE containment banfer due to: Containment barrier Containment Banter

• Imminent banter degradation

• Degraded fission banter monitoring capability From SU5 CUt CUI

• Consistent withbcurrent NEI EAL RCS Leakage RCS Leakage RCS Leakage Op. Modes: Run, Startup, Hot SJD Operating Modes: Cold S/D Operating Modes: Cold SID Unidentified or pressure boundary leakage greater Unidentifled or pressure boundary leakage greater CU .1 Unidentified or pressure boundary leakage than 10 GPM. than 10 gpm. GREATER THAN 10 GPM. NONE OR OR OR Identified leakage greater than 25 GPM. Identified leakage greater than 25 gprn. CUI.2 Identified leakage GREATER THAN 25 GPM.

OR Main steam line break as determined from annunciators or plant personnel report.

I

Last SER Approved EAL s NEI 9901 RvK

• EAL's Proposed EAL's Dmnco Deviation Epanatn atnfo CU2 CU2

• Change In format to ensure consistent understanding UNPLANNED Loss of RCS Inventory with Unplanned Loss of RCS Inventory with Irradiated that these are two separate ICs Irradiated Fuel In the RPV. Fuel In the RPV.

• Consistent with current NEI EAL Operating Modes: Refueling Operating Modes: Refueling

• Tank levels not normally monitored by Operations.

For BWR's. the sumps are monitored via sump timers and run times, which are used to calculate rate. For UNPLANNED RCS level decrease BELOW the RPV CU2.1 Unplanned RCS level decrease BELOW the BWR s. sump and Torus level increases would be NEW flange for 115 minutes RPV flange for 15 minutes or longer DIFFERENCE potentially Indicative of RCS level loss If no other OR OR method existed for monitoring RCS lel.

Loss of RPV Inventory as Indicated by unexplained CU2.2 RPV Level cannot be monitored

• tor 15 minutes or longer' Is equivalent to _ 15 (site-specific) sump and tank level Increase AND minutes and minimizes the use of symbols.

AND Loss of RPV Inventory as Indicated by RPV level cannot be monitored unexplained DryweltlReactor Building Equipment or Floor Drain sump, or Torus, level Increase.

From SU1 CU3 CU3

• Consistent with current NEI EAL Loss of All Offsite Power to Essential Busses for Loss of All OffsKe Power to Essential Busses for Loss of Alt Offite Power to Essential Busses for

• Defined emergency busses as 1A3 and 1A4.

Greater Than 15 Minutes Greater Than 15 Minutes. Greater Than 15 Minutes Op Modes: ALL Operating Modes: Cold StD, Refueling Operating Modes: Cold SID, Refueling Loss of Offsite Power Lasting More Than 15 Minutes. Loss of power to (site-speciftic) transfomners for CU3.1 Loss of al otfste power to Emergency Busses DIFFERENCE greater than 15 minutes. 1A3 and 1A4 Is expected to tast forgreater than AND 15 minutes.

At least (site-specific) emergency generators are AND supplying power to emergency busses. At least one Emergency Bus, 1A3 or 1A4, Is powered by ft's Standby Diesel Generator.

CU4 CU4

• Consistent with current NEI EAL UNPLANNED Loss of Decay Heat Removal Unplanned Loss of Decay Heat Removal

• exceeding the Technical Specification cold shutdown Capability with Irradiated Fuel In the RPV. Capability with Irradiated Fuel In the RPV. temperature Itmr Is equivalent to GREATER THAN Operating Modes: Cold S/0, Refueling Operating Modes: Cold StD, Refueling 212 F. This Is DAECs Technical Specification imit.

GREATER THANW Is equivalent to ' and minimizes NEW An UNPLANNED event results In RCS temperature CU4.1 An unplanned event results In RCS DIFFERENCE the use of symbols.

exceeding the Technical Specification cold shutdown temperature GREATER THAN 212 OF temperature limit OR OR CU4.2 Loss of all RCS temperature and RPV level Loss of all RCS temperature and RPV level Indcation Indication for GREATER THAN 15 minutes.

for ' 15 minutes.

I

Last SER Approved E4L's NEI 99-01 Rev. 4 EAl's Proposed EAL's Difference orExpfanon DeviationEpaato From SU4 CUS CU5

• 2.0 pCVgm dose equivalent 1-131 Is the maximum Fuel Clad Degradation Fuel Clad Degradation Fuel Clad Degradation concentation In DAECs Technical Speciftcations.

Op. Modes: Run, Startup. Hot SID Operating Modes: Cold SID, Refueling Operating Modes: Cold StD8Refueling f The calculated radlationtrionotor' setpoIntIndicative of fuel clad degradation greater than TechnIcal Specification allowable limits Is greater than the Valid Pretreat RM-4104 red monitor reading ABOVE (Ste-specific) radiation monitor readings Indicating CU5.1 Reactor Coolant sample activity Indlcating fuel DIFFERENCE radiation i set oinbaIn an r n place vmieAL 4E+3 mRflmr fuel dad degradation greater than Technical dad degradaiion GREATER THAN 2.0 Therefore, EAL RA2 wil bound the expected NEI IC for OR Specification allowable imits dose equivalent 1-131.

pCVgm EAL CU5 requiring a radiation monitor value Indicative CoolantacivityABOVE12 pClmlDOSE OR of fue dad degradation when Inthe Cold S/D or EOUIVALENT 1-131 (Site-spiedifc) coolant sample activity value Indicating R e operati modes. Exposurs Catcslat fuel dad degradation greab fueidaddegradaton greater ler thnthan TSpent Technical No. 04-004-A. Rev.1, 'Radiation Fuel PootARM Exposure 9178 Following Rates at with Gap Release Speciication allowable limits. th Cavity Flooded and DE/ = 2 rrno-Cm7 From SU6 CU6 CU6 Consistent wth current NEI EAL Unplanned Loss of All Onsite or OffsIte UNPLANNED Loss of All Onsite or Offhite Unplanned Loss of All Onolte or OffsIte Communications Capabilities Communications CapabIlities. Communications Capabilities Op. Modes: ALL Operating Modes5 Cold SID, Refueling Operating Modes: Cold SID, Refueling Loss of ALL onsite telephone and radio Loss of all (site-speciflc fst) onsite communications CU6.1 Loss of ALL of the following onsite communication methods (PABX. direct-rIng. UHF, and capability affecting the ability to perform routine communIcation capabilities affecting the ability radiological survey radio systems). operations. to perform routine operation:

OR OR

• Plant Operations Radio System Loss of ALL electronic communication methods with Loss of all (site-spedfle list) offsite communications

• In-Plant Telephones NONE government agencies (PABX. direct-ring. ENS. capablilty.

micaowave and police radio). Plant Paging System OR CU6.2 Loss of ALL ofthe following offsite communications capability:

• Al telephone ilnes (commercial)

• Microwave Phone System

• FTS Phone System From SU? CUT CUT

• Consistent wtth last SER Approved EAL and current NEI EAL Unplanned Loss of Required DC Power During UNPLANNED Loss of Required DC Power for Unplanned Loss of Required DC Power For Cold Shutdown or Refuel Mode For Greater Than Greater than 15 Minutes. Greater Than 15 Minutes

• Slight change Inwording, no change to Intent.

15 MInutes Operating Modes: Cold SID, Refueling Operating Modes: Cold SID, Refuel

• 105 VDC Is the setpoint for a loss of 125 VDC busses.

Op. Modes: Cold SID, Refuel UNPLANNED Loss of Vital DC power to required DC CU7.1 Unplanned Loss of Vital DC power to required DIFFERENCE Unplanned Loss of Div l and Div 2 125 VDC busses busses based on (site-specific) bus voltage DC busses based on bus voltage LESS THAN based on bus voltage less than 105 VDC Indicated. Indications. 105 VDC Indicated.

AND AND AND Failure to restore power to at least one required 125 Failure to restore power to at least one required DC Failure to restore power to at least one VDC bus withIn 15 minutes from time of loss. bus within 15 minutes from the lime of loss. required DC bus within 15 minutes from the time of loss.

j

LastSERApprovedEAL's NEI99-1R v.4EAL's Proposed'EAL's Differenceaort Ixplanation CUe CUS

• Consistent with current NEI EAL Inadvertent Criticality. Inadvertent Criticality.

• Used BWR specific EAL NEW Operating Modes: Cold S/D, Refueling Operating Modes: Cold S/D, RefuelIng DIFFERENCE An UNPLANNED extended positive period observed CU8.1 An unplanned extended positive period on nuclear Instrunentation. observed on nuclear Instrumentation.

CAt CAI

• Consistent wlth current NEI EAL Loss of RCS Inventory. Loss of RCS Inventory.

• GREATER THAN' Is equivalent to' and minimizes Operating Modes: Cold S/D Operating Modes: Cold S/D the use of symbols.

• 119.5 Inches Is DAEC's low-low ECCS actuation Loss of RCS Inventory as Indicated by RPV level less CAl.1 Loss of RCS Inventory as Indicated by RPV setpolnt NEW than {site-speciitc leve). (ow-low ECCS actuation level LESS THAN 119.5 Inches. DIFFERENCE

• Tank levels not normally monitored by Operations. For setpolnt) BWR's. the sumps are monitored via sump tfiers and OR run times. which are used to calculate rate. For ORCA1.2Loss of RCS hventory as indicated by BWR's sump and Torus level Increases would be Loss of RCS Inventory as indicated by unexpaianed unexplaoned DrywentReactor Buindic g potentially Indicative of RCS level loss If no other (site-specitic) sump and tank level Increase and RCS Equipment or Floor Drain sump, or Torus. method existed for monitoring RCS level.

level cannot be monitored for > 15 minutes level Increase and RCS level cannot be monitored for GREATER THAN 15 mInutes CA2 CA2

• 119.5 Inches Is DAEC's low-low ECCS actuation Loss of RPV Inventory with Irradiated Fuel In the Loss of RPV Inventory with Irradiated Fuel In the selpotnt.

RPV RPV

• Consistent with current NEI EAL Operating Modes: Refueling Operating Modes: Refueling * 'GREATER THiAN Is equivalent to >' and minimizes the use of symbols.

as byRPV ndiate DIFERECE Tank levels not normally monitored by Operations. For NEW Loss of RPV Inventory as Indicated by RPV level less CA2.1 Loss of RPV Inventory as by RPV

• Tan hetedicated sumps are monitored via sump timers and than (site-specific level). (low-low ECCS actuation level LESS THAN 119.5 Inches.

seipoint) run times. which are used to calculate rate. For BWR's, OR sump and Torus level Increases would be potentially OR CA22 Loss of RPV Inventory as Indicated by Indicative of RCS level loss If no other method existed Loss of RPV Inventory as Indicated by unexplained unexplained Dryweli/Reador BuidA.g for monitoring RCS evel.

(site-specific) sump and tank level Increase and RPV Equipment or Floor Drain sump, or Torus level cannot be monitored for > 15 minutes level Increase and RPV level cannot be monitored for GREATER THAN 15 mInutes

Last SER Approved EAL's NEI 99-0l Rv.4 EAL's Proposed EAL's Difference Deviationor Explanton From SA1 CA3 CA3

• Consistent with current NEI EAL.

Loss of All Offslte Power and Loss of All Onslte AC Loss of All Offaste Power and Loss of All OnsIte Loss of All Offsite Power and Loss of All Onsfte

• Defined emergency busses as 1A3 and 1A4.

Power to Essential Busses During Cold Conditions AC Power to Essential Busses. AC Power to Essential Busses. DAEC states the emergency busses (ste-specIfic Op. Modes: Cold SID. Refuel, Defueled Operating Modes: Cold SID, Refueling, Defueled Operating Modes: Cold SID, Refueling, Defueled Information) powered by the applicable transformers to minimize any potential confusion whereby the transformer would have power, but the essential bus Loss of Voltage on Buses 1A3 and 1A4 lasting more Loss of powerto (sie-specific) transformers. CA3.1 Loss of alioffsite powerto Emergency Busses would not The Intent of the EAL Is a loss of offbite than 15 mInutes AND 1A3 and 1A4. power to the essential busses, both Diesel Generators Faiiure of (site-specific) emergency generators to AND DIFFERENCE fall, and one essential bus cannot be restored.

supply power to emergency busses. Fal"ure of A Diesel Generator (1G-31) and B AND busses(1G-21)

Diesel Generator emergency 1A3 Wanto 1A4.

supply power to Failure to restore power to at least one emergency AND bus within 15 minutes from the time of loss of both ofsfite and onsite AC power. Failure to restore power to at least one emergency bus. 1A3or1A4, within 15 minutes from the time of loss of both offsite and onsite AC power.

From SA3 CA4 CA4

• 10 pslg owest value that can beread on meter.

Inability to MaIntaIn Plant In Cold Shutdown Inability to MaIntaIn Plant In Cold Shutdown with Inability to MaIntaIn Plant In Cold Shutdown with Reference NEP 2004-0034.

Op. Modes: Cold SID, Refuel Irradiated Fuel In the RPV. Irradiated Fuel In the RPV

• Consistent with current NEI EAL Operating Modes: Cold SID, RefuelIng Operating Modes: Cold SID, Refueling

• Added elther to IC #2 to carify the logic as NEI wording could be misinterpreted.

Loso etrmvlssesrqie ea o With CONTAINMENT CLOSURE and RCS Integrity CA4.1 With Secondary Containment and RCS * 'exceeding the Technical Specification cold shutdown Lomtof decay htrdov t u nol established an UNPLANNED event results InRCS Integrity ntrestablished, an unplanned event temperature lmtr Is equivalent to GREATER THAN miticodsudw.temperature exceeding the Technical Specification results In RCS temperature GREATER THAN 212 FP AND cold shutdown temperature limit 212 IF

• Clarified Containment as being Secondary Temperature rise that exceeds 212 F. OR OR Containment In accordance with NEI guidance for OR With CONTAINMENT CLOSURE established a CA4.2 With Secondary Containment established BWRas RCS Integrity not established gr RCS Inventory either RCS Inlegrity not1 established gr RCS

• Added note from NEI Bases to the EAL to clarify Uncontrolled temperature rise approaching 212F. reduced an UNPLANNED event results In RCS Inventory reduced, an unplanned event results DIFFERENCE expectations, (Note: If an RCS heat removal temperature exceeding the Technical Specification In RCS temperature GREATER THAN 212 IF system Itnoperation within this time frame and cold shutdown temperature limit for greater than 20 for GREATER THAN 20 minutes. (Note: f an RCS temperature Is being reduced then this EAL minutes. Noe: iffan RCS heatremovalsystem is in RCS heet removal system Is In operation Is not appllcable)y operation within this time frame and RCS temperature within this time frame and RCS Is being reduced then this EAL Is not applcable.) temperature Is being reduced then this OR EAL Is not applicable.)

An UNPLANNED event resuits InRCS temperature OR exceeding the Technical Specification cold shutdown CA4.3 An unplanned event results In RCS temperature limit for greater than 60 minutes or temperature GREATER THAN 212 F for results In an RCS pressure Increase of greater than GREATER THAN 60 minutes or results Inan (site specific) psig. (Note ff an RCS heat removai RCS pressure Increase of GREATER THAN system Is Inoperation within this time frame and RCS 10 psg. (Noe: ff an RCS heat removai temperature Is being reduced then this EAL Is not system Is In operation within this time appficable.) frame and RCS temperature Is being reduced then this EAL Is not applicable.)

I

D/ifference or xlnto Last SER Approved EAL's NEI 99-01 Rev. 4 EAL's Proposed EAL's ODevilton Explanation From SS5 CS1 CS1

• 113.5 inches b derved fom subtracting 6inchesfrom Loss of Water Level In the Reactor Vessel That Loss of RPV InventoryAffecting Core Decay Heat Loss of RPV Inventory Affecting Core Decay Heat the low-low ECCS actuation setpoint. At DAEC, the Has or Will Uncover Fuel In the Reactor Vessel Removal Capability. Removal Capability. lowlow ECCS actuation setpointa 119.5 Inches.

Op. Modes: Cold SID. Refuel Operating Modes: Cold SID Operating Modes: Cold S/D

• Consistent with current NEI EAL

• +15 Inches Is DAEC's TOAF NO cooling method lined up or avallable AND RPV 1. With CONTAINMENT CLOSURE iW established: CS1 .1 With Secondary ContainrmentQ nestablshed:

• Clarifted Containment as being Secondary Level BELOW 15 Inhs Containment In accordance with NEI guIdance for

.a RPV inventory as Indicated by RPV a. RPV inventory as indicated by RPV level BWR's.

level less than (site-speciffc level) (6 Is LESS THAN 113.5 Inches BELOW the low-tow ECCS actuation OR* GREATER THAN' Is equivalent to and minimizes setpoint) (BWR) the use of symbois.

OR b. RPV level cannot be monitored for

• Tank levels not normmaliymonitored by Operations.

GREATER THAN 30 minutes with a loss For BWRTsithe sumps are monitoredvia sump timers

b. RPV level cannot be monitored for > of RPV Inventory as Indicated by and run times, which are used to calculate rate. For 30 minutes with a loss of RPV unexplained Drywell/Reactor Building BWRs sump and Torus ievei hcreases would be Inventory as Indicated by unexplalned Equipment or Floor Drain sump, or Torus. potentialy Indicat of RCS level ess w no other citea-spedic) sump and tank level level Increase DIFFERENCE method existed for monitoring RCS level.

Increase OR OR CS1.2 Wth Secondary Containment established:

2. With CONTAINMENT CLOSURE established a RPVlhventoryas hdicatedbyRPV hvei

a. RPV inventory as indicated by RPV Is LESS THAN +15 inches level less than TOAF OR OR b. RPV level cannot be monitored for

b. RPV level cannot be monitored for > GREATER THAN 30 minutes with a loss 30 minutes with a loss of RPV of RPV inventory as Indicated by either Inventory as Indicated by either

• Unexplained Drywell/Reactor

• Unexplained (site. Building Equipment or Floor Drain specific) sump and tank sump, or Torus, level Increase level increase

• Erratic Source Range Monitor

• Erratic Source Range Indication Monitor Indication i

Last SERApprovedcEAL' s NEt 99-01 Rev. 4EAL's ProposedtEAL's lDOee or J Explanation From SS5 CS2 CS2

• 113.5 Inches Is derived from subtracting 6 Inches from Loss of Water Level In the Reactor Vessel That Loss of RPV Inventory Affecting Core Decsy Het Loss of RPV inventory Affecting Core Decay Heat the low-low ECCS actuation setpoint. Al DAEC, the low-Has or Will Uncover Fuel In the Reactor Vessel Removal Capability with Irradiated Fuel In the Removal Capability with Irradiated Fuel In the low ECCS actuation setpoint 1195 Inches.

Op. Modes: Cold SID. Refuel RPV. RPV.

• Consistent with current NEI EAL.

Operating Modes: Refueling Operating Modes: Refueling

• +15 Inches Is DAEC's TOAF.

NO cooling method lined up or available AND RPV

• Clarified Containment as being Secondary Containment Level BELOW 15 Inches. 1. With CONTAINMENT CLOSURE nOiestablished: CS2.1 Wth SECONDARY CONTAINMENT NQI In accordance with NEI guidance for BWRs.

a. RPV Inventory as Indicated by RPV ESTABLISHED, EITHER of the following level less than {site-spedfic level) (6c occurs: * 'GREATER THAN' Is equivalent to ' and minimizes BELOW the Fwlow ECCS actuation (a) RPV Inventory as Indicated by RPV level the use of symbols.

setpoint) Is LESS THAN 113.5 inches

• As water level In the RPV lowers, the dose rate above OR (b) RPV level cannot be monitored with the core Wil increase. The dose rate due to this core Indication of core uncovery as evidenced shine will result In slgnificantly Increased Containment

b. RPV level cannot be monitored with High Range Radiation Monitor readings. An Indication of core uncovery as by one or more of the following:

unexplained reading of greater than 10 Rem/hr may be evidenced by one or more of the

• Containment High Range Rad Indicative offuel damage. The basis for 10 Rem/hr Is followIng: Monitor reading GREATER that it Is suffidently above the normal shutdown levels to THAN 10Rem/hr. avoid an unnecessary entry Into the EAL The 10

• Containment High Range Radiation Monitor reading

• Erratic Source Range Monitor Rem/hr Is also well below the containment radiation

> {slte-spedtic) setpoint monitor reading of 2E+2 R/hr that would be Indicative of Indication 1% dad failure found Inthe following calculation:

• Enatic Source Range OR Monitor Indication Calculation of Drywall Radiation Monitor Reading CS22 With SECONDARY CONTAINMENT DIFFERENCE ESTABLiSHED, EITHER of the following Assurning 1%Gap Release

• Other sie-spedfic)

Indications occurs:

NG-834966 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 (a) RPV Inventory as Indicated by RPV level drywel

• 2.9E+3Renvhr Is LESS THAN +15 Inches OR Drywell reading a 2.gE+3Remlhr x 11% 120 %] =

(b) RPV level cannot be monitored with Indication of core uncovery as evidenced 1A5E+2 Remlhr, round off as 2E+2 Rem/hr

2. With CONTAINMENT CLOSURE established by one or more of the following:

a. RPV inventory as indlicated by RPV

• Containment High Range Rad level less than TOAF Monitor reading GREATER THAN 1ORem/hr.

OR

b. RPV level cannot be monitored with

• Erratic Source Range Monitor Indication of core uncovery as Indication evidenced by one or more of the following:

• Containment High Range Radiation Monitor reading

> (site-spectilc) setpoint

• Erratic Source Range Monitor Indication

• Otherfsite-specIfic)

Indications n I A

Last SER Approved EAL's NE? 99-01 Rev. 4 EAL's J ProposedEAL's Deffermon lEo CGt CG1

• 113.5 Inches h derived from subtracting 6 Inches from Loss of RPV Inventory Affecting Fuel Clad Loss of RPV Inventory Affecting Fuel Clad the low-low ECCS actuation setpoint. At DAEC, the Integrity with Containment Challenged with Integrity with Containment Challenged with low-low ECCS actuation setpoint 119.5 inches.

Irradiated Fuel In the RPV Irradiated Fuel In the RPV

• Consistent with current NEI EAL Operating Modes: Cold S/D, Refueling Operating Modes: Cold SID, Refueling * +15 inches Is DAEC's TOAF.

• Tank levels not normally monitored by Operations.

1. Loss of RPV Inventory as Indicated by CG1.1 For BWR's, the sumps are monitored via sump timers unexplained (site-specific) sump and tank level and run times, which are used to calculate rate. For (1) Loss of RPV Inventory as Indicated by BWR's, sump and Torus ilv increases would be Increase unexplained DrywealReactor Building potentially Indicative of RCS level oss If no other AND Equipment or Floor Drain sump, or Torus, method existed for monitoring RCS level.

level Increase

2. RPV Level:

• DAEC chose to define explosive mixture.

a. less than TOAF for > 30 minutes AND

• Changed format for IC #3 to ensure consistent (2) RPV Level: understanding of logic. No change to Intent.

OR

b. cannot be monitored with Indication of (a) LESS THAN +15 Inches for * 'GREATER THAN' Is equivalent to > and minimizes GREATER THAN 30 minutes the use of symbols.

core uncovery for,> 30 minutes as evidenced by one or more of the OR

• 53 psig Is DAECs primary containment pressure nimit.

folwing:

(b) Cannot be monitored with Indication

• As water level Inthe RPV lowers, the dose rate above

• Containment High Range of core uncovery for GREATER the core will increase. The dose rate due to this core Radiation Monitor reading * (site. THAN 30 minutes as evidenced by shine will result In significantily Increased Containment specifc) selpoint one or more of the following:

NEW DIFFERENCE High Range Radiation Monitor readings. An

• Erratic Source Range Monitor

• Containment High Range unexplained reading of greater than 10 Rem/hr may be Indication Red Monitor reading Indicative of Medamage. The basis for 10 Rem/hr Is GREATER THAN 1ORemlhr. that I Is sufficientiy above the normal shutdown levels to

• Other (site-specific) Indications avoid an unnecessary entry Into the EAL The 10

• Erratic Source Range Monitor Rem/hr Is also well below the containment radiation AND Indication monitor reading of 2E+2 Rfhr that would be Indicative of

3. (Site specific) Indication of CONTAINMENT 1%dad failure found In the following calculation:

AND challenged as Indicated by one or more of the following: (3) Indication of Secondary Containment Calculation of Drywell Radiation Monitor Reading challenged as Indicated by one or more of Assuming 1%Gap Release

• Explosive mixture inside the following:

containment

• Drywall Hydrogen or Torus NG4880966 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

• Pressure ABOVE {site specific) Hydrogen GREATER THAN drywell

• 2.9E.3Rem/hr value 6% AND Drywall Oxygen or

• CONTAINMENT CLOSURE md] Torus Oxygen GREATER Drywell reading

• 2.9E+3Rem/hrx 11% 120 %] -

established THAN 5% 1A5E+2 Rem/hr. round off as 2E+2 Rem/hr

• Secondary Containment radiation

• Containment Pressure monitors ABOVE (site specifc) GREATER THAN 53 psig value (BWR only)

• Secondary Containment n2t established

• Two or more Reactor Building areas exceed Max Safe Radiation Levels

Last SER Approved AL's EUI J NFI 99-01 Rev. 4 EAL's E-HUI Proposed EALs EU1 lDier or Explnation Consistent with last SER Approved EAL and current Damage To A Loaded Cask Confinement Boundary Damage to a loaded cask CONFINEMENT Damage To A Loaded Cask Confinement Boundary NEI EAL BOUNDARY.

• Changed from ECIOSM to Emergency Director to be consistent with NEI.

Operating Modes: Not applicable Operating Modes: Not applicable Operating Modes: Not applicable

• Used EU1 Instead of E-HU1 for consistency on our wallboard.

Any one of the folowing natural phenomena events wth Natural phenomena events affecting a loaded cask EU1.1 Any one of the following natural phenomena CONFINEMENT BOUNDARY.(sftespecfl list)

• Added with resultant visible... In place of esfectlng resultant visible damage to or loss of a baoded cask events with resultant visible damage to or loss to enhance clarity.

confinement boundary OR of a loaded cask confinement boundary:

• Report by plant personnel of a tornado strke. Accident conditlons affecting a loaded cask

• Report by plant personnel of a tomado

• Report by plant personnel of a seismic event CONFINEMENT BOUNDARY.(sIte-pecflc list) strike.

OR

• Report byplant personnel of a selsmIc OR event.

The folowing addent condition wIth resulant visible damage to or loss ofa boded cask confilnement OR boundary. Any condition in the opinion of the Emergency Director that Indicates loss of baded fuel storage cask EU12 The following accident condition with resultant DIFFERENCE

• A loaded transfer cask Is dropped as a result of CONFINEMENT BOUNDARY. visible damage to or loss of a loaded cask normal handling or transporting. confinement boundary OR

• A baded transfer cask dropped as a Any condieon in the opinion of the EC/OSM that result of normal handlng or transpoutng.

Indicates loss of loaded fu storage cask conflnement bou~ndary, OR EU13 Any conditin i the opinlon of the Emergency Director that Indicates loss of baded fue storage cask confinement boundary.

Last SER Approved EAL's NEI 99-01 Rev. 4 EAL's Proposed EAL's Dferenve or Exptanation EU2 E-HU2 EU2

• Consistent with last SER Approved EAL and current Confirmed Security Event With Potential Loss Of Confirmed Security Event with potential loss of Confirmed Secur ty EventWith Potential Loss Of NEI EAL Level Of Safety Of The ISFSI level of safety of the ISFSI. Level Of Safety Of The ISFSI

• Used EU2 Instead of E-HU2 for consistency on our Operating Modes: Not applicable wallboard

• Added *whkhaffects the isFsr to ensure no confusion Operating Modes: Not epplicable Operating Modes: Not applicable exists with HA4.

EU2.1 DAEC Security Supervision reports ANY Of

• Added DAEC Security Supervision as ths position Is Suspected sabotage devise effectng a horizontal Security Event as determined from (site-specific) the flowing: responsible for determining security events.

storage module. dry shielded canister or transfer cask Security Plan and reprvted by the (si-specfic) Suspected sabotage device affecting a . Revised format to be more consistent with NEI.

or found Inside ISFSI Protected aR . security s horizontal storage module, dry shielded Confirmed tampering with a horizontal storage canister or traner cask, or found Inside module, dry shielded canister or transfer cask. ISFSI protected area OR

• Confirmed tampering with a horizontal A hostage sfiuation that disrupbt normal ISFSI storage module, dry shielded canister or operations, transfer cask.

OR ClM disturbance or strike that cspa nomal ISFSI

• A hostage situation that disrupts normal emtkn ISFSI operations.

OR CM disturbance or strike that disrupts DIFFERENCE Internal disturbance that Isnot short Ived or Is not a normal ISFSI operations.

harmless outburst involving one or noe Individuals within the ISFSI protected area.

• Intemal disturbance that Is not short lived OR or is not a harmless outburst Invving Intrusion Into the ISFSI protected area by a hostie one or more indviduals within the ISFSI hteeprotected area.

OR OR 0 Intrusion Into the iSFSI protected area by Any security event of Increasing severity that persists for a hostile force.

230 minutes.

• Credible bomb threats

• Any security event of Increasing severity

• Ethat persists for 30 minutes, or greater,

• Extotious fwhich affects the ISFSI:

• SuspdcousfAre orExplosion

• Significant Security System Hardware Failure o Credible bomb threats

• Loss of Guard Post Contact o HostageExtortion o Suspiclous ire or Explosion o Significant Security System Hardware Falur o Loss of Guard Post Contact

, ENCLOSURE 5 COMPACT DISK OF ENCLOSURES, REFERENCES, AND SUPPORTING DOCUMENTATION Contents Justification Matrix Bases Documents - Index, EAL Supporting Reference Information, Regulatory Context,

-ISFSI Abnormal Events Category, Abnormal Radiation Levels/Radiological Effluent Category, Cold Shutdown/Refueling, Fission Product Barrier Degradation Category, Hazards & Other Conditions Affecting Plant Safety, System Malfunction Category State and Local Government Official Agreement Documentation DAEC 4160 VAC Essential Power Distribution Diagram State and Local Government Official Training Information NEP 2004-0034 Letter for EAL CA4.3 Pressure Indicator Engineering Calculation - CAL-R04-002 "Radiation Exposure Rates at Spent Fuel Pool RE9178 Following Gap Release with the Reactor Cavity Flooded and DEI=2micro-Ci/gm" Plant Chemistry Procedure, PCP 8.6, "Alarm Setpoints and Efficiency For OG Pretreatment" Emergency Operating Procedures Bases Document, "EOP Breakpoints" Annunciator Response Procedure, ARP 1C03A, "Reactor and Containment Cooling and Isolation" EAL Wallchart - Cold EAL Wallchart - Hot Technical Specifications Page 3.4-13 Action Request File Containing; CAP028632, CA029218 and CA029219 I disk follows